APPENDIX F: FACILITY ACCIDENTS F.1 Evaluation Methodologies And Assumptions F.1.1 Introduction The potential for facility accidents and the magnitudes of their consequences are important factors in the evaluation of No Action and tritium supply technologies and recycling facilities addressed in this Programmatic Environmental Impact Statement (PEIS). The health risk issues are twofold and consider: Whether potential accidents for any tritium supply technologies or recycling facility pose unacceptable health risks to workers or the general public; and Whether alternative locations for tritium supply technologies and recycling facilities can provide lesser public or worker health risks. These lesser risks may arise either from a greater isolation of the site from the public, or from a reduced frequency of such external accident initiators as seismic events, aircraft crashes, and other initiating events that are external to the facility. Public comments received during the Draft PEIS reviews clearly indicated the public concern with facility safety and consequent health risks, and the need to address these concerns in the decisionmaking process. F.1.2 Safety Design Process The tritium supply and recycling facilities would be designed to comply with current Federal, state, and local laws, Department of Energy (DOE) orders, and industrial codes and standards. This would provide a plant that is highly resistant to the effects of natural phenomena, including earthquake, flood, tornado, and high wind, as well as credible events as appropriate to the site, such as fire and explosions and man-made threats to its continuing structural integrity for containing hazardous materials. The facilities would be designed to maintain their continuing structural integrity in the event of any credible accident or event, including an aircraft crash, at these sites. The design process for the facilities would comply with the requirements for safety analysis and evaluation in DOE Orders 4700.1 Project Management System and 5480.23 Nuclear Safety Analysis Reports. These require that the safety assessment be an integral part of the design process to ensure compliance with all DOE safety criteria by the time the facilities are constructed and in operation. The safety analysis process begins early in conceptual design with identification of hazards having the potential to produce unacceptable safety consequences to workers or the public. The Preliminary Hazards Assessment determines whether the operations that take place in the facility represent enough of a risk to warrant a Safety Analysis Report. As the design develops, failure mode and effects analyses are performed to identify events that have the potential to release hazardous and/or radioactive material. The kinds of events considered include equipment failure, spills, human error, fire and explosions, criticality, earthquake, electrical storms, tornado, flood, and aircraft crash. These postulated events become focal points for design changes or improvements to prevent unacceptable accidents. These analyses continue as the design progresses to assess the need for safety equipment and to assess the performance of this equipment in accident mitigation. Eventually, the safety analyses are formally documented in a Safety Analysis Report. A detailed comprehensive preliminary Safety Analysis Report is issued upon completion of preliminary design and provides a broad assessment of the range of design-basis accident scenarios and the performance of equipment provided in the facility specifically for accident consequence mitigation. The Safety Analysis Report continues to be developed during detailed design. The review of the Safety Analysis Report is completed and safety issues resolved before initiation of construction of the facility. Final approval of the preliminary Safety Analysis Report is required before construction can commence on the new facility. A Final Safety Analysis Report is also produced that includes documentation of safety-related design changes during construction and the impact of those changes on the safety assessment. It also includes the results of any safety-related research and development that has been performed to support the safety assessment of the facility. Final approval of the Final Safety Analysis Report is required before the facility is allowed to commence operation. F.1.3 Analysis Methodology F.1.3.1 Introduction The GENII computer code was used to estimate the consequences of all tritium supply and recycling facilities design-basis accidents. For beyond design-basis accidents at tritium production facilities, which include reactors, accelerators, and support facilities, the MACCS computer code was used. A discussion of the GENII code is provided in appendixE. A discussion of the MACCS computer code is provided in section F.1.3.2. A detailed description of the model is available in a 3-volume report: MELCOR Accident Consequence Code System (MACCS) (NUREG/CR-4691 SAND 86-1562). F.1.3.2 MELCOR Accident Consequence Code System Overview The MACCS computer code models the offsite consequences of an accident that releases a plume of radioactive materials to the atmosphere. Should such an accidental release occur, the radioactive gases and aerosols in the plume would be transported by the pre- vailing wind while dispersing in the atmosphere. The environment would be contaminated by radioactive materials deposited from the plume and the population would be exposed to radiation. An estimation of the range and probability of the health effects induced by the radiation exposures not avoided by protective actions and the economic costs and losses that would result from the contamination of the environment are the objectives of a MACCS calculation. There are two fundamental aspects of the organization of MACCS which are basic to its understanding: the time scale after the accident is divided into various "phases" and the region surrounding the reactor is divided into a polar-coordinate grid. The time scale after the accident is divided into three phases: emergency phase, intermediate phase, and long-term phase. The emergency phase begins immediately after the accident and could last up to 7 days following the accident. In this period, the exposure of population to both radioactive clouds and contaminated ground is modeled. Various protective measures can be specified for this phase, including evacuation, sheltering, and dose-dependent relocation. The intermediate phase can be used to represent a period in which evaluations are performed and decisions are made regarding the type of protective actions which need to be taken. In this period, the radioactive clouds are assumed to be gone and the only exposure pathways are those from the contaminated ground. The protective measure which can be taken during this period is temporary relocation. The long-term phase represents all time subsequent to the intermediate phase. The only exposure pathways considered here are those resulting from the contaminated ground. A variety of protective measures can be taken in the long-term phase in order to reduce doses to acceptable levels: decontamination, interdiction, and condemnation of property. The spatial grid used to represent the region is centered on the facility itself. The user specifies the number of radial divisions as well as their endpoint distances. Up to 35 of these divisions may be defined, extending out to a maximum distance of 6,200 miles (9,999 kilometers). The angular divisions used to define the spatial grid correspond to the 16 directions of the compass. The emergency phase calculations utilizing dose-response models for early fatality and early injury are performed on a finer grid than the calculations of the intermediate and long-term phases. For this phase, the 16 compass sectors are divided into three, five, or seven user-specified subdivisions in the calculations. The increased likelihood of latent cancer fatality to a member of the public is taken as 5.0x10-4 times the dose in rem for values of dose less than 20 rem. For larger doses, when the rate of exposure is greater than 10 rads per hour, the increased likelihood of latent cancer fatality is doubled. MACCS incorporates this by assuming that the rate of exposure during the accident emergency phase is greater than 10 rads per hour if the individual dose received during this phase is greater than 20 rem. Subsequent to the emergency phase (intermediate and long-term phases) the exposure rate is assumed to be less than 10 rads per hour (NUREG/CR- 6059, SAND92-2146:3). The MACCS code was applied in a probabilistic manner using a weather bin sampling technique. Centerline doses as a function of distance were calculated for each of 150 meteorological sequence samples; the mean value of these doses and increased likelihoods of cancer fatality for the distance corresponding to the location of the maximum offsite individual at each site were reported for that individual. Doses to uninvolved workers were calculated similarly, except that these workers will experience an increased likelihood of cancer fatality of 4.0x10-4 times the dose in rem for doses less than 20 rem or exposure rates less than 10rads per hour. For larger doses, when the rate of exposure is greater than 10 rads per hour, the increased likelihood of latent cancer fatality is doubled. The workers were placed at 1,000 and 2,000 meters from the release. It should be noted that since the doses and cancer fatalities for the maximum offsite individual and the workers reported in the high consequence/low probability accident tables are "mean" values based on 150 meteorological sequence samples, there is no direct correlation between the mean value of dose and the mean value of cancer fatalities. For example; high mean doses, in excess of 1,000 rad for the maximum offsite individual or 1,250 rad for workers during the emergency phase of an accident, will not result in an increased likelihood of cancer fatality mean value of 1.0 unless the individual doses resulting from all 150 meteorological sequence samples exceeds he emergency phase threshold values of 1,000 rad for the maximum exposed individual or 1,250 rad for the worker. Offsite population doses and latent cancer fatalities are calculated by MACCS using a similar methodology to that described for the maximum offsite individual. In the case of the population, each of the sampled meteorological sequences was applied to each of the 16 sectors (accounting for the frequency of occurrence of the wind blowing in that direction). Population doses are the sum of the individual doses in each sector. Once again, the mean value of the calculated population doses and latent cancer fatalities for each of these trials are reported. F.1.3.3 Application to Tritium Production For the analysis of high consequence accidents at tritium supply facilities, the MACCS calculations used the source term data presented in section F.2.1 and modeled the dispersion and deposition of radionuclides released from the reactor or accelerator containments to the atmosphere with a straight-line Gaussian plume. Plume rise and dry and wet deposition were taken into consideration. One year of hourly onsite meteorological data and a weather bin sampling technique were used to represent the dispersion process according to each site's characteristic weather. Downwind concentrations of radionuclides up to a distance of 50 miles (80 kilometers) were calculated for each of 16 directional sectors around the reactor or accelerator. Radiation doses to an offsite population were calculated in the dosimetry models using the concentrations of radionuclides obtained from the dispersion models. Dose conversion factors were used to convert the radionuclide concentrations to organ dose equivalents and whole-body effective dose equivalents. Exposure pathways considered in the MACCS for calculating doses received during the period following an accident were direct radiation from the passing plume and from radioactive material deposited on the ground, inhalation from the plume, deposition on skin, and inhalation of resuspended ground contamination. Long-term exposure pathways and liquid exposure pathways were not considered. No credit was taken for short-term actions such as evacuation, sheltering, and relocation. F.2 Tritium Supply and Recycling Accidents The tritium supply facility can be configured as a reactor or as an accelerator. The reactor configuration includes the reactor, reactor fuel/target fabrication facilities, and target extraction facilities. The Heavy Water Reactor (HWR), Modular High Temperature Gas-Cooled Reactor (MHTGR), and Advanced Light Water Reactor (ALWR) are candidate reactor technologies for tritium supply. Four ALWR configurations; the AP600, Simplified Boiling Water Reactor, Advanced Boiling Water Reactor, and CE System 80+; are under consid- eration for the ALWR tritium supply technology. The candidate ALWR configurations have been classified into two groups, Large ALWRs and Small ALWRs. The Advanced Boiling Water Reactor and CE System 80+ configurations are designated Large ALWRs and the AP600 and Simplified Boiling Water Reactor configurations are designated Small ALWRs. The Acceler- ator Production of Tritium (APT) facility configuration is associated with the linear accelerator and target areas of the facility. Two target designs are under consider- ation, the helium-3 target system and the spallation-induced lithium conversion target system. For the helium-3 target design, tritium would be continuously removed from the target and packaged without any additional target processing. For the spallation-induced lithium conversion target design, production targets will be processed at a tritium recycling facility collocated with the APT. The tritium recycling facility design and operation is similar for all reactor technologies and the spallation induced lithium conversion target system. F.2.1 Tritium Supply Facility High Consequence Accidents High consequence accidents for candidate tritium supply technologies and recycling facilities at five potential sites, (Idaho National Engineering Laboratory (INEL), Nevada Test Site (NTS), Oak Ridge Reservation (ORR), Pantex Plant (Pantex), and Savannah River Site (SRS)), have been evaluated using the MACCS computer code. The MACCS computer code is described in section F.1.3.3. The report, MELCOR Accident Consequence Code System, presents additional details on the computer code. F.2.1.1 Heavy Water Reactor Previous studies performed for the HWR developed a spectrum of severe accidents and their respective source terms (DOE 1995d). The release frequencies were in the range of 1.0x10-8 to 2.0x10-6 per reactor year. In order to provide a reasonably similar basis for compari- sons, five accidents with an annual frequency of occurrence equal to or greater than 1.0x10-7 were selected for evaluation in this PEIS. The selected combination of release category and frequency is representative of accident conditions at the low frequency end of the credible range for beyond design-basis accidents. Core Melt with Containment Spray System and Containment Functioning Scenario. The HWR high consequence accident postulated an internally initiated core melt event. The containment spray system functioned and the containment did not fail (DOE 1995d). The source term is presented in table F.2.1.1-1. The annual frequency of occurrence for this accident is 5.0x10-6 per year (DOE 1995d). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.1-2 through F.2.1.1-6 for public consequences and in tables F.2.1.1-7 through F.2.1.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.1-1 using the MACCS computer code. Seismically Induced Core Melt with Containment Spray System Failure and Containment Functioning Scenario. The HWR high consequence accident postulated a seismically induced core melt event. The containment spray system failed but the containment did not fail (DOE 1995d). The source term is presented in table F.2.1.1-1. The annual frequency of occurrence for this accident is 2.0x10-6 per year (DOE 1995d). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.1-2 through F.2.1.1-6 for public consequences and in tables F.2.1.1-7 through F.2.1.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.1-1 using the MACCS computer code. Core Melt with Containment Spray System Failure and Containment Functioning Scenario. The HWR high consequence accident postulated an internally initiated core melt event. The containment spray system failed but the containment did not fail (DOE 1995d). The source term is presented in table F.2.1.1-1. The annual frequency of occurrence for this accident is 2.0x10-6 per year (DOE 1995d). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.1-2 through F.2.1.1-6 for public consequences and in tables F.2.1.1-7 through F.2.1.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.1-1 using the MACCS computer code. Seismically Induced Core Melt with Containment Spray System Failure and Early Containment Failure Scenario. The HWR high consequence accident postulated a seismically induced core melt event. The containment spray system failed and the containment failed early (DOE 1995d). The source term is presented in table F.2.1.1-1. The annual frequency of occurrence for this accident is 1.0x10-7 per year (DOE 1995d). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.1-2 through F.2.1.1-6 for public consequences and in tables F.2.1.1-7 through F.2.1.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.1-1 using the MACCS computer code. Core Melt with Early Containment Spray System and Containment Failure Scenario. The HWR high consequence accident postulated an internally initiated core melt event. The containment spray system and the containment failed early (DOE 1995d). The source term is presented in table F.2.1.1-1. The annual frequency of occurrence for this accident is 1.0x10-7 per year (DOE 1995d). Consequences. The estimated consequences of the postulated accident with at each site are shown in tables F.2.1.1-2 through F.2.1.1-6 for public consequences and in tables F.2.1.1-7 through F.2.1.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.1-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Five Heavy Water Reactor High Consequence Accidents Figure F.2.1.1-1 shows the annual probability that, in the event of any accident in the composite set of HWR high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.1-2 through F.2.1.1-11. Figure (Page F-5) Figure F.2.1.1-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for the Heavy Water Reactor. Table F.2.1.1-1.-Heavy Water Reactor High Consequence Accident Source Terms [Page 1 of 5] Isotope Released Activity (curies) - Core Melt with Containment Seismically Induced Core Core Melt with Containment Seismically Induced Core Core Melt with Early Spray System and Melt with Containment Spray Spray System Failure and Melt with Containment Spray Containment Spray System Containment Functioning System Failure and Containment Functioning System Failure and Early and Containment Failure Containment Functioning Containment Failure H-3 3.8x107 3.8x107 3.8x107 3.8x107 3.8x107 Se-84 0.071 14 14 7.1x104 7.1x104 Se-85 0.045 9 9 4.5x104 4.5x104 Se-86 0.094 19 19 9.4x104 9.4x104 Se-87 0.07 14 14 7.0x104 7.0x104 Br-84 0.22 36 36 2.2x104 2.2x104 Br-85 0.27 45 45 2.7x104 2.7x104 Br-86 0.2 34 34 2.0x104 2.0x104 Br-86m 0.21 34 34 2.1x104 2.1x104 Br-87 0.47 78 78 4.7x104 4.7x104 Br-88 0.53 89 89 5.3x104 5.3x104 Br-89 0.4 67 67 4.0x104 4.0x104 Br-90 0.27 45 45 2.7x104 2.7x104 Kr-85 6.0x102 6.0x102 6.0x102 2.0x104 2.0x104 Kr-85m 2.7x104 2.7x104 2.7x104 9.1x106 9.1x106 Kr-87 5.5x104 5.5x104 5.5x104 1.8x107 1.8x107 Kr-88 7.8x104 7.8x104 7.8x104 2.6x107 2.6x107 Kr-89 9.9x104 9.9x104 9.9x104 3.3x107 3.3x107 Kr-90 9.8x104 9.8x104 9.8x104 3.3x107 3.3x107 Kr-91 7.3x104 7.3x104 7.3x104 2.4x107 2.4x107 Kr-92 3.2x104 3.2x104 3.2x104 1.1x106 1.1x106 Rb-88 0.79 130 130 7.9x104 7.9x104 Rb-89 1 170 170 1.0x106 1.0x106 Rb-90 1 170 170 1.0x106 1.0x106 Rb-90m 0.21 35 35 2.1x104 2.1x104 Rb-91 1.2 200 200 1.2x106 1.2x106 Rb-92 1 170 170 1.0x106 1.0x106 Rb-93 0.76 130 130 7.6x104 7.6x104 Rb-94 0.38 63 63 3.8x104 3.8x104 Rb-95 0.19 32 32 1.9x104 1.9x104 Sr-89 0.35 110 110 3.5x104 3.5x104 Sr-90 0.016 4.8 4.8 1.6x104 1.6x104 Sr-91 0.42 130 130 4.2x104 4.2x104 Sr-92 0.43 130 130 4.3x104 4.3x104 Sr-93 0.46 140 140 4.6x104 4.6x104 Sr-94 0.42 130 130 4.2x104 4.2x104 Sr-95 0.39 120 120 3.9x104 3.9x104 Sr-96 0.27 81 81 2.7x104 2.7x104 Sr-97 0.14 41 41 1.4x104 1.4x104 Sr-98 0.05 15 15 5.0x104 5.0x104 Y-90 8.4x10-6 1.7x10-3 1.7x10-3 8.4 8.4 Y-91 2.1x10-4 0.043 0.043 210 210 Y-91m 1.2x10-4 0.024 0.024 120 120 Y-92 2.1x10-4 0.043 0.043 210 210 Y-93 2.3x10-4 0.046 0.046 230 230 Y-94 2.2x10-4 0.045 0.045 220 220 Y-95 2.3x10-4 0.046 0.046 230 230 Y-96 2.2x10-4 0.043 0.043 220 220 Y-97 1.8x10-4 0.036 0.036 180 180 Y-98 1.3x10-4 0.026 0.026 130 130 Y-99 7.2x10-4 0.014 0.014 72 72 Y-100 3.2x10-4 6.3x10-3 6.3x10-3 32 32 Zr-95 2.4x10-4 0.047 0.047 240 240 Zr-97 2.1x10-4 0.042 0.042 210 210 Zr-98 2.1x10-4 0.042 0.042 210 210 Zr-99 2.1x10-4 0.041 0.041 210 210 Zr-100 1.9x10-4 0.038 0.038 190 190 Zr-101 1.2x10-4 0.025 0.025 120 120 Zr-102 6.4x10-5 0.013 0.013 64 64 Ru-103 1.2x10-4 0.024 0.024 120 120 Ru-105 3.9x10-5 7.8x10-3 7.8x10-3 39 39 Ru-106 1.0x10-5 2.1x10-3 2.1x10-3 10 10 Rh-103m 1.1x10-4 0.021 0.021 110 110 Rh-104 4.7x10-5 9.4x10-3 9.4x10-3 47 47 Rh-105 3.7x10-5 7.3x10-3 7.3x10-3 37 37 Rh-106 1.2x10-5 2.4x10-3 2.4x10-3 12 12 Sb-129 2.5x10-5 5.0x10-3 5.0x10-3 25 25 Sb-130m 4.5x10-5 9.0x10-3 9.0x10-3 45 45 Sb-131 9.3x10-5 0.019 0.019 93 93 Sb-132 6.0x10-5 0.012 0.012 60 60 Sb-132m 3.9x10-5 7.8x10-3 7.8x10-3 39 39 Sb-133 8.1x10-5 0.016 0.016 81 81 Te-129 0.05 9.9 9.9 5.0x104 5.0x104 Te-129m 7.6x10-3 1.5 1.5 7.6x102 7.6x102 Te-131 0.19 37 37 1.9x104 1.9x104 Te-132 0.31 62 62 3.1x104 3.1x104 Te-133 0.28 56 56 2.8x104 2.8x104 Te-133m 0.21 42 42 2.1x104 2.1x104 Te-134 0.48 96 96 4.8x104 4.8x104 Te-135 0.24 47 47 2.4x104 2.4x104 Te-136 0.13 27 27 1.3x104 1.3x104 Sn-130 3.4x10-5 6.8x10-3 6.8x10-3 34 34 Sn-131 3.5x10-5 7.0x10-3 7.0x10-3 35 35 Sn-132 2.1x10-5 4.2x10-3 4.2x10-3 21 21 I-131 0.63 110 110 6.3x104 6.3x104 I-132 0.93 160 160 9.3x104 9.3x104 I-133 1.5 240 240 1.5x106 1.5x106 I-134 1.6 270 270 1.6x106 1.6x106 I-135 1.4 230 230 1.4x106 1.4x106 I-136 0.66 110 110 6.6x104 6.6x104 I-136m 0.42 70 70 4.2x104 4.2x104 I-137 0.7 120 120 7.0x104 7.0x104 I-138 0.36 60 60 3.6x104 3.6x104 I-139 0.16 26 26 1.6x104 1.6x104 Xe-133 1.4x106 1.4x106 1.4x106 4.8x107 4.8x107 Xe-135 3.5x104 3.5x104 3.5x104 1.2x107 1.2x107 Xe-135m 2.5x104 2.5x104 2.5x104 8.2x106 8.2x106 Xe-137 1.3x106 1.3x106 1.3x106 4.3x107 4.3x107 Xe-138 1.3x106 1.3x106 1.3x106 4.5x107 4.5x107 Xe-139 1.1x106 1.1x106 1.1x106 3.6x107 3.6x107 Xe-140 7.6x104 7.6x104 7.6x104 2.5x107 2.5x107 Xe-141 2.5x104 2.5x104 2.5x104 8.4x106 8.4x106 Cs-134 0.043 7.1 7.1 4.3x104 4.3x104 Cs-137 0.05 8.4 8.4 5.0x104 5.0x104 Cs-138 1.4 240 240 1.4x106 1.4x106 Cs-139 1.4 230 230 1.4x106 1.4x106 Cs-140 1.2 210 210 1.2x106 1.2x106 Cs-141 0.95 160 160 9.5x104 9.5x104 Cs-142 0.61 100 100 6.1x104 6.1x104 Cs-143 0.33 55 55 3.3x104 3.3x104 Ba-137m 0.016 4.7 4.7 1.6x104 1.6x104 Ba-139 0.46 140 140 4.6x104 4.6x104 Ba-140 0.46 140 140 4.6x104 4.6x104 Ba-141 0.42 130 130 4.2x104 4.2x104 Ba-142 0.42 120 120 4.2x104 4.2x104 Ba-143 0.38 110 110 3.8x104 3.8x104 Ba-144 0.3 91 91 3.0x104 3.0x104 Ba-145 0.14 43 43 1.4x104 1.4x104 Ba-146 0.048 15 15 4.8x104 4.8x104 La-140 2.3x10-4 0.047 0.047 230 230 La-141 2.1x10-4 0.042 0.042 210 210 La-142 2.1x10-4 0.042 0.042 210 210 La-143 2.1x10-4 0.042 0.042 210 210 La-144 1.9x10-4 0.038 0.038 190 190 La-145 1.3x10-4 0.026 0.026 130 130 La-146 8.3x10-5 0.017 0.017 83 83 La-147 3.8x10-5 7.7x10-3 7.7x10-3 38 38 Ce-141 2.2x10-4 0.043 0.043 220 220 Ce-143 2.1x10-4 0.043 0.043 210 210 Ce-144 1.5x10-4 0.029 0.029 150 150 Ce-145 1.4x10-4 0.028 0.028 140 140 Ce-146 1.1x10-4 0.021 0.021 110 110 Ce-147 8.0x10-5 0.016 0.016 80 80 Ce-148 5.5x10-6 1.1x10-3 1.1x10-3 5.5 5.5 Ce-149 2.8x10-5 5.6x10-3 5.6x10-3 28 28 Pr-143 2.1x10-4 0.042 0.042 210 210 Pr-144 1.5x10-4 0.029 0.029 150 150 Pr-144m 1.7x10-6 3.5x10-4 3.5x10-4 1.7 1.7 Pr-145 1.4x10-4 0.028 0.028 140 140 Pr-146 1.1x10-4 0.021 0.021 110 110 Pr-147 8.2x10-5 0.016 0.016 82 82 Pr-148 6.1x10-5 0.012 0.012 61 61 Pr-149 3.9x10-5 7.7x10-3 7.7x10-3 39 39 Pr-150 2.2x10-5 4.3x10-3 4.3x10-3 22 22 Nd-147 8.4x10-5 0.017 0.017 84 84 Nd-149 4.0x10-5 8.0x10-3 8.0x10-3 40 40 Pm-147 1.8x10-6 3.6x10-4 3.6x10-4 1.8 1.8 Pm-148 2.8x10-5 5.7x10-3 5.7x10-3 28 28 Pm-148m 3.5x10-6 7.1x10-4 7.1x10-4 3.5 3.5 Pm-149 4.6x10-5 9.1x10-3 9.1x10-3 46 46 Sm-153 2.5x10-5 5.1x10-3 5.1x10-3 25 25 U-234 3.0x10-10 6.0x10-8 6.0x10-8 3.0x10-4 3.0x10-4 U-237 9.9x10-5 0.02 0.02 99 99 U-239 5.5x10-5 0.011 0.011 55 55 Np-238 1.9x10-5 3.7x10-3 3.7x10-3 19 19 Np-239 5.5x10-5 0.011 0.011 55 55 Pu-238 6.3x10-8 1.3x10-5 1.3x10-5 0.063 0.063 Pu-239 8.0x10-8 1.6x10-7 1.6x10-7 8.0x10-4 8.0x10-4 Pu-240 5.0x10-10 9.9x10-8 9.9x10-8 5.0x10-4 5.0x10-4 Pu-241 2.3x10-7 4.7x10-5 4.7x10-5 0.23 0.23 Pu-243 4.0x10-7 8.0x10-5 8.0x10-5 0.4 0.4 Cm-242 3.7x10-8 7.4x10-6 7.4x10-6 0.037 0.037 Cm-244 6.7x10-10 1.3x10-7 1.3x10-7 6.7x10-4 6.7x10-4 Note: Tritium source term based on an assumed production design goal of 32 million curies, tritium inventory in coolant of 6 million curies, and a release fraction of 1. Se is assumed to have the Te release fraction. Br is assumed to have the I release fraction. Sn, Pm, Sm, and U are assumed to have the "Other" release fraction. Source: Source term derived from core inventory and accident release fractions (DOE 1995d: HNUS 1995c:3). Table F.2.1.1-2.-Heavy Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Core melt with containment spray system and containment functioning 0.35 1.8x10-4 371 0.19 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 0.36 1.8x10-4 394 0.2 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 0.36 1.8.x10-4 394 0.2 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 41 0.025 1.3x104 64 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 41 0.025 1.3x104 64 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 7.1x10-4 - 1.6 - Expected risk for composite set of accidents (per year) - 6.5x10-9 - 1.4x10-5 - Table F.2.1.1-3.-Heavy Water Reactor High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Core melt with containment spray system and containment functioning 0.9 4.5x10-4 36 0.018 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 0.91 4.6x10-4 38 0.019 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 0.91 4.6x10-4 38 0.019 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 117 0.071 1.2x104 6.1 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 117 0.071 1.2x104 6.1 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 2.0x10-3 - 0.15 - Expected risk (per year) - 1.8x10-8 - 1.4x10-6 - Table F.2.1.1-4.-Heavy Water Reactor High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Core melt with containment spray system and containment functioning 6.2 3.1x10-3 3.8x103 1.9 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 6.4 3.2x10-3 4.0x103 2 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 6.4 3.2x10-3 4.0x103 2 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 1.0x103 0.54 9.9x105 496 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 1.0x103 0.54 9.9x105 496 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 0.015 - 13 - Expected risk (per year) - 1.4x10-7 - 1.2x10-4 - Table F.2.1.1-5.-Heavy Water Reactor High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Core melt with containment spray system and containment functioning 4 2.0x10-3 480 0.24 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 4.1 2.1x10-3 504 0.25 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 4.1 2.1x10-3 504 0.25 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 685 0.38 1.3x105 65 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 685 0.38 1.3x105 65 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 0.01 - 1.7 - Expected risk (per year) - 9.5x10-8 - 1.5x10-5 - Table F.2.1.1-6.-Heavy Water Reactor High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Core melt with containment spray system and containment functioning 0.29 1.5x10-4 1.4x103 0.71 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 0.3 1.5x10-4 1.5x103 0.75 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 0.3 1.5x10-4 1.5x103 0.75 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 44 0.024 4.4x105 222 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 44 0.024 4.4x105 222 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 6.6x10-4 - 5.5 - Expected risk (per year) - 6.0x10-9 - 5.1x10-5 - Table F.2.1.1-7.-Heavy Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Core melt with containment spray system and containment functioning 32 0.018 13 5.6x10-3 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 33 0.018 13 5.7x10-3 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 33 0.018 13 5.7x10-3 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 5.6x103 0.77 2.0x103 0.55 - failure Core melt with early containment spray system failure and containment failure 5.6x103 0.77 2.0x103 0.55 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 0.034 - 0.017 - Expected risk (per year) - 3.2x10-7 - 1.6x10-7 - Table F.2.1.1-8.-Heavy Water Reactor High Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Core melt with containment spray system and containment functioning 23 0.012 9.5 4.1x10-3 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 24 0.013 9.8 4.3x10-3 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 24 0.03 9.8 4.3x10-3 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 4.3x103 0.86 1.6x103 0.52 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 4.3x103 0.86 1.6x103 0.52 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 0.034 - 0.015 - Expected risk (per year) - 3.2x10-7 - 1.4x10-7 - Table F.2.1.1-9.-Heavy Water Reactor High Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Core melt with containment spray system and containment functioning 31 0.017 12 5.3x10-3 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 32 0.018 12 5.4x10-3 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 32 0.018 12 5.4x10-3 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 6.1x103 0.82 2.1x103 0.63 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 6.1x103 0.82 2.1x103 0.63 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 0.035 - 0.019 - Expected risk (per year) - 3.2x10-7 - 1.4x10-7 - Table F.2.1.1-10.-Heavy Water Reactor High Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Core melt with containment spray system and containment functioning 14 6.1x10-3 5.5 2.3x10-3 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 14 6.3x10-3 5.7 2.3x10-3 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 14 6.3x10-3 5.7 2.3x10-3 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 2.6x103 0.81 974 0.39 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 2.6x103 0.81 974 0.39 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 0.024 - 0.011 - Expected risk (per year) - 2.2x10-7 - 1.0x10-7 - Table F.2.1.1-11.-Heavy Water Reactor High Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Core melt with containment spray system and containment functioning 14 6.4x10-3 5.5 2.3x10-3 5.0x10-6 Seismically induced core melt with containment spray system failure and containment 14 6.6x10-3 5.7 2.4x10-3 2.0x10-6 functioning Core melt with containment spray system failure and containment functioning 14 6.6x10-3 5.7 2.4x10-3 2.0x10-6 Seismically induced core melt with containment spray system failure and containment 2.7x103 0.74 983 0.37 1.0x10-7 failure Core melt with early containment spray system failure and containment failure 2.7x103 0.74 983 0.37 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 0.023 - 0.01 - Expected risk (per year) - 2.1x10-7 - 9.5x10-8 - F.2.1.2 Modular High Temperature Gas-Cooled Reactor Previous studies performed for the MHTGR developed a spectrum of severe accidents and their respective source terms. The release frequencies were in the range of 1.0x10-9 to 6.0x10-6 per reactor year (DOE 1995e). In order to provide a reasonably similar basis for comparisons with other technologies, four accidents with an annual frequency of occurrence greater than 1.0x10-7 were selected for evaluation in this PEIS. The selected combination of release category and frequency is representative of accident conditions at the low frequency end of the credible range for beyond design-basis accidents. Depressurized Conduction Cooldown with Reactor Cavity Cooling System Functioning Scenario. The MHTGR high consequence accident postulated a depressurized reactor cooldown event. The reactor cavity cooling system was functioning and containment leak rate was 100 percent per day. The source term is presented in table F.2.1.2-1. The annual frequency of occurrence for this accident is 6.0x10-6 per year (DOE 1995e). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.2-2 through F.2.1.2-6 for public consequences and in tables F.2.1.2-7 through F.2.1.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.1-1 or F.2.1.2-1 using the MACCS computer code. Depressurized Conduction Cooldown Without Reactor Cavity Cooling System Functioning Scenario. The MHTGR high consequence accident postulated a depressurized reactor cooldown event. The reactor cavity cooling system was not functioning and containment leak rate was 1 percent per day. The source term is presented in table F.2.1.2-1. The annual frequency of occurrence for this accident is 6.0x10-6 per year (DOE 1995e). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.2-2 through F.2.1.2-6. The dose estimates are based on analysis of the source terms in table F.2.1.2-1 using the MACCS computer code. Air Ingress Scenario. The MHTGR high consequence accident postulated an air ingress event with the containment leak rate at 100 percent per day. The source term is presented in table F.2.1.2-1. The annual frequency of occurrence for this accident is 2.0x10-6 per year (DOE 1995e). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.2-2 through F.2.1.2-6 for public consequences and in tables F.2.1.2-7 through F.2.1.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.2-1 using the MACCS computer code. Moisture Ingress Scenario. The MHTGR high consequence accident postulated a moisture ingress event with the containment leak rate at 1 percent per day. The source term is presented in table F.2.1.2-1. The annual frequency of occurrence for this accident is 2.0x10- 6 per year (DOE 1995e). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.2-2 through F.2.1.2-6 for public consequences and in tables F.2.1.2-7 through F.2.1.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.2-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Four Modular High Temperature Gas-Cooled Reactor High Consequence Accidents Figure F.2.1.2-1 shows the annual probability that, in the event of any accident in the composite set of MHTGR high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence, as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.2-2 through F.2.1.2-11. Figure (Page F-17) Figure F.2.1.2-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for the Modular High Temperature Gas-Cooled Reactor Table F.2.1.2-1.-Modular High Temperature Gas-Cooled Reactor High Consequence Accident Source Terms [Page 1 of 2] Isotope Released Activity (curies) - Depressurized Conduction Depressurized Conduction Air Ingress Moisture Ingress Cooldown with Reactor Cavity Cooldown without Reactor (leakage 100 percent per day) (leakage 1 percent per day) Cooling System Functioning Cavity Cooling System (leakage 100 percent per day) Functioning (leakage 1 percent per day) H-3 7.3x105 2.8x106 7.3x105 2.8x106 Kr-85 0.014 0.014 0.014 0.014 Kr-85m 1.3 0.017 1.3 0.017 Kr-87 0.89 9.6x10-3 0.89 9.6x10-3 Kr-88 2.7 0.031 2.7 0.031 Rb-86 1.6 0.028 1.6 0.028 Sr-89 1.2x103 52 1.2x103 52 Sr-90 74 3.1 74 3.1 Sr-91 2.4 0.69 2.4 0.69 Y-90 34 1.9 34 1.9 Y-91 1.5x103 65 1.5x103 65 Zr-95 1.7x103 72 1.7x103 72 Zr-97 41 4.9 41 4.9 Nb-95 1.6x103 7.0x103 1.6x103 7.0x103 Mo-99 660 36 660 36 Tc-99m 0.067 0.59 0.067 0.59 Ru-103 960 41 960 41 Ru-105 6.2x10-4 2.0x10-3 6.2x10-4 2.0x10-3 Ru-106 69 3 69 3 Rh-105 70 4.8 70 4.8 Sb-127 28 1.4 28 1.4 Sb-129 5.3x10-8 3.9x10-7 5.3x10-8 3.9x10-7 Te-127 9.9 0.17 9.9 0.17 Te-127m 0.81 0.014 0.81 0.014 Te-129 0.076 8.1x10-4 0.076 8.1x10-4 Te-129m 2.1 0.037 2.1 0.037 Te-131m 1.2 0.019 1.2 0.019 Te-132 80 1.4 80 1.4 I-131 28 0.48 28 0.48 I-132 21 0.23 21 0.23 I-133 15 0.22 15 0.22 I-134 0.51 5.3x10-3 0.51 5.3x10-3 I-135 5 0.064 5 0.064 Xe-133 2.5 0.2 2.5 0.2 Xe-135 2.1 0.033 2.1 0.033 Cs-134 250 9.1 250 9.1 Cs-136 170 6.5 170 6.5 Cs-137 110 4 110 4 Ba-140 1.1x103 48 1.1x103 48 La-140 430 28 430 28 Ce-141 1.5x103 66 1.5x103 66 Ce-143 240 17 240 17 Ce-144 540 24 540 24 Pr-143 1.3x103 58 1.3x103 58 Nd-147 510 27 510 23 Source: DOE 1995e calculated using the source terms in table F.2.1.1-1 and the MACCS computer code. Table F.2.1.2-2.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 0.17 8.5x10-5 554 0.28 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling system 0.065 3.2x10-5 170 0.085 6.0x10-6 functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 0.17 8.5x10-5 554 0.28 2.0x10-6 Moisture ingress (leakage 1 percent per day) 0.065 3.2x10-5 170 0.085 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 5.9x10-5 - 0.18 - Expected risk (per year) - 9.4x10-10 - 2.9x10-6 - Table F.2.1.2-3.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 0.49 2.5x10-4 53 0.026 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling system 0.18 9.0x10-5 16 8.1x10-3 6.0x10-6 functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 0.49 2.5x10-4 53 0.026 2.0x10-6 Moisture ingress (leakage 1 percent per day) 0.18 9.0x10-5 16 8.1x10-3 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 1.7x10-4 - 0.017 - Expected Risk (per year) - 2.7x10-9 - 2.8x10-7 - Table F.2.1.2-4.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 4.4 2.2x10-3 4.3x103 2.2 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling 1.5 7.7x10-4 1.4x103 0.68 6.0x10-6 system functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 4.4 2.2x10-3 4.3x103 2.2 2.0x10-6 Moisture ingress (leakage 1 percent per day) 1.5 7.7x10-4 1.4x103 0.68 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 1.5x10-3 - 1.4 - Expected risk (per year) - 2.4x10-8 - 2.3x10-5 - Table F.2.1.2-5.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 3 1.5x10-3 570 0.29 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling 1 5.2x10-4 178 0.089 6.0x10-6 system functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 3 1.5x10-3 570 0.29 2.0x10-6 Moisture ingress (leakage 1 percent per day) 1 5.2x10-4 178 0.089 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 1.0x10-3 - 0.19 - Expected risk (per year) - 1.6x10-8 - 3.0x10-6 - Table F.2.1.2-6.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 0.19 9.3x10-5 1.9x103 0.96 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling system 0.066 3.3x10-5 596 0.3 6.0x10-6 functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 0.19 9.3x10-5 1.9x103 0.96 2.0x10-6 Moisture ingress (leakage 1 percent per day) 0.066 3.3x10-5 596 0.3 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 6.3x10-5 - 0.63 - Expected risk (per year) - 1.0x10-9 - 1.0x10-5 - Table F.2.1.2-7.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 25 9.9x10-3 8.5 3.4x10-3 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling system 8.7 3.5x10-3 3.1 1.2x10-3 6.0x10-6 functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 25 9.9x10-3 8.5 3.4x10-3 2.0x10-6 Moisture ingress (leakage 1 percent per day) 8.7 3.5x10-3 3.1 1.2x10-3 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 6.7x10-3 - 2.3x10-3 - Expected risk (per year) - 1.1x10-7 - 3.7x10-8 - Table F.2.1.2-8.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 19 7.5x10-3 6.8 2.7x10-3 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling system 6.5 2.6x10-3 2.4 9.6x10-4 6.0x10-6 functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 19 7.5x10-3 6.8 2.7x10-3 2.0x10-6 Moisture ingress (leakage 1 percent per day) 6.5 2.6x10-3 2.4 9.6x10-4 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 5.0x10-3 - 1.8x10-3 - Expected risk (per year) - 8.1x10-8 - 3.0x10-8 - Table F.2.1.2-9.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 27 0.011 9.1 3.6x10-3 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling system 9 3.6x10-3 3.1 1.3x10-3 6.0x10-6 functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 27 0.011 9.1 3.6x10-3 2.0x10-6 Moisture ingress (leakage 1 percent per day) 9 3.6x10-3 3.1 1.3x10-3 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 7.1x10-3 - 2.4x10-3 - Expected risk (per year) - 1.1x10-7 - 3.9x10-8 - Table F.2.1.2-10.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 12 4.6x10-3 4.2 1.7x10-3 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling system 3.9 1.6x10-3 1.5 5.8x10-4 6.0x10-6 functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 12 4.6x10-3 4.2 1.7x10-3 2.0x10-6 Moisture ingress (leakage 1 percent per day) 3.9 1.6x10-3 1.5 5.8x10-4 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 3.1x10-3 - 1.1x10-3 - Expected risk (per year) - 5.0x10-8 - 1.8x10-8 - Table F.2.1.2-11.-Modular High Temperature Gas-Cooled Reactor High Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Depressurized conduction cooldown with reactor cavity cooling system 12 4.8x10-3 4.3 1.7x10-3 6.0x10-6 functioning (leakage 100 percent per day) Depressurized conduction cooldown without reactor cavity cooling system 4 1.6x10-3 1.5 5.9x10-4 6.0x10-6 functioning (leakage 1 percent per day) Air ingress (leakage 100 percent per day) 12 4.8x10-3 4.3 1.7x10-3 2.0x10-6 Moisture ingress (leakage 1 percent per day) 4 1.6x10-3 1.5 5.9x10-4 2.0x10-6 Evaluation of Composite Set of Accidents Expected consequences - 3.2x10-3 - 1.1x10-3 - Expected risk (per year) - 5.1x10-8 - 1.8x10-8 - F.2.1.3 Advanced Light Water Reactor Previous studies performed for the ALWR developed a spectrum of severe accidents and their respective source terms (ABB 1994a; DOE 1992t; GE 1993a; GE nda; TTI 1995b). The studies of the four ALWR technologies were for the Advanced Boiling Water Reactor , CE System 80+, AP600, and the Simplified Boiling Water Reactor; and were performed independently by their respective vendors for licensing purposes. Because they were performed independently, the modeling assumptions, techniques, and resulting source terms and consequences do not have uniform bases. Although the results are considered adequate for comparisons with other non-ALWR technologies, they should not be used for comparisons among the four ALWR technologies without further analyses using uniform bases. The release frequencies for the four ALWR release categories were in the range of 5.0x10-11 to 1.0x10-6 per reactor year. In order to provide a reasonably similar basis for comparisons with other technologies, a release category and corresponding frequency, out of several available, were chosen to represent the consequences and risks associated with each ALWR tech- nology at each of the five candidate sites. The selected combination of release category and frequency for each technology is representative of accident conditions at the low frequency end of the credible range for beyond design-basis accidents. F.2.1.3.1 Advanced Boiling Water Reactor Chapter 19 of the Advanced BWR Standard Safety Analysis Report, evaluated beyond design-basis accidents that were initiated by either internal events (e.g., a sequence of equipment failures) or external events (e.g., severe natural phenomena such as beyond design-basis earthquakes). The evaluation of external event initiated accidents did not present accident frequency data, release fractions, or source term data that could be used to analyze the accident consequences and risks for this class of accident in this PEIS. Numerous internal event initiated accidents were evaluated in Chapter 19. The accidents that had a common source term were binned or grouped together and evaluated as a single accident and a single total annual frequency of occurrence was defined for the group. Release fractions and the annual frequency of occurrence were defined for ten accidents. The annual frequency of occurrence for these ten accidents ranged from 7.0x10-8 per year to less than 1.0x10-10 per year (GE nda). Two of the accidents had an annual frequency of occurrence greater than 1.0x10-8 per year. These two accidents were selected for evaluation in this PEIS. Accident No. 1 Scenario. The postulated accident is an anticipated transient without scram with the loss of core cooling. Due to the loss of core cooling, core damage results, the vessel fails in approximately 1 hour, and the containment fails in approximately 19 hours (GE nda). The source term is presented in table F.2.1.3.1-1. The annual frequency of occurrence for this accident is 1.3x10-7 per year (GE nda). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.3.1-2 through F.2.1.3.1-6 for public consequences and in tables F.2.1.3.1-7 through F.2.1.3.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.1-1 using the MACCS computer code. Accident No. 2 Scenario. Accident No. 2 is represented by a source term that is common for a group of accidents. The group of accidents include the following: Loss of all core cooling, vessel failure at high pressure, firewater addition system switched to drywell spray mode, containment overpressure protection system rupture disk ruptures, and release negligible - less than 0.1 percent volatile fission products. Loss of all core cooling, vessel failure at high pressure, passive flooder and drywell spray available, containment overpressure protection system rupture disk ruptures, and release negligible - less than 0.1 percent volatile fission products. Large break loss of coolant accident, loss of all core cooling, firewater addition system switched to drywell spray mode, containment overpressure protection system rupture disk ruptures, and release negligible - less than 0.1 percent volatile fission products. Station blackout with RCIC operating for 8 hours, offsite power restored at 8 hours, firewater addition system switched to drywell spray mode, containment over-pressure protection system rupture disk ruptures, and release negligible - less than 0.1 percent volatile fission products. Loss of all core cooling, vessel failure at low pressure, passive flooder available, containment overpressure protection system rupture disk ruptures, and release negligible - less than 0.1 percent volatile fission products. Loss of all core cooling, vessel failure at low pressure, firewater addition system switched to drywell spray mode, containment overpressure protection system rupture disk ruptures, and release negligible - less than 0.1 percent volatile fission products (GE nda). The source term is presented in table F.2.1.3.1-1. The annual frequency of occurrence for the group of accidents is 2.1x10-8 per year (GE nda). Consequences. The estimated consequences of Accident No. 2 at each site are shown in tables F.2.1.3.1-2 through F.2.1.3.1-6 for public consequences and in tables F.2.1.3.1-7 through F.2.1.3.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.1-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Two High Consequence Accidents Figure F.2.1.3.1-1 shows the annual probability that, in the event of any accident in the composite set of Advanced Boiling Water Reactor ALWR high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence, as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.3.1-2 through F.2.1.3.1-11. Figure (Page F-27) Figure F.2.1.3.1-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for the Advanced Boiling Water Reactor. Table F.2.1.3.1-1.- Advanced Boiling Water Reactor High Consequence Accident Source Terms Isotope Released Activity (curies) - Accident No. 1 Accident No. 2 H-3 1.4x106 3.2x107 Kr-85 4.4x104 1.0x106 Kr-85m 1.6x106 3.6x107 Kr-87 2.9x106 6.6x107 Kr-88 3.9x106 9.0x107 Rb-86 1.3 0.73 I-131 2.4x103 16 I-132 3.5x103 23 I-133 5.0x103 33 I-134 5.5x103 36 I-135 4.7x103 31 Xe-133 9.6x106 2.2x108 Xe-135 2.3x106 5.2x107 Cs-134 390 220 Cs-136 100 59 Cs-137 230 130 Source: Source term derived from accident release fractions (GE nda) and core inventory (TTI 1995b). Table F.2.1.3.1-2.- Advanced Boiling Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) No. 1 0.86 4.3x10-4 640 0.32 1.3x10-7 No. 2 16 0.14 1.3x103 0.64 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 2.3x10-3 - 0.36 - Expected risk (per year) - 3.5x10-10 - 5.5x10-8 - Table F.2.1.3.1-3.-Advanced Boiling Water Reactor High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) No. 1 2 1.0x10-3 61 0.03 1.3x10-7 No. 2 37 0.033 126 0.063 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 5.5x10-3 - 0.035 - Expected risk (per year) - 8.3x10-10 - 5.3x10-9 - Table F.2.1.3.1-4.-Advanced Boiling Water Reactor High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) No. 1 12 7.4x10-3 6.6x103 3.3 1.3x10-7 No. 2 186 0.099 4.9x104 24 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.02 - 6.2 - Expected risk (per year) - 3.1x10-9 - 9.4x10-7 - Table F.2.1.3.1-5.-Advanced Boiling Water Reactor High Consequence Accidents at Pantex Plant-Public Consequence - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) No. 1 7.3 3.8x10-3 819 0.41 1.3x10-7 No. 2 102 0.084 5.3x103 2.6 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.015 - 0.72 - Expected risk (per year) - 2.3x10-9 - 1.1x10-7 - Table F.2.1.3.1-6.-Advanced Boiling Water Reactor High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) No. 1 0.66 3.3x10-4 2.4x103 1.2 1.3x10-7 No. 2 11 7.6x10-3 9.5x103 4.7 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 1.3x10-3 - 1.7 - Expected risk (per year) - 2.0x10-10 - 2.6x10-7 - Table F.2.1.3.1-7.-Advanced Boiling Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Accident No. 1 49 0.028 22 0.013 1.3x10-7 No. 2 562 0.066 311 0.059 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.033 - 0.019 - Expected risk (per year) - 5.0x10-9 - 2.9x10-9 - Table F.2.1.3.1-8.-Advanced Boiling Water Reactor High Consequence Accidents at Nevada Test Site-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) No. 1 36 0.019 17 8.4x10-3 1.3x10-7 No. 2 409 0.093 230 0.075 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.03 - 0.018 - Expected risk (per year) - 4.5x10-9 - 2.7x10-9 - Table F.2.1.3.1-9.-Advanced Boiling Water Reactor High Consequence Accidents at Oak Ridge Reservation-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) No. 1 51 0.029 22 0.012 1.3x10-7 No. 2 561 0.056 295 0.073 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.032 - 0.02 - Expected risk (per year) - 4.9x10-9 - 3.0x10-9 - Table F.2.1.3.1-10.-Advanced Boiling Water Reactor High Consequence Accidents at Pantex Plant-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) No. 1 22 0.01 9.7 4.2x10-3 1.3x10-7 No. 2 239 0.14 127 0.075 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.023 - 0.014 - Expected risk (per year) - 3.5x10-9 - 2.1x10-9 - Table F.2.1.3.1-11.-Advanced Boiling Water Reactor High Consequence Accidents at Savannah River Site-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) No. 1 22 0.011 9.9 4.5x10-3 1.3x10-7 No. 2 246 0.097 130 0.066 2.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.023 - 0.013 - Expected risk (per year) - 3.4x10-9 - 2.0x10-9 - F.2.1.3.2 CE System 80+ Advanced Light Water Reactor Chapter 19 of the CESSAR Design Certification - System 80+ Standard Design evaluated beyond design-basis accidents that were initiated by internal events (e.g., a sequence of equipment failures). The accidents that had a common source term were binned or grouped together and evaluated as a single accident and a single total annual frequency of occur- rence was defined for the group. Release fractions and the annual frequency of occurrence were defined for 23 accident groupings. The annual frequency of occurrence for these 23 accident groupings ranged from 1.4x10-6 to 5.1x10-10 (ABB 1994a). Two of the accidents had an annual frequency of occurrence greater than 1.0x10-8. These two accidents were selected for evaluation in this PEIS. Tornado Strike Disables Switchyard and Both Emergency Diesel Generators Failed Scenario. The analysis postulated that a tornado struck the switchyard. As a result of loss of load, the turbine tripped and the reactor tripped. The analysis postulated that both diesels failed to start and a station blackout condition existed at the site. When the emergency batteries were depleted, the core would overheat, the core would fail, and the vessel would fail. Ultimately the containment would fail (ABB 1994a:19.7-23,19.7-24,19.12-121). If all sites were assumed to be located in tornado region B, the region with the highest tornado frequency, the tornado strike frequency for the plant is 1.07x10-5 per year (ABB 1994a). Based on NRC licensing requirements, the minimum acceptable emergency diesel generator target reliability is 95 percent (NCR1988a.155:1.155-3). The annual frequency of a tornado striking the plant switchyard and the failure of both emergency diesel generators is 2.7x10-8. Consequences. The annual frequency of occurrence for the postulated accident is less than 1.0x10-7 and thus the accident consequence is considered beyond the scope of this PEIS and was not analyzed (DOE1993z:28). Loss of Coolant Accident, Failure of Safety Systems, and Containment Failure Scenario. A spectrum of beyond design-basis loss of coolant accidents were postulated. The individual accident scenarios postulated the failure of safety systems that mitigate the accident consequences. Due to the failure of the safety systems, core damage resulted, the containment may overpressurize and fail or the containment may fail due to basemat melt- through. The annual frequency of occurrence for the spectrum of beyond design-basis loss of coolant accidents is in the range of 3.8x10-8 for release class RC2.4E to 1.8x10-9 for release class RC4.22E (ABB 1994a:19.12-116-19.12-129). Consequences. The annual frequency of occurrence for the each of the of loss of coolant accidents sequences was less than 1.0x10-7 and thus the accident consequences are considered beyond the scope of this PEIS and were not analyzed (DOE 1993z). Loss of Feedwater, Loss of Emergency Feedwater, and Failure to Bleed System Scenario. The accident is initiated by loss of feedwater followed by the loss of emergency feedwater and the failure to bleed the system preventing feed and bleed cooling. Core damage is assumed to occur at 4 hours with vessel failure at 5 hours. Containment spray and containment heat removal are assumed operational and the cavity is flooded. The releases are assumed to start at the time of vessel breach at 4 hours and continue for 24 hours. The release occurs at an elevation 16.6 meters above grade. The source term is presented in table F.2.1.3.2-1. The annual frequency of occurrence for this accident is 1.4x10-6 per year (ABB1994a:19.12-115). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.3.2-2 through F.2.1.3.2-6 for public consequences and in tables F.2.1.3.2-7 through F.2.1.3.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.2-1 using the MACCS computer code. Loss of Feedwater and Failure of Long-Term Decay Heat Removal Scenario. The accident is initiated by loss of feedwater. The emergency feedwater is initially successful but there is a failure of long-term decay heat removal in the 8- to 24-hour period. Core damage is assumed to occur at 16 hours with vessel failure at 17 hours. The cavity is assumed flooded. The releases are assumed to start at the time of vessel breach at 17 hours and continue for 24 hours. The release occurs at an elevation 16.6 meters above grade. The source term is presented in table F.2.1.3.2-1. The annual frequency of occurrence for this accident is 3.8x10-7 per year (ABB 1994a:19.12-115,19.12-116). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.3.2-2 through F.2.1.3.2-6 for public consequences and in tables F.2.1.3.2-7 through F.2.1.3.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.2-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Two Loss of Feedwater High Consequence Accidents Figure F.2.1.3.2-1 shows the annual probability that, in the event of any accident in the composite set of CE System 80+ ALWR high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.3.2-2 through F.2.1.3.2-11. Figure (Page F-33) Figure F.2.1.3.2-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for the CE System 80+ Reactor. Table F.2.1.3.2-1.-CE System 80+ Advanced Light Water Reactor High Consequence Accident Source Terms Isotope Released Activity (curies) - Released Activity (curies) - Loss of Feedwater, Loss of Feedwater Isotope Loss of Feedwater, Loss of Feedwater Loss of Emergency and Failure of Long- Loss of Emergency and Failure of Long- Feedwater, and Term Decay Heat Feedwater, and Term Decay Heat Failure to Bleed Removal Failure to Bleed Removal System System H-3 1.6x105 1.6x105 Xe-133 1.2x106 1.2x106 Kr-85 5.7x103 5.7x103 Xe-135 3.2x105 3.2x105 Kr-85m 1.8x105 1.8x105 Cs-134 2.1 1 Kr-87 3.6x105 3.6x105 Cs-136 0.8 0.4 Kr-88 5.1x105 5.1x105 Cs-137 2.1 1 Rb-86 0.03 0.015 Ba-139 5.2 2.5 Sr-89 0.59 0.26 Ba-140 5.1 2.5 Sr-90 0.038 0.017 La-140 1.4 0.69 Sr-91 0.72 0.32 La-141 1.3 0.63 Sr-92 0.76 0.34 La-142 1.2 0.61 Y-90 0.056 0.028 Ce-141 5.4 2.3 Y-91 1.1 0.53 Ce-143 5 2.2 Y-92 1.1 0.55 Ce-144 4 1.8 Y-93 1.2 0.61 Pr-143 1.2 0.59 Zr-95 5.6 2.4 Nd-147 0.5 0.25 Zr-97 5.3 2.3 Np-239 46 20 Nb-95 0.3 0.19 Pu-238 2.8x10-3 1.2x10-3 Mo-99 0.3 0.2 Pu-239 8.7x10-4 3.8x10-4 Tc-99m 0.27 0.17 Pu-240 1.1x10-3 4.7x10-4 Ru-103 0.22 0.14 Pu-241 0.2 0.088 Ru-105 0.13 0.085 Te-127 1 0.5 Ru-106 0.054 0.035 Te-127m 0.14 0.066 Rh-105 0.12 0.08 Te-129 3.3 1.6 Sb-127 1 0.51 Te-129m 0.49 0.24 Sb-129 3.4 1.6 Te-131m 1.6 0.76 I-131 27 2.4x103 Te-132 16 7.8 I-132 39 3.5x103 Am-241 3.8x10-5 1.9x10-5 I-133 57 5.1x103 Cm-242 8.3x10-3 4.1x10-3 I-134 63 5.7x103 Cm-244 1.2x10-4 5.8x10-5 I-135 53 4.8x103 - - - Table F.2.1.3.2-2.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 0.094 4.7x10-5 11 5.6x10-3 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 0.11 5.3x10-5 18 9.0x10-3 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 4.8x10-5 - 6.3x10-3 - Expected risk (per year) - 8.6x10-11 - 1.1x10-8 - Table F.2.1.3.2-3.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 0.21 1.0x10-4 1.1 5.4x10-4 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 0.24 1.2x10-4 1.8 8.7x10-4 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.1x10-4 - 6.1x10-4 - Expected risk (per year) - 1.9x10-10 - 1.1x10-9 - Table F.2.1.3.2-4.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 1.1 5.4x10-4 308 0.15 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 1.3 6.5x10-4 394 0.2 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 5.6x10-4 - 0.16 - Expected risk (per year) - 1.0x10-9 - 2.9x10-7 - Table F.2.1.3.2-5.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 0.6 3.0x10-4 34 0.017 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 0.75 3.8x10-4 45 0.023 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 3.1x10-4 - 0.018 - Expected risk (per year) - 5.6x10-10 - 3.2x10-8 - Table F.2.1.3.2-6.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 0.063 3.2x10-5 67 0.034 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 0.073 3.7x10-5 97 0.049 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 3.3x10-5 - 0.037 - Expected risk (per year) - 5.8x10-11 - 6.5x10-8 - Table F.2.1.3.2-7.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 3.4 1.3x10-3 1.8 7.3x10-4 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 4.7 1.9x10-3 2.4 9.4x10-4 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.5x10-3 - 7.7x10-4 - Expected risk (per year) - 2.6x10-9 - 1.4x10-9 - Table F.2.1.3.2-8.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 2.5 9.8x10-4 1.4 5.4x10-4 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 3.4 1.3x10-3 1.7 7.0x10-4 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.1x10-3 - 5.7x10-4 - Expected risk (per year) - 1.9x10-9 - 1.0x10-9 - Table F.2.1.3.2-9.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 3.4 1.3x10-3 1.7 6.9x10-4 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 4.5 1.8x10-3 2.2 8.7x10-4 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.4x10-3 - 7.3x10-4 - Expected risk (per year) - 2.6x10-9 - 1.3x10-9 - Table F.2.1.3.2-10.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 1.4 5.7x10-4 0.75 3.0x10-4 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 1.9 7.8x10-4 0.96 3.8x10-4 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 6.2x10-4 - 3.2x10-4 - Expected risk (per year) - 1.1x10-9 - 5.6x10-10 - Table F.2.1.3.2-11.-CE System 80+ Advanced Light Water Reactor High Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of feedwater, loss of emergency feedwater, and failure to bleed system 1.5 5.9x10-4 0.76 3.0x10-4 1.4x10-6 Loss of feedwater and failure of long-term decay heat removal 2 7.9x10-4 0.97 3.9x10-4 3.8x10-7 Evaluation of Composite Set of Accidents Expected consequences - 6.3x10-4 - 3.2x10-4 - Expected risk (per year) - 1.1x10-9 - 5.6x10-10 - F.2.1.3.3 AP600 Advanced Light Water Reactor The AP600 Standard Safety Analysis Report (DOE 1992t), evaluated beyond design-basis accidents that were initiated by either internal events (e.g., a sequence of equipment failures) or external events (e.g., severe natural phenomena such as beyond design basis earthquakes). The evaluation of external event initiated accidents did not present accident frequency data, release fractions, or source term data that could be used to analyze the accident consequences and risks for this class of accident in this PEIS. Numerous internal event initiated accidents were evaluated in the Safety Analysis Report. The accidents that had a common source term or release category were binned or grouped together and evaluated as a single accident and a single total annual frequency of occurrence was defined for the group. Release fractions and the annual frequency of occurrence were defined for four accidents. The annual frequency of occurrence for these four accidents ranged from 2.5x10-7 per year to 7.6x10-10 per year (DOE 1992t). Two of the accident groups with an annual frequency of occurrence greater than 5.0x10-8 per year were selected for evaluation in this PEIS. A representative accident within each group was used to define a typical accident sequence for the group. Loss of Coolant Accident with Failure of Refueling Water Storage Tank and Residual Heat Removal Scenario. The representative accident sequence for the OK release category has an initiating event which is a 4-inch diameter loss of coolant accident with a failure of the in-containment refueling water storage tank check valves and normal residual heat removal injection. Core damage begins 2 hours into the accident. The in-containment refueling water and storage tank is not drained into the containment cavity to provide external cooling to the reactor vessel. The vessel fails at 11.8 hours and the molten core drains into the containment at low pressure. The debris is quenched and cooled in the reactor cavity. The passive containment cooling system and hydrogen igniters are available and containment pressure remains below design pressure. The final source term at 24 hours after core damage is presented in table F.2.1.3.3-1. The annual frequency of occurrence for this accident is 2.5x10-7 per year (DOE 1992t). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.3.3-2 through F.2.1.3.3-6 for public consequences and in tables F.2.1.3.3-7 through F.2.1.3.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.3-1 using the MACCS computer code. Loss of Coolant Accident with Failure of Refueling Water Storage Tank, Residual Heat Removal, and Passive Containment Cooling System Cooling Water Scenario. The representative accident sequence for the OKP release category is initiated by a 4-inch diameter loss of cooling accident with a failure of the in-containment refueling water and storage tank check valves, normal residual heat removal injection, and passive containment cooling system cooling water. Four of the four core makeup tanks and accumulators are available. Core damage occurs at 2.5hours and the vessel fails at 15.8hours. The debris is quenched and cooled in the reactor cavity. The containment pressure is elevated over the long term, but it equilibrates at a pressure well below the ultimate capacity of the shell so containment integrity is maintained. The final source term, at 24hours after core damage is presented in table F.2.1.3.3-1. The annual frequency of occurrence for this accident is 5.6x10-8 per year (DOE1992t:1B-4,1B-5). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.3.3-2 through F.2.1.3.3-6 for public consequences and in tables F.2.1.3.3-7 through F.2.1.3.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.3-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Two High Consequence Accidents Figure F.2.1.3.3-1 shows the annual probability that, in the event of any accident in the composite set of AP600 ALWR high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.3.3-2 through F.2.1.3.3-11. Figure (Page F-41) Figure F.2.1.3.3-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for the AP600 Reactor. Table F.2.1.3.3-1.-AP600 Advanced Light Water Reactor High Consequence Accident Source Terms Isotope Released Activity (curies) - Released Activity (curies) - Loss of Cooling Accident with Loss of Cooling Accident with Isotope Loss of Cooling Accident with Loss of Cooling Accident with Failure of Refueling Water Storage Failure of Refueling Water Storage Failure of Refueling Water Storage Failure of Refueling Water Storage Tank and Residual Heat Removal Tank, Residual Heat Removal, and Tank and Residual Heat Removal Tank, Residual Heat Removal, and (OK) Possible Containment Cooling (OK) Passive Containment Cooling System Cooling Water(OKP) System Cooling Water (OKP) H-3 1.3x103 3.2x103 Te-132 79 380 Kr-85 23 54 I-131 31 110 Kr-85m 590 1.4x103 I-132 45 160 Kr-87 1.1x103 2.7x103 I-133 62 220 Kr-88 1.6x103 3.8x103 I-134 67 240 Sr-89 1.7 4.2 I-135 56 200 Sr-90 0.14 0.35 Xe-133 4.6x103 1.1x104 Ru-103 50 85 Xe-135 1.5x103 3.6x103 Ru-105 0 0 Cs-134 5.2 18 Ru-106 16 28 Cs-136 1.6 5.6 Te-129m 4.7 23 Cs-137 3.5 12 Te-131m 8.4 40 - - - Note: OK and OKP - release category codes for composite set of accident sequences. Source: Derived from TTI 1995b. Table F.2.1.3.3-2.-AP600 Advanced Light Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of cooling accident with failure of refueling water and storage tank and residual heat 4.0x10-3 2.0x10-6 12 5.8x10-3 2.5x10-7 removal (OK) Loss of cooling accident with failure of refueling water and storage tank, residual heat 0.012 6.1x10-6 37 0.018 5.6x10-8 removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 2.8x10-6 - 8.1x10-3 - Expected risk (per year) - 8.5x10-13 - 2.5x10-9 - Table F.2.1.3.3-3.-AP600 Advanced Light Water Reactor High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of cooling accident with failure of refueling water storage tank and residual 0.011 5.7x10-6 1.1 5.5x10-4 2.5x10-7 heat removal (OK) Loss of cooling accident with failure of refueling water storage tank, residual 0.035 1.7x10-5 3.5 1.7x10-3 5.6x10-8 heat removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 7.8x10-6 - 7.7x10-4 - Expected risk (per year) - 2.4x10-12 - 2.4x10-10 - Table F.2.1.3.3-4.-AP600 Advanced Light Water Reactor High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of cooling accident with failure of refueling water storage tank and residual 0.099 4.9x10-5 92 0.046 2.5x10-7 heat removal (OK) Loss of cooling accident with failure of refueling water storage tank, residual 0.3 1.5x10-4 286 0.14 5.6x10-8 heat removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 6.8x10-5 - 0.064 - Expected risk (per year) - 2.1x10-11 - 2.0x10-8 - Table F.2.1.3.3-5.-AP600 Advanced Light Water Reactor High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of cooling accident with failure of refueling water storage tank and residual 0.066 3.3x10-5 12 6.1x10-3 2.5x10-7 heat removal (OK) Loss of cooling accident with failure of refueling water storage tank, residual 0.2 1.0x10-4 38 0.019 5.6x10-8 heat removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 4.5x10-5 - 8.4x10-3 - Expected risk (per year) - 1.4x10-11 - 2.6x10-9 - Table F.2.1.3.3-6.-AP600 Advanced Light Water Reactor High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of cooling accident with failure of refueling water storage tank and residual 4.2x10-3 2.1x10-6 41 0.02 2.5x10-7 heat removal (OK) Loss of cooling accident with failure of refueling water storage tank, residual 0.013 6.5x10-6 128 0.064 5.6x10-8 heat removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 2.9x10-6 - 0.028 - Expected risk (per year) - 8.9x10-13 - 8.7x10-9 - Table F.2.1.3.3-7.-AP600 Advanced Light Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of cooling accident with failure of refueling water storage tank and residual 0.55 2.2x10-4 0.19 7.6x10-5 2.5x10-7 heat removal (OK) Loss of cooling accident with failure of refueling water storage tank, residual 1.7 6.7x10-4 0.59 2.3x10-4 5.6x10-8 heat removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 3.0x10-4 - 1.1x10-4 - Expected risk (per year) - 9.2x10-11 - 3.2x10-11 - Table F.2.1.3.3-8.-AP600 Advanced Light Water Reactor High Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of cooling accident with failure of refueling water storage tank and residual 0.41 1.6x10-4 0.15 6.1x10-5 2.5x10-7 heat removal (OK) Loss of cooling accident with failure of refueling water storage tank, residual 1.3 5.1x10-4 0.47 1.9x10-4 5.6x10-8 heat removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 2.3x10-4 - 8.4x10-5 - Expected risks (per year) - 7.0x10-11 - 2.6x10-11 - Table F.2.1.3.3-9.-AP600 Advanced Light Water Reactor High Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of cooling accident with failure of refueling water storage tank and residual heat 0.58 2.3x10-4 0.2 8.1x10-5 2.5x10-7 removal (OK) Loss of cooling accident with failure of refueling water storage tank, residual heat 1.8 7.2x10-4 0.62 2.5x10-4 5.6x10-8 removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 3.2x10-4 - 1.1x10-4 - Expected risk (per year) - 9.8x10-11 - 3.4x10-11 - Table F.2.1.3.3-10.-AP600 Advanced Light Water Reactor High Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of cooling accident with failure of refueling water storage tank and residual heat 0.25 1.0x10-4 0.094 3.7x10-5 2.5x10-7 removal (OK) Loss of cooling accident with failure of refueling water storage tank, residual heat 0.78 3.1x10-4 0.29 1.2x10-4 5.6x10-8 removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 1.4x10-4 - 5.2x10-5 - Expected risk (per year) - 4.3x10-11 - 1.6x10-11 - Table F.2.1.3.3-11.-AP600 Advanced Light Water Reactor High Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of cooling accident with failure 0.26 1.0x10-4 0.094 3.8x10-5 2.5x10-7 of refueling water storage tank and residual heat removal (OK) Loss of cooling accident with failure 0.8 3.2x10-4 0.29 1.2x10-4 5.6x10-8 of refueling water storage tank, residual heat removal, and passive containment cooling system and cooling water (OKP) Evaluation of Composite Set of Accidents Expected consequences - 1.4x10-4 - 5.2x10-5 - Expected risk (per year) - 4.4x10-11 - 1.6x10-11 - F.2.1.3.4 Simplified Boiling Water Reactor Chapter 19 of the Simplified BWR Standard Safety Analysis Report, evaluated beyond design-basis accidents that were initiated by either internal events (e.g., a sequence of equipment failures) or external events (e.g., severe natural phenomena such as beyond design-basis earthquakes). The evaluation of external event initiated accidents did not present accident frequency data, release fractions, or source term data that could be used to analyze the accident consequences and risks for this class of accident in this PEIS. Fourteen internal event initiated accidents were evaluated in Chapter 19. The annual frequency of occurrence for these accidents ranged from 7.0x10-8 per year to 1.0x10-10 per year (GE 1993a). Four of the accidents had an annual frequency of occurrence greater than 1.0x10-8 per year. These four accidents were selected for evaluation in this PEIS. Low Pressure Core Melt with Loss of Short-Term Coolant Makeup, Failure of the Drywell Sprays to Operate, and Normal Containment Leakage Scenario. The postulated accident is initiated by the inadvertently open relief valve that depressurizes the reactor. The reactor scrams, the main steam isolation valves close, the feedwater pumps trip, and the automatic depressurizing system actuates. All high and low pressure injection systems are assumed to fail. Approximately 1 hour into the accident, the core is uncovered and fuel rods melt. The reactor lower vessel head penetrations fail at approximately 4.5hours. Local temperatures cause the flooder to open and the gravity driven cooling system pool water drains into the lower drywell. The debris is quenched and the long-term containment pressure is less than the suppression chamber vent pressure setpoint. Normal containment leakage is the only mode of fission product release (GE1993a:19B.6-819B.6-9). The source term is presented in tableF.2.1.3.4-1. The annual frequency of occurrence for this accident is 7.0x10-8 per year (GE1993a). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.3.4-2 through F.2.1.3.4-6 for public consequences and in tables F.2.1.3.4-7 through F.2.1.3.4-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.4-1 using the MACCS computer code. Low Pressure Core Melt with Loss of Long-Term Coolant Makeup, Failure of the Drywell Sprays to Operate, and Normal Containment Leakage Scenario. The postulated accident is initiated by the inadvertently open relief valve that depressurizes the reactor. The reactor scrams, the main steam isolation valves close, the feedwater pumps trip, and the automatic depressurizing system actuates. One gravity driven cooling system pool injects water into the reactor vessel. Approximately 7 hours into the accident, the core is uncovered and fuel rods melt. The reactor lower vessel head penetrations fail at approximately 12.5 hours. Local temperatures cause the flooder to open and the gravity driven cooling system pool water drains into the lower drywell. The debris is quenched and the long-term containment pressure is less than the suppression chamber vent pressure setpoint. Normal containment leakage is the only mode of fission product release (GE 1993a:19B.6-8,19B.6-9). The source term is presented in table F.2.1.3.4-1. The annual frequency of occurrence for this accident is 6.4x10-8 per year (GE 1993a). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.3.4-2 through F.2.1.3.4-6 for public consequences and in tables F.2.1.3.4-7 through F.2.1.3.4-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.4-1 using the MACCS computer code. Low Pressure Core Melt with Loss of Short-Term Coolant Makeup, Failure of the Drywell Sprays to Operate, and Containment Vented Scenario. The postulated accident is initiated by the inadvertently open relief valve that depressurizes the reactor. The reactor scrams, the main steam isolation valves close, the feedwater pumps trip, and the automatic depressurizing system actuates. All high and low pressure injection systems are assumed to fail. Approximately 1 hour into the accident, the core is uncovered and fuel rods melt. The reactor lower vessel head penetrations fail at approximately 4.5 hours. Local temperatures cause the flooder to open and the gravity driven cooling system pool water drains into the lower drywell. Relocation of the debris causes the long-term containment pressure to increase to the suppression chamber vent pressure setpoint and the containment is breached at approximately 29 hours. The fission product release is complete after the containment is vented (GE 1993a:19B.6-8-19B.6-10). The source term is presented in table F.2.1.3.4-1. The annual frequency of occurrence for this accident is 1.1x10-8 per year (GE 1993a). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.3.4-2 through F.2.1.3.4-6 for public consequences and in tables F.2.1.3.4-7 through F.2.1.3.4-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.4-1 using the MACCS computer code. Low Pressure Core Melt with Loss of Long-Term Coolant Makeup, Failure of the Drywell Sprays to Operate, and Containment Vented Scenario. The postulated accident is initiated by the inadvertently open relief valve that depressurizes the reactor. The reactor scrams, the main steam isolation valves close, the feedwater pumps trip, and the automatic depressurizing system actuates. One gravity driven cooling system pool injects water into the reactor vessel. Approximately 7 hours into the accident, the core is uncovered and fuel rods melt. The reactor lower vessel head penetrations fail at approximately 12.5 hours. Local temperatures cause the flooder to open and the gravity driven cooling system pool water drains into the lower drywell. Relocation of the debris causes the long-term containment pressure to increase to the suppression chamber vent pressure setpoint and the containment is breached at approximately 36.5 hours. The fission product release is complete after the containment is vented (GE 1993a:19B.6-8,19B.6-11). The source term is presented in table F.2.1.3.4-1. The annual frequency of occurrence for this accident is 1.1x10-8 per year (GE 1993a). Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.1.3.4-2 through F.2.1.3.4-6 for public consequences and in tables F.2.1.3.4-7 through F.2.1.3.4-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.3.4-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Four High Consequence Accidents Figure F.2.1.3.4-1 shows the annual probability that, in the event of any accident in the composite set of Simplified Boiling Water Reactor ALWR high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.3.4-2 through F.2.1.3.4-11. Figure (Page F-49) Figure F.2.1.3.4-1.-Simplified Boiling Water Reactor Cancer Fatalities Complementary Cumulative Distribution Functions for High Consequence Accidents. Table F.2.1.3.4-1.-Simplified Boiling Water Reactor High Consequence Accident Source Terms [Page 1 of 3] Isotope Released Activity (curies) - Low Pressure Core Melt Low Pressure Core Melt Low Pressure Core Melt Low Pressure Core Melt with Loss of Short-Term with Loss of Long-Term with Loss with Loss of Long-Term Coolant Makeup and Coolant Makeup and of Short-Term Coolant Coolant Makeup and Normal Containment Normal Containment Makeup and Containment Containment Vented Leakage Leakage Vented H-3 4.5x104 5.7x104 3.2x107 3.2x107 Co-58 12 14 14 15 Co-60 14 17 17 18 Kr-85 720 920 5.1x105 5.1x105 Kr-85m 2.6x104 3.3x104 1.9x107 1.9x107 Kr-87 4.7x104 6.1x104 3.4x107 3.4x107 Kr-88 6.4x104 8.2x104 4.6x107 4.6x107 Rb-86 2.7 3.7 26 160 Sr-89 62 90 90 110 Sr-90 4.4 6.4 6.4 7.6 Sr-91 81 120 120 140 Sr-92 84 120 120 150 Y-90 1.7 0.47 2.7 1.2 Y-91 27 7.6 43 19 Y-92 30 8.5 49 22 Y-93 34 9.6 55 25 Zr-95 150 64 260 140 Zr-97 150 66 270 140 Nb-95 3.3x103 3.9x103 3.9x103 4.2x103 Mo-99 3.8x103 4.5x103 4.5x103 4.9x103 Tc-99m 3.3x103 3.9x103 3.9x103 4.2x103 Ru-103 2.9x103 3.4x103 3.4x103 3.7x103 Ru-105 1.9x103 2.3x103 2.3x103 2.5x103 Ru-106 780 920 920 1.0x103 Rh-105 1.4x103 1.7x103 1.7x103 1.8x103 Sb-127 1.2x103 1.3x103 5.2x104 1.1x105 Sb-129 4.3x103 4.4x103 1.8x105 4.0x105 I-131 4.6x103 6.9x103 3.2x104 8.4x104 I-132 6.7x103 1.0x104 4.6x104 1.2x105 I-133 9.6x103 1.4x104 6.6x104 1.8x105 I-134 1.1x104 1.6x104 7.2x104 1.9x105 I-135 9.1x103 1.4x104 6.2x104 1.7x105 Xe-133 1.5x105 2.0x105 1.1x108 1.1x108 Xe-135 3.7x104 4.7x104 2.6x107 2.6x107 Cs-134 800 1.1x103 7.8x103 4.8x104 Cs-136 220 300 2.1x103 1.3x104 Cs-137 480 670 4.7x103 2.9x104 Ba-139 570 710 670 800 Ba-140 560 700 660 790 La-140 40 11 65 29 La-141 37 10 60 26 La-142 36 10 57 26 Ce-141 150 65 260 140 Ce-143 140 63 260 130 Ce-144 95 42 170 89 Pr-143 34 9.6 55 24 Nd-147 15 4.3 25 11 Np-239 1.9x103 820 3.4x103 1.7x103 Pu-238 0.13 0.057 0.23 0.12 Pu-239 0.033 0.014 0.059 0.031 Pu-240 0.041 0.018 0.074 0.038 Pu-241 7 3.1 13 6.6 Te-127 0.064 0.064 2.1x104 4.6x104 Te-127m 8.7x10-3 8.7x10-3 2.8x103 6.2x103 Te-129 0.22 0.22 7.1x104 1.5x105 Te-129m 0.057 0.057 1.9x104 4.1x104 Te-131m 0.11 0.11 3.6x104 7.8x104 Te-132 1.1 1.1 3.5x105 7.6x105 Am-241 1.7x10-3 4.9x10-4 2.8x10-3 1.2x10-3 Cm-242 0.46 0.13 0.74 0.33 Cm-244 0.025 7.0x10-3 0.04 0.018 Table F.2.1.3.4-2.-Simplified Boiling Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 0.44 2.2x10-4 1.5x103 0.75 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 0.58 2.9x10-4 2.0x103 1 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 13 9.8x10-3 1.5x104 7.6 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 31 0.02 8.0x104 40 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 2.3x10-3 - 4.1 - Expected risk (per year) - 3.6x10-10 - 6.4x10-7 - Table F.2.1.3.4-3.-Simplified Boiling Water Reactor High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 1.3 6.4x10-4 142 0.071 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 1.7 8.4x10-4 190 0.095 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 33 0.026 1.5x103 0.72 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 84 0.054 7.6x103 3.8 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 6.3x10-3 - 0.39 - Expected risk (per year) - 9.8x10-10 - 6.1x10-8 - Table F.2.1.3.4-4.-Simplified Boiling Water Reactor High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 12 5.8x10-3 1.2x104 5.8 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 15 7.6x10-3 1.6x104 7.7 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 219 0.14 1.4x105 69 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 692 0.38 6.3x105 315 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.042 - 33 - Expected risk (per year) - 6.6x10-9 - 5.1x10-6 - Table F.2.1.3.4-5.-Simplified Boiling Water Reactor High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 7.8 3.9x10-3 1.5x103 0.77 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 10 5.2x10-3 2.0x103 1 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 135 0.099 1.8x104 8.8 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 454 0.26 8.3x104 41 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.029 - 4.3 - Expected risk (per year) - 4.6x10-9 - 6.7x10-7 - Table F.2.1.3.4-6.-Simplified Boiling Water Reactor High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 0.49 2.4x10-4 5.2x103 2.6 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 0.65 3.2x10-4 7.0x103 3.5 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 11 6.4x10-3 5.6x104 28 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 31 0.017 2.8x105 139 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 1.9x10-3 - 14 - Expected risk (per year) - 2.9x10-10 - 2.3x10-6 - Table F.2.1.3.4-7.-Simplified Boiling Water Reactor High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 65 0.028 22 9.0x10-3 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 86 0.037 29 0.012 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 992 0.28 409 0.15 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 3.7x103 0.66 1.3x103 0.39 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequence - 0.094 - 0.047 - Expected risk (per year) - 1.5x10-8 - 7.3x10-9 - Table F.2.1.3.4-8.-Simplified Boiling Water Reactor High Consequence Accidents at Nevada Test Site-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 49 0.02 18 7.2x10-3 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 65 0.027 23 9.5x10-3 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 734 0.25 313 0.14 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 2.8x103 0.69 1.0x103 0.36 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.087 - 0.042 - Expected risk - 1.4x10-8 - 6.6x10-9 - Table F.2.1.3.4-9.-Simplified Boiling Water Reactor High Consequence Accidents at Oak Ridge Reservation-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 70 0.029 24 9.7x10-3 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 93 0.039 32 0.013 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 1.0x103 0.3 408 0.17 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 3.9x103 0.72 1.4x103 0.45 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.1 - 0.054 - Expected risk (per year) - 1.6x10-8 - 8.3x10-9 - Table F.2.1.3.4-10.-Simplified Boiling Water Reactor High Consequence Accidents at Pantex Plant-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 31 0.012 11 4.5x10-3 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 41 0.016 15 5.9x10-3 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 442 0.2 184 0.1 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 1.7x103 0.61 641 0.28 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.07 - 0.031 - Expected risk (per year) - 1.1x10-8 - 4.9x10-9 - Table F.2.1.3.4-11.-Simplified Boiling Water Reactor High Consequence Accidents at Savannah River Site-Worker Consequences Accident Worker at 1,000 meters Worker at 2,000 meters - - Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Low pressure core melt with loss of short-term coolant makeup and normal containment 31 0.013 11 4.5x10-3 7.0x10-8 leakage Low pressure core melt with loss of long-term coolant makeup and normal containment 42 0.017 15 6.0x10-3 6.4x10-8 leakage Low pressure core melt with loss of short-term coolant makeup and containment vented 453 0.2 187 0.095 1.1x10-8 Low pressure core melt with loss of long-term coolant makeup and containment vented 1.8x103 0.58 648 0.26 1.1x10-8 Evaluation of Composite Set of Accidents Expected consequences - 0.067 - 0.03 - Expected risk (per year) - 1.1x10-8 - 4.6x10-9 - F.2.1.4 Accelerator Production of Tritium A study of the APT performed by Sandia National Laboratories, New Mexico, for DOE (SNL1995a:8-1,8-2) has evaluated the hazards associated with the APT accelerator and beam transport system and has judged them to be a Category 3 hazard per DOE Order 5480.23. (A Category 3 hazard has the potential for only significant, but localized onsite consequences.) The spallation-induced lithium conversion and helium-3 target systems have been judged to be a Category 2 hazard. (A Category 2 hazard has the potential for significant onsite consequences, but does not have the potential for significant offsite consequences.) The helium-3 target tritium extraction has been judged a Category 3 hazard because only 15 grams of tritium are expected to be contained in the helium-3 blanket and in the target extraction facility. The spallation-induced lithium conversion target tritium extraction has been judged a Category 2 hazard. F.2.1.4.1 Accelerator and Beam Transport System Scenario. The only beyond design-basis event currently identified for the accelerator and beam transport system that has any significant probability involves misdirection or misfocusing of the beam. In this scenario, the beam is not terminated rapidly by the fast protection system, leading to vacuum seal failure, outright breaching of the vacuum system envelope, and/or partial melting of critical accelerator structures (SNL 1995a:8-9). Consequences. The major consequence of this accident would be lost production time (SNL 1995a:8-9). F.2.1.4.2 Helium-3 Target System Loss of Coolant Accident with Loss of Emergency Cooling and Heat Sink but Confinement Operational Scenario. The postulated high consequence accident for the Full and Phased APT is a large break loss of coolant accident with total failure of the active emergency cooling system and loss of heat sink. The postulated accident sequence assumed that the confinement system remained operational. A source term release to the environment was determined. Table F.2.1.4.2-1 presents the source term released by the Full APT during the accident and table F.2.1.4.2-2 presents the source term released by the Phased APT during the accident (SNL 1995a:8-18, 9- 9). The accident annual frequency of occurrence is estimated at 7.0x10-7 per year (SNL1995b:1). Consequences. The estimated consequences to the public for the postulated Full APT with the helium-3 target system accidents for each site are shown in tables F.2.1.4.2-3 through F.2.1.4.2-7. Consequences to the worker are shown in tables F.2.1.4.2-8 through F.2.1.4.2-12. The estimated consequences for the Phased APT with the helium-3 target system are shown for the public in table F.2.1.4.2-13 and for the worker in table F.2.1.4.2-14. Comparison of tables F.2.1.4.2-3 through F.2.1.4.2-14 indicates that the resultant doses and cancer risks are identical for the Full and the Phased APT beyond design-basis accidents. Review of the source terms for both accidents (tables 1 and 2) indicates that the tritium component of the source term is identical for both accidents. Review of the MACCS computer code output data for each accident analysis indicated that the tritium component of the source term dominated the dose calculation results. The impact of the other source term isotopes on the dose calculation results was negligible. Loss of Coolant Accident with Loss of Emergency Cooling, Heat Sink, and Confinement Scenario. The postulated bounding high consequence accident for the Full APT is a large break loss of coolant accident with total failure of the active emergency cooling system, loss of heat sink, and loss of confinement. The source term is presented in table F.2.1.4.2-1. The annual frequency of occurrence for this accident is 1.0x10-8 per year (SNL 1995b:1). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.4.2-3 through F.2.1.4.2-7 for public consequences and in tables F.2.1.4.2-8 through F.2.1.4.2-12 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.4.2-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Two Full Accelerator Production of Tritium High Consequence Accidents Figure F.2.1.4.2-1 shows the annual probability that, in the event of any accident in the composite set of Full APT high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.4.2-3 through F.2.1.4.2-12. Figure (Page F-58) Figure F.2.1.4.2-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for the Full Size Accelerator Production of Tritium with Helium-3 Target. Table F.2.1.4.2-1.-Source Term for Full Accelerator Production of Tritium with Helium-3 Target System High Consequence Accidents Isotope Release Activity (curies) - Loss of Coolant Accident with Loss of Loss of Coolant Accident with Loss Emergency Cooling and Heat Sink, but of Emergency Cooling, Heat Sink, Confinement Operational and Confinement H-3 1,500 48,000 W-185 14,500 925,000 W-187 10,600 675,000 W-181 2,850 181,000 W-178 910 57,900 Xe-127 51 1,640 W-177 47 3,000 W-176 42 2,660 Cs-131 38 2,440 Xe-125 29 930 Cs-129 25 1,610 Cs-128 22 1,380 I-125 21 1,340 Ar-37 11 340 P-32 11 710 Cs-127 10 670 Te-121 9 290 I-123 8 510 Kr-79 7 220 Re-186 7 420 Xe-122 4 115 Source: SNL 1995a; SNL 1995b:1. Table F.2.1.4.2-2.-Source Term for Phased Accelerator Production of Tritium with Helium-3 Target System High Consequence Accidents Isotope Released Activity (curies) Isotope Released Activity (curies) H-3 1,500 Cs-128 13 W-185 8,700 I-125 13 W-187 6,400 Ar-37 6.6 W-181 1,700 P-32 6.6 W-178 550 Cs-127 6.0 Xe-127 31 Te-121 5.4 W-177 28 I-123 4.8 W-176 25 Kr-79 4.2 Cs-131 23 Re-186 4.2 Xe-125 17 Xe-122 <1 Cs-129 15 - - Source: SNL 1995a. Table F.2.1.4.2-3.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 8.7x10-6 4.3x10-9 0.014 7.2x10-6 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 2.8x10-4 1.4x10-7 0.46 2.3x10-4 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 6.2x10-9 - 1.0x10-5 - Expected risk (per year) - 4.4x10-15 - 7.4x10-12 - Table F.2.1.4.2-4.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Nevada Test Site- Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 2.4x10-5 1.2x10-8 1.4x10-3 6.9x10-7 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 7.5x10-4 3.8x10-7 0.044 2.2x10-5 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 1.7x10-8 - 9.9x10-7 - Expected risk (per year) - 1.2x10-14 - 7.0x10-13 - Table F.2.1.4.2-5.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 1.9x10-4 3.0x10-8 0.13 6.7x10-5 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 5.9x10-3 3.0x10-6 4.3 2.2x10-3 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 1.3x10-7 - 9.6x10-5 - Expected risk (per year) - 9.5x10-14 - 6.8x10-11 - Table F.2.1.4.2-6.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 1.2x10-4 6.2x10-8 0.017 8.7x10-6 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 4.0x10-3 2.0x10-6 0.56 2.8x10-4 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 9.0x10-8 - 1.3x10-5 - Expected risk (per year) - 6.4x10-14 - 8.9x10-12 - Table F.2.1.4.2-7.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 8.0x10-6 4.0x10-9 0.054 2.7x10-5 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 2.6x10-4 1.3x10-7 1.7 8.6x10-4 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 5.7x10-9 - 3.9x10-5 - Expected risk (per year) - 4.1x10-15 - 2.8x10-11 - Table F.2.1.4.2-8.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 1.1x10-3 4.3x10-7 3.9x10-4 1.6x10-7 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.034 1.4x10-5 0.013 5.0x10-6 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 6.1x10-7 - 2.3x10-7 - Expected risk (per year) - 4.4x10-13 - 1.6x10-13 - Table F.2.1.4.2-9.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 7.8x10-3 3.1x10-7 3.0x10-4 1.2x10-7 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.025 1.0x10-5 9.6x10-3 3.9x10-6 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 4.5x10-7 - 1.7x10-7 - Expected risk (per year) - 3.2x10-13 - 1.2x10-13 - Table F.2.1.4.2-10.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 1.0x10-3 4.2x10-7 3.7x10-4 1.5x10-7 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.034 1.3x10-5 0.012 4.8x10-6 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 6.0x10-7 - 2.2x10-7 - Expected risk (per year) - 4.3x10-13 - 1.5x10-13 - Table F.2.1.4.2-11.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 4.6x10-4 1.8x10-7 1.7x10-4 7.0x10-8 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.015 5.9x10-6 5.6x10-3 2.2x10-6 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 2.6x10-7 - 1.0x10-7 - Expected risk (per year) - 1.9x10-13 - 7.1x10-14 - Table F.2.1.4.2-12.-Full Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 4.7x10-4 1.9x10-7 1.7x10-4 7.0x10-8 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.015 5.9x10-6 5.6x10-3 2.2x10-6 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 2.7x10-7 - 1.0x10-7 - Expected risk (per year) - 1.9x10-13 - 7.1x10-14 - Table F.2.1.4.2-13.-Phased Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accident-Public Consequences - Individual at Site Boundary Population to 50 Miles - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering Laboratory 8.7x10-6 4.3x10-9 0.014 7.2x10-6 7.0x10-7 Nevada Test Site 2.4x10-5 1.2x10-8 1.4x10-3 6.9x10-7 7.0x10-7 Oak Ridge Reservation 1.9x10-4 9.3x10-8 0.13 6.7x10-5 7.0x10-7 Pantex Plant 1.2x10-4 6.2x10-8 0.017 8.7x10-6 7.0x10-7 Savannah River Site 8.0x10-6 4.0x10-9 0.054 2.7x10-5 7.0x10-7 Expected Risk of Cancer Fatality (per year) Idaho National Engineering Laboratory - 3.0x10-15 - 5.0x10-12 - Nevada Test Site - 8.3x10-15 - 4.8x10-13 - Oak Ridge Reservation - 6.5x10-14 - 4.7x10-11 - Pantex Plant - 4.4x10-14 - 6.1x10-12 - Savannah River Site - 2.8x10-15 - 1.9x10-11 - Table F.2.1.4.2-14.-Phased Accelerator Production of Tritium with the Helium-3 Target System High Consequence Accident-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering Laboratory 1.1x10-3 4.3x10-7 3.9x10-4 1.6x10-7 7.0x10-7 Nevada Test Site 7.8x10-4 3.1x10-7 3.0x10-4 1.2x10-7 7.0x10-7 Oak Ridge Reservation 1.0x10-3 4.2x10-7 3.7x10-4 1.5x10-7 7.0x10-7 Pantex Plant 4.6x10-4 1.8x10-7 1.7x10-4 7.0x10-8 7.0x10-7 Savannah River Site 4.7x10-4 1.9x10-7 1.7x10-4 7.0x10-8 7.0x10-7 Expected Risk of Cancer Fatality (per year) Idaho National Engineering Laboratory - 3.0x10-13 - 1.1x10-13 - Nevada Test Site - 2.2x10-13 - 8.4x10-14 - Oak Ridge Reservation - 2.9x10-13 - 1.0x10-13 - Pantex Plant - 1.3x10-13 - 4.9x10-14 - Savannah River Site - 1.3x10-13 - 4.9x10-14 - F.2.1.4.3 Spallation-Induced Lithium Conversion Target System Loss of Coolant Accident with Loss of Emergency Cooling and Natural Circulation, but Confinement Operational Scenario. The postulated high consequence accident for the Full APT with the spallation-induced lithium conversion target system configuration is a large break loss of coolant accident, followed by a successful beam trip, but total failure of the active and passive cooling systems. This scenario would lead to partial melting of the target. Based on these analyses, a bounding source term release to the environment was determined. Table F.2.1.4.3-1 presents the source term released during the accident. The analysis did not estimate the accident annual frequency of occurrence (SNL 1995a:8-12-8-14). The postulated accident sequence assumed that the only safety system to function is the passive water dump tank that floods the target room in the event of a loss of coolant accident. The postulated accident sequence assumed that the confinement system remained operational. The probability of the accident is in the residual risk category, but it is within the design basis of confinement (SNL 1995a:8-12). The accident annual frequency of occurrence is estimated at 7.0x10-7 per year (SNL1995b:1). Consequences. The estimated consequences of the postulated accident with at each site are shown in tables F.2.1.4.3-2 through F.2.1.4.3-6 for public consequences and in tables F.2.1.4.3-7 through F.2.1.4.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.4.3-1 using the MACCS computer code. Loss of Coolant Accident with Loss of Emergency Cooling, Natural Circulation, and Confinement Scenario. The postulated bounding high consequence accident for the Full APT with the spallation-induced lithium conversion target system is a large break loss of coolant accident with total failure of the active emergency cooling system, loss of natural cir- culation, and loss of confinement. The source term is presented in table F.2.1.4.3-1. The annual frequency of occurrence for this accident is 1.0x10-8 per year (SNL 1995b:1). Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.4.3-2 through F.2.1.4.3-6 for public consequences and in tables F.2.1.4.3-7 through F.2.1.4.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.4.3-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Two Full Accelerator Production of Tritium with Spallation-Induced Lithium Conversion Target System High Consequence Accidents Figure F.2.1.4.3-1 shows the annual probability that, in the event of any accident in the composite set of Full APT with spallation-induced lithium conversion target system high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.4.3-2 through F.2.1.4.3-11. Figure (Page F-67) Figure F.2.1.4.3-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for the Full Accelerator Production of Tritium with Spallation-Induced Lithium Conversion Target System. Table F.2.1.4.3-1.-Source Term for Full Accelerator Production of Tritium with Spallation-Induced Lithium Conversion Target System High Consequence Accident - Released Activity (curies) Isotope Loss of Coolant Accident with Loss Loss of Coolant Accident with Loss of Emergency Cooling and Natural of Emergency Cooling, Circulation, but Confinement Natural Circulation, and Operational Confinement H-3 1,900 61,000 Hg-197 1,065 68,000 F-18 1,039 66,000 Kr-83m 1,039 33,000 Hg-195 518 33,000 Kr-79 477 15,000 Xe-125 465 15,000 Xe-127 320 10,000 Kr-88 259 8,200 Kr-85m 258 8,200 Br-83 243 16,000 Kr-87 221 7,000 Hg-193 211 13,000 Br-82 193 12,000 Br-76 177 11,000 Hg-203 136 8,600 Hg-192 115 7,300 I-125 113 7,200 I-123 101 6,400 I-126 84 5,400 Br-84 83 5,300 Br-77 79 5,000 Xe-122 77 2,400 I-121 76 4,900 I-124 64 4,100 I-120 55 3,500 I-130 54 3,500 I-128 45 2,900 Hg-197m 40 2,500 I-122 38 2,500 I-131 27 1,700 Hg-195m 20 1,300 Hg-190 14 910 I-133 13 820 I-135 12 760 Source: SNL 1995a; SNL 1995b:1. Table F.2.1.4.3-2.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 1.5x10-4 7.6x10-8 0.11 5.2x10-5 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 7.8x10-3 3.9x10-6 6.0 3.0x10-3 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 1.3x10-7 - 9.4x10-5 - Expected risk (per year) - 9.2x10-14 - 6.7x10-11 - Table F.2.1.4.3-3.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 3.8x10-4 1.9x10-7 0.01 5.0x10-6 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.02 1.0x10-5 0.58 2.9x10-4 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 3.3x10-7 - 9.0x10-6 - Expected risk (per year) - 2.3x10-13 - 6.4x10-12 - Table F.2.1.4.3-4.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 2.5x10-3 1.3x10-6 1.2 5.9x10-4 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.14 6.9x10-5 66 0.033 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 2.2x10-6 - 1.0x10-3 - Expected risk for (per year) - 1.6x10-12 - 7.4x10-10 - Table F.2.1.4.3-5.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 1.6x10-3 8.2x10-7 0.15 7.6x10-5 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.09 4.5x10-5 8.5 4.3x10-3 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 1.4x10-6 - 1.3x10-4 - Expected risk (per year) - 1.0x10-12 - 9.6x10-11 - Table F.2.1.4.3-6.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 1.2x10-4 6.0x10-8 0.43 2.1x10-4 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 6.3x10-3 3.1x10-6 24 0.012 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences for composite set of accidents - 1.0x10-7 - 3.8x10-4 - Expected risk (per year) - 7.3x10-14 - 2.7x10-10 - Table F.2.1.4.3-7.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 0.013 5.3x10-6 5.4x10-3 2.2x10-6 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.75 3.0x10-4 0.3 1.2x10-4 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 9.4x10-6 - 3.8x10-6 - Expected risk (per year) - 6.7x10-12 - 2.7x10-12 - Table F.2.1.4.3-8.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 9.4x10-3 3.8x10-6 4.0x10-3 1.6x10-6 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.53 2.1x10-4 0.22 8.9x10-5 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected Consequences - 6.7x10-6 - 2.8x10-6 - Expected Risk (per year) - 4.8x10-12 - 2.0x10-12 - Table F.2.1.4.3-9.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 0.012 4.9x10-6 4.8x10-3 1.9x10-6 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.69 2.8x10-4 0.26 1.1x10-4 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 8.7x10-6 - 3.4x10-6 - Expected risk (per year) - 6.2x10-12 - 2.4x10-12 - Table F.2.1.4.3-10.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 5.3x10-3 2.1x10-6 2.2x10-3 8.9x10-7 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.3 1.2x10-4 0.12 4.9x10-5 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 3.8x10-6 - 1.6x10-6 - Expected risk (per year) - 2.7x10-12 - 1.1x10-12 - Table F.2.1.4.3-11.-Full Accelerator Production of Tritium with the Spallation-Induced Lithium Conversion Target System High Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loss of coolant accident with loss of emergency cooling and heat sink, but 5.4x10-3 2.2x10-6 2.2x10-3 8.9x10-7 7.0x10-7 confinement operational Loss of coolant accident with loss of emergency cooling, heat sink, and 0.3 1.2x10-4 0.12 4.9x10-5 1.0x10-8 confinement Evaluation of Composite Set of Accidents Expected consequences - 3.8x10-6 - 1.6x10-6 - Expected risk (per year) - 2.7x10-12 - 1.1x10-12 - F.2.1.5 Multipurpose Reactor Facility The multipurpose reactor facility consists of three elements. (1) The reactor element that burns the plutonium or mixed-oxide fuel can be either a Modular High Temperature Gas-Cooled Reactor or an Advanced Light Water Reactor. (2) The fuel fabrication element produces the fuel for use in the reactor. (3) The pit disassembly and conversion element disassembles plutonium pits and converts the plutonium in the pit to plutonium-oxide which is used in the production of plutonium or mixed-oxide fuel. F.2.1.5.1 Multipurpose Reactor Modular High Temperature Gas-Cooled Reactor The use of plutonium in the plutonium-oxide fueled MHTGR will not have a significant effect on the source term for high consequence accidents generated for the uranium fueled MHTGR because no fuel failures are expected (HNUS 1995c:1). The accident consequences estimated for the uranium fueled MHTGR are applicable for the plutonium fueled MHTGR. Refer to section F.2.1.2 for the applicable accident consequences of the plutonium fueled MHTGR. Advanced Light Water Reactor The use of plutonium in the mixed-oxide fueled ALWR, as compared to the uranium-fueled ALWR, will not significantly affect the consequence of radioactivity releases for high consequence accidents. While there will be some small changes in the source term release spectrum and frequency, the changes will not have a significant effect on the accident consequences (HNUS 1995c:2). The accident consequences estimated for the uranium-fueled ALWR are applicable for the mixed-oxide fueled ALWR. Refer to section F.2.1.3 for the applicable accident consequences of the mixed-oxide fueled ALWR. F.2.1.5.2 Mixed-Oxide and Plutonium-Oxide Fuel Fabrication Criticality Scenario. The postulated solid criticality accident is the result of accidental improper stacking of items. There will not be sufficient quantities of plutonium solutions in the fuel fabrication area to cause a liquid criticality accident if mishandled. It is assumed that the postulated solid criticality incident would not exceed 5.0x1017 fissions. Table F.2.1.5.2-2 presents the source term for important nuclides released to the environment during the postulated criticality accident. The annual frequency of occurrence for the criticality accident is estimated to be less than 1.0x10-7 per year (LANL 1995d). For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-7 per year. Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.5.2-3 through F.2.1.5.2-7 for public consequences and in tables F.2.1.5.2-8 through F.2.1.5.2-12 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.5.2-2 using the MACCS computer code. Beyond Design-Basis Fire Scenario. The accident postulated is a fire in a process cell area with coincident failure of major safety systems. It is assumed that the process cell contains a glovebox used for milling plutonium powder. The gloves have become coated with a layer of plutonium dust. The analysis estimated the glove loading at 2 grams of plutonium per glove. Each of the 12 gloves is assumed to be stowed outside of the glovebox. A flammable cleaning liquid such as acetone or isopropyl alcohol is brought into the process cell in violation of operating procedures, spills, and ignites. All gloves are incinerated, but the sprinkler system does not activate to protect the glovebox from further damage. The ventilation system and HEPA filters are also assumed inoperative. Normally closed doors are assumed to remain closed except during personnel evacuation from the area. The analysis using the LANL computer code known as GASFLOW was used to model the dispersion of the fire products. The analysis estimated that 0.034 gram of plutonium is released to the environment. The annual frequency of occurrence is estimated to be less than 1.0x10-7 per year (LANL 1995d). For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-7 per year. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the fuel fabrication facility. Table F.2.1.5.2-1 presents the source term, by isotope, for the 0.034 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.5.2-3 through F.2.1.5.2-7 for public consequences and in tables F.2.1.5.2-8 through F.2.1.5.2-12 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.5.2-1 using the MACCS computer code. Beyond Design-Basis Explosion Scenario. The explosion of an oxyacetylene bottle in a process cell has been postulated as a beyond design-basis explosion. The explosion has the potential to blow out the HEPA filters and cause significant damage to the ventilation system and nearby equipment. The explosion is postulated to occur in a process cell near a glovebox. The glovebox identi- fied as having the most material-at-risk contains the milling operation where plutonium-oxide is milled to a fine powder prior to mixing with uranium dioxide. Based on a LANL TA-55 standard operating procedure, the criticality limit for plutonium-oxide in a dry atmosphere is assumed to be 10 kg. The analysis assumed the glovebox contains 10 kg of plutonium-oxide. The analysis estimated that 50 grams of plutonium are released up the stack. Sufficient control on the use of oxyacetylene welding equipment in process cells ensures that the probability of an accident occurring is less than 1.0x10-7 per year (LANL 1995d). For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-7 per year. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the fuel fabrication facility. Table F.2.1.5.2-1 presents the source term, by isotope, for the 50 grams of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.5.2-3 through F.2.1.5.2-7 for public consequences and in tables F.2.1.5.2-8 through F.2.1.5.2-12 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.5.2-1 using the MACCS computer code. Beyond Design-Basis Earthquake Scenario. The following assumptions are made for a beyond design-basis earthquake analysis: (1) the ventilation system is disabled, (2) there is significant structural damage but the building does not totally collapse, (3) a ceiling slab falls on a glovebox with the most material-at-risk (10 kg of plutonium-oxide powder) and the glovebox is significantly damaged, (4) the process cell with the glovebox has one wall on the outside of the building, (5) this outside wall cracks and the cracks have a total length of 10 meters and a 1-mm width, (6) the wind is blowing at 10 m/s, and (7) the cracks are located on the lee side of the building. The analysis estimated that 25 grams of plutonium were released at the building level. The annual frequency of occurrence is estimated to be less than 1.0x10-7 per year (LANL 1995d). For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-7 per year. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the fuel fabrication facility. Table F.2.1.5.2-1 presents the source term, by isotope, for the 25 grams of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.5.2-3 through F.2.1.5.2-7 for public consequences and in tables F.2.1.5.2-8 through F.2.1.5.2-12 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.5.2-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Four Multipurpose Fuel Fabrication High Consequence Accidents Figure F.2.1.5.2-1 shows the annual probability that, in the event of any accident in the composite set of mixed-oxide fuel fabrication high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.5.2-3 through F.2.1.5.2-12. Figure (Page F-76) Figure F.2.1.5.2-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for the Multipurpose Reactor Fuel Fabrication Facility. Table F.2.1.5.2-1.-Multipurpose Reactor Fuel Fabrication High Consequence Accident Source Terms - Released Activity (curies) Isotope Criticality Beyond Design- Beyond Design- Beyond Design- Basis Fire Basis Explosion Basis Earthquake Pu-238 0 2.9x10-4 0.42 0.21 Pu-239 0 2.0x10-3 2.9 1.4 Pu-240 0 4.6x10-4 0.67 0.34 Pu-241 0 0.014 20 10 Am-241 0 4.4x10-4 0.64 0.32 Kr-83m 2.8 0 0 0 Kr-85m 1.8 0 0 0 Kr-85 2.0x10-4 0 0 0 Kr-87 11 0 0 0 Kr-88 5.8 0 0 0 Kr-89 325 0 0 0 Xe-131m 2.5x10-3 0 0 0 Xe-133m 0.05 0 0 0 Xe-133 0.75 0 0 0 Xe-135m 83 0 0 0 Xe-135 10 0 0 0 Xe-137 1.2x103 0 0 0 Xe-138 275 0 0 0 I-131 0.025 0 0 0 I-132 3 0 0 0 I-133 0.4 0 0 0 I-134 11 0 0 0 I-135 1.1 0 0 0 Source: Derived from LANL1995 and table F.2.1.5.2-1. Table F.2.1.5.2-2.-Isotopic Distribution for a Plutonium Release Isotope Isotope/ Plutonium Specific Activity Specific Activity (gram) of Isotope (Ci Isotope/g Plutonium) (Ci/g) Pu-238 5.0x10-4 16.8 8.4x10-3 Pu-239 0.933 0.0616 0.0575 Pu-240 0.059 0.227 0.0134 Pu-241 3.5x10-3 115 0.403 Am-241 4.0x10-3 3.2 0.0128 Total Specific Activity - - 0.495 (Ci/g plutonium) Source: Derived from LANL1995i:1. Table F.2.1.5.2-3.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 3.1x10-6 1.5x10-9 6.7x10-4 3.4x10-7 1.0x10-7 Beyond design-basis fire 1.0x10-4 5.1x10-8 0.18 8.9x10-5 1.0x10-7 Beyond design-basis explosion 0.15 7.4x10-5 258 0.13 1.0x10-7 Beyond design-basis earthquake 0.073 3.6x10-5 127 0.063 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 2.8x10-5 - 0.048 - Expected risk (per year) - 1.1x10-11 - 1.9x10-8 - Table F.2.1.5.2-4.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 8.4x10-6 4.2x10-9 6.2x10-5 3.1x10-8 1.0x10-7 Beyond design-basis fire 2.8x10-4 1.4x10-7 0.017 8.5x10-6 1.0x10-7 Beyond design-basis explosion 0.4 2.0x10-4 25 0.012 1.0x10-7 Beyond design-basis earthquake 0.2 9.9x10-5 12 6.1x10-3 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 7.5x10-5 - 4.6x10-3 - Expected risk (per year) - 3.0x10-11 - 1.8x10-9 - Table F.2.1.5.2-5.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 5.5x10-5 2.7x10-8 0.012 6.2x10-6 1.0x10-7 Beyond design-basis fire 2.2x10-3 1.1x10-6 1.6 8.2x10-4 1.0x10-7 Beyond design-basis explosion 3.2 1.6x10-3 2.4x103 1.2 1.0x10-7 Beyond design-basis earthquake 1.6 7.9x10-4 1.2x103 0.58 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 6.0x10-4 - 0.44 - Expected risk (per year) - 2.4x10-10 - 1.8x10-7 - Table F.2.1.5.2-6.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 3.9x10-5 1.9x10-8 1.7x10-3 8.5x10-7 1.0x10-7 Beyond design-basis fire 1.5x10-3 7.4x10-7 0.21 1.1x10-4 1.0x10-7 Beyond design-basis explosion 2.2 1.1x10-3 308 0.15 1.0x10-7 Beyond design-basis earthquake 1.1 5.3x10-4 152 0.076 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 4.0x10-4 - 0.057 - Expected risk (per year) - 1.6x10-10 - 2.3x10-8 - Table F.2.1.5.2-7.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 2.6x10-6 1.3x10-9 3.4x10-3 1.7x10-6 1.0x10-7 Beyond Design Basis Fire 9.5x10-5 4.8x10-8 0.66 3.3x10-4 1.0x10-7 Beyond Design Basis Explosion 0.14 6.9x10-5 959 0.48 1.0x10-7 Beyond Design Basis Earthquake 0.068 3.4x10-5 471 0.24 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 2.6x10-5 - 0.18 - Expected risk (per year) - 1.0x10-11 - 7.2x10-8 - Table F.2.1.5.2-8.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 2.7x10-4 1.1x10-7 1.2x10-4 4.7x10-8 1.0x10-7 Beyond design-basis fire 0.013 5.1x10-6 4.7x10-3 1.9x10-6 1.0x10-7 Beyond design-basis explosion 18 9.1x10-3 6.8 2.9x10-3 1.0x10-7 Beyond design-basis earthquake 9 3.6x10-3 3.3 1.3x10-3 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 3.2x10-3 - 1.1x10-3 - Expected risk (per year) - 1.3x10-9 - 4.2x10-10 - Table F.2.1.5.2-9.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 2.0x10-4 7.9x10-8 8.9x10-5 3.5x10-8 1.0x10-7 Beyond design-basis fire 9.3x10-3 3.7x10-6 3.6x10-3 1.4x10-6 1.0x10-7 Beyond design-basis explosion 14 6.2x10-3 5.2 2.2x10-3 1.0x10-7 Beyond design-basis earthquake 6.6 2.7x10-3 2.6 1.0x10-3 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 2.2x10-3 - 8.0x10-4 - Expected risk (per year) - 8.9x10-10 - 3.2x10-10 - Table F.2.1.5.2-10.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 2.5x10-4 1.0x10-7 1.0x10-4 4.1x10-8 1.0x10-7 Beyond design-basis fire 0.013 5.0x10-6 4.5x10-3 1.8x10-6 1.0x10-7 Beyond design-basis explosion 18 8.5x10-3 6.5 2.7x10-3 1.0x10-7 Beyond design-basis earthquake 9 3.6x10-3 3.2 1.3x10-3 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 3.0x10-3 - 1.0x10-3 - Expected risk (per year) - 1.2x10-9 - 4.0x10-10 - Table F.2.1.5.2-11.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 1.2x10-4 4.6x10-8 5.1x10-5 2.1x10-8 1.0x10-7 Beyond design-basis fire 5.5x10-3 2.2x10-6 2.1x10-3 8.4x10-7 1.0x10-7 Beyond design-basis explosion 8 3.3x10-3 3 1.2x10-3 1.0x10-7 Beyond design-basis earthquake 3.9 1.6x10-3 1.5 6.0x10-4 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.2x10-3 - 4.6x10-4 - Expected risk (per year) - 4.9x10-10 - 1.8x10-10 - Table F.2.1.5.2-12.-Multipurpose Reactor Fuel Fabrication High Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 1.2x10-4 4.6x10-8 5.0x10-5 2.0x10-8 1.0x10-7 Beyond design-basis fire 5.6x10-3 2.2x10-6 2.1x10-3 8.3x10-7 1.0x10-7 Beyond design-basis explosion 8.1 3.5x10-3 3 1.2x10-3 1.0x10-7 Beyond design-basis earthquake 4 1.6x10-3 1.5 5.9x10-4 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.3x10-3 - 4.6x10-4 - Expected risk (per year) - 5.0x10-10 - 1.8x10-10 - F.2.1.5.3 Pit Disassembly and Conversion Criticality Scenario. The postulated solid criticality accident is the result of accidental improper stacking of items. There will not be sufficient quantities of plutonium solutions in the mixed-oxide fuel fabrication area to cause a liquid criticality accident if mishandled. It is assumed that the postulated solid criticality incident would not exceed 5.0x1017 fissions. Table F.2.1.5.3-1 presents the source term for important nuclides released to the environment during the postulated criticality accident. The annual frequency of occurrence for the criticality accident is estimated to be less than 1.0x10-7 per year (LANL 1995b:1). For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-7 per year. Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.5.3-2 through F.2.1.5.3-6 for public consequences and in tables F.2.1.5.3-7 through F.2.1.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.5.3-1 using the MACCS computer code. Beyond Design-Basis Fire Scenario. The accident postulated is a fire in a process cell area with coincident failure of major safety systems. It is assumed that the process cell contains a glovebox used for milling plutonium powder. The gloves have become coated with a layer of plutonium dust. The analysis estimated the glove loading at 2 grams of plutonium per glove. Each of the 12 gloves is assumed to be stowed outside of the glovebox. A flammable cleaning liquid such as acetone or isopropyl alcohol is brought into the process cell in violation of operating procedures, spills and ignites. All gloves are incinerated, but the sprinkler system does not activate to protect the glovebox from further damage. The ventilation system and HEPA filters are also assumed inoperative. Normally, closed doors are assumed to remain closed except during personnel evacuation from the area. The analysis using the LANL computer code known as GASFLOW was used to model the dispersion of the fire products. The analysis estimated that 0.034 gram of plutonium is released to the environment. The annual frequency of occurrence is estimated to be less than 1.0x10-7 per year. (LANL1995b:1) For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-7 per year. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.1.5.3-1 presents the source term, by isotope, for the 0.034gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.5.3-2 through F.2.1.5.3-6 for public consequences and in tables F.2.1.5.3-7 through F.2.1.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.5.3-1 using the MACCS computer code. Beyond Design-Basis Explosion Scenario. The explosion of an oxyacetylene bottle in a process cell has been postulated as a beyond design-basis explosion. The explosion has the potential to blow out the HEPA filters and cause significant damage to the ventilation system and nearby equipment. The explosion is postulated to occur in a process cell near a glovebox. The glovebox identi- fied as having the most material at risk contains the milling operation where plutonium-oxide is milled to a fine powder prior to mixing with uranium dioxide. Based on a LANL TA-55 standard operating procedure, the criticality limit for plutonium-oxide in a dry atmosphere is assumed to be 4.5 kg. The analysis assumed the glovebox contains 4.5 kg of plutonium-oxide. The analysis estimated that 22.5 grams of plutonium is released up the stack. Sufficient control on the use of oxyacetylene welding equipment in process cells ensures that the probability of an accident occurring is less than 1.0x10-7 per year. (LANL 1995b:1) For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-7 per year. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.1.5.3-1 presents the source term, by isotope, for the 22.5grams of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.5.3-2 through F.2.1.5.3-6 for public consequences and in tables F.2.1.5.3-7 through F.2.1.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.5.3-1 using the MACCS computer code. Beyond Design-Basis Earthquake Scenario. The following assumptions are made for a beyond design-basis earthquake analysis: (1) the ventilation system is disabled, (2) there is significant structural damage but the building does not totally collapse, (3) a ceiling slab falls on a glovebox with the most material at risk (4.5 kg of plutonium-oxide powder) and the glovebox is significantly damaged, (4) the process cell with the glovebox has one wall on the outside of the building, (5) this outside wall cracks and the cracks have a total length of 10 meters and a 1-mm width, (6) the wind is blowing at 10 m/s, and (7) the cracks are located on the lee side of the building. The analysis estimated that 11.3 g of plutonium was released at the building level. The annual frequency of occurrence is estimated to be less than 1.0x10-7 per year. (LANL1995b:1) For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-7 per year. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.1.5.3-1 presents the source term, by isotope, for the 11.3 grams of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident at each site are shown in tables F.2.1.5.3-2 through F.2.1.5.3-6 for public consequences and in tables F.2.1.5.3-7 through F.2.1.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.1.5.3-1 using the MACCS computer code. Integrated Cancer Fatalities Complementary Cumulative Distribution Function for the Four Pit Disassembly and Conversion High Consequence Accidents Figure F.2.1.5.3-1 shows the annual probability that, in the event of any accident in the composite set of mixed-oxide fuel fabrication high consequence accidents at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.5.3-2 through F.2.1.5.3-11. Figure (Page F-85) Figure F.2.1.5.3-1.-High Consequence Accident-Cancer Fatality Frequency Distributions Functions for the Disassembly and Conversion Facility. Table F.2.1.5.3-1.-Pit Disassembly and Conversion High Consequence Accident Source Terms Isotope Released Activity (curies) - Criticality Beyond Design-Basis Beyond Design-Basis Beyond Design-Basis Fire Explosion Earthquake Pu-238 0 2.9x10-4 0.19 0.095 Pu-239 0 2.0x10-3 1.3 0.65 Pu-240 0 4.6x10-4 0.3 0.15 Pu-241 0 .014 9.1 4.6 Am-241 0 4.4x10-4 0.29 0.14 Kr-83m 5.5 0 0 0 Kr-85m 3.6 0 0 0 Kr-85 4.1x10-5 0 0 0 Kr-87 22 0 0 0 Kr-88 12 0 0 0 Kr-89 650 0 0 0 Xe-131m 5.0x10-3 0 0 0 Xe-133m 0.11 0 0 0 Xe-133 1.4 0 0 0 Xe-135m 165 0 0 0 Xe-135 21 0 0 0 Xe-137 2.5x103 0 0 0 Xe-138 550 0 0 0 I-131 0.14 0 0 0 I-132 15 0 0 0 I-133 2 0 0 0 I-134 54 0 0 0 I-135 5.4 0 0 0 Source: Derived from LANL 1995def and table F.2.1.5.3-1. Table F.2.1.5.3-2.-Pit Disassembly and Conversion High Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 1.1x10-5 5.7x10-9 3.2x10-3 1.6x10-6 1.0x10-7 Beyond design-basis fire 1.0x10-4 5.1x10-8 0.18 8.9x10-5 1.0x10-7 Beyond design-basis explosion 0.067 3.3x10-5 116 0.058 1.0x10-7 Beyond design-basis earthquake 0.033 1.7x10-5 58 0.029 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.2x10-5 - 0.022 - Expected risk (per year) - 5.0x10-12 - 8.7x10-9 - Table F.2.1.5.3-3.-Pit Disassembly and Conversion High Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 3.3x10-5 1.6x10-8 3.0x10-4 1.5x10-7 1.0x10-7 Beyond design-basis fire 2.8x10-4 1.4x10-7 0.017 8.5x10-6 1.0x10-7 Beyond design-basis explosion 0.18 9.1x10-5 11 5.5x10-3 1.0x10-7 Beyond design-basis earthquake 0.091 4.5x10-5 5.5 2.8x10-3 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 3.4x10-5 - 2.1x10-3 - Expected risk (per year) - 1.4x10-11 - 8.3x10-10 - Table F.2.1.5.3-4.-Pit Disassembly and Conversion High Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 2.3x10-4 1.1x10-7 0.052 2.6x10-5 1.0x10-7 Beyond design basis fire 2.2x10-3 1.1x10-6 1.6 8.2x10-4 1.0x10-7 Beyond design basis explosion 1.4 7.2x10-4 1.1x103 0.53 1.0x10-7 Beyond design basis earthquake 0.72 3.6x10-4 530 0.27 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 2.7x10-4 - 0.20 - Expected risk (per year) - 1.1x10-10 - 8.0x10-8 - Table F.2.1.5.3-5.-Pit Disassembly and Conversion High Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 1.7x10-4 8.4x10-8 7.4x10-3 3.7x10-6 1.0x10-7 Beyond design-basis fire 1.5x10-3 7.4x10-7 0.21 1.1x10-4 1.0x10-7 Beyond design-basis explosion 0.97 4.9x10-4 139 0.069 1.0x10-7 Beyond design-basis earthquake 0.48 2.4x10-4 69 0.035 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.8x10-4 - 0.026 - Expected risk (per year) - 7.3x10-11 - 1.0x10-8 - Table F.2.1.5.3-6.-Pit Disassembly and Conversion High Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Criticality 1.0x10-5 5.1x10-9 0.016 7.7x10-6 1.0x10-7 Beyond design-basis fire 9.5x10-5 4.8x10-8 0.66 3.3x10-4 1.0x10-7 Beyond design-basis explosion 0.062 3.1x10-5 432 0.22 1.0x10-7 Beyond design-basis earthquake 0.032 1.5x10-5 215 0.11 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.2x10-5 - 0.081 - Expected risk (per year) - 4.6x10-12 - 3.2x10-8 - Table F.2.1.5.3-7.-Pit Disassembly and Conversion High Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 1.2x10-3 4.9x10-7 5.0x10-4 2.0x10-7 1.0x10-7 Beyond design-basis fire 0.013 5.1x10-6 4.7x10-3 1.9x10-6 1.0x10-7 Beyond design-basis explosion 8.3 3.3x10-3 3.0 1.2x10-3 1.0x10-7 Beyond design-basis earthquake 4.1 1.7x10-3 1.5 6.1x10-4 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.2x10-3 - 4.6x10-4 - Expected risk (per year) - 5.0x10-10 - 1.8x10-10 - Table F.2.1.5.3-8.-Pit Disassembly and Conversion High Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 8.8x10-4 3.5x10-7 3.8x10-4 1.5x10-7 1.0x10-7 Beyond design-basis fire 9.3x10-3 3.7x10-6 3.6x10-3 1.4x10-6 1.0x10-7 Beyond design-basis explosion 6.1 2.4x10-3 2.3 9.3x10-4 1.0x10-7 Beyond design-basis earthquake 3. 1.2x10-3 1.2 4.6x10-4 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 9.1x10-4 - 3.5x10-4 - Expected risk (per year) - 3.6x10-10 - 1.4x10-10 - Table F.2.1.5.3-9.-Pit Disassembly and Conversion High Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 1.1x10-3 4.5x10-7 4.4x10-4 1.8x10-7 1.0x10-7 Beyond design-basis fire 0.013 5.0x10-6 4.5x10-3 1.8x10-6 1.0x10-7 Beyond design-basis explosion 8.2 3.3x10-3 2.9 1.2x10-3 1.0x10-7 Beyond design-basis earthquake 4.1 1.6x10-3 1.5 5.8x10-4 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 1.2x10-3 - 4.4x10-4 - Expected risk (per year) - 4.9x10-10 - 1.8x10-10 - Table F.2.1.5.3-10.-Pit Disassembly and Conversion High Consequence Accidents at Pantex Plant -Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 5.2x10-4 2.1x10-7 2.2x10-4 9.0x10-8 1.0x10-7 Beyond design-basis fire 5.5x10-3 2.2x10-6 2.1x10-3 8.4x10-7 1.0x10-7 Beyond design-basis explosion 3.6 1.4x10-3 1.4 5.5x10-4 1.0x10-7 Beyond design-basis earthquake 1.8 7.1x10-4 0.68 2.7x10-4 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 5.4x10-4 - 2.0x10-4 - Expected risk (per year) - 2.1x10-10 - 8.2x10-11 - Table F.2.1.5.3-11.-Pit Disassembly and Conversion High Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Criticality 5.2x10-4 2.1x10-7 2.2x10-4 8.7x10-8 1.0x10-7 Beyond design-basis fire 5.6x10-3 2.2x10-6 2.1x10-3 8.3x10-7 1.0x10-7 Beyond design-basis explosion 3.6 1.5x10-3 1.4 5.4x10-4 1.0x10-7 Beyond design-basis earthquake 1.8 7.2x10-4 0.68 2.7x10-4 1.0x10-7 Evaluation of Composite Set of Accidents Expected consequences - 5.4x10-4 - 2.0x10-4 - Expected risk (per year) - 2.2x10-10 - 8.2x10-11 - F.2.1.6 Tritium Target Extraction Facility Scenario. A tritium extraction facility removes tritium from targets. The accidents for the tritium extraction facility are based on the analysis of tritium operation at SRS. The high consequence accident for the facility postulated a beyond design-basis earthquake that caused the release of major portions of the process vessel tritium inventory. Approximately 2.4x106 Ci of tritium in oxide form could be released to the environment. The accident annual frequency of occurrence is estimated at 1.4x10-4 per year at SRS (DOE 1995d). The accident annual frequency of occurrence for new facilities at the other candidate sites will be less than the frequency for existing facilities at SRS. It is assumed that the process systems, tanks, and confinement systems will be designed to maintain functional integrity following a design-basis earthquake or a safe shutdown earthquake with a return frequency of 1.0x10-4 per year. The evaluation also assumed that the process system pressure boundary and/or some of the active or passive safety systems may survive an earthquake with a return frequency of 1.0x10-5 per year but catastrophic failure of the facility could be expected after an earthquake with a return frequency of 1.0x10-6 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence for all new facilities is assumed to be 1.0x10-6 per year. Consequences. The estimated consequences of the postulated tritium target extraction facility accident for each site are shown for the public in table F.2.1.6-1, and for the worker in table F.2.1.6-2. NTS can be seen to have the lowest number of cancer fatal- ities in the event of this accident. The dose and cancer fatality estimates are based on analysis to the release of 2.4x106 Ci of tritium in the oxide form using the MACCS computer code. Cancer Fatalities Complementary Cumulative Distribution Function for the Tritium Target Extraction Facility High Consequence Accident Figure F.2.1.6-1 shows the annual probability that, in the event of the tritium target extraction facility high consequence accident at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.1.6-1 and F.2.1.6-2. Figure (Page F-93) Figure F.2.1.6-1.-High Consequence Accident-Cancer Fatality Frequency Distribution Functions for Tritium Extraction. Table F.2.1.6-1.-Tritium Target Extraction Facility High Consequence Accident -Public Consequences - Individual at Site Boundary Population to 50 Miles - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 0.014 6.9x10-6 23 0.012 1.0x10-6 Laboratory Nevada Test Site 0.038 1.9x10-5 2.2 1.1x10-3 1.0x10-6 Oak Ridge Reservation 0.3 1.5x10-4 215 0.11 1.0x10-6 Pantex Plant 0.2 1.0x10-4 28 0.014 1.0x10-6 Savannah River Site 0.013 6.4x10-6 86 0.043 1.4x10-4 Expected Risk of Cancer Fatality (per year) Idaho National Engineering - 6.9x10-12 - 1.2x10-8 - Laboratory Nevada Test Site - 1.9x10-11 - 1.1x10-9 - Oak Ridge Reservation - 1.5x10-10 - 1.1x10-7 - Pantex Plant - 1.0x10-10 - 1.4x10-8 - Savannah River Sitec - 9.0x10-10 - 6.0x10-6 - Table F.2.1.6-2.-Tritium Target Extraction Facility High Consequence Accident -Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 1.7 6.8x10-4 0.63 2.5x10-4 1.0x10-6 Laboratory Nevada Test Site 1.2 5.0x10-4 0.48 1.9x10-4 1.0x10-6 Oak Ridge Reservation 1.7 6.7x10-4 0.6 2.4x10-4 1.0x10-6 Pantex Plant 0.73 2.9x10-4 0.28 1.1x10-4 1.0x10-6 Savannah River Site 0.74 3.0x10-4 0.28 1.1x10-4 1.4x10-4 Expected Risk of Cancer Fatality (per year) Idaho National Engineering - 6.8x10-10 - 2.5x10-10 - Laboratory Nevada Test Site - 5.0x10-10 - 1.9x10-10 - Oak Ridge Reservation - 6.7x10-10 - 2.4x10-10 - Pantex Plant - 2.9x10-10 - 1.1x10-10 - Savannah River Sitec - 4.2x10-8 - 1.5x10-8 - F.2.2 Tritium Supply and Recycling Facility Low-to-Moderate Consequence Accidents Low-to-moderate consequence accidents for candidate tritium supply technologies and recycling facilities at potential sites (INEL, NTS, ORR, Pantex, and SRS) have been evaluated using the GENII computer code. The consequences are based on inhalation and external dose pathways. Ingestion pathways are modeled but not included because it is assumed the food and water supply will be interdicted. The details of the evaluation are presented in sections F.2.2.1 through F.2.2.6. F.2.2.1 Heavy Water Reactor Scenario. The HWR low-to-moderate consequence accident occurs due to a charge-and-discharge mishap. During refueling operations, an irradiated fuel assembly containing tritium targets is assumed to melt in air in the hot cell refueling canyon, due to an assumed failure of the crane motive systems and the water-delivery systems that are used to cool the fuel assembly. Initially, the hot cell vents to the environment through filters. After 1 minute, the hot cell is isolated and leaks into the containment, which in turn leaks to the environment at the rate of 0.1 percent of its volume per day. Table F.2.2.1-1 presents the source term released during the accident. The analysis did not estimate the accident annual frequency (DOE 1995d:B-3). Consequences. The estimated consequences of the postulated accident for 50 percent meteorology conditions, to the public and worker at each site, are shown in tables F.2.2.1-2 and F.2.2.1-3, respectively. The dose estimates are based on analysis of the source term in table F.2.2.1-1 using the GENII computer code. Table F.2.2.1-1.-Source Term for Heavy Water Reactor Charge/Discharge Accident Isotope Released Activity Isotope Released Activity (curies) (curies) H-3 2.61x103 Xe-135 24.7 Br-83 0.0615 I-129 2.69x10-6 Br-83 4.21x10-8 I-131 70.5 Kr-83m 1.90 I-132 0.145 Kr-85 11.9 I-133 10.4 Kr-85m 54.0 I-134 2.47x10-4 Kr-87 0.45 I-135 3.11 Kr-88 41.3 Cs-134 26.5 Rb-86 0.124 Cs-136 0.94 Xe-131m 15.4 Cs-137 9.69 Xe-133 2.90x103 Cs-138 1.00x10-7 Xe-133m 558 Cs-139 1.62x10-26 Source: DOE 1995d. Table F.2.2.1-2.-Heavy Water Reactor Charge/Discharge Accident-Public Consequences - Individual at Site Boundary Population to 50 Miles - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 0.016 8.1x10-6 150 0.074 Nevada Test Site Laboratory 8.4x10-3 4.2x10-6 2.4 1.2x10-3 c Oak Ridge Reservation 0.14 6.8x10-5 1.5x103 0.75 c Pantex Plant 0.012 6.2x10-6 52 0.026 c Savannah River Site 0.046 Idaho National Engineering - 8.1x10-9 - 7.4x10-5 - Laboratory Nevada Test Site - 4.2x10-9 - 1.2x10-6 - Oak Ridge Reservation - 6.8x10-8 - 7.5x10-4 - Pantex Plant - 6.2x10-9 - 2.6x10-5 - Savannah River Site - 2.3x10-8 - 7.3x10-4 - Table F.2.2.1-3.-Heavy Water Reactor Charge/Discharge Accident-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 0.27 1.1x10-4 0.088 3.5x10-5 b Laboratory Nevada Test Site 0.07 2.8x10-5 0.025 9.8x10-6 c Oak Ridge Reservation 0.41 1.6x10-4 0.13 5.3x10-5 c Pantex Plant 0.031 1.2x10-5 8.7x10-3 3.5x10-6 c Savannah River Site 0.72 2.9x10-4 0.25 9.8x10-5 c Expected Risk of Cancer Fatality per year) Idaho National Engineering - 1.1x10-7 - 3.5x10-8 - Laboratory Nevada Test Site - 2.8x10-8 - 9.8x10-9 - Oak Ridge Reservation - 1.6x10-7 - 5.3x10-8 - Pantex Plant - 1.2x10-8 - 3.5x10-9 - Savannah River Site - 2.9x10-7 - 9.8x10-8 - F.2.2.2 Modular High Temperature Gas-Cooled Reactor The Draft of the PEIS, issued for review and comment in February 1995, considered a large break in the primary coolant system as the bounding MHTGR low-to-moderate consequence accident. The actual source term used in the analysis assumed the failure of redundant trains of safety class systems. The calculated accident consequences were significantly higher than the consequences for equivalent design-basis or evaluation-basis accidents where appropriate credit was taken for safety class systems to mitigate the accident consequences. The postulated bounding MHTGR low-to-moderate consequence accident was actually a beyond design-basis accident with high consequences and has been dropped from consideration as an MHTGR low-to-moderate consequence accident. A spectrum of low- to-moderate consequence accidents for MHTGRs was reviewed and two accidents were selected for evaluation in this document. Small Primary System Break Scenario. The accident postulated is a small break in the primary system piping that results in the release of the circulating radioactive material in the primary coolant into the containment. The containment leak rate to the environment is assumed at the rate of 1percent per day (DOE1992i:I-13). Table F.2.2.2-1 presents the source term released to the environment. The accident annual frequency is in the moderate class (DOE1992i). Moderate frequency events are events that would reasonably be expected to occur once during any year of reactor operations (DOE1992i:I-8). For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.2-2 through F.2.2.2-6 for public consequences and in tables F.2.2.2-7 through F.2.2.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.2-1 using the GENII computer code. Moderate Primary System Break Scenario. The accident postulated is a moderate break in the primary system piping that results in the release of the circulating radioactive material in the primary coolant into the containment. The shear force of the coolant as it escapes the primary system is assumed to lift off any radioactive material deposited on the inside surfaces of the primary system and cause it to be released to the containment along with the circulating radioactive material. The containment leak rate to the environment is assumed at the rate of 1 percent per day (DOE1992i:I-13). Table F.2.2.2-1 presents the source term released to the environment. The accident annual frequency is in the infrequent class (DOE1992i). Infrequent frequency events are events that would reasonably be expected to occur once during the plant's lifetime (DOE1992i:I-8). For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 2.5x10-2 per year (1 time in 40 years). Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.2-2 through F.2.2.2-6 for public consequences and in tables F.2.2.2-7 through F.2.2.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.2-1 using the GENII computer code. Table F.2.2.2-1.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accident Source Terms Isotope Released Activity (curies) Isotope Released Activity (curies) - Small Primary Moderate - Small Primary Moderate System Break Primary System System Break Primary System Break Break H-3 2.3 2.3 Sb-129 4.5x10-10 4.5x10-8 Kr-85m 0.017 1.3 Te-127m 1.4x10-6 8.0x10-5 Kr-85 0.014 0.014 Te-127 4.7x10-6 2.7x10-4 Kr-87 9.6x10-3 0.89 Te-129m 1.2x10-5 6.8x10-4 Kr-88 0.031 2.7 Te-129 1.5x10-4 0.014 Kr-89 2.2x10-4 0.022 Te-131m 1.5x10-4 9.5x10-3 Kr-90 1.6x10-5 1.6x10-3 Te-132 4.6x10-4 0.026 Rb-86 1.6x10-5 9.3x10-4 I-131 6.1x10-4 0.036 Sr-89 1.1x10-5 6.5x10-4 I-132 1.2x10-3 0.11 Sr-90 2.3x10-8 1.3x10-6 I-133 2.7x10-3 0.18 Sr-91 1.3x10-8 9.6x10-7 I-134 1.2x10-3 0.12 Y-90 2.3x10-10 1.4x10-8 I-135 2.4x10-3 0.19 Y-91 2.2x10-10 1.3x10-8 Xe-133 0.2 2.5 Zr-95 2.2x10-10 1.3x10-8 Xe-135 0.033 2.1 Zr-97 1.1x10-8 7.5x10-7 Cs-134 3.2x10-10 1.8x10-8 Nb-95 4.0x10-10 2.3x10-8 Cs-136 8.9x10-11 5.1x10-9 Mo-99 4.2x10-9 2.6x10-7 Cs-137 1.0x10-10 5.9x10-9 Tc-99m 1.5x10-8 1.2x10-6 Ba-140 8.0x10-10 4.6x10-8 Ru-103 2.1x10-10 1.2x10-8 La-140 7.3x10-9 4.6x10-7 Ru-105 5.3x10-9 4.4x10-7 Ce-141 4.2x10-10 2.4x10-8 Ru-106 2.0x10-12 1.2x10-10 Ce-143 6.7x10-9 4.3x10-7 Rh-105 1.8x10-9 1.1x10-7 Ce-144 1.9x10-11 1.1x10-9 Ag-110m 2.2x10-14 1.3x10-12 Pr-143 9.2x10-10 5.3x10-8 Sb-127 1.0x10-10 6.3x10-9 Nd-147 4.6x10-10 2.6x10-8 Source: DOE1992i. Table F.2.2.2-2.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Primary System Break Small 3.1x10-7 1.5x10-10 2.2x10-3 1.1x10-6 1 Moderate 1.0x10-5 5.1x10-9 0.04 2.0x10-5 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 1.5x10-10 - 1.1x10-6 - Moderate - 1.3x10-10 - 5.0x10-7 - Table F.2.2.2-3.-Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Primary System Break Small 1.3x10-7 6.6x10-11 4.3x10-5 2.1x10-8 1 Moderate 4.4x10-6 2.2x10-9 1.4x10-3 6.8x10-7 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 6.6x10-11 - 2.1x10-8 - Moderate - 5.5x10-11 - 1.7x10-8 - Table F.2.2.2-4.-Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Primary System Break Small 2.5x10-6 1.2x10-9 0.028 1.4x10-5 1 Moderate 8.7x10-5 4.4x10-8 0.86 4.3x10-4 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 1.2x10-9 - 1.4x10-5 - Moderate - 1.1x10-9 - 1.1x10-5 - Table F.2.2.2-5.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Primary System Break Small 2.3x10-7 1.2x10-10 8.8x10-4 4.4x10-7 1 Moderate 7.9x10-6 4.0x10-9 0.025 1.2x10-5 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 1.2x10-10 - 4.4x10-7 - Moderate - 1.0x10-10 - 3.0x10-7 - Table F.2.2.2-6.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Primary System Break Small 7.9x10-7 4.0x10-10 0.022 1.1x10-5 1 Moderate 2.4x10-5 1.2x10-8 0.5 2.5x10-4 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 4.0x10-10 - 1.1x10-5 - Moderate - 3.0x10-10 - 6.3x10-6 - Table F.2.2.2-7.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Primary System Break Small 1.1x10-5 4.5x10-9 3.7x10-6 1.5x10-9 1 Moderate 3.2x10-4 1.3x10-7 1.1x10-4 4.2x10-8 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 4.5x10-9 - 1.5x10-9 - Moderate - 3.3x10-9 - 1.1x10-9 - Table F.2.2.2-8.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Primary System Break Small 3.0x10-6 1.2x10-9 1.1x10-6 4.2x10-10 1 Moderate 8.2x10-5 3.3x10-8 3.0x10-5 1.2x10-8 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 1.2x10-9 - 4.2x10-10 - Moderate - 8.3x10-10 - 3.0x10-10 - Table F.2.2.2-9.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Primary System Break Small 1.7x10-5 6.9x10-9 5.8x10-6 2.3x10-9 1 Moderate 4.8x10-4 1.9x10-7 1.6x10-4 6.5x10-8 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 6.9x10-9 - 2.3x10-9 - Moderate - 4.8x10-9 - 1.6x10-9 - Table F.2.2.2-10.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Primary System Break Small 1.3x10-6 5.3x10-10 3.7x10-7 1.5x10-10 1 Moderate 3.8x10-5 1.5x10-8 1.0x10-5 4.2x10-9 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 5.3x10-10 - 1.5x10-10 - Moderate - 3.8x10-10 - 1.1x10-10 - Table F.2.2.2-11.- Modular High Temperature Gas-Cooled Reactor Low-to-Moderate Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Primary System Break Small 3.0x10-5 1.2x10-8 1.0x10-5 4.2x10-9 1 Moderate 8.5x10-4 3.4x10-7 2.9x10-4 1.2x10-7 2.5x10-2 Expected Risk of Cancer Fatality (per year) Primary System Break Small - 1.2x10-8 - 4.2x10-9 - Moderate - 8.5x10-9 - 3.0x10-9 - F.2.2.3 Advanced Light Water Reactor The Draft of this PEIS, issued for review and comment in February 1995, considered a large break in the primary coolant system as the bounding ALWR low-to-moderate consequence accident. The actual source term used in the analysis assumed the failure of redundant trains of safety class systems. The calculated accident consequences were significantly higher than the consequences for equivalent design-basis or evaluation basis accidents where appropriate credit was taken for safety class systems to mitigate the accident consequences. The postulated bounding ALWR low-to-moderate consequence accident was actually a beyond design-basis accident with high consequences and has been dropped from consideration as an ALWR low-to-moderate consequence accident. A spectrum of low-to-moderate consequence accidents for Large and Small ALWRs are evaluated in this document. The evaluation considered the AP600 Reactor, the Simplified Boiling Water Reactor, and the Advanced Boiling Water Reactor options. Data was not available for the CE System 80+ ALWR option. Sections F.2.2.3.1 through F.2.2.3.3 present the evaluations for the AP600 Reactor, the Simplified Boiling Water Reactor, and the Advanced Boiling Water Reactor options. F.2.2.3.1 AP600 Reactor Reactor Coolant Pump Shaft Seizure (Locked Rotor) Scenario. The accident postulated is an instantaneous seizure of a reactor coolant pump rotor. The reactor will trip on a low-flow signal and the turbine will trip. Following the reactor trip, heat stored in the fuel rods and the target assemblies continues to be transferred to the coolant, causing the coolant to expand and the reactor coolant system to pressurize. The pressurizer safety valves open to control the overpressure transient. There are two components to the radioactive releases to the environment; the activity initially in the secondary coolant at the time of the accident and the activity from the reactor coolant leaking into the steam generator is assumed to mix with the secondary coolant. Radioactive releases to the environment will continue as long as the secondary coolant steam releases continue. Table F.2.2.3.1-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence (TTI1995b). It is expected that the postulated annual frequency of occurrence would range from 1.0x10-4 to 1.0x10-6 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-5 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.1-2 through F.2.2.3.1-6 for public consequences and in tables F.2.2.3.1-7 through F.2.2.3.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.1-1 using the GENII computer code. Rod Cluster Control Assembly Ejection Scenario. The accident postulated the continuous withdrawal of a single rod control cluster assembly. This results in an increase in core power and coolant temperature. The reactor ultimately trips. Following reactor trip, normal reactor shutdown procedures are followed. Table F.2.2.3.1-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence (TTI 1995b). It is expected that the postulated annual frequency of occurrence would range from 0.01 to 1.0x10-4 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-3 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.1-2 through F.2.2.3.1-6 for public consequences and in tables F.2.2.3.1-7 through F.2.2.3.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.1-1 using the GENII computer code. Failure of Small Primary Coolant Line Outside of Containment Scenario. The accident postulated the failure of a sample line between the isolation valve outside of containment and the sample panel. The sample line includes a flow restrictor at the sample point to limit the break flow to less than 130 gpm. The analysis assumed that the flow from the break was isolated after 30 minutes. Table F.2.2.3.1-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence. It is expected that the postulated annual frequency of occurrence would range from 0.01 to 1.0x10-4 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-3 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.1-2 through F.2.2.3.1-6 for public consequences and in tables F.2.2.3.1-7 through F.2.2.3.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.1-1 using the GENII computer code. Steam Generator Tube Rupture Scenario. The accident postulated the complete severance of a single steam generator tube which leads to an increase in contamination of the secondary system due to leakage of radioactive coolant from the reactor coolant system. Continued loss of reactor coolant inventory leads to a reactor trip. The analysis assumed that the accident occurred coincident with the loss of offsite power and the high steam generator pressure causes a steam discharge to the atmosphere. The analysis also assumed that the initial iodine concentrations are those associated with the design fuel defect level. The iodine spike is assumed to be initiated by the accident. Table F.2.2.3.1-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence. It is expected that the postulated annual frequency of occurrence would range from 1.0x10-3 to 1.0x10-5 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-4 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.1-2 through F.2.2.3.1-6 for public consequences and in tables F.2.2.3.1-7 through F.2.2.3.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.1-1 using the GENII computer code. Fuel Handling Accident Scenario. The accident postulated that a spent fuel/target assembly dropped outside of containment in the auxiliary building fuel handling area. The analysis assumed that the assembly was dropped in such a way that every rod/target in the assembly has its cladding breached. The analysis also assumed that subsequent to the fuel handling accident, there was a loss of spent fuel cooling capability for up to 72 hours resulting in boiling water in the spent fuel pool. Table F.2.2.3.1-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence. It is expected that the postulated annual frequency of occurrence would range from 1.0x10-4 to 1.0x10-6 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-5 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.1-2 through F.2.2.3.1-6 for public consequences and in tables F.2.2.3.1-7 through F.2.2.3.1-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.1-1 using the GENII computer code. Table F.2.2.3.1-1.-AP600 Low-to-Moderate Consequence Accident Source Terms - Released Activity (curies) Isotope Reactor Coolant Rod Cluster Failure of Small Steam Generator Fuel Pump Shaft Control Primary Coolant Tube Rupture Handling Seizure Assembly Line Outside Ejection Containment H-3 3.8x104 9.2x104 1.4x103 1.2x104 0 I-131 63 60 22 15 230 I-132 24 45 120 75 3.7 I-133 110 100 47 33 9.6 I-134 47 41 54 30 0 I-135 150 77 45 31 0 Xe-131m 12 29 5.8 32 130 Xe-133m 470 330 53 290 2.3x103 Xe-133 3.3x103 4.6x103 820 4.7x10 2.7x104 Xe-135m 11 6.2 0.49 1.7 0 Xe-135 820 180 24 130 51 Xe-138 41 25 0.83 2.8 0 Kr-85m 240 39 5.9 31 0 Kr-85 58 150 21 120 610 Kr-87 140 24 2.9 14 0 Kr-88 450 67 9.7 51 0 Source: TTI 1995b. Table F.2.2.3.1-2.-AP600 Reactor Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Reactor coolant pump shaft seizure 9.3x10-3 4.7x10-6 67 0.034 b Rod cluster control assembly ejection 0.012 6.0x10-6 100 0.051 c Failure of small primary coolant line outside 2.6x10-3 1.3x10-6 16 8.1x10-3 d containment Steam Generator Tube Rupture 2.6x10-3 1.3x10-6 19 9.3x10-3 d Fuel handling 0.014 6.8x10-6 120 0.062 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 4.7x10-11 - 3.4x10-7 - Rod cluster control assembly ejection - 6.0x10-9 - 5.1x10-5 - Failure of small primary coolant line outside - 1.3x10-9 - 8.1x10-6 - containment Steam generator tube rupture - 1.3x10-10 - 9.3x10-7 - Fuel handling - 6.8x10-11 - 6.2x10-7 - Table F.2.2.3.1-3.-AP600 Reactor Low-to-Moderate Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Reactor coolant pump shaft seizure 4.1x10-3 2.1x10-6 1.3 6.7x10-4 b Rod cluster control assembly ejection 5.4x10-3 2.7x10-6 1.8 9.0x10-4 c Failure of small primary coolant line outside 1.2x10-3 5.9x10-7 0.35 1.8x10-4 d containment Steam generator tube rupture 1.2x10-3 5.9x10-7 0.38 1.9x10-4 d Fuel handling 6.0x10-3 3.0x10-6 2.1 1.0x10-3 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 2.1x10-11 - 6.7x10-9 - Rod cluster control assembly ejection - 2.7x10-9 - 9.0x10-7 - Failure of small primary coolant line outside - 5.9x10-10 - 1.8x10-7 - containment Steam generator tube rupture - 5.9x10-11 - 1.9x10-8 - Fuel handling - 3.0x10-11 - 1.0x10-8 - Table F.2.2.3.1-4.-AP600 Reactor Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Reactor coolant pump shaft seizure 0.079 4.0x10-5 840 0.42 b Rod cluster control assembly ejection 0.1 5.1x10-5 1.1x103 0.55 c Failure of small primary coolant line outside 0.023 1.2x10-5 230 0.11 d containment Steam generator tube rupture 0.023 1.2x10-5 240 0.12 d Fuel handling 0.12 5.8x10-5 1.3x103 0.64 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 4.0x10-10 - 4.2x10-6 - Rod cluster control assembly ejection - 5.1x10-8 - 5.5x10-4 - Failure of small primary coolant line outside - 1.2x10-8 - 1.1x10-4 - containment Steam generator tube rupture - 1.2x10-9 - 1.2x10-5 - Fuel handling - 5.8x10-10 - 6.4x10-6 - Table F.2.2.3.1-5.-AP600 Reactor Low-to-Moderate Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Reactor coolant pump shaft seizure 7.2x10-3 3.6x10-6 27 0.013 b Rod cluster control assembly ejection 9.3x10-3 4.6x10-6 37 0.018 c Failure of small primary coolant line outside 2.1x10-3 1.1x10-6 7.1 3.6x10-3 d containment Steam generator tube rupture 2.1x10-3 1.1x10-6 7.8 3.9x10-3 d Fuel handling 0.01 5.2x10-6 42 0.021 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 3.6x10-11 - 1.3x10-7 - Rod cluster control assembly ejection - 4.6x10-9 - 1.8x10-5 - Failure of small primary coolant line outside - 1.1x10-9 - 3.6x10-6 - containment Steam generator tube rupture - 1.1x10-10 - 3.9x10-7 - Fuel handling - 5.2x10-11 - 2.1x10-7 - Table F.2.2.3.1-6.-AP600 Reactor Low-to-Moderate Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Reactor coolant pump shaft seizure 0.025 1.3x10-5 690 0.34 b Rod cluster control assembly ejection 0.032 1.6x10-5 980 0.49 c Failure of small primary coolant line outside 6.6x10-3 3.3x10-6 170 0.084 d containment Steam generator tube rupture 7.0x10-3 3.5x10-6 190 0.097 d Fuel handling 0.039 2.0x10-5 1.2x103 0.6 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 1.3x10-10 - 3.4x10-6 - Rod cluster control assembly ejection - 1.6x10-8 - 4.9x10-4 - Failure of small primary coolant line outside - 3.3x10-9 - 8.4x10-5 - containment Steam generator tube rupture - 3.5x10-10 - 9.7x10-6 - Fuel handling - 2.0x10-10 - 6.0x10-6 - Table F.2.2.3.1-7.-AP600 Reactor Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Reactor coolant pump shaft seizure 0.29 1.2x10-4 0.096 3.8x10-5 b Rod cluster control assembly ejection 0.41 1.6x10-4 0.14 5.4x10-5 c Failure of small primary coolant line outside 0.067 2.7x10-5 0.022 8.9x10-6 d containment Steam generator tube rupture 0.084 3.4x10-5 0.029 1.1x10-5 d Fuel handling 0.33 1.3x10-4 0.11 4.5x10-5 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 1.2x10-9 - 3.8x10-10 - Rod cluster control assembly ejection - 1.6x10-7 - 5.4x10-8 - Failure of small primary coolant line outside - 2.7x10-8 - 8.9x10-9 - containment Steam generator tube rupture - 3.4x10-9 - 1.1x10-9 - Fuel handling - 1.3x10-9 - 4.5x10-10 - Table F.2.2.3.1-8.-AP600 Reactor Low-to-Moderate Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Reactor coolant pump shaft seizure 0.085 3.4x10-5 0.029 1.1x10-5 b Rod cluster control assembly ejection 0.12 5.0x10-5 0.041 1.6x10-5 c Failure of small primary coolant line outside 0.02 8.0x10-6 6.6x10-3 2.6x10-6 d containment Steam generator tube rupture 0.025 9.9x10-6 8.3x10-3 3.3x10-6 d Fuel handling 0.097 3.9x10-5 0.033 1.3x10-5 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 3.4x10-10 - 1.1x10-10 - Rod cluster control assembly ejection - 5.0x10-8 - 1.6x10-8 - Failure of small primary coolant line outside - 8.0x10-9 - 2.6x10-9 - containment Steam generator tube rupture - 9.9x10-10 - 3.3x10-10 - Fuel handling - 3.9x10-10 - 1.3x10-10 - Table F.2.2.3.1-9.-AP600 Reactor Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Reactor coolant pump shaft seizure 0.44 1.8x10-4 0.15 6.0x10-5 b Rod cluster control assembly ejection 0.63 2.5x10-4 0.21 8.6x10-5 c Failure of small primary coolant line outside 0.1 4.0x10-5 0.034 1.4x10-5 d containment Steam generator tube rupture 0.13 5.2x10-5 0.043 1.7x10-5 d Fuel handling 0.52 2.1x10-4 0.17 6.7x10-5 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 1.8x10-9 - 6.0x10-10 - Rod cluster control assembly ejection - 2.5x10-7 - 8.6x10-8 - Failure of small primary coolant line outside - 4.0x10-8 - 1.4x10-8 - containment Steam generator tube rupture - 5.2x10-9 - 1.7x10-9 - Fuel handling - 2.1x10-9 - 6.7x10-10 - Table F.2.2.3.1-10.-AP600 Reactor Low-to-Moderate Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Reactor coolant pump shaft seizure 0.033 1.3x10-5 9.4x10-3 3.8x10-6 b Rod cluster control assembly ejection 0.049 1.9x10-5 0.014 5.4x10-6 c Failure of small primary coolant line outside 7.9x10-3 3.2x10-6 2.2x10-3 8.8x10-7 d containment Steam generator tube rupture 9.9x10-3 4.0x10-6 2.8x10-3 1.1x10-6 d Fuel handling 0.039 1.6x10-5 0.011 4.4x10-6 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 1.3x10-10 - 3.8x10-11 - Rod cluster control assembly ejection - 1.9x10-8 - 5.4x10-9 - Failure of small primary coolant line outside - 3.2x10-9 - 8.8x10-10 - containment Steam generator tube rupture - 4.0x10-10 - 1.1x10-10 - Fuel handling - 1.6x10-10 - 4.4x10-11 - Table F.2.2.3.1-11.-AP600 Reactor Low-to-Moderate Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Reactor coolant pump shaft seizure 0.76 3.0x10-4 0.27 1.1x10-4 b Rod cluster control assembly ejection 1.1 4.4x10-4 0.38 1.5x10-4 c Failure of small primary coolant line outside 0.18 7.3x10-5 0.059 2.4x10-5 d containment Steam generator tube rupture 0.23 9.2x10-5 0.076 3.0x10-5 d Fuel handling 0.9 3.6x10-4 0.31 1.2x10-4 c Expected Risk of Cancer Fatality (per year) Reactor coolant pump shaft seizure - 3.0x10-9 - 1.1x10-9 - Rod cluster control assembly ejection - 4.4x10-7 - 1.5x10-7 - Failure of small primary coolant line outside - 7.3x10-8 - 2.4x10-8 - containment Steam generator tube rupture - 9.2x10-9 - 3.0x10-9 - Fuel handling - 3.6x10-9 - 1.2x10-9 - F.2.2.3.2 Simplified Boiling Water Reactor Failure of Small Primary Coolant Line Outside of Containment Scenario. The accident postulated the rupture of an instrument line outside the drywell but inside the reactor building. The leak is not isolatable. The flow from the instrument line is limited by a one-quarter inch diameter flow restricting orifice inside the drywell. The total integrated mass of fluid released into the reactor building is 13,000 kg with approximately 5,000 kg being flashed into steam. The accident sequence is terminated by the orderly shutdown and depressurization of the reactor. Table F.2.2.3.2-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence. It is expected that the postulated annual frequency of occurrence would range from 0.01 to 1.0x10-4 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-3 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.2-2 through F.2.2.3.2-6 for public consequences and in tables F.2.2.3.2-7 through F.2.2.3.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.2-1 using the GENII computer code. Steam System Piping Break Outside Containment Scenario. The accident postulated a large steam line break outside of containment downstream of the outermost isolation valve. The plant is designed to immediately detect the break and initiate isolation of the broken line. Table F.2.2.3.2-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence (TTI 1995b). It is expected that the postulated annual frequency of occurrence would range from 1.0x10-4 to 1.0x10-6 per year. For the purpose of cal- culating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-5 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.2-2 through F.2.2.3.2-6 for public consequences and in tables F.2.2.3.2-7 through F.2.2.3.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.2-1 using the GENII computer code. Feedwater Line Break Outside of Containment Scenario. The accident postulated a feedwater line break outside of containment. Feedwater line check valves isolate the leak from the reactor. The total mass of fluid released is 320,000 kg with approximately 10,000 kg being flashed into steam. The reactor core remains covered during the accident and no core heatup occurs. Table F.2.2.3.2-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence. It is expected that the postulated annual frequency of occurrence range from the 1.0x10-4 to 1.0x10-6 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-5 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.2-2 through F.2.2.3.2-6 for public consequences and in tables F.2.2.3.2-7 through F.2.2.3.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.2-1 using the GENII computer code. Fuel Handling Accident Scenario. The accident postulated that a spent fuel/target assembly dropped into the reactor core. The analysis assumed that some rods/targets in the dropped assembly and in the struck assembly fail. Table F.2.2.3.2-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence (TTI1995b). It is expected that the postulated annual frequency of occurrence would range from 1.0x10-4 to 1.0x10-6 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-5 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.2-2 through F.2.2.3.2-6 for public consequences and in tables F.2.2.3.2-7 through F.2.2.3.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.2-1 using the GENII computer code. Table F.2.2.3.2-1.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accident Source Terms Isotope Released Activity (curies) - Failure of Small Steam System Piping Feedwater Line Fuel Handling Primary Coolant Break Outside Break Outside Line Outside Containment Containment Containment H-3 1.0x103 1.1x103 2.0x104 0.044 I-131 30 20 2.7x10-3 150 I-132 46 0.84 0.020 1.6x10-6 I-133 71 22 0.019 25 I-134 77 0.027 0.032 2.3x10-28 I-135 68 5.1 0.024 0.046 Xe-131m 0 1.5x10-4 0 110 Xe-133m 0 0.084 0 730 Xe-133 0 3.0x10-3 0 3.0x104 Xe-135m 0 0.22 0 0 Xe-135 0 0.26 0 86 Xe-137 0 1.4 0 1.4 Xe-138 0 0.89 0 0 Xe-139 0 2.3 0 0 Kr-83m 0 0.035 0 8.6x10-10 Kr-85m 0 0.062 0 7.0x10-3 Kr-85 0 2.0x10-4 0 300 Kr-87 0 0.2 0 3.5x10-17 Kr-88 0 0.2 0 7.6x10-6 Kr-89 0 1.2 0 0 Kr-90 0 2.2 0 0 Source: TTI 1995b. Table F.2.2.3.2-2.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside 3.3x10-3 1.6x10-6 21 0.011 b containment Steam system piping break outside 1.4x10-3 7.1x10-7 12 5.9x10-3 c containment Feedwater line break outside containment 1.4x10-3 7.0x10-7 14 7.0x10-3 d Fuel handling 9.8x10-3 4.9x10-6 89 0.044 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 1.6x10-9 - 1.1x10-5 - containment Steam system piping break outside - 7.1x10-12 - 5.9x10-8 - containment Feedwater line break outside containment - 7.0x10-12 - 7.0x10-8 - Fuel handling - 4.9x10-11 - 4.4x10-7 - Table F.2.2.3.2-3.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside 1.4x10-3 7.2x10-7 0.45 2.2x10-4 b containment Steam system piping break outside 6.3x10-4 3.1x10-7 0.21 1.0x10-4 c containment Feedwater line break outside containment 6.2x10-4 3.1x10-7 0.22 1.1x10-4 d Fuel handling 4.0x10-3 2.0x10-6 1.5 7.5x10-4 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 7.2x10-10 - 2.2x10-7 - containment Steam system piping break outside - 3.1x10-12 - 1.0x10-9 - containment Feedwater line break outside containment - 3.1x10-12 - 1.1x10-9 - Fuel handling - 2.0x10-11 - 7.5x10-9 - Table F.2.2.3.2-4.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside 0.028 1.4x10-5 290 0.15 b containment Steam system piping break outside 0.012 6.1x10-6 130 0.066 c containment Feedwater line break outside containment 0.012 6.0x10-6 140 0.07 d Fuel handling 0.083 4.2x10-5 940 0.47 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 1.4x10-8 - 1.5x10-4 - containment Steam system piping break outside - 6.1x10-11 - 6.6x10-7 - containment Feedwater line break outside containment - 6.0x10-11 - 7.0x10-7 - Fuel handling - 4.2x10-10 - 4.7x10-6 - Table F.2.2.3.2-5.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside 2.5x10-3 1.2x10-6 9.2 4.6x10-3 b containment Steam system piping break outside 1.1x10-3 5.5x10-7 4.3 2.2x10-3 c containment Feedwater line break outside containment 1.1x10-3 5.5x10-7 4.5 2.3x10-3 d Fuel handling 7.6x10-3 3.8x10-6 31 0.016 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 1.2x10-9 - 4.6x10-6 - containment Steam system piping break outside - 5.5x10-12 - 2.2x10-8 - containment Feedwater line break outside containment - 5.5x10-12 - 2.3x10-8 - Fuel handling - 3.8x10-11 - 1.6x10-7 - Table F.2.2.3.2-6.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside 8.4x10-3 4.2x10-6 230 0.11 b containment Steam system piping break outside 7.9x10-3 4.0x10-6 250 0.12 c containment Feedwater line break outside containment 4.0x10-3 2.0x10-6 130 0.065 d Fuel handling 0.028 1.4x10-5 850 0.43 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 4.2x10-9 - 1.1x10-4 - containment Steam system piping break outside - 4.0x10-11 - 1.2x10-6 - containment Feedwater line break outside containment - 2.0x10-11 - 6.5x10-7 - Fuel handling - 1.4x10-10 - 4.3x10-6 - Table F.2.2.3.2-7.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside 0.08 3.2x10-5 0.027 1.1x10-5 b containment Steam system piping break outside 0.034 1.4x10-5 0.011 4.6x10-6 c containment Feedwater line break outside containment 0.061 2.4x10-5 0.021 8.4x10-6 d Fuel handling 0.26 1.0x10-4 0.085 3.4x10-5 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 3.2x10-8 - 1.1x10-8 - containment Steam system piping break outside - 1.4x10-10 - 4.6x10-11 - containment Feedwater line break outside containment - 2.4x10-10 - 8.4x10-11 - Fuel handling - 1.0x10-9 - 3.4x10-10 - Table F.2.2.3.2-8.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside 0.023 9.4x10-6 8.0x10-3 3.2x10-6 b containment Steam system piping break outside 0.01 4.0x10-6 3.4x10-3 1.4x10-6 c containment Feedwater line break outside containment 0.018 7.2x10-6 6.0x10-3 2.4x10-6 d Fuel handling 0.075 3.0x10-5 0.025 9.9x10-6 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 9.4x10-9 - 3.2x10-9 - containment Steam system piping break outside - 4.0x10-11 - 1.4x10-11 - containment Feedwater line break outside containment - 7.2x10-11 - 2.4x10-11 - Fuel handling - 3.0x10-10 - 9.9x10-11 - Table F.2.2.3.2-9.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside 0.12 5.0x10-5 0.04 1.6x10-5 b containment Steam system piping break outside 0.051 2.0x10-5 0.018 7.0x10-6 c containment Feedwater line break outside containment 0.09 3.6x10-5 0.032 1.3x10-5 d Fuel handling 0.39 1.6x10-4 0.13 5.2x10-5 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 5.0x10-8 - 1.6x10-8 - containment Steam system piping break outside - 2.0x10-10 - 7.0x10-11 - containment Feedwater line break outside containment - 3.6x10-10 - 1.3x10-10 - Fuel handling - 1.6x10-9 - 5.2x10-10 - Table F.2.2.3.2-10.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside 9.4x10-3 3.8x10-6 2.7x10-3 1.1x10-6 b containment Steam system piping break outside 4.1x10-3 1.6x10-6 1.1x10-3 4.4x10-7 c containment Feedwater line break outside containment 7.2x10-3 2.9x10-6 2.0x10-3 8.0x10-7 d Fuel handling 0.03 1.2x10-5 8.5x10-3 3.4x10-6 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 3.8x10-9 - 1.1x10-9 - containment Steam system piping break outside - 1.6x10-11 - 4.4x10-12 - containment Feedwater line break outside containment - 2.9x10-11 - 8.0x10-12 - Fuel handling - 1.2x10-10 - 3.4x10-11 - Table F.2.2.3.2-11.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside 0.22 8.6x10-5 0.072 2.9x10-5 b containment Steam system piping break outside 0.25 1.0x10-4 0.088 3.5x10-5 c containment Feedwater line break outside containment 0.16 6.4x10-5 0.057 2.3x10-5 d Fuel handling 0.7 2.8x10-4 0.23 9.3x10-5 d Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside - 8.6x10-8 - 2.9x10-8 - containment Steam system piping break outside - 1.0x10-9 - 3.5x10-10 - containment Feedwater line break outside containment - 6.4x10-10 - 2.3x10-10 - Fuel handling - 2.8x10-9 - 9.3x10-10 - F.2.2.3.3 Advanced Boiling Water Reactor Failure of Small Primary Coolant Line Outside of Containment Scenario. The accident postulated the rupture of an instrument line outside the drywell but inside the reactor building. The leak is not isolatable. The flow from the instrument line is limited by a 0.64 cm diameter flow restricting orifice inside the drywell. The total integrated mass of fluid released into the reactor building is 5,442 kg with approximately 2,270 kg being flashed into steam. The accident sequence is terminated by the orderly shutdown and depressurization of the reactor. Table F.2.2.3.3-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence (TTI 1995b). It is expected that the postulated annual frequency of occurrence would be in the 0.01 to 1.0x10-4 per year range. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-3 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.3-2 through F.2.2.3.3-6 for public consequences and in tables F.2.2.3.3-7 through F.2.2.3.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.3-1 using the GENII computer code. Steam System Piping Break Outside Containment Scenario. The accident postulated a large steam line break outside of containment downstream of the outermost isolation valve. The plant is designed to immediately detect the break and initiate isolation of the broken line. Table F.2.2.3.3-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence (TTI 1995b). It is expected that the postulated annual frequency of occurrence would be in the 1.0x10-4 to 1.0x10-6 per year range. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-5 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.3-2 through F.2.2.3.3-6 for public consequences and in tables F.2.2.3.3-7 through F.2.2.3.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.3-1 using the GENII computer code. Cleanup Water Line Break Outside Containment Scenario. The accident postulated a large cleanup water line break outside of containment. The analysis assumed that the non-filtered inventory in both the regenerative and non-regenerative heat exchangers is released through the break. The leak is automatically isolated approximately 75 seconds after the break. Table F.2.2.3.3-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence (TTI 1995b). It is expected that the postulated annual frequency of occurrence would be in the 1.0x10-4 to 1.0x10-6 per year range. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-5 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.3-2 through F.2.2.3.3-6 for public consequences and in tables F.2.2.3.3-7 through F.2.2.3.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.3-1 using the GENII computer code. Fuel Handling Accident Scenario. The accident postulated that a spent fuel/target assembly dropped into the reactor core. The analysis assumed that some rods/targets in the dropped assembly and in the struck assembly fail. Table F.2.2.3.3-1 presents the source term released to the environment. The analysis did not estimate the accident annual frequency of occurrence (TTI1995b). It is expected that the postulated annual frequency of occurrence would be in the 1.0x10-4 to 1.0x10-6 per year range. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence is assumed to be 1.0x10-5 per year. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.3.3-2 through F.2.2.3.3-6 for public consequences and in tables F.2.2.3.3-7 through F.2.2.3.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.3.3-1 using the GENII computer code. Table F.2.2.3.3-1.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accident Source Terms Isotope Released Activity (curies) - Failure of Small Steam System Piping Cleanup Water Line Fuel Handling Primary Coolant Break Outside Break Outside Line Outside Containment Containment Containment H-3 890 1.5x103 1.3x103 0.037 I-131 3.8 39 2.2 120 I-132 32 380 5.1 150 I-133 26 270 6.2 130 I-134 51 750 8.6 6.2x10-6 I-135 36 390 6.8 21 Xe-131m 0 2.9x10-4 0 84 Xe-133m 0 5.5x10-3 0 1.1x103 Xe-133 0 0.15 0 2.8x104 Xe-135m 0 0.47 0 220 Xe-135 0 0.44 0 6.4x103 Xe-137 0 2 0 2.1x10-10 Xe-138 0 1.5 0 4.3x10-10 Xe-139 0 0.7 0 0 Kr-83m 0 0.066 0 6.4 Kr-85m 0 0.12 0 85 Kr-85 0 3.7x10-4 0 480 Kr-87 0 0.4 0 0.012 Kr-88 0 0.4 0 24 Kr-89 0 1.6 0 8.1x10-11 Kr-90 0 0.42 0 0 Source: TTI 1995b. Table F.2.2.3.3-2.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside containment 9.6x10-4 4.8x10-7 4.7 2.4x10-3 b Steam system piping break outside containment 9.9x10-3 5.0x10-6 42 0.021 c Cleanup water line break outside containment 3.9x10-4 1.9x10-7 2.8 1.4x10-3 d Fuel handling 9.9x10-3 5.0x10-6 76 0.038 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 4.8x10-10 - 2.4x10-6 - Steam system piping break outside containment - 5.0x10-11 - 2.1x10-7 - Cleanup water line break outside containment - 1.9x10-12 - 1.4x10-8 - Fuel handling - 5.0x10-11 - 3.8x10-7 - Table F.2.2.3.3-3.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside containment 4.3x10-4 2.2x10-7 0.13 6.4x10-5 b Steam system piping break outside containment 6.5x10-5 3.3x10-8 1.3 6.4x10-4 c Cleanup water line break outside containment 1.7x10-4 8.7x10-8 0.056 2.8x10-5 d Fuel handling 4.4x10-3 2.2x10-6 1.5 7.3x10-4 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 2.2x10-10 - 6.4x10-8 - Steam system piping break outside containment - 3.3x10-13 - 6.4x10-9 - Cleanup water line break outside containment - 8.7x10-13 - 2.8x10-10 - Fuel handling - 2.2x10-11 - 7.3x10-9 - Table F.2.2.3.3-4.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside containment 8.6x10-3 4.3x10-6 81 0.041 b Steam system piping break outside containment 0.091 4.6x10-5 820 0.41 c Cleanup water line break outside containment 3.4x10-3 1.7x10-6 35 0.018 d Fuel handling 0.086 4.3x10-5 910 0.46 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 4.3x10-9 - 4.1x10-5 - Steam system piping break outside containment - 4.6x10-10 - 4.1x10-6 - Cleanup water line break outside containment - 1.7x10-11 - 1.8x10-7 - Fuel handling - 4.3x10-10 - 4.6x10-6 - Table F.2.2.3.3-5.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside containment 7.8x10-4 3.9x10-7 2.5 1.2x10-3 b Steam system piping break outside containment 8.2x10-3 4.1x10-6 24 0.012 c Cleanup water line break outside containment 3.1x10-4 1.5x10-7 1.1 5.7x10-4 d Fuel handling 7.7x10-3 3.9x10-6 29 0.015 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 3.9x10-10 - 1.2x10-6 - Steam system piping break outside containment - 4.1x10-11 - 1.2x10-7 - Cleanup water line break outside containment - 1.5x10-12 - 5.7x10-9 - Fuel handling - 3.9x10-11 - 1.5x10-7 - Table F.2.2.3.3-6.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Failure of small primary coolant line outside containment 2.3x10-3 1.2x10-6 53 0.027 b Steam system piping break outside containment 0.023 1.2x10-5 500 0.25 c Cleanup water line break outside containment 1.0x10-3 5.2x10-7 28 0.014 d Fuel handling 0.027 1.3x10-5 760 0.038 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 1.2x10-9 - 2.7x10-5 - Steam system piping break outside containment - 1.2x10-10 - 2.5x10-6 - Cleanup water line break outside containment - 5.2x10-12 - 1.4x10-7 - Fuel handling - 1.3x10-10 - 3.8x10-6 - Table F.2.2.3.3-7.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside containment 0.026 1.0x10-5 8.5x10-3 3.4x10-6 b Steam system piping break outside containment 0.26 1.0x10-4 0.087 3.5x10-5 c Cleanup water line break outside containment 0.012 4.8x10-6 4.0x10-3 1.6x10-6 d Fuel handling 0.26 1.0x10-4 0.089 3.6x10-5 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 1.0x10-8 - 3.4x10-9 - Steam system piping break outside containment - 1.0x10-9 - 3.5x10-10 - Cleanup water line break outside containment - 4.8x10-11 - 1.6x10-11 - Fuel handling - 1.0x10-9 - 3.6x10-10 - Table F.2.2.3.3-8.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside containment 7.5x10-3 3.0x10-6 2.5x10-3 1.0x10-6 b Steam system piping break outside containment 0.077 3.1x10-5 0.026 1.0x10-5 c Cleanup water line break outside containment 3.6x10-3 1.4x10-6 1.2x10-3 4.7x10-7 d Fuel handling 0.078 3.1x10-5 0.026 1.0x10-5 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 3.0x10-9 - 1.0x10-9 - Steam system piping break outside containment - 3.0x10-10 - 1.0x10-10 - Cleanup water line break outside containment - 1.4x10-11 - 4.7x10-12 - Fuel handling - 3.1x10-10 - 1.0x10-10 - Table F.2.2.3.3-9.-Advanced Boiling Water Reactor Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside containment 0.039 1.6x10-5 0.013 5.1x10-6 b Steam system piping break outside containment 0.4 1.6x10-4 0.13 5.2x10-5 c Cleanup water line break outside containment 0.018 7.2x10-6 6.1x10-3 2.5x10-6 d Fuel handling 0.4 1.6x10-4 0.14 5.4x10-5 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 1.6x10-8 - 5.1x10-9 - Steam system piping break outside containment - 1.6x10-9 - 5.2x10-10 - Cleanup water line break outside containment - 7.2x10-11 - 2.5x10-11 - Fuel handling - 1.6x10-9 - 5.4x10-10 - Table F.2.2.3.3-10.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside containment 3.0x10-3 1.2x10-6 8.3x10-4 3.3x10-7 b Steam system piping break outside containment 0.03 1.2x10-5 8.4x10-3 3.4x10-6 c Cleanup water line break outside containment 1.4x10-3 5.6x10-7 4.0x10-4 1.6x10-7 d Fuel handling 0.031 1.2x10-5 8.4x10-3 3.4x10-6 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 1.2x10-9 - 3.3x10-10 - Steam system piping break outside containment - 1.2x10-10 - 3.4x10-11 - Cleanup water line break outside containment - 5.6x10-12 - 1.6x10-12 - Fuel handling - 1.2x10-10 - 3.4x10-11 - Table F.2.2.3.3-11.-Simplified Boiling Water Reactor Low-to-Moderate Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Failure of small primary coolant line outside containment 0.066 2.6x10-5 0.022 8.9x10-6 b Steam system piping break outside containment 0.67 2.7x10-4 0.22 8.8x10-5 c Cleanup water line break outside containment 0.032 1.3x10-5 0.011 4.4x10-6 d Fuel handling 0.71 2.8x10-4 0.24 9.6x10-5 c Expected Risk of Cancer Fatality (per year) Failure of small primary coolant line outside containment - 2.6x10-8 - 8.9x10-9 - Steam system piping break outside containment - 2.7x10-9 - 8.8x10-10 - Cleanup water line break outside containment - 1.3x10-10 - 4.4x10-11 - Fuel handling - 2.8x10-9 - 9.6x10-10 - F.2.2.4 Accelerator Production of Tritium F.2.2.4.1 Accelerator and Beam Transport System One design-basis accident for the accelerator was considered. Incorrect administrative procedures and control for maintenance access to activated accelerator components could result in higher than permitted dose levels to service personnel. The consequences of the accident are limited to the dose received by service personnel. No lost production time or equipment replacement expense would be incurred. Based on operating APT experience, the annual frequency of occurrence is estimated at 1 time per year (SNL 1995a:8-5). F.2.2.4.2 Helium-3 Target System Scenario. The low-to-moderate consequence accident for this APT technology is a double-ended guillotine cold leg break near the pump discharge. The plant protection and safety systems performed as designed. The analysis assumed the most limiting single failure was the loss of power to one of the residual heat removal pumps. During this accident, the rod temperatures flatten out at approximately 340 Kelvin (152 F) and would be expected to decrease in time as the power decays. The source term to the confinement for this design basis accident is judged to be similar to and bounded by the source term for the beyond design-basis accident (large break low-to-moderate consequence) presented in section F.2.1.4.2. The analysis did not estimate the accident annual frequency of occurrence (SNL 1995a:8-8). Consequences. The estimated consequences of the postulated APT with helium-3 target system low-to-moderate consequence accident are bounded by the beyond design-basis accident presented in section F.2.1.4.2. F.2.2.4.3 Spallation-Induced Lithium Conversion Target System The low-to-moderate consequence accident for the APT technology is a large break in the primary coolant system. The analysis assumed that all plant protection systems functioned as designed. The worst single failure responding to the initiating event was assumed. The source term for this accident will consist of a small fraction of the radioactivity inventory released from the heavy-water coolant that is expelled into the confinement. The radionuclides released to the confinement are minimal. The analysis did not estimate the accident annual frequency of occurrence (SNL 1995a:8-6). F.2.2.5 Multipurpose Reactor Facility The multipurpose fuel reactor facility consists of three elements. 1) The reactor element that burns the plutonium mixed oxide fuel can be either a modular high temperature gas cooled reactor or an advanced light water reactor. 2) The fuel fabrication element produces the fuel for use in the reactor. 3) The pit disassembly and conversion element disassembles plutonium pits and converts the plutonium in the pit to plutonium-oxide which is used in the production of the plutonium or mixed-oxide fuel. F.2.2.5.1 Multipurpose Reactor Modular High Temperature Gas-Cooled Reactor The use of plutonium oxide as the fuel in the MHTGR will not have a significant effect on the source term for low-to-moderate consequence accidents generated for the uranium fueled MHTGR because no fuel failures are expected (HNUS1995c:1). The accident consequences estimated for the uranium fueled MHTGR are applicable for the plutonium-oxide fueled MHTGR. Refer to section F.2.2.2 for the applicable accident consequences of the plutonium-oxide fueled MHTGR. Advanced Light Water Reactor The use of plutonium-oxide in the fuel in Large and Small ALWRs will have a significant effect on the source term for low-to-moderate consequence accidents generated for the uranium fueled ALWRs because of increased gap inventories in the mixed-oxide fuels. Tables F.2.2.3.1-1, F.2.2.3.2-1, and F.2.2.3.3-1 present the low-to-moderate consequence accident source terms for the AP600, Simplified Boiling Water Reactor and Advanced Boiling Water Reactor ALWRs. When the accident source terms are adjusted for the increased gap inventory of gasses (ORNL 1995b:B-13) and the typical core inventory isotope ratios for the mixed-oxide core (ORNL 1995c) are considered (HNUS1995c:2), it is estimated that the consequences for uranium fueled ALWR low-to-moderate accident consequences should be increased by an approximate factor of 1.5 to 2 to obtain the consequences for equivalent mixed-oxide fueled ALWR accidents. Refer to section F.2.2.3 and apply an approximate correction factor of 1.5 to 2 to assess the increased consequences of mixed-oxide fueled ALWR low-to-moderate consequence accidents. F.2.2.5.2 Multipurpose Reactor Fuel Fabrication Loading Dock Fire Scenario. The accident postulated is a fire on an open loading dock caused by welding, cleaning solvents, electrical shorts, or other miscellaneous causes. A single drum of combustible waste, containing 18grams of plutonium, is involved in the fire. The analysis estimated that 0.077 gram of plutonium was released directly to the environment by the fire. The annual frequency of occurrence is estimated to be in the range of 1.0x10-3 to 1.0x10-4 per year (LANL1995d). For calculational purposes, the annual frequency of occurrence is assumed to be 5.0x10-4 per year, the mid point of the estimated range. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.2-1 presents the source term, by isotope, for the 0.077 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.2-2 through F.2.2.5.2-6 for public consequences and in tables F.2.2.5.2-7 through F.2.2.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.2-1 using the GENII computer code. Process Cell Fire Scenario. The accident postulated is a fire in a process cell area. It is assumed that the process cell contains a glovebox used for milling plutonium powder. The gloves have become coated with a layer of plutonium dust. The analysis estimated the glove loading at 2 grams of plutonium per glove. Each of the 12 gloves is assumed to be stowed outside of the glovebox. A flammable cleaning liquid such as acetone or isopropyl alcohol is brought into the process cell in violation of operating procedures, spills and ignites. All gloves are incinerated, the sprinkler system activates and protects the glovebox from further damage. The ventilation system and HEPA filters continue to function through the accident. It is estimated that 4.8x10-6 grams of plutonium are released to the environment. The annual frequency of occurrence is estimated to be in the range of 1.0x10-3 to 1.0x10-5 per year (LANL1995d). For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-4 per year, the mid point of the estimated range. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.2-1 presents the source term, by isotope, for the 4.8x10-6 grams of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.2-2 through F.2.2.5.2-6 for public consequences and in tables F.2.2.5.2-7 through F.2.2.5.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.2-1 using the GENII computer code. Spill Scenario. The most catastrophic case of leak or spill of plutonium would result from a fork lift or other large vehicle running over a package of plutonium-oxide and breaching the package. The analysis postulated that the package contained 4 kg of pluto- nium-oxide and that 0.4 gram would become airborne after the accident. During cleanup operations, the analysis assumed that an additional 0.04gram would be resuspended for a total airborne release to the room of 0.44 gram of plutonium-oxide. After three stage HEPA filtration of the facility exhaust, the total release to the environment is estimated to be 1.7x10-9 gram of plutonium. The probability calculated from the event tree for this scenario is 4.5x10-5 per year (LANL1995d). Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.2-1 presents the source term, by isotope, for the 1.7x10-9 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.2-2 through F.2.2.5.2-6 for public consequences and in tables F.2.2.5.2-7 through F.2.2.5.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.2-1 using the GENII computer code. Glovebox Explosion Scenario. The bounding design-basis accident explosion is a conflagration for a flammable gas mixture inside a glovebox. The glovebox identified as having the most material at risk contains the milling operation where plutonium-oxide is milled to a fine powder prior to mixing with uranium dioxide. Based on a LANL TA-55 standard operating procedure, the criticality limit for plutonium-oxide in a dry atmosphere is assumed to be 10 kg. The analysis assumed the glovebox contains 10 kg of plutonium-oxide and through some unforeseen set of failures, a combustible gas mixture accumulates inside a glovebox and is ignited, possibly by an electrical spark from an operating electrical device. The conflagration blows out the HEPA filter from the glovebox ventilation system exit. In addition, gloves may also be blown out. The building HEPA filters and ventilation system continue to operate during the accident. The analysis estimated that 1.0x10-3 gram of plutonium is released up the stack. The annual frequency of occurrence is estimated to be in the range of 1.0x10-3 to 1.0x10-5 per year (LANL1995d). For calculational purposes, the annual frequency of occupance is assumed to be 1.0x10-4 per year, the mid point of the estimated range. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.2-1 presents the source term, by isotope, for the 1.0x10-3 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.2-2 through F.2.2.5.2-6 for public consequences and in tables F.2.2.5.2-7 through F.2.2.5.2-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.2-1 using the GENII computer code. Table F.2.2.5.2-1.-Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accident Source Terms Isotope Released Activity (curies) - Loading Dock Fire Process Cell Fire Spill Glovebox Explosion Pu-238 6.5x10-4 4.0x10-8 1.4x10-11 8.4x10-6 Pu-239 4.5x10-3 2.8x10-7 9.8x10-11 5.8x10-5 Pu-240 1.0x10-3 6.4x10-8 2.3x10-11 1.3x10-5 Pu-241 0.031 1.9x10-6 6.9x10-10 4.0x10-4 Am-241 9.9x10-4 6.1x10-8 2.2x10-11 1.3x10-5 Source: Derived from LANL 1995d and table F.2.1.5.2-2. Table F.2.2.5.2-2.-Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 1.9x10-3 9.5x10-7 18 9.0x10-3 5.0x10-4 Process cell fire 1.2x10-7 6.0x10-11 1.1x10-3 5.5x10-7 1.0x10-4 Spill 4.2x10-11 2.1x10-14 4.0x10-7 2.0x10-10 4.5x10-5 Glovebox explosion 2.5x10-5 1.3x10-8 0.24 1.2x10-4 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 4.8x10-10 - 4.5x10-6 - Process cell fire - 6.0x10-15 - 5.5x10-11 - Spill - 9.5x10-19 - 9.0x10-15 - Glovebox explosion - 1.3x10-12 - 1.2x10-8 - Table F.2.2.5.2-3.- Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 8.3x10-4 4.2x10-7 0.29 1.5x10-4 5.0x10-4 Process cell fire 5.2x10-8 2.6x10-11 1.8x10-5 9.0x10-9 1.0x10-4 Spill 1.9x10-11 9.5x10-15 6.5x10-9 3.3x10-12 4.5x10-5 Glovebox explosion 1.1x10-5 5.5x10-9 3.7x10-3 1.9x10-6 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 2.1x10-10 - 7.5x10-8 - Process cell fire - 2.6x10-15 - 9.0x10-13 - Spill - 4.3x10-19 - 1.5x10-16 - Glovebox explosion - 5.5x10-13 - 1.9x10-10 - Table F.2.2.5.2-4.-Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 0.016 8.0x10-6 180 0.09 5.0x10-4 Process cell fire 1.0x10-6 5.0x10-10 0.011 5.5x10-6 1.0x10-4 Spill 3.6x10-10 1.8x10-13 4.0x10-6 2.0x10-9 4.5x10-5 Glovebox explosion 2.1x10-4 1.1x10-7 2.4 1.2x10-3 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 4.0x10-9 - 4.5x10-5 - Process cell fire - 5.0x10-14 - 5.5x10-10 - Spill - 8.1x10-18 - 9.0x10-14 - Glovebox explosion - 1.1x10-11 - 1.2x10-7 - Table F.2.2.5.2-5.- Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 1.5x10-3 7.5x10-7 6 3.0x10-3 5.0x10-4 Process cell fire 9.2x10-8 4.6x10-11 3.8x10-4 1.9x10-7 1.0x10-4 Spill 3.2x10-11 1.6x10-14 1.3x10-7 6.5x10-11 4.5x10-5 Glovebox explosion 1.9x10-5 9.5x10-9 0.077 3.9x10-5 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 3.8x10-10 - 1.5x10-6 - Process cell fire - 4.6x10-15 - 1.9x10-11 - Spill - 7.2x10-19 - 2.9x10-15 - Glovebox explosion - 9.5x10-13 - 3.9x10-9 - Table F.2.2.5.2-6.-Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 5.4x10-3 2.7x10-6 170 0.085 5.0x10-4 Process cell fire 3.4x10-7 1.7x10-10 0.011 5.5x10-6 1.0x10-4 Spill 1.2x10-10 6.0x10-14 3.7x10-6 1.9x10-9 4.5x10-5 Glovebox explosion 7.1x10-5 3.6x10-8 2.2 1.1x10-3 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 1.4x10-9 - 4.3x10-5 - Process cell fire - 1.7x10-14 - 5.5x10-10 - Spill - 2.7x10-18 - 8.6x10-14 - Glovebox explosion - 3.6x10-12 - 1.1x10-7 - Table F.2.2.5.2-7.-Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 0.083 3.3x10-5 0.028 1.1x10-5 5.0x10-4 Process cell fire 5.2x10-6 2.1x10-9 1.7x10-6 6.8x10-10 1.0x10-4 Spill 1.8x10-9 7.2x10-13 6.1x10-10 2.4x10-13 4.5x10-5 Glovebox explosion 1.1x10-3 4.4x10-7 3.6x10-4 1.4x10-7 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 1.7x10-8 - 5.5x10-9 - Process cell fire - 2.1x10-13 - 6.8x10-14 - Spill - 3.2x10-17 - 1.1x10-17 - Glovebox explosion - 4.4x10-11 - 1.4x10-11 - Table F.2.2.5.2-8.-Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 0.024 9.6x10-6 8.1x10-3 3.2x10-6 5.0x10-4 Process cell fire 1.5x10-6 6.0x10-10 5.0x10-7 2.0x10-10 1.0x10-4 Spill 5.4x10-10 2.2x10-13 1.8x10-10 7.2x10-14 4.5x10-5 Glovebox explosion 3.2x10-4 1.3x10-7 1.0x10-4 4.0x10-8 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 4.8x10-9 - 1.6x10-9 - Process cell fire - 6.0x10-14 - 2.0x10-14 - Spill - 9.9x10-18 - 3.2x10-18 - Glovebox explosion - 1.3x10-11 - 4.0x10-12 - Table F.2.2.5.2-9.-Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 0.13 5.2x10-5 0.042 1.7x10-5 5.0x10-4 Process cell fire 8.1x10-6 3.2x10-9 2.6x10-6 1.0x10-9 1.0x10-4 Spill 2.8x10-9 1.1x10-12 9.3x10-10 3.7x10-13 4.5x10-5 Glovebox explosion 1.6x10-3 6.4x10-7 5.4x10-4 2.2x10-7 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 2.6x10-8 - 5.0x10-9 - Process cell fire - 3.2x10-13 - 1.0x10-13 - Spill - 5.0x10-17 - 1.7x10-17 - Glovebox explosion - 6.4x10-11 - 2.2x10-11 - Table F.2.2.5.2-10.-Mixed Oxide Fuel Fabrication Low-to-Moderate Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 9.8x10-3 3.9x10-6 2.7x10-3 1.1x10-6 5.0x10-3 Process cell fire 6.0x10-7 2.4x10-10 1.7x10-7 6.8x10-11 1.0x10-4 Spill 2.2x10-10 8.8x10-14 6.1x10-11 2.4x10-14 4.5x10-5 Glovebox explosion 1.3x10-4 5.2x10-8 3.6x10-5 1.4x10-8 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 2.0x10-8 - 5.5x10-9 - Process cell fire - 2.4x10-14 - 6.8x10-15 - Spill - 4.0x10-18 - 1.1x10-18 - Glovebox explosion - 5.2x10-12 - 1.4x10-12 - Table F.2.2.5.2-11.- Multipurpose Reactor Fuel Fabrication Low-to-Moderate Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 0.22 8.8x10-5 0.076 3.0x10-5 5.0x10-4 Process cell fire 1.4x10-5 5.6x10-9 4.8x10-6 1.9x10-9 1.0x10-4 Spill 4.9x10-9 2.0x10-12 1.7x10-9 6.8x10-13 4.5x10-5 Glovebox explosion 2.9x10-3 1.2x10-6 9.9x10-4 4.0x10-7 1.0x10-4 Expected Risk of Cancer Fatality (per year) Loading dock fire - 4.4x10-8 - 1.5x10-7 - Process cell fire - 5.6x10-13 - 1.9x10-13 - Spill - 9.0x10-17 - 3.1x10-17 - Glovebox explosion - 1.2x10-10 - 4.0x10-11 - F.2.2.5.3 Pit Disassembly and Conversion Loading Dock Fire Scenario. The accident postulated is a fire on an open loading dock caused by welding, cleaning solvents, electrical shorts, or other miscellaneous causes. A single drum of combustible waste, containing 18 grams of plutonium, is involved in the fire. The analysis estimated that 0.077 gram of plutonium was released directly to the environment by the fire. The annual frequency of occurrence is estimated to be in the range of 1.0x10-4 to 1.0x10-5 per year. (LANLdef) For calculational purposes, the annual frequency of occurrence is assumed to be 5.0x10-4 per year, the mid point of the estimated range. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.3-1 presents the source term, by isotope, for the 0.077 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.3-2 through F.2.2.5.3-6 for public consequences and in tables F.2.2.5.3-7 through F.2.2.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.3-1 using the GENII computer code. Process Cell Fire Scenario. The accident postulated is a fire in a process cell area. It is assumed that the process cell contains a glovebox used for milling plutonium powder. The gloves have become coated with a layer of plutonium dust. The analysis estimated the glove loading at 2 grams of plutonium per glove. Each of the 12 gloves is assumed to be stowed outside of the glovebox. A flammable cleaning liquid such as acetone or isopropyl alcohol is brought into the process cell in violation of operating procedures, spills and ignites. All gloves are incinerated, the sprinkler system activates and protects the glovebox from further damage. The ventilation system and HEPA filters continue to function through the accident. It is estimated that 4.8x10-6 gram of plutonium is released to the environment. The annual frequency of occurrence is estimated to be in the range of 1.0x10-3 to 1.0x10-5 per year. (LANLdef) For calculational purposes, the annual frequency of occurrence is assumed to be 1.0x10-4 per year, the mid point of the estimated range. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.3-1 presents the source term, by isotope, for the 4.8x10-6 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.3-2 through F.2.2.5.3-6 for public consequences and in tables F.2.2.5.3-7 through F.2.2.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.3-1 using the GENII computer code. Spill Scenario. The most catastrophic case of leak or spill of plutonium would result from a fork lift or other large vehicle running over a package of plutonium-oxide and breaching the package. The analysis postulated that the package contained 4 kg of plutonium-oxide and that 0.4 gram would become airborne after the accident. During cleanup operations, the analysis assumed that an additional 0.04 gram would be resuspended for a total airborne release to the room of 0.44gram of plutonium-oxide. After three stage HEPA filtration of the facility exhaust, the total release to the environment is estimated to be 1.7x10-9 gram of plutonium. The probability calculated from the event tree for this scenario is 4.5x10-5 per year (LANL 1995 def). Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.3-1 presents the source term, by isotope, for the 1.7x10-9 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.3-2 through F.2.2.5.3-6 for public consequences and in tables F.2.2.5.3-7 through F.2.2.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.3-1 using the GENII computer code. Glovebox Explosion Scenario. The bounding design-basis accident explosion is a conflagration for a flammable gas mixture inside a glovebox. The glovebox identified as having the most material at risk contains the milling operation where plutonium-oxide is milled to a fine powder prior to mixing with uranium dioxide. Based on a LANL TA-55 standard operating procedure, the criticality limit for plutonium-oxide in a dry atmosphere is assumed to be 4.5 kg. The analysis assumed the glovebox contains 4.5 kg of plutonium-oxide and through some unforeseen set of failures, a combustible gas mixture accumulates inside a glovebox and is ignited, possibly by an electrical spark from an operating electrical device. The con- flagration blows out the HEPA filter from the glovebox ventilation system exit. In addition, gloves may also be blown out. The building HEPA filters and ventilation system continue to operate during the accident. The analysis estimated that 4.5x10-4 gram of plutonium is released up the stack. The annual frequency of occurrence is estimated to be in the range of 1.0x10-3 to 1.0x10-5 per year. (LANL 1995 def) For calculational purposes, the annual frequency of occupance is assumed to be 1.0x10-4 per year, the mid point of the estimated range. Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.3-1 presents the source term, by isotope, for the 4.5x10-4 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.3-2 through F.2.2.5.3-6 for public consequences and in tables F.2.2.5.3-7 through F.2.2.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.3-1 using the GENII computer code. Ion Exchange Column Explosion Scenario. The postulated accident assumed the processing of the maximum possible plutonium load and 223 grams of material would be released to the room due to the explosion. 45 grams of the material would be aerosol consisting of 20 grams per liter of plutonium nitrate solution. The quantity of soluble plutonium released would be 0.75 gram. A total of 3 grams of plutonium would be released to the room. The ventilation system continues to operate and the aerosol would be carried through the ventilation system to the HEPA filters. The final environmental release was estimated to be 3.0x10-9 gram of plutonium. The accident frequency is estimated to be 7.0x10-4 per year. (LANL 1995 def) Table F.2.1.5.2-2 presents the isotopic distribution for a plutonium release at the mixed-oxide fuel reactor facility. Table F.2.2.5.3-1 presents the source term, by isotope, for the 3.0x10-9 gram of plutonium released to the environment during the postulated accident. Consequences. The estimated consequences of the postulated accident with 50 percent meteorology at each site are shown in tables F.2.2.5.3-2 through F.2.2.5.3-6 for public consequences and in tables F.2.2.5.3-7 through F.2.2.5.3-11 for worker consequences. The dose estimates are based on analysis of the source terms in table F.2.2.5.3-1 using the GENII computer code. Table F.2.2.5.3-1.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accident Source Terms Isotope Released Activity (curies) - Loading dock fire Process Spill Glovebox explosions Ion Exchange Cell Fire Column Explosion Pu-238 6.5x10-4 4.0x10-8 1.4x10-11 3.8x10-6 2.5x10-11 Pu-239 4.5x10-3 2.8x10-7 9.8x10-11 2.6x10-5 1.7x10-10 Pu-240 1.0x10-3 6.4x10-8 2.3x10-11 6.0x10-6 4.0x10-11 Pu-241 0.031 1.9x10-6 6.9x10-10 1.8x10-4 1.2x10-9 Pu-241 9.9x10-4 6.1x10-8 2.2x10-11 5.8x10-6 3.8x10-11 Source: Derived from LANL 1995d and table F.2.1.5.2-2.. Table F.2.2.5.3-2.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 1.9x10-3 9.5x10-7 0.18 9.0x10-3 5.0x10-4 Process cell fire 1.2x10-7 6.0x10-11 1.1x10-3 5.5x10-7 1.0x10-4 Spill 4.21x10-11 2.1x10-14 4.0x10-7 2.0x10-10 4.5x10-5 Glovebox explosion 1.1x10-5 5.5x10-8 0.10 5.0x10-5 1.0x10-4 Ion exchange column explosion 7.3x10-11 3.7x10-14 7.1x10-7 3.6x10-10 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 4.8x10-10 - 4.5x10-6 - Process cell fire - 6.0x10-15 - 5.5x10-11 - Spill - 9.5x10-19 - 9.0x10-15 - Glovebox explosion - 5.5x10-12 - 5.0x10-9 - Ion exchange column explosion - 2.6x10-17 - 2.5x10-13 - Table F.2.2.5.3-3.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Nevada Test Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 8.3x10-4 4.2x10-7 0.29 1.5x10-4 5.0x10-4 Process cell fire 5.2x10-8 2.6x10-11 1.8x10-5 9.0x10-9 1.0x10-4 Spill 1.9x10-11 9.5x10-15 6.5x10-9 3.3x10-12 4.5x10-5 Glovebox explosion 4.9x10-6 2.5x10-9 1.7x10-3 8.5x10-7 1.0x10-4 Ion exchange column explosion 3.2x10-11 1.6x10-14 1.1x10-8 5.5x10-12 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 2.1x10-10 - 7.5x10-8 - Process cell fire - 2.6x10-15 - 9.0x10-13 - Spill - 4.3x10-19 - 1.5x10-16 - Glovebox explosion - 2.5x10-13 - 8.5x10-11 - Ion exchange column explosion - 1.1x10-17 - 3.9x10-15 - Table F.2.2.5.3-4.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 0.016 8.0x10-6 180 0.090 5.0x10-4 Process cell fire 1.0x10-6 5.0x10-10 0.011 5.5x10-6 1.0x10-4 Spill 3.6x10-10 1.8x10-13 4.0x10-6 2.0x10-9 4.5x10-5 Glovebox explosion 9.4x10-5 4.7x10-8 1.0 5.0x10-4 1.0x10-4 Ion exchange column explosion 6.1x10-10 3.1x10-13 7.1x10-6 3.6x10-9 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 4.0x10-9 - 4.5x10-5 - Process cell fire - 5.0x10-14 - 5.5x10-10 - Spill - 8.1x10-18 - 9.0x10-14 - Glovebox explosion - 4.7x10-12 - 5.0x10-8 - Ion exchange column explosion - 2.2x10-16 - 2.5x10-12 - Table F.2.2.5.3-5.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Pantex Plant-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 1.5x10-3 7.5x10-7 6.0 3.0x103- 5.0x10-4 Process cell fire 9.2x10-8 4.6x10-11 3.8x10-4 1.9x10-7 1.0x10-4 Spill 3.2x10-11 1.6x10-14 1.3x10-7 6.5x10-11 4.5x10-5 Glovebox explosion 8.7x10-6 4.4x10-9 0.035 1.8x10-5 1.0x10-4 Ion exchange column explosion 5.6x10-11 2.8x10-14 2.3x10-7 1.2x10-10 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 3.8x10-10 - 1.5x10-6 - Process cell fire - 4.6x10-15 - 1.9x10-11 - Spill - 7.2x10-19 - 2.9x10-15 - Glovebox explosion - 4.4x10-13 - 1.8x10-9 - Ion exchange column explosion - 2.0x10-17 - 8.4x10-14 - Table F.2.2.5.3-6.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Savannah River Site-Public Consequences - Maximum Offsite Individual Population to 50 Miles - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Loading dock fire 5.4x10-3 2.7x10-6 170 0.085 5.0x10-4 Process cell fire 3.4x10-7 1.7x10-10 0.011 5.5x10-6 1.0x10-4 Spill 1.2x10-10 6.0x10-14 3.7x10-6 1.9x10-9 4.5x10-5 Glovebox explosion 3.2x10-5 1.6x10-8 1.0 5.0x10-4 1.0x10-4 Ion exchange column explosion 2.1x10-5 1.1x10-13 6.6x10-6 3.3x10-9 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 1.4x10-9 - 4.3x10-5 - Process cell fire - 1.7x10-14 - 5.5x10-10 - Spill - 2.7x10-18 - 8.6x10-14 - Glovebox explosion - 1.6x10-12 - 5.0x10-8 - Ion exchange column explosion - 7.7x10-17 - 2.3x10-12 - Table F.2.2.5.3-7.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Idaho National Engineering Laboratory-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 0.083 3.3x10-5 0.028 1.1x10-5 5.0x10-4 Process cell fire 5.2x10-6 2.1x10-9 1.7x10-6 6.8x10-10 1.0x10-4 Spill 1.8x10-9 7.2x10-13 6.1x10-10 2.4x10-13 4.5x10-5 Glovebox explosion 4.8x10-4 1.9x10-7 1.6x10-4 6.4x10-8 1.0x10-4 Ion exchange column explosion 3.2x10-9 1.3x10-12 1.1x10-9 4.4x10-13 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 1.7x10-8 - 5.5x10-9 - Process cell fire - 2.1x10-13 - 6.8x10-14 - Spill - 3.2x10-17 - 1.1x10-17 - Glovebox explosion - 1.9x10-11 - 6.4x10-12 - Ion exchange column explosion - 9.1x10-16 - 3.1x10-16 - Table F.2.2.5.3-8.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Nevada Test Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 0.024 9.6x10-6 8.1x10-3 3.2x10-6 5.0x10-4 Process cell fire 1.5x10-6 6.0x10-10 5.0x10-7 2.0x10-10 1.0x10-4 Spill 5.4x10-10 2.2x10-13 1.8x10-10 7.2x10-14 4.5x10-5 Glovebox explosion 1.4x10-4 5.6x10-8 4.8x10-5 1.9x10-8 1.0x10-4 Ion exchange column explosion 9.4x10-10 3.8x10-13 3.1x10-10 1.2x10-13 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 4.8x10-9 - 1.6x10-9 - Process cell fire - 6.0x10-14 - 2.0x10-14 - Spill - 9.9x10-18 - 3.2x10-18 - Glovebox explosion - 5.6x10-12 - 1.9x10-12 - Ion exchange column explosion - 2.7x10-16 - 8.4x10-17 - Table F.2.2.5.3-9.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Oak Ridge Reservation-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 0.13 5.2x10-5 0.042 1.7x10-5 5.0x10-4 Process cell fire 8.1x10-6 3.2x10-9 2.6x10-6 1.0x10-9 1.0x10-4 Spill 2.8x10-9 1.1x10-12 9.3x10-10 3.7x10-13 4.5x10-5 Glovebox explosion 7.2x10-4 2.9x10-7 2.5x10-4 1.0x10-7 1.0x10-4 Ion exchange column explosion 4.9x10-9 2.0x10-12 1.6x10-9 6.4x10-13 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 2.6x10-8 - 8.5x10-9 - Process cell fire - 3.2x10-13 - 1.0x10-13 - Spill - 5.0x10-17 - 1.7x10-17 - Glovebox explosion - 2.9x10-11 - 1.0x10-11 - Ion exchange column explosion - 1.4x10-15 - 4.5x10-16 - Table F.2.2.5.3-10.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Pantex Plant-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 9.8x10-3 3.9x10-6 2.7x10-3 1.1x10-6 5.0x10-4 Process cell fire 6.0x10-7 2.4x10-10 1.7x10-7 6.8x10-11 1.0x10-4 Spill 2.2x10-10 8.8x10-14 6.1x10-11 2.4x10-14 4.5x10-5 Glovebox explosion 5.6x10-5 2.2x10-8 1.6x10-5 6.4x10-9 1.0x10-4 Ion exchange column explosion 3.7x10-10 1.5x10-13 1.0x10-10 4.0x10-14 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 2.0x10-9 - 5.5x10-10 - Process cell fire - 2.4x10-14 - 6.8x10-15 - Spill - 4.0x10-18 - 1.1x10-18 - Glovebox explosion - 2.2x10-12 - 6.4x10-13 - Ion exchange column explosion - 1.1x10-16 - 2.8x10-17 - Table F.2.2.5.3-11.-Pit Disassembly and Conversion Low-to-Moderate Consequence Accidents at Savannah River Site-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Accident Description Dose Cancer Dose Cancer Accident (rem) Fatality (rem) Fatalitya Frequency (per year) Loading dock fire 0.22 8.8x10-5 7.6x10-2 3.0x10-5 5.0x10-4 Process cell fire 1.4x10-5 5.6x10-9 4.8x10-6 1.9x10-9 1.0x10-4 Spill 4.9x10-9 2.0x10-12 1.7x10-9 6.8x10-13 4.5x10-5 Glovebox explosion 1.3x10-3 5.2x10-7 4.5x10-4 1.8x10-7 1.0x10-4 Ion exchange column explosion 8.4x10-9 3.4x10-12 2.9x10-9 1.2x10-12 7.0x10-4 Expected Risk of Cancer Fatality (cancer fatalities per year) Loading dock fire - 4.4x10-8 - 1.5x10-8 - Process cell fire - 5.6x10-13 - 1.9x10-13 - Spill - 9.0x10-17 - 3.1x10-17 - Glovebox explosion - 5.2x10-11 - 1.8x10-11 - Ion exchange column explosion - 2.4x10-15 - 8.4x10-16 - F.2.2.6 Tritium Target Extraction Facility Scenario. A tritium target extraction facility removes tritium from the targets. The bounding accidents for the tritium extraction facility are based on the analysis of tritium operations at SRS. The bounding low-to-moderate consequence accident for the facility postulated an explosion in the extraction facility. The explosion was initiated by air leakage from furnace leaks, tank leaks, connection leaks, pump leaks, valve leaks or during process maintenance. The air leakage formed a flammable mixture that subsequently ignites. Approximately 1.4x106 Ci of tritium in oxide form could be released to the material handling room and subsequently to the environment. The accident annual frequency of occurrence is estimated at 2.0x10-5 per year at SRS (DOE 1994a). Consequences. The estimated consequences of the postulated tritium target extraction facility bounding accident for each site are shown for the public in table F.2.2.6-1 and for the worker in table F.2.2.6-2 for 50percent meteorology conditions. The estimates are based on the postulated release of 1.4x106 Ci of tritium in the oxide form directly to the environment during the accident using the GENII computer code. Table F.2.2.6-1.-Tritium Target Extraction Facility Bounding Low-to-Moderate Consequence Accident-Public Consequences - Individual at Site Boundary Population to 50 Miles - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 0.099 5.0x10-5 900 0.45 2.0x10-5 Laboratory Nevada Test Site 0.043 2.2x10-5 15 7.5x10-3 2.0x10-5 Oak Ridge Reservation 0.84 4.2x10-4 9.0x103 4.5 2.0x10-5 Pantex Plant 0.077 3.9x10-5 320 0.16 2.0x10-5 Savannah River Site 0.23 1.2x10-4 1.2x104 6 2.0x10-5 Expected Risk of Cancer Fatality (per year) Idaho National Engineering - 1.0x10-9 - 9.0x10-6 - Laboratory Nevada Test Site - 4.4x10-10 - 1.5x10-7 - Oak Ridge Reservation - 8.4x10-9 - 9.0x10-5 - Pantex Plant - 7.8x10-10 - 3.2x10-6 - Savannah River Site - 2.4x10-9 - 1.2x10-4 - Table F.2.2.6-2.-Tritium Target Extraction Facility Bounding Low-to-Moderate Consequence Accident-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 4.3 1.7x10-3 1.4 5.6x10-4 2.0x10-5 Laboratory Nevada Test Site 1.3x10-4 5.2x10-8 4.2x10-5 1.7x10-8 2.0x10-5 Oak Ridge Reservation 6.6 2.6x10-3 2.2 8.8x10-4 2.0x10-5 Pantex Plant 0.51 2.4x10-4 0.14 5.6x10-5 2.0x10-5 Savannah River Site 12 4.8x10-3 4 1.6x10-3 2.0x10-5 Expected Risk of Cancer Fatality (per year) Idaho National Engineering - 3.4x10-8 - 1.1x10-8 - Laboratory Nevada Test Site - 1.0x10-12 - 3.4x10-13 - Oak Ridge Reservation - 5.2x10-8 - 1.8x10-8 - Pantex Plant - 4.8x10-9 - 1.1x10-9 - Savannah River Site - 9.6x10-8 - 3.2x10-8 - F.2.3 Tritium Recycling Facility High Consequence Accident The bounding accidents selected for the tritium recycling facility are based on the analysis of tritium operations at SRS. While the spectrum of accidents is representative of the types of accidents to be considered in the design, development, and analysis of the plant, the estimated consequences of the accidents may be conservative because they are based on analyses of facilities that may not all meet the general design and safety requirements that will be implemented for new tritium supply facilities. If the tritium supply facility is located at either INEL, NTS, ORR, or Pantex, the tritium recycling facility could be collocated at the same site. If the tritium supply facility is located at SRS, the existing tritium recycling facilities at SRS would be upgraded. Both high consequence accidents and design-basis/operational accidents are considered. High consequence accidents include accidents caused by natural phenomena (i.e., earthquake, flood, tornado, tornado-driven debris, and high winds) in excess of the module design basis for safety systems. Operational accidents include fire, explosion, and spills. All upgraded or new tritium recycling facility safety-class structures and safety systems will be designed and installed to meet the design-basis earthquake, flood, tornado, tornado driven debris, and wind natural phenomena requirements. Scenario. The postulated bounding high consequence accident is a beyond design-basis earthquake that results in the spontaneous ignition of tritium released from ruptured reservoirs stored in the facility unloading station. The analysis postulated that the accident source term released to the environment during the accident is 8.4x106 Ci of tritium in oxide form. The accident annual frequency of occurrence at SRS is 2.0x10-5 per year (DOE 1995g). The accident annual frequency of occurrence for new tritium recycling facilities at the other candidate sites will be less than the frequency for existing facilities at SRS. It is assumed that the storage and confinement systems will be designed to maintain functional integrity following a design-basis earthquake or a safe shutdown earthquake with a return frequency of 1.0x10-4 per year. The evaluation also assumed that the storage and confinement systems may survive an earthquake with a return frequency of 1.0x10-5 per year but catastrophic failure of the facility could be expected after an earthquake with a return frequency of 1.0x10-6 per year. For the purpose of calculating the point estimate of risk for the postulated accident, the accident annual frequency of occurrence for all new facilities is assumed to be 1.0x10-6 per year. Consequences. The estimated consequences of the postulated high consequence accident for each of the four sites and for the SRS upgrade are shown for the public in table F.2.3-1 and for the worker in table F.2.3-2. The dose and latent cancer fatality estimates were generated using the MACCS computer code and the postulated release of 8.4x106 Ci of tritium in the oxide form directly to the environment during the accident. Cancer Fatalities Complementary Cumulative Distribution Function for the Tritium Recycling Facility High Consequence Accident Figure F.2.3-1 shows the annual probability that, in the event of the tritium recycling facility high consequence accident at one of the sites, the number of cancer fatalities exceeds the value N indicated on the horizontal axis. The curves, technically referred to as complementary cumulative distribution functions, reflect the probability of the accident's occurrence as well as the variability in the magnitude of its consequences. Generally, a curve that extends the farthest to the right has the highest accident consequences while a curve that is nearest to the left has the lowest accident consequences. A comparison of alternatives should include the information provided by these curves in conjunction with the point values shown in tables F.2.3-1 and F.2.3-2. Figure (Page F-145) Figure F.2.3-1.-Tritium Recycling Facility Cancer Fatalities Complementary Cumulative Distribution Functions for High Consequence Accident. Table F.2.3-1.-Tritium Recycling Facility High Consequence Accident-Public Consequences - Individual at Site Boundary Population to 50 Miles - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 0.048 2.4x10-5 81 0.04 1.0x10-6 Laboratory Nevada Test Site 0.13 6.6x10-5 7.7 3.9x10-3 1.0x10-6 Oak Ridge Reservation 1 5.2x10-4 751 0.38 1.0x10-6 Pantex Plant 0.7 3.5x10-4 98 0.049 1.0x10-6 Savannah River Site 0.045 2.2x10-5 302 0.15 2.0x10-5 Expected Risk of Cancer Fatality (per year) Idaho National Engineering - 2.4x10-11 - 4.0x10-8 - Laboratory Nevada Test Site - 6.6x10-11 - 3.9x10-9 - Oak Ridge Reservation - 5.2x10-10 - 3.8x10-7 - Pantex Plant - 3.5x10-10 - 4.9x10-8 - Savannah River Site - 4.4x10-10 - 3.0x10-6 - Table F.2.3-2.-Tritium Recycling Facility High Consequence Accident-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 6 2.4x10-3 2.2 8.8x10-4 1.0x10-6 Laboratory Nevada Test Site 4.4 1.7x10-3 1.7 6.7x10-4 1.0x10-6 Oak Ridge Reservation 5.9 2.3x10-3 2.1 8.4x10-4 1.0x10-6 Pantex Plant 2.6 1.0x10-3 0.98 3.9x10-4 1.0x10-6 Savannah River Site 2.6 1.0x10-3 0.98 3.9x10-4 2.0x10-5 Expected Risk of Cancer Fatality (per year) Idaho National Engineering - 2.4x10-9 - 8.8x10-10 - Laboratory Nevada Test Site - 1.7x10-9 - 6.7x10-10 - Oak Ridge Reservation - 2.3x10-9 - 8.4x10-10 - Pantex Plant - 1.0x10-9 - 3.9x10-10 - Savannah River Sitec - 2.0x10-8 - 7.8x10-9 - F.2.4 Tritium Recycling Facility Low-to-Moderate Consequence Accident Scenario. The postulated bounding low-to-moderate consequence accident is the overheating and rupture of a hydride bed. Hydride beds are capable of being overheated to rupture due to equipment failures. Approximately 6,000 Ci of tritium in oxide form could be released to the environment. The accident annual frequency of occurrence is estimated at 2.0x10-4 per year at SRS (DOE 1995g). Consequences. The estimated consequences of the postulated hydride bed rupture accident for each of the four tritium supply technologies and recycling sites and for the SRS recycling facilities upgrade option are shown for the public in table F.2.4-1 and for the workers in table F.2.4-2 for 50 percent meteorology conditions. The estimates are based on the analysis of the postulated release of 6,000 Ci of tritium in oxide form directly to the environment during the accident using the GENII computercode. Table F.2.4-1.-Tritium Recycling Facility Hydride Bed Rupture Accident-Public Consequences - Individual at Site Boundary Population to 50 Miles - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 4.2x10-4 2.1x10-7 4.1 2.1x10-3 2.0x10-4 Laboratory Nevada Test Site 1.9x10-4 9.5x10-8 0.064 3.2x10-5 2.0x10-4 Oak Ridge Reservation 3.6x10-3 1.8x10-6 41 0.021 2.0x10-4 Pantex Plant 3.3x10-4 1.7x10-7 1.4 7.0x10-4 2.0x10-4 Savannah River Site 9.9x10-4 4.9x10-7 49 0.025 2.0x10-4 Expected Risk of Cancer Fatality (per year) Idaho National Engineering - 4.2x10-11 - 4.2x10-7 - Laboratory Nevada Test Site - 1.9x10-11 - 6.4x10-9 - Oak Ridge Reservation - 3.6x10-10 - 4.2x10-6 - Pantex Plant - 3.4x10-11 - 1.4x10-7 - Savannah River Siteb - 9.8x10-11 - 5.0x10-6 - Table F.2.4-2.-Tritium Recycling Facility Hydride Bed Rupture Accident-Worker Consequences - Worker at 1,000 meters Worker at 2,000 meters - Site Dose Cancer Dose Cancer Accident (rem) Fatality (person-rem) Fatality Frequency (per year) Idaho National Engineering 1.8x10-6 7.2x10-10 6.1x10-7 2.4x10-10 2.0x10-4 Laboratory Nevada Test Site 5.4x10-3 2.2x10-6 1.8x10-3 7.2x10-7 2.0x10-4 Oak Ridge Reservation 0.028 1.1x10-5 9.0x10-3 3.6x10-6 2.0x10-4 Pantex Plant 2.2x10-3 8.8x10-7 6.0x10-8 2.4x10-11 2.0x10-4 Savannah River Site 0.049 2.0x10-5 1.7x10-2 6.8x10-6 2.0x10-4 Expected Risk of Cancer Fatality (per year) Idaho National Engineering - 1.4x10-13 - 4.8x10-14 - Laboratory Nevada Test Site - 4.4x10-10 - 1.4x10-10 - Oak Ridge Reservation - 2.2x10-9 - 7.2x10-10 - Pantex Plant - 1.8x10-10 - 4.8x10-15 - Savannah River Sitec - 4.0x10-9 - 1.4x10-9 - F.3 Secondary Impacts of Accidents The primary impacts of accidents are measured in terms of public and worker exposures to radiation and toxic chemicals. The secondary impacts of accidents affect elements of the environment other than humans. For example, a radiological release may contaminate farmland, surface and underground water, recreational areas, industrial parks, historical sites, or the habitat of an endangered species. As a result, farm products may have to be destroyed; the supply of drinking water may be lowered; recreational areas may be closed; industrial parks may suffer economic losses during shutdown for decontamination; historical sites may have to be closed to visitors; and the endangered species may move closer to extinction. This section addresses the secondary impacts of a reactor charge/discharge design-basis accident in the region of a radiological release. This accident was selected as representative of a design-basis accident although another accident for any other technology could also have been selected to illustrate the secondary effects. Other design-basis accidents with greater source terms could also be found that would show secondary effects extending over a larger region than shown in figures F.3.1-1 through F.3.5-1. The source term for the HWR charge/discharge accident is shown in table F.2.2.1-1. The level of exposure estimates are based on analysis of the source term in table F.2.2.1-1 using the GENII computer code with 50 percent meteorology conditions for each site. The region of secondary effects extends out from the point of release in a pattern formed by dispersion parameters such as meteorology. The level of exposure is generally decreasing with increasing distance from the release point. Figures F.3.1-1 through F.3.5-1 show the shapes of patterns for each site at a distance at which the level of exposure from the accidental release would be equivalent to the level of annual exposure from natural background radiation at each site. Levels of exposure that are less than natural background can be expected in areas outside of the shaded pattern. These results are useful for comparing the sensitivity of sites with respect to the secondary impacts for an accidental radiological release from a reactor. In reviewing the results, it is useful to note whether the impacted area extends beyond the site boundary where the economic impacts would be larger than if the area were contained within the site boundary. It is also useful to note the size of the contaminated area in which the level of exposure exceeds exposures from natural background. F.3.1 Idaho National Engineering Laboratory In the region of INEL, the natural background level of radiation (excluding radon) is 113 mrem per year. The results shown in figure F.3.1-1 indicate that, for an accidental release, the radiation levels exceeding 113 mrem per year (shaded area bounded by a bold line) are well within the site boundary. The size of the area in which exposure levels would exceed exposures from natural background radiation is 6.7x107 square meters (16,556 acres). Section 4.2 describes the land, water, biotic, cultural, paleontological, and socioeconomic resources in the INEL environment that may receive secondary impacts from a design-basis accident. F.3.2 Nevada Test Site In the region of NTS, the natural background level of radiation (excluding radon) is 78 mrem per year. The results shown in figure F.3.2-1 indicate that, for an accidental release, the radiation levels exceeding 78 mrem per year (shaded area bounded by a bold line) are well within the site boundary. The size of the area in which exposure levels would exceed exposures from natural background radiation is 9.1x105 square meters (225 acres). Section 4.3 describes the land, water, biotic, cultural, paleontological, and socioeconomic resources in the NTS environment that may receive secondary impacts from a design-basis accident. F.3.3 Oak Ridge Reservation In the region of ORR, the natural background level of radiation (excluding radon) is 67 mrem per year. The results shown in figure F.3.3-1 indicate that, for an accidental release, the radiation levels exceeding 67 mrem per year (shaded area bounded by a bold line) are well within the site boundary. The size of the area in which exposure levels would exceed exposures from natural background radiation is 1.4x107 square meters (3,459 acres). Section 4.4 describes the land, water, biotic, cultural, paleontlogical, and socioeconomic resources in the ORR environment that may receive secondary impacts from a design-basis accident. F.3.4 Pantex Plant In the region of Pantex, the natural background level of radiation (excluding radon) is 107 mrem per year. The results shown in figure F.3.4-1 indicate that, for an accidental release, the radiation levels exceeding 107 mrem per year (shaded area bounded by a bold line) extend beyond the site boundary. The size of the area in which exposure levels would exceed exposures from natural background radiation is 9.3x107 square meters (22,980 acres). Section 4.5 describes the land, water, biotic, cultural, paleontological, and socioeconomic resources in the Pantex environment that may receive secondary impacts from a design-basis accident. F.3.5 Savannah River Site In the region of the SRS, the natural background level of radiation (excluding radon) is 76 mrem per year. The results shown in figure F.3.5-1 indicate that, for an accidental release, the radiation levels exceeding 76 mrem per year (shaded area bounded by a bold line) are well within the site boundary. The size of the area in which exposure levels would exceed exposures from natural background radiation is 2.9x107 square meters (7,166 acres). Section 4.6 describes the land, water, biotic, cultural, paleontological, and socioeconomic resources in the SRS environment that may receive secondary impacts from a design basis accident. Figure (Page F-150) Figure F.3.1-1.-Design-Basis Accident for Typical Reactor at Idaho National Engineering Laboratory (ground surface exposure-113 mrem per year). Figure (Page F-151) Figure F.3.2-1.-Design-Basis Accident for Typical Reactor at Nevada Test Site (ground surface exposure-78 mrem per year). Figure (Page F-152) Figure F.3.3-1.-Design-Basis Accident for Typical Reactor at Oak Ridge Reservation (ground surface exposure-67 mrem per year). Figure (Page F-153) Figure F.3.4-1.-Design-Basis Accident for Typical Reactor at Pantex Plant, Texas (ground surface exposure-107 mrem per year). Figure (Page F-154) Figure F.3.5-1.-Design-Basis Accident for Typical Reactor at Savannah River Site (ground surface exposure-76 mrem per year).
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