APPENDIX E. ACCIDENTS
This appendix summarizes accidents that could involve nuclear material management. It provides consequences (e.g., resulting doses) from potential releases of specific nuclear materials for each alternative discussed in this EIS.
In preparing this environmental impact statement, DOE reviewed safety analysis reports and supporting accident analyses for facilities that the alternatives described in Chapter 2 could involve. There are no accident analyses for alternatives that would involve new facilities or extensive modifications to existing facilities. In such cases, DOE used accident analyses for existing facilities at SRS that perform similar operations or that process and handle forms of nuclear material that are more hazardous than those being considered in this EIS. DOE believes that the types of accidents evaluated for such existing facilities would be comparable to those for new or modified facilities.
E.1 General Accident Information
An "accident," as discussed in this appendix, is an unplanned release of radioactive or hazardous materials resulting from "initiating" events and the additional failures resulting from the initiating event. In this case, an accident is an inadvertent release of radioactive or hazardous materials from their containers or confinement to the environment.1 Initiating events are typically defined in three broad categories:
- External initiators originate outside the facility and potentially affect the ability of the facility to maintain confinement of its materials. Examples of external initiators include aircraft crashes, nearby explosions, and hazardous material releases from nearby facilities that could affect the ability of personnel to manage the facility and its materials properly.
- Internal initiators originate within a facility and are usually the result of facility operation. Examples of internal initiators include equipment failures and human errors.
- Natural phenomena initiators are natural occurrences such as weather-related (e.g., floods and tornadoes) and seismic events (i.e., earthquakes).
The likelihood of an accident occurring and its consequences usually depend on the type of initiator(s) causing the accident, the frequency at which that initiator occurs, and the frequency of conditions that will lead to a release caused by the initiating event. Accidents can be grouped into four categories -- anticipated accidents, unlikely accidents, extremely unlikely accidents, and not reasonably foreseeable accidents -- based on their estimated frequency or likelihood of occurrence. Table E-1 lists these accident categories and their corresponding frequency ranges.
Table E-1. Accident frequency categories.a
| Frequency category | Frequency range (incidents per year) | Description |
| 1. Anticipated accidents | Less than once in 10 years but greater than once in 100 years | Accidents that might occur several times during the lifetime of the facility. |
| 2. Unlikely accidents | Less than once in 100 years but greater than once in 10,000 years | Accidents that are not likely to occur during the lifetime of the facility; natural phenomena of this probability class include Uniform Building Code-level earthquake, 100-year flood, maximum wind gust, etc. |
| 3. Extremely unlikely accidents | Less than once in 10,000 years but greater than once in 1,000,000 years | Accidents that probably will not occur during the life cycle of the facility; this includes a severe tornado, airplane crash, etc. |
| 4. Not reasonably foreseeable accidents | Less than once in 1,000,000 years | All other accidents (e.g., a direct meteorite strike) |
a Source: DOE (1994a).
This EIS evaluation examined a full spectrum of accidents; the tables in this appendix reflect the bounding (risk or consequence) event for the frequency ranges listed in Table E-1, in which risk is defined as the product of the frequency (events per year) and the consequence of an event.
The bounding consequence events would result in the largest projected increases in latent cancer fatalities, were these accidents to occur. The bounding risk events would represent the highest individual likelihood of contracting a fatal cancer, or the highest incremental cancer fatality rate in an exposed population, expressed in units of latent cancer fatalities per year. The tables in Section E.3 present the highest point estimate of risk to the maximally exposed offsite individual for each phase in bold type. This bolded number, when compared to the 1990 United States annual average risk of dying of cancer of about 0.002 (DOC 1992), provides a perspective on whether the event would be likely to increase an individual's lifetime cancer risk due to an accident dose received in that year.
E.2 Accident Analysis Method
The accidents analyzed in this EIS would result from events that are considered "reasonably foreseeable" (expected to occur at least once in 1,000,000 years). The frequencies listed in the tables in this appendix are usually associated with the initial event that leads to a release of radioactive material. In most cases, this is a conservative frequency (i.e., it overestimates the risk) because in reality a chain of events, each with its own frequency, must occur; this includes the unlikely and highly unfavorable meteorological conditions assumed to prevail at the time of the accident. In addition, the analysis might have used conservative release assumptions to calculate potential consequences (doses) that could result from such accidents. These consequences are conservative because the release of radioactivity from the facility associated with the initiating event (e.g., earthquake) could occur only after the failure of a number of safety systems.
For example, a release of radioactive material from a chemical separations facility (e.g., F-Canyon) could occur in the following manner: An earthquake occurs during a tank-to-tank transfer of radioactive solution in the canyon. The transfer pipe fails or ruptures but the transfer continues and half the contents of the tank spill to the floor of the canyon. Simultaneous with the pipe rupture, the walls of the canyon crack, providing a release pathway to the environment. In addition, the canyon ventilation system fails. (The ventilation system normally maintains the interior of the canyon at a lower pressure than the outside environment. In this way, air leaks are normally into rather than out of the canyon.) After the radioactive material spills, a fraction becomes airborne and passes through the cracks in the canyon walls. This airborne radioactivity is blown off the Site.
This scenario is conservative because tank-to-tank transfers do not occur on a continuous basis, and the earthquake would have to occur during a transfer. DOE assumes that the following failures would allow the release to reach the offsite population at the projected dose levels: (1) the transfer pipe fails, (2) operators fail to respond or are unable to stop the transfer, (3) the canyon walls crack sufficiently to allow the escape of 10 percent of the airborne radioactive material, and (4) power distribution and electrical relays associated with the ventilation system fail. In addition, all released material escapes the facility in the first 2 hours and the meteorological conditions are such that only limited dispersion (or dilution) of the material has occurred by the time it reaches the SRS boundary. Figure E-1 is a sample event tree that shows the effects of this hypothetical earthquake.
The analytical method described in the following sections did not include emergency response actions to accident situations (e.g., evacuation of personnel to a safe distance or notification of the public to perform such response actions as taking shelter) in its determination of potential impacts on workers or members of the public. To minimize potential human exposures and impacts on the environment from postulated accidents, the SRS has established an Emergency Plan (WSRC 1994a) that governs responses to potential accidents.
Figure E-1. Example of a fault tree.
The presentation of data in this appendix uses an alternative scientific notation that facilitates comparisons of the results in tables that sometimes cover several pages. This notation is explained below:
7.1E-01 = 7.1×10-1
or = 0.71
2.4E+3 = 2.4×10+3
or = 2,400
The use of this notation shows the relative magnitude of any data entry. The absence of an "E" notation indicates an actual number without the need for a multiple of 10.
To approximate the potential accident impact contribution for each material (or group of materials) of interest, DOE created a flow diagram showing the location, condition, and chemical or physical form of the material. If a safety analysis report provided different frequencies for an event depending on location in a facility (e.g., a fire is more likely in a glovebox than in a dissolver tank), the analysis used the appropriate frequency for each location housing the material. In some cases, the current forms of the materials differ greatly, although two material groups both might contain primarily plutonium-239. At some point in the processing of both materials (under the preferred alternative), the original form would be lost and the newly generated form would be virtually identical. An example would be Mark-31 plutonium targets and H-Canyon plutonium solutions. After dissolution and processing, the Mark-31 targets would have formed "newly generated" plutonium solutions. These solutions would not pose the same level of concern as those in F-Canyon, which have been stored for several years longer than planned.
In addition to customizing the frequency by location, the analysis customized the source term and composition of the material to the extent possible. For example, if a solution has been processed through a canyon, concentrated, and purified by removing fission products, the source term was adjusted to the maximum concentration with fission product contributions subtracted. The effect of this type of customization is evident in the tables that list the impacts from earthquakes. The frequency remains constant, but both the quantity (in terms of curies) and the isotopic composition (e.g., more americium-241 than plutonium-239) vary by material. These variations enable discrimination of the impacts from one material to another. This discrimination can determine the potential risk reduction if the material of interest is stabilized.
If it was not possible to customize the frequency or source term for a material, the results from the safety analysis report were used. These results represent the bounding accident analysis and are useful for predictions of the impacts from a common mode failure (i.e., a severe earthquake). Table E-5 lists F-Canyon bounding severe earthquake impacts under the heading "F-Canyon (full operation)." Table E-6 lists H-Canyon bounding severe earthquake impacts under the heading "H-Canyon (limiting solution source term)."] Because severe earthquake impacts from the canyons would far exceed those from other facilities involved in the interim management of nuclear materials, a total impact due to a severe earthquake could be approximated by adding the individual impacts from F- and H-Canyons. This cumulative impact is conservative because it is unlikely that both canyons would experience the maximum effects from a severe earthquake.
E.2.1 AFFECTED FACILITIES
Appendix C discusses the facilities used for nuclear material management activities within the scope of this EIS. In addition to the primary areas that house nuclear material, other SRS facilities contain nuclear materials (e.g., the TNX facility has two tanks of depleted uranyl nitrate solution and N-Area has drums of depleted uranium oxide). DOE has evaluated these facilities for their potential hazards and has determined that safety analysis reports were not required due to the low hazards posed by the facilities. This means that a total release of materials without mitigation would result in consequences below the threshold requiring detailed analysis. As a result, the extent of quantitative impact data is limited. In most cases, the impacts will be compared to known impacts that bound those from secondary facilities. To determine the types of accident scenarios this appendix would present, DOE performed an extensive review of existing safety documentation for facilities that either perform or support activities that could be involved with management of nuclear material.
E.2.2 RADIOLOGICAL ACCIDENT ANALYSIS METHOD
DOE used computer models to determine the consequences resulting from the release of radioactivity. This evaluation assumed the release of 1 curie of pertinent isotopes to a surface stream (for liquids) or to the atmosphere at ground level and at an elevated level, such as through an exhaust stack for the various facilities involved in the alternatives discussed in this EIS. Using the computer models, the evaluation calculated doses to an uninvolved worker, the maximally exposed offsite individual, and the offsite population within 80 kilometers (50 miles) of the Site (Simpkins 1994a,b).
DOE used two SRS-specific computer codes -- AXAIR89Q and LADTAP XL -- to calculate the doses from each of the 1-curie releases postulated. Both codes perform accident analyses described in facility safety analysis reports and postulated accident impacts presented in other EISs developed for the SRS.
The AXAIR89Q computer code (WSRC 1994b), which was developed in accordance with guidelines established by the U.S. Nuclear Regulatory Commission for modeling atmospheric releases, models the doses from airborne constituents of postulated accidental releases of radionuclides to the environment. The modeling of the various accidents postulated for the facilities associated with the different alternatives assumed conservative (99.5 percentile) meteorological conditions (e.g., direction and speed of prevailing wind). "Conservative meteorological conditions" are those for which, for a given release, the concentration of radionuclides (and the resulting doses) at a fixed downwind location will not be exceeded 99.5 percent of the time. Usually, this means a highly stable-low wind speed weather condition where the wind provides only limited dilution of the material released. Use of these meteorological conditions results in consequences approximately three to four times higher for onsite workers and between 10 and 100 times higher for the offsite population than those that would occur during average (50 percentile) meteorological conditions.
The LADTAP XL computer code was developed to model aqueous (i.e., liquid) releases of radionuclides during routine operations and potential accidents. The modeling of the aqueous releases associated with the postulated accidents described in this appendix took no credit for the holdup of radionuclides within the soils surrounding the area where the accidents would occur. In other words, the modeling assumed that the entire release would discharge directly as a liquid to the ground, migrate to the Savannah River either directly or through Fourmile Branch, Pen Branch, etc., and enter the drinking water supply.
DOE calculated most of the impacts (e.g., exposure, expressed as millirem or projected cancer incidence) to individuals from postulated accidental releases of radionuclides to the environment for the various facilities by multiplying the quantity of each isotope in the source term release (in curies per isotope) presented in the safety analysis documents by the doses calculated for a 1-curie release, as discussed in the previous paragraphs. For example, if a facility safety analysis report stated that 0.00044 curie of strontium-90 was released at ground level in the F-Area, and the projected dose to the maximally exposed offsite individual from a 1-curie release of strontium-90 at ground level in the F-Area is 0.1 millirem, then the dose to the maximally exposed offsite individual from the release would be determined by multiplying 0.00044 curie by 0.1 millirem per curie, resulting in a dose of 0.000044 millirem. The total projected dose would then equal the sum of the doses received from each radionuclide (isotope) released during the accident. This approach was not used for impacts already presented in other NEPA documents (e.g., high-level waste tank accidents or plutonium-238 accidents); in such cases, the impacts were obtained from those documents. Section E.3 presents the doses to uninvolved workers, maximally exposed offsite individuals, and the offsite population postulated for the facility radiological accidents evaluated in this appendix.
Each table in Section E.3 reflects the projected consequences in terms of dose (millirem or person-rem), point estimate of risk (dose × frequency, in units of millirem per year or person-rem per year), and latent cancer fatalities based on projections using guidelines developed by the International Commission of Radiation Protection (see Chapter 4). These guidelines, which are based on several decades of statistical analyses, provide a projection of an individual's chance of developing a cancer that proves to be fatal over time or a projection of the number of fatal cancers that would be likely to result from a population of individuals receiving a collective dose. These numbers enable comparisons of the highest consequence accidents among alternatives and among the phases of an alternative. The projections do not reflect the actual risk to an individual or population because the analysis does not consider the frequency of the accident (likelihood of occurrence). The risk of developing cancer resulting from SRS activities to manage nuclear material would be very low because accidents with large consequences from radioactive materials have not occurred historically and are unlikely to occur in the future. Each table also contains a column listing the total number of released curies estimated for each accident. The variations in dose estimates from similar release amounts is due to the varying impacts of different radioactive isotopes (e.g., 1 curie of plutonium-239 has almost five times the impact of 1 curie of plutonium-241).
As discussed above, this appendix describes risks to uninvolved workers and members of the public from radiological accidents involving nuclear materials in a quantitative fashion using such parameters as dose, accident frequency, and latent cancer fatalities in the population (as discussed in Section E.3). However, it presents potential impacts to involved, or "close-in" workers, from postulated accidents in a qualitative rather than a quantitative (in numerical terms) fashion, primarily because there is no adequate method for calculating meaningful consequences at or near the location where the accidental release occurs (DOE 1994a). The following example illustrates this concept.
A typical method for attempting to calculate the dose to an involved worker is to assume that the material is released in a room occupied by the worker and that the material instantly disperses throughout the room. Because the worker would be in the room when the release occurred, that individual probably would breathe some fraction of the radioactive materials for a given number of seconds before leaving the room. Typically, estimates of exposure time are based on assumptions about worker response to the incident (e.g., how long before the worker left the room, or whether the worker evacuated the room through an area of higher airborne concentrations). For example, consider the instance in which the worker drops a container with 2,000 grams (4.4 pounds) of plutonium oxide powder. Depending on the size of the room where the release occurred, the assumptions made on how much of the released powder became airborne and respirable, and the length of time the exposed worker remained in the room, the calculated dose to the worker could be anywhere between 80 and 78,000 rem (DOE 1994a). The uncertainty of this estimate is large, and no additional insight into the activity is available because the occurrence is accepted as undesirable without needing to perform the calculations. Historic evidence (DOE 1994a) indicates that this would be a nonfatal accident resulting in room contamination with the potential for minor personnel contamination and assimilation. Presenting this wide range of worker dose is not helpful in comparisons of impacts among alternatives. Section E.3.2 discusses potential radiological impacts to facility workers from accidents in a facility.
E.2.3 HAZARDOUS MATERIAL ACCIDENT ANALYSIS METHODOLOGY
A full understanding of the hazards associated with SRS nuclear facilities under the alternatives considered in this EIS requires analyses of potential accidents involving both hazardous and radiological materials. For chemically toxic materials, several government agencies recommend quantifying the health effects that cause short-term consequences as threshold values of concentrations in air. Because the long-term health consequences of human exposure to hazardous materials are not as well understood as those related to radiation exposure, a determination of potential health effects from exposures to hazardous materials is more subjective than a determination of health effects from exposure to radiation. Therefore, the consequences from accidents involving hazardous materials postulated in this appendix are in terms of airborne concentrations at various distances from the accident location, rather than dose or latent cancer fatalities. Because hazardous materials are used during the operations of each facility, the actual quantity associated with a particular alternative for the materials discussed in this EIS cannot be determined. For example, if a chemical is used to prevent microbiological growth in service water for a facility, then that chemical's tank or vessel must be assumed to be present for the duration of any facility function. Some or all of the hazardous substances could be eliminated if the mission of the facility were completed. None of the primary facilities involved in the storage or management of nuclear material is likely to complete its total mission within the period covered by this EIS.
To determine potential health effects to workers and members of the public that could result from accidents involving hazardous materials, DOE determined the airborne concentrations of such materials released during an accident where the uninvolved worker and offsite individual would be [i.e., 640 meters (2,100 feet) and the nearest SRS boundary, respectively] and compared them to the Emergency Response Planning Guideline (ERPG) values (AIHA 1991). The American Industrial Hygiene Association established these values, which depend on the material or chemical being considered, for three general severity levels to ensure that the necessary emergency actions occur to minimize worker and public exposures after accidents. These severity levels include the following:
- ERPG-1 Values. Exposure to airborne concentrations greater than ERPG-1 values for a period greater than 1 hour results in an unacceptable likelihood that a person would experience mild transient adverse health effects or perception of a clearly defined objectionable odor.
- ERPG-2 Values. Exposure to airborne concentrations greater than ERPG-2 values for a period greater than 1 hour results in an unacceptable likelihood that a person would experience or develop irreversible or other serious health effects or symptoms that could impair one's ability to take protective action.
- ERPG-3 Values. Exposure to airborne concentrations greater than ERPG-3 values for a period greater than 1 hour results in an unacceptable likelihood that a person would experience or develop life-threatening health effects.
Because all hazardous materials do not have ERPG values, DOE could not use such values to estimate potential impacts on the public from each hazardous material accident postulated for the SRS facilities discussed in this appendix. For chemicals that do not have ERPG values, this assessment compared airborne concentrations of hazardous materials resulting from postulated accidents to the most restrictive available exposure limits established by other guidelines (WSRC 1992) to control worker exposures to hazardous materials. Table E-2 lists the hierarchy of exposure limits that DOE used to evaluate potential health effects resulting from postulated hazardous material accidents.
DOE used a bounding approach to determine the potential impacts on individuals at different positions (e.g., uninvolved workers and the maximally exposed offsite individual) from postulated accidents in each facility area from Extremely Hazardous Substances; the amounts of such substances and their locations were determined from the SRS Tier Two Emergency and Hazardous Chemical Inventory Report (WSRC 1994c). This annual report identifies the chemicals at the Site that are
Table E-2. Hierarchy of established limits and guidelines used to determine impacts from postulated hazardous material accidents.
| Primary airborne concentration guideline | Hierarchy of alternative guidelines (if primary guidelines are unavailable) | Reference of alternative guideline |
| ERPG-3 |
EEGLa (30-minute exposure) IDLHb |
NAS (1985) NIOSH (1990) |
| ERPG-2 |
EEGL (60-minute exposure) LOCc PEL-Cd TLV-Ce TLV-TWAf multiplied by 5 |
NAS (1985) EPA (1987) 29 CFR Part 1910.1000, Subpart Z ACGIH (1992) ACGIH (1992 |
| ERPG-1 |
TWA-STELg TLV-STELh TLV-TWA multiplied by 3 |
29 CFR Part 1910.100, Subpart Z ACGIH (1992) ACGIH (1992) |
a Emergency Exposure Guidance Level (EEGL): "A
concentration of a substance in air (as a gas, vapor, or aerosol) that may be
judged by the Department of Defense to be acceptable for the performance of
specific tasks during emergency conditions lasting for a period of 1 to 24
hours. Exposure at an EEGL might produce reversible effects that do not impair
judgment and do not interfere with proper responses to an emergency." The
EEGL is "...a ceiling guidance level for a single emergency exposure,
usually lasting from 1 to 24 hours -- an occurrence expected to be infrequent in
the lifetime of a person."
b Immediately Dangerous to Life
and Health (IDLH): "The maximum concentration from which, in the event of
respirator failure, one could escape within 30 minutes without a respirator and
without experiencing any escape-impairing (e.g., severe eye irritation) or
irreversible health effects."
c Level of Concern (LOC): "The
concentration of an extremely hazardous substance in air above which there may
be serious irreversible health effects or death as a result of a single exposure
for a relatively short period of time."
d Permissible
Exposure Limit - Ceiling (C): "The employee's exposure which shall not be
exceeded during any part of the work day."
e Threshold
Limit Value - Ceiling (TLV-C): "The concentration that should not be
exceeded during any part of the working exposure."
f
Threshold Limit Value - Time Weighted Average (TLV-TWA): "The
time-weighted average concentration for a normal 8-hour workday and a 40-hour
workweek, to which nearly all workers may be repeatedly exposed, day after day,
without adverse effect."
g Time Weighted Average -
Short-Term Exposure Limit (TWA-STEL): "The employee's 15-minute time
weighted average exposure which shall not be exceeded at any time during a work
day unless another time limit is specified...."
h
Threshold Limit Value - Short-Term Exposure Limit (TLV-STEL): "The
concentration to which workers can be exposed continuously for a short period of
time without suffering from (1) irritation, (2) chronic or irreversible tissue
damage, or (3) narcosis of sufficient degree to increase the likelihood of
accidental injury, impair self-rescue, or materially reduce work efficiency, and
provided that the daily TLV-TWA is not exceeded."
hazardous or that require the establishment of emergency response procedures. Following identification of the amounts and locations of the Extremely Hazardous Substances (see Section E.4) in each area, DOE calculated the airborne concentrations at 640 meters (2,100 feet) from the point of release and the nearest SRS boundary (i.e., locations of the uninvolved worker and maximally exposed offsite individual, respectively) that would be likely from a release of the maximum inventory of each Extremely Hazardous Substance in a single location. EPICode™ (Emergency Prediction and Information Code), a commercially available computer code for modeling routine or accidental releases of hazardous chemicals to the environment (Homann 1988), calculated the airborne concentrations at the different locations.
E.3 Postulated Accidents Involving Radioactive Materials
E.3.1 Impacts to Uninvolved Workers and Members of the Public
This EIS presents the consequences and risks of bounding accidents. In this EIS, the term "bounding accident" represents postulated events or accidents that have higher consequences or risks (i.e., consequences × frequencies) than other accidents postulated in the same frequency range. A consideration of the risks associated with bounding events or accidents for a facility can establish an understanding of the overall risk to workers, members of the public, and the environment from nuclear material management activities. In addition, the risks of different alternatives can be compared relatively by comparing the risks associated with the bounding accidents for the phases of each alternative. Figure E-2 shows the concept of bounding risk accidents. The accident impact tables in this section list the bounding events for each pertinent frequency range. These tables list in bold type the highest overall point estimate of risk for the maximally exposed offsite individual and the highest consequence to the population for each phase. Some tables also list a representative selection from the full spectrum of accidents to aid in comparisons among alternatives or to demonstrate the elimination of some accidents for specific materials.
Table E-3 is a summary matrix of the facilities used for each phase of the alternatives considered for each material category. The No-Action alternative column lists the facility where the material is currently stored; this alternative has no phases. The "conversion" phase refers to any initial treatment; it is not limited to processing in a canyon. Not all alternatives have all phases (e.g., the additional conversion phase could be beyond the timeframe of this EIS).
Table E-3 is intended for use in conjunction with Tables E-4 through E-12, which list accident analysis data for each material and the facilities that could be involved in a specific phase for the corresponding material. Table E-3 can be used to determine the facility accidents analyzed that would be applicable to a specific phase. However, because the canyons and their support facilities are similar, conversion activities could occur in either area. As stated above, the tables list in bold type the maximum point estimate of risk for the maximally exposed offsite individual and the highest consequence to the population for each phase. Because an alternative might not involve every facility listed in each phase, these maximum values would not necessarily apply to all alternatives. For example, the highest point estimate of risk for the conversion phase of the H-Canyon uranium solutions (0.0000036 latent fatal cancers per year) would occur for H-Canyon. However, the Low Enriched Uranium Alternative for this material would use FA-Line for the processing phase; therefore, the maximum point estimate of risk for this alternative during processing would be 0.00000000018 latent fatal cancer per year. As noted above, the accident consequences have been tailored to the extent possible to reflect consequences attributable to the specific material.
Table E-2
a. MEI = maximally exposed individual.
b. These data were not available.
Section E.8 includes a glossary of accident descriptions. These descriptions describe the events listed in the tables. The tables use titles that indicate the facility mode as used throughout the tables [e.g., "F-Canyon (without dissolver)]." This entry means the action of dissolving would not be part of the management alternative for this material; the safety analysis report data for this mode or condition has not been used.
E.3.2 Impacts to Facility Workers from Postulated Facility Accidents
E.3.2.1 F-Canyon and H-Canyon
No fatalities to involved or "close-in" workers from the accident scenarios postulated under current or full operations in the F- or H-Canyon are a likely result of exposure to radiation. Releases from most accidents would be contained in the processing area and filtered through the canyon ventilation system. Because the ventilation system flows from areas of lowest to highest radioactivity, and because releases flow through an exhaust stack after passing through a filtration system, the doses received by workers from these accidents are not likely to be substantially larger than those received during routine operations. For postulated accidents in which the release was not likely to be maintained within the ventilation system (i.e., airborne releases from the ground level or liquid releases), involved worker exposures would be unlikely to result in adverse health effects. For an inadvertent nuclear criticality in the processing vessels, the doses to involved workers would likely be minimized due to the shielding between the vessels and the locations a worker could occupy.
E.3.2.2 FB-Line Facility
With the exception of an inadvertent nuclear criticality during processing, no fatalities to involved workers from the accident scenarios postulated under current or full operations in the FB-Line would be likely as a result of exposure to radiation (see Section E.7). Current operations primarily involve storage activities in the FB-Line vaults. Because access to storage areas in the FB-Line is limited, only a small number of individuals could receive impacts from an accidental release of material or an inadvertent nuclear criticality in a storage vault. Under full operations, potential accidents resulting from processing, such as a fire or uncontrolled chemical reaction, would not result in substantial exposures because most work would occur inside gloveboxes. Based on historic accident information, exposures to involved workers would be within limits established for routine operations if the implementation of emergency response actions occurred. Of the approximately 74 persons who could be in the FB-Line facility during processing activities, about 56 would be in areas where they could receive substantial doses from a criticality. Of the 56, an estimated 4 workers could receive lethal doses of radiation, while the other individuals would receive varying nonlethal levels.
E.3.2.3 FA-Line
For accidents postulated for FA-Line, with the exception of a red-oil explosion or a severe earthquake, no substantial injuries to involved workers are likely. The force of the explosion or flying debris initiated by the red-oil explosion could result in physical injuries to involved workers. Although the likelihood for an involved worker fatality due to radiation exposure alone after a severe earthquake is minimal, the earthquake itself could result in significant injuries or death for involved workers.
E.3.2.4 235-F Storage Vaults
With the exception of an inadvertent nuclear criticality in the storage vaults, no fatalities to involved workers from the accident scenarios postulated for the 235-F facility are likely as a result of exposure to radiation. Section E.7 discusses the criticality safety program. Because the number of persons permitted in the 235-F storage vaults is limited, the number of individuals who could be impacted from an inadvertent nuclear criticality would be limited. No more than two involved workers would be likely to receive lethal doses of radiation, with a limited number of additional individuals receiving exposures significantly above the annual administrative limit established for routine operations. For other postulated accident scenarios for the 235-F facility, exposures to involved workers are likely to be within limits established for routine operations, even if the inventories of materials within the vaults increased as a result of stabilization of materials at other SRS facilities.
E.3.2.5 HB-Line Facility
Fatalities to involved or close-in workers from the accident scenarios postulated for full operation of the HB-Line facility are not a likely result of exposure to radiation. For many of the accidents, releases would be contained in the gloveboxes and filtered through the process system and canyon ventilation systems. Because the ventilation system flows from areas of lowest to highest radioactivity, and because releases flow through an exhaust stack after passing through a filtration system, the worker doses from these accidents are not likely to be substantially larger than those received during routine operations. For postulated accidents in which the release is not likely to remain in the ventilation system, such as a ground-level airborne release initiated by a severe earthquake, involved worker exposures would be unlikely to result in adverse health effects. An inadvertent nuclear criticality is not considered credible in the HB-Line, either during current or full operations, due to the forms and isotopes of the materials. Therefore, exposures or fatalities are not likely from inadvertent nuclear criticalities.
E.3.2.6 Uranium Solidification Facility
With the exception of an inadvertent nuclear criticality during processing, no fatalities to involved workers from the accident scenarios postulated for the Uranium Solidification Facility are likely as a result of exposure of radiation. Section E.7 discusses the criticality safety program. If an inadvertent nuclear criticality occurred, either during processing (criticality in a liquid) or packaging and storage (criticality in a powder), the radiation field generated by the criticality could lead to involved worker fatalities. However, DOE expects that the number of fatalities would be limited to two; additional individuals in the facility could receive doses that significantly exceeded their annual administrative exposure limits.
E.3.2.7 H-Area Receiving Basin for Offsite Fuels
No fatalities are likely to involved workers from the radiological accident scenarios postulated for the Receiving Basin for Offsite Fuels. Worker doses for all postulated basin accidents would be minimal.
E.3.2.8 Reactor Disassembly Basins
No fatalities are likely to involved workers from the radiological accident scenarios postulated for the reactor disassembly basins. Worker doses for all postulated basin accidents would be minimal. This conclusion is based on the fact that the fuels and targets stored in each basin are maintained at a distance below the surface level of the water sufficient to minimize involved worker exposures. In addition, in events that involved a substantial loss of basin water after which fuels and targets could be exposed to the air (e.g., draindown of half the basin water or discharge of all basin water following a severe earthquake), sufficient time would be available to allow involved workers to take the precautions necessary to evacuate the area or implement other actions to minimize exposures.
E.3.2.9 Other Facilities
In addition to the facilities discussed above, M-Area buildings, the Savannah River Technology Center, the TNX facility, and the high-level waste tanks contain nuclear materials addressed by this EIS.
No fatalities to involved workers from the accident scenarios postulated for M-Area are likely as a result of exposure to radiation. DOE anticipates that involved worker doses received from accidents would be minimal because the area serves as a storage vault for stable materials and involves only routine monitoring and maintenance activities.
No fatalities to involved workers from the accident scenarios postulated for the Savannah River Technology Center are likely as a result of exposure to radiation from accidents involving these materials, and DOE anticipates that involved worker doses received from accidents would be minimal. This conclusion is based on the very small amount of irradiated, aluminum-clad fuel assembly pieces, which would be a candidate for further stabilization in other facilities. The only alternative proposed for this material in the Savannah River Technology Center is No Action.
DOE anticipates no radiation-induced fatalities would result from accidents in the TNX facility or the waste tanks. The tanks in both areas store liquid radioactive materials and involve routine monitoring or remote transfers. The high-level waste tanks are in F- and H-Areas.
E.3.3 STABLE MATERIALS
Although this EIS considers no alternatives other than Continued Storage (No Action) for stable materials, this section summarizes the accident analyses presented in the safety analysis reports for the facilities housing these materials. These documents discuss accident impacts for an uninvolved worker and the maximally exposed individual off the Site.
E.3.3.1 Postulated Radiological Accidents for the M-Area Reactor Materials Facilities
The primary purpose of the M-Area facilities was to manufacture fuel and target assemblies. The enriched uranium storage vault is constructed of reinforced concrete with walls and roof 30 centimeters (12 inches) thick. The four walls extend 1.8 meters (6 feet) into the ground and rest on 0.6-meter (2-foot)-thick footings. The storage vault was constructed to be a "maximum resistance" area [able to withstand a Fujita Intensity Five (F-5) tornado or a Modified Mercalli VIII (MM VIII) earthquake with little or no damage]. The SRS document explaining the limited continued operations in M-Area contains accident analyses for the facilities containing the nuclear materials addressed by this EIS. The bounding event for impact on the maximally exposed individual is an explosion in Building 320-M, which would result in a risk of 0.00014 rem per year and a latent cancer fatality projection of 0.00000007. For the uninvolved worker for the same event, the estimated risk would be 0.00044 rem per year and the latent cancer fatality projection would be 0.00000018. This accident is representative of bounding events related to the storage of a variety of materials for which further stabilization is not required. This group contains all material in the Reactor Material Area, including miscellaneous depleted uranium and uranium metal, oxide, slugs, cores, sludges, enriched uranium residues, lithium aluminum control rods, spargers, targets, unirradiated Mark-22s with lithium target tubes, natural and enriched lithium metal in cans, Mark-16 and Mark-22 tubes, Mark-31 slugs, and neptunium targets. Stable material is stored in Buildings 313-M, 315-M, 320-M, 321-M, 322-M, and 341-1M. Unirradiated Mark-31 slugs (depleted uranium in aluminum housings) constitute most of the inventory. The No-Action Alternative is proposed for the materials currently stored in M-Area.
E.3.3.2 Postulated Radiological Accidents for Savannah River Technology Center
Nuclear material used or stored in the Savannah River Technology Center includes a small amount of americium and curium solution and targets; americium-241 scrap; depleted uranium slurry, metal, and oxide; enriched uranium sweepings; etc.
Under the No-Action Alternative, current research activities at the Savannah River Technology Center would continue, and DOE would continue to store equivalent amounts and types of material in Building 773-A laboratories and cells. These materials are generally stored in limited-quantity cans, bottles, or sample carriers. Most are contained further in laboratory hoods, gloveboxes, or cells. These items, or equivalent new sample quantities, would be in a safe stable form for storage for several years.
The Savannah River Technology Center Safety Analysis Report summarizes consequences from postulated accidents at the center involving areas that contain the materials listed above. The actual contribution to the accident scenarios from these materials would be negligible, but these events are bounding for all alternatives for stable materials (i.e., the No-Action Alternative). An earthquake with a magnitude of 0.2 times gravity poses the highest risk for the maximally exposed individual. The risk associated with this event would be 0.00023 rem per year and the latent cancer fatality projection would be 0.00000012. In the highly unlikely event that this accident occurred, it would cause a projected increase of 0.48 in latent cancer fatalities. From the same event, the uninvolved worker risk would be 0.0043 rem per year and the latent cancer fatality projection would be 0.0000017.
E.3.3.3 Postulated Radiological Accidents for the TNX Research Facility
The TNX facility is a "radiological facility," as determined by the quantity of nuclear material present (DOE 1992). This hazard classification is the lowest for a facility that contains radioactive materials and requires no safety analysis report. This assessment does not summarize accident analyses for this facility because the impacts are bounded by those for several other facilities; only the No-Action Alternative would apply.
E.4 Postulated Accidents Involving Extremely Hazardous Substances
Because of the many types of materials and chemicals at the Site and the varying quantities of these materials in different locations, the analysis of potential accident scenarios involving hazardous materials was limited to substances categorized by the U.S. Environmental Protection Agency as "Extremely Hazardous Substances" (40 CFR Part 355), as designated under the Emergency Planning and Community Right-to-Know Act of 1986. Although materials not categorized as Extremely Hazardous Substances can affect the health and safety of workers and the public if released in sufficient quantities and forms, the Site has implemented programs in accordance with DOE Order requirements (e.g., DOE 1985, 1993, 1994b) that incorporate programmatic and management requirements of other government agencies, such as the Occupational Safety and Health Administration. While these materials might present hazards to workers or the public if accidentally released to the environment, their impacts are likely to be bounded by potential impacts from accidents involving Extremely Hazardous Substances; therefore, this appendix does not analyze them.
This section presents potential impacts from postulated chemical accidents at facilities that are or could be involved with safely managing or stabilizing SRS nuclear materials. For each area, it presents potential impacts of the bounding hypothetical chemical accident scenarios (as calculated using the method described in Section E.2.4).
Substances present in bulk quantities can, in some cases, be reduced or eliminated after stabilization of the associated nuclear material. In other cases (e.g., the Receiving Basin for Offsite Fuels), the chemicals support long-term facility functions independent of the interim management of the nuclear materials covered in this EIS. The accident consequences presented in this section assume a maximum chemistry inventory and are bounding for all alternatives.
E.4.1 POSTULATED CHEMICAL ACCIDENTS FOR F-AREA FACILITIES
Based on a review of current inventories at the facilities in the F-Area (DOE 1994c), DOE determined that seven Extremely Hazardous Substances are in use in the area. Table E-13 lists the maximum amounts of each substance in a single location in the F-Area.
Table E-13. Inventories of Extremely Hazardous Substancesa in F-Area.
| Substance | Maximum amount in a single location (kilograms)b,c |
| Hydrochloric acid | 34.0 |
| Hydrogen fluoride | 1,174.8 |
| Hydrogen peroxide | 122.5 |
| Nitric acid | 65,771.6 |
| Phenol | 0.9 |
| Phosphorous pentoxide | 0.9 |
| Sulfuric acid | 3,823.8 |
a Materials categorized as Extremely Hazardous Substances (40
CFR Part 355), as designated under the Emergency Planning and Community
Right-to-Know Act of 1986.
b To determine the quantity in
pounds, multiply by 2.2046.
c Amounts are based on 1993
(1-year) values.
To determine airborne concentrations at 640 meters (2,100 feet) and the nearest SRS boundary (the locations of the uninvolved worker and maximally exposed offsite individual, respectively), DOE assumed an inadvertent release to the environment of the maximum amount of each material in a single location. This method enables a comparison of the impacts of the various substances as well as impacts at the facilities housing these substances. These impacts are conservative because the analysis does not consider the frequency of an initiating event that could lead to the release of this maximum amount.
DOE used the EPICode™ computer code (see Section E.2.6) to model the release of each material. Table E-14 lists the results of the analyses and compares expected airborne concentrations at the uninvolved worker and maximally exposed individual locations to the different threshold Emergency Response and Planning Guidelines or their equivalents.
| Substance released | Maximum amount in F-Area (kg)a | Airborne concentration (milligram per cubic meter)b | ||||
| At 640mc | At Site boundaryd | ERPG-1e | ERPG-2e | ERPG-3e | ||
| Hydrochloric acid | 3.4E+01 | 6.3E-03 | 8.5E-05 | 4.5 | 3.0E+01 | 1.5E+02 |
| Hydrogen fluoride | 1.2E+03 | 2.2E+02 | 2.9 | 4.0 | 1.6E+01 | 4.1E+01 |
| Hydrogen peroxide | 1.2E+02 | 2.3E-02 | 3.1E-04 | 1.4 | -- | 1.1E+02 |
| Nitric acid | 6.6E+04 | 1.4E+01 | 3.6 | 5.2 | 3.9E+01 | 7.7E+01 |
| Phenol | 9.1E-01 | 1.5E-04 | 1.7E-06 | 3.9E+01 | 1.9E+02 | 7.7E+02 |
| Phosphorous pentoxide | 9.1E-01 | 1.5E-04 | 1.7E-06 | 5.0 | 2.5E+01 | 1.0E+02 |
| Sulfuric acid | 3.8E+03 | 2.2E-07 | 3.7E-09 | 2.0 | 1.0E+01 | 3.0E+01 |
a To determine the quantity in pounds, multiply by 2.2046.
b Airborne concentrations derived assuming conservative (99.5
percentile) meteorological conditions for the Site.
c Location
of the uninvolved worker, assumed to be located 640 meters (2,100 feet) downwind
from the release.
d Location of the maximally exposed offsite
individual, assumed to reside at the nearest SRS boundary downwind from the
point of release at 10.6 kilometers (6.6 miles).
e Either the
Emergency Response Planning Guidelines value or most restrictive exposure
guideline available, as discussed in Section E.2.4 and listed in
Table E-2. For substances with limits established in terms
of parts per million, the value in milligrams per cubic meter was determined
using the following equation: milligrams per cubic meter = (limit in parts per
million) × (gram molecular weight of substance) / 24.45.
Because a severe seismic event has the potential to initiate the release of the same material from different locations in the F-Area, DOE analyzed a release of the maximum daily inventory. Table E-15 lists the results of these analyses. A total release of the entire inventory of a particular material from the F-Area to the environment is extremely unlikely, especially if the material is in several different locations, facilities, or buildings in the area. However, the assumption of a total release of the maximum inventories in the area provides a bounding estimate for the largest airborne concentrations DOE could expect following a severe earthquake.
As listed in Tables E-14 and E-15, the airborne concentrations for a gaseous release of hydrogen fluoride (hydrofluoric acid) would exceed the ERPG-3 threshold limits at 640 meters (2,100 feet) from the point of release. As explained in Section E.2.4, ERPG-3 threshold values represent concentrations at which an individual would experience or develop life-threatening health effects if exposed for longer than 1 hour. Because individuals could be notified and evacuated to a safe location (e.g., inside a building with adequate ventilation) within 1 hour of an inadvertent release of hydrogen fluoride, DOE does not expect any life-threatening or long-term health effects to uninvolved workers. Uninvolved workers could experience mild burning of the lungs from inhaling hydrogen fluoride, burning of the eyes, and mild skin irritations. In addition, because the airborne concentrations at the nearest SRS boundary would be below ERPG-1 threshold values, no measurable health effects are likely to members of the public. However, for involved workers, there is a potential for serious worker injury and potential fatalities because of the large concentrations expected at locations close to the point of release, which could hinder personnel from taking appropriate emergency response actions.
| Substance released | Maximum daily amount in entire F-Area (kg)a | Airborne concentration (milligram per cubic meter)b | ||||
| At 640mc | At Site boundaryd | ERPG-1e | ERPG-2e | ERPG-3e | ||
| Hydrochloric acid | 1.0E+02 | 1.9E-02 | 2.6E-04 | 4.5 | 3.0E+01 | 1.5E+02 |
| Hydrogen fluoride | 1.2E+03 | 2.2E+02 | 2.9 | 4.0 | 1.6E+01 | 4.1E+01 |
| Hydrogen peroxide | 1.2E+02 | 2.3E-02 | 3.1E-04 | 1.4 | -- | 1.1E+02 |
| Nitric acid | 2.7E+05 | 3.9E+02 | 1.4E+01 | 5.2 | 3.9E+01 | 7.7E+01 |
| Phenol | 1.4 | 2.3E-04 | 2.6E-06 | 3.9E+01 | 1.9E+02 | 7.7E+02 |
| Phosphorous pentoxide | 9.1E-01 | 1.5E-04 | 1.7E-06 | 5.0 | 2.5E+01 | 1.0E+02 |
| Sulfuric acid | 4.0E+03 | 2.3E-07 | 4.0E-09 | 2.0 | 1.0E+01 | 3.0E+01 |
a To determine the quantity in pounds, multiply by 2.2046.
b Airborne concentrations derived assuming conservative (99.5
percentile) meteorological conditions for the Site.
c Location
of the uninvolved worker, assumed to be located 640 meters (2,100 feet) downwind
from the release.
d Location of the maximally exposed offsite
individual, assumed to reside at the nearest SRS boundary downwind from the
point of release at 10.6 kilometers (6.6 miles).
e Either the
Emergency Response Planning Guidelines value or most restrictive exposure
guideline available, as discussed in Section E.2.4 and listed in
Table E-2. For substances with limits established in terms
of parts per million, the value in milligrams per cubic meter was determined
using the following equation: milligrams per cubic meter = (limit in parts per
million) × (gram molecular weight of substance) / 24.45.
Table E-15 indicates that, in the event of a severe earthquake, a release of the total quantity of nitric acid in the F-Area would exceed ERPG-3 values at a distance of 640 meters (2,100 feet) and ERPG-1 values at the nearest SRS boundary. As explained in Section E.2.4, the health effects from being exposed to ERPG-1 threshold values for greater than 1 hour are minor (e.g., irritation of the eyes and objectionable odor). For uninvolved and involved workers, although the release would exceed ERPG-3 threshold values, no worker fatalities from exposure to airborne acid concentrations would be likely; some individuals could experience significant short-term health effects, such as burning of the lungs and irritation of the skin. Because this scenario assumes that all nitric acid in the F-Area would be released from a single location during a severe earthquake, airborne concentrations would be lower than those listed in Table E-15.
E.4.2 POSTULATED CHEMICAL ACCIDENTS FOR H-AREA FACILITIES
Based on a review of current inventories at the various H-Area facilities (DOE 1994b), DOE determined that seven Extremely Hazardous Substances are in use in the H-Area. Table E-16 lists the maximum amounts of each substance in a single location in the H-Area.
Table E-16. Inventories of Extremely Hazardous Substancesa in H-Area.
| Substance | Maximum amount in a single location (kilograms)b,c |
| Ammonia | 27.2 |
| Hydrochloric acid | 2.7 |
| Hydrogen fluoride | 2.3 |
| Nitric acid | 39,814.7 |
| Nitric oxide | 1,315.4 |
| Phosphorous pentoxide | 1.4 |
| Sulfuric acid | 0.9 |
a Materials categorized as Extremely Hazardous Substances (40
CFR Part 355), as designated under the Emergency Planning and Community
Right-to-Know Act of 1986.
b To determine the quantity in
pounds, multiply by 2.2046.
c Amounts are based on 1993
(1-year) values.
Table E-17 lists the results of the analyses and compares the expected airborne concentrations at the uninvolved worker and maximally exposed individual locations to the different threshold Emergency Response and Planning Guidelines or their equivalents. Because a severe seismic event has the potential to initiate the release of the same material from different locations within the H-Area, DOE analyzed a release of the maximum daily inventory to the environment. Table E-18 lists the results of these analyses.
As listed in Tables E-17 and E-18, the airborne concentrations for a gaseous release of nitric oxide would exceed the ERPG-3 threshold limits at a distance of 640 meters (2,100 feet) from the point of release. Table E-18 indicates that, in a severe earthquake, a release of the total quantities of nitric acid in the H-Area would exceed ERPG-3 values at a distance of 640 meters (2,100 feet) and ERPG-1 values at the nearest SRS boundary. For uninvolved and involved workers, although the release would exceed ERPG-3 threshold values, no worker fatalities from exposure to the airborne acid concentrations would be likely; some individuals could experience significant short-term health effects, such as burning of the lungs and irritation of the skin. Because this scenario assumes that all nitric acid in the H-Area would be released from a single location during a severe earthquake, airborne concentrations would be lower than those listed in Table E-18.
| Substance released | Maximum amount in a single H-Area location (kg)a | Airborne concentration (milligram per cubic meter)b | ||||
| At 640mc | At Site boundaryd | ERPG-1e | ERPG-2e | ERPG-3e | ||
| Ammonia | 2.7 | 5.1E-03 | 5.8E-05 | 2.5E+01 | 2.0E+02 | 1.0E+03 |
| Hydrochloric acid | 2.7 | 5.0E-04 | 5.7E-06 | 4.5 | 3.0E+01 | 1.5E+02 |
| Hydrogen fluoride | 2.3 | 4.3E-04 | 4.9E-06 | 4.0 | 1.6E+01 | 4.1E+01 |
| Nitric acid | 4.0E+04 | 9.5E+01 | 1.9 | 5.2 | 3.9E+01 | 7.7E+01 |
| Nitric Oxide | 1.3E+03 | 4.9E+03 | 4.4 | 9.3E+01 | 1.2E+02f | 1.2E+02 |
| Phosphorous pentoxide | 1.4 | 1.2E-01 | 1.1E-03 | 5.0 | 2.5E+01 | 1.0E+02 |
| Sulfuric acid | 9.0E-01 | 1.7E-04 | 1.9E-06 | 2.0 | 1.0E+01 | 3.0E+01 |
a To determine the quantity in pounds, multiply by 2.2046.
b Airborne concentrations derived assuming conservative (99.5
percentile) meteorological conditions for the Site.
c Location
of the uninvolved worker, assumed to be located 640 meters (2,100 feet) downwind
from the release.
d Location of the maximally exposed offsite
individual, assumed to reside at the nearest SRS boundary downwind from the
point of release at 10.6 kilometers (6.6 miles).
e Either the
Emergency Response Planning Guidelines value or most restrictive exposure
guideline available, as discussed in Section E.2.4 and listed in
Table E-2. For substances with limits established in terms
of parts per million, the value in milligrams per cubic meter was determined
using the following equation: milligrams per cubic meter = (limit in parts per
million) × (gram molecular weight of substance) / 24.45.
f
Alternative concentration limit guideline for ERPG-2 value (TWA × 5) was
adjusted down to the next higher range value (IDLH).
| Substance released | Maximum daily amount in entire H-Area (kg)a | Airborne concentration (milligram per cubic meter)b | ||||
| At 640mc | At Site boundaryd | ERPG-1e | ERPG-2e | ERPG-3e | ||
| Ammonia | 2.7E+01 | 5.1E-03 | 5.8E-05 | 2.5E+01 | 2.0E+02 | 1.0E+03 |
| Hydrochloric acid | 1.1E+01 | 2.1E-03 | 2.4E-05 | 4.5 | 3.0E+01 | 1.5E+02 |
| Hydrogen fluoride | 3.6 | 6.7E-04 | 7.6E-06 | 4.0 | 1.6E+01 | 4.1E+01 |
| Nitric acid | 1.2E+05 | 2.3E+02 | 5.7 | 5.2 | 3.9E+01 | 7.7E+01 |
| Nitric Oxide | 1.3E+03 | 4.9E+03 | 4.4 | 9.3E+01 | 1.2E+02f | 1.2E+02 |
| Phosphorous pentoxide | 1.4 | 1.2E-01 | 1.1E-03 | 5.0 | 2.5E+01 | 1.0E+02 |
| Sulfuric acid | 2.7 | 5.0E-04 | 5.7E-06 | 2.0 | 1.0E+01 | 3.0E+01 |
a To determine the quantity in pounds, multiply by 2.2046.
b Airborne concentrations derived assuming conservative (99.5
percentile) meteorological conditions for the Site.
c Location
of the uninvolved worker, assumed to be located 640 meters (2,100 feet) downwind
from the release.
d Location of the maximally exposed offsite
individual, assumed to reside at the nearest SRS boundary downwind from the
point of release at 10.6 kilometers (6.6 miles).
e Either the
Emergency Response Planning Guidelines value or most restrictive exposure
guideline available, as discussed in Section E.2.4 and listed in
Table E-2. For substances with limits established in terms
of parts per million, the value in milligrams per cubic meter was determined
using the following equation: milligrams per cubic meter = (limit in parts per
million) × (gram molecular weight of substance) / 24.45.
f
Alternative concentration limit guideline for ERPG-2 value (TWA × 5) was
adjusted down to the next higher range value (IDLH).
E.4.3 POSTULATED CHEMICAL ACCIDENTS FOR K-, L-, AND P-REACTOR BASINS
Based on a review of the chemical inventory that supports the water chemistry in the L-Reactor basin, DOE determined that the only identified Extremely Hazardous Substance was a small quantity of nitric acid. For 45.6 kilograms (100 pounds) of nitric acid modeled as a liquid spill (the maximum daily amount in the basin), the airborne concentration at 640 meters (2,100 feet) would be several orders of magnitude lower than the ERPG-1 concentration limit. DOE assumed that this was typical for all SRS reactor basins that store nuclear material.
In addition, because the airborne concentrations at the nearest SRS boundary would be considerably below ERPG-1 threshold values, no measurable health effects to members of the public would be likely. No impacts would hinder involved workers from taking appropriate emergency response actions.
E.4.4 POSTULATED CHEMICAL ACCIDENTS FOR M-AREA FACILITIES
Based on a review of current inventories at the various facilities in the M-Area (DOE 1994b), DOE determined that five Extremely Hazardous Substances are in use in the area. Table E-19 lists the maximum amounts of each substance in a single location in the M-Area. However, M-Area contains nuclear materials that require no further stabilization. Therefore, this EIS proposes no alternatives for the safe management of nuclear materials in M-Area. As a result, no further chemical accident analysis is required.
Table E-19. Inventories of Extremely Hazardous Substancesa in M-Area.
| Substance | Maximum amount in a single location (kilograms)b,c |
| Hydrochloric acid | 34.0 |
| Hydrofluoric acid | 2.27 |
| Nitric acid | 34,807.5 |
| Phenol | 2.27 |
| Sulfuric acid | 15,241.0 |
a Materials categorized as Extremely Hazardous Substances (40
CFR Part 355), as designated under the Emergency Planning and Community
Right-to-Know Act of 1986.
b To determine the quantity in
pounds, multiply by 2.2046.
c Amounts are based on 1993
(1-year) values.
E.4.5 SAVANNAH RIVER technology center
Based on a review of current inventories at the various facilities in the Savannah River Technology Center (DOE 1994d), DOE determined that eight Extremely Hazardous Substances are in use in SRTC facilities. Table E-20 lists the total annual maximum and average daily quantities of these substances based on 1993 (1-year) inventories. In addition, Table E-20 lists the maximum amounts of each substance in a single location in the SRTC. However, the Center contains nuclear materials that require no further stabilization. Therefore, this EIS proposes no alternatives for the safe management of the nuclear materials in SRTC facilities. As a result, no further chemical accident analysis is required.
Table E-20. Inventories of Extremely Hazardous Substancesa in Savannah River Technology Center.
| Substance | Maximum amount in a single location (kilograms)b,c |
| Ammonia | 0.5 |
| Hydrochloric acid | 2,215.4 |
| Hydrogen fluoride | 38.1 |
| Nitric acid | 3,864.2 |
| Nitric oxide | 0.9 |
| Phenol | 4.5 |
| Phosphorous pentoxide | 3.18 |
| Sulfuric acid | 13.6 |
a Materials categorized as Extremely Hazardous Substances (40
CFR Part 355), as designated under the Emergency Planning and Community
Right-to-Know Act of 1986.
b To determine the quantity in
pounds, multiply by 2.2046.
c Amounts are based on 1993
(1-year) values.
E.4.6 POSTULATED CHEMICAL ACCIDENTS FOR THE TNX AREA
Based on a review of chemical usage in the TNX area, DOE determined that no chemicals in the area were required to support the continued safe management of nuclear materials. As a result, no further chemical accident analysis was performed for the TNX area.
E.5 Environmental Justice
When the 99.5 percent meteorology model is used, the SRS sector most affected by accidents is the Northwest. Although this is not typical of weather conditions (e.g., not the prevailing wind direction), the model calculated the highest impact to an individual at the SRS boundary.
Figures 3-7 and 3-8 show the distributions, by census tracts, of people of color and low-income populations, respectively. Parts of two census tracts in the Northwest sector adjoin the SRS. Neither tract is a low-income community or a community comprised of 50 percent or more of people of color, although one of the tracts contains between 35 and 50 percent people of color.
Farther from the SRS in the Northwest sector are low-income communities and communities that contain 50 percent or more of people of color. However, other communities in the sector are not low-income and contain fewer than 35 percent people of color, and they are as close as, or closer to, the SRS boundaries than the low-income communities or the communities of people of color.
Based on the distribution of types of communities and on the low dose received by the maximally exposed individual (see tables in this appendix), the accident scenarios would not result in disproportionately high or adverse human health and environmental impacts on people of color or low-income populations.
E.6 Accident Mitigation
Although DOE expends extensive efforts and large amounts of money to prevent accidents involving radioactive and hazardous materials, accidents and inadvertent releases to the environment can still occur. Therefore, an important part of the accident analysis process is the identification of actions that can mitigate consequences from accidents if they occur.2 This section summarizes the SRS Emergency Plan, which governs responses to accident situations that could affect Site employees or the offsite population.
The Savannah River Site Emergency Plan (WSRC 1994a) defines appropriate response measures for the management of SRS emergencies (e.g., radiological or hazardous material accidents). It incorporates into one document the entire process designed to respond to and mitigate the consequences of a potential accident. For example, it establishes protective action guidelines for accidents involving chemical releases to keep onsite and offsite exposures as low as possible. It accomplishes minimization or prevention of exposures by minimizing time spent in the vicinity of the hazard or the release plume, keeping personnel as far from the hazard or plume as possible (e.g., using physical barricades and evacuation), and taking advantage of available shelter.
Emergencies that could cause activation of all or portions of this plan and the SRS Emergency Response Office include the following:
- Events (operational, transportation, etc.) with the potential to cause releases above allowable limits of radiological or hazardous materials.
- Events (fires, explosions, tornadoes, hurricanes, earthquakes, dam failures, etc.) that affect or could affect safety systems designed to protect Site and offsite populations and the environment. The effectiveness of the emergency plan would depend on the severity of the event and the impact on the Site and local infrastructure.
- Events (bomb threats, hostage situations, etc.) that reduce the security posture of the Site.
- Events created by proximity to other facilities such as the Vogtle Electric Generating Plant (a commercial nuclear utility across the Savannah River from the Site) or nearby commercial chemical facilities.
Depending on the types of postulated accidents and the potential impacts that could result from those accidents, emergencies are classified in several categories in accordance with requirements defined in the DOE 5500 Series of Orders, as follows:
- Alerts are confined within the affected facility boundary; no measurable impacts to workers or members of the public outside the facility boundary are likely.
- Site Area Emergencies are events that are in progress or that have occurred involving actual or likely major failures of facility safety or safeguards systems needed for the protection of onsite personnel, the public, the environment, or national security; because they have the potential to impact workers at colocated facilities or members of the public in the SRS vicinity, these situations require notification of and coordination of responses with the appropriate local authorities.
- General Emergencies produce consequences that require the implementation of protective actions to minimize impacts to both workers and the public; full mobilization of all available onsite and offsite resources is usually required to deal with the event and its consequences.
In accordance with the Site Emergency Plan, DOE conducts periodic drills and exercises at the SRS to develop, maintain, and test response capabilities, and validate the adequacy of emergency facilities, equipment, communications, procedures, and training. For example, drills occur for the following accident scenarios in the facilities or facility areas: facility or area evacuations, shelter protection, toxic gas releases, nuclear incident monitor alarms (following an inadvertent nuclear criticality), fire alarms, medical emergencies, and personnel accountability (to ensure that all personnel have safely evacuated a facility or area following an emergency). DOE and Westinghouse Savannah River Company conduct and evaluate periodic drills with the following organizations or groups to ensure that they continue to maintain (from both a personnel and an equipment standpoint) the capability to respond adequately to emergency situations: first aid teams; rescue teams; fire wardens, fire response and firefighting teams; SRS medical and Health Protection personnel and personnel from the Eisenhower Army Medical Center; SRS and local communications personnel and systems; SRS security forces; and SRS Health Protection agencies.
E.7 Nuclear Criticality Safety Program
As discussed above, with the exception of an inadvertent nuclear criticality, no fatalities to involved workers from accident scenarios postulated for the management, stabilization, or storage of nuclear material would be likely to result from exposure to radiation. A criticality occurs when a neutron fissions the nucleus of a fissionable material to produce energy, fission fragments, neutrons, and various radiations. While nuclear reactors are specifically designed to produce energy from fission by controlling this neutron chain reaction, nonreactor nuclear facilities at the SRS do not generally provide the same control, shielding, and containment characteristics. Thus, an inadvertent fission chain reaction (nuclear criticality) in an SRS nonreactor nuclear facility could produce harmful radiation-related effects on nearby personnel.
As a result, nuclear criticality safety has been defined as "the prevention or termination of inadvertent nuclear chain reactions in nonreactor environments." In practice, the first concept--prevention--is by far the primary goal. As a consequence, SRS maintains a nuclear criticality safety program that establishes and defines the principles, practices, and controls to be used for the prevention of criticality accidents. When it has been determined that the potential for an inadvertent nuclear criticality accident exists for a facility, the design of criticality controls, including equipment and procedures, shall meet, at a minimum, the requirements described in the WSRC Nuclear Criticality Safety Manual. For a new facility, the use of physical design features to prevent criticality would be preferable. To ensure the successful implementation of this program, a training policy recently adopted at the SRS supports the goal that all reasonable efforts shall be taken to reduce or eliminate the potential for, and consequences of, a criticality accident. Nuclear criticality safety training programs at the SRS are developed to be consistent with DOE Orders 5480.20 and 5480.24 for operating facility personnel and all other personnel requiring criticality safety training.
Positive identification of fissionable material, particularly fissile material, is essential to criticality safety. Adequate labeling of fissionable material and clear posting of work and storage areas in which fissionable materials are present are important in avoiding the accumulation of unsafe quantities of such materials. Appropriate fissionable material labeling and area posting are maintained at SRS nonreactor nuclear facilities specifying material identification and all parameter limits subject to procedural control. Storage requirements include minimum spacing distances to prevent sufficient material from being in close proximity. Criticality "poisons," such as boron, are often used in storage racks or packaging for material.
Written plans and procedures govern operations at SRS in which criticality safety is a consideration. These plans and procedures cover startup, operations, and any modifications that might affect criticality safety. Procedures clearly specify all controlled parameters and limits related to criticality safety. All criticality safety-related limits contained in the operating procedures are based on Nuclear Criticality Safety Evaluations (NCSEs). New or revised procedures containing nuclear safety steps, criticality safety limits, or criticality safety requirements undergo review and approval by a Criticality Safety Engineering Group (CSEG) before implementation. In the event of a criticality limit violation, SRS procedures specifically govern actions to be taken in the event of an undesirable situation; the objective of such procedures is to place the operation into as stable and safe a condition as possible until a criticality safety engineer or specialist can conduct an evaluation.
Water, the most often used firefighting agent, is an efficient moderator and reflector of neutrons (i.e., it can contribute to a criticality). In the absence of moderating materials such as water, relatively large masses of dry fissile materials such as powders or metals can be handled safely. In the event of a fire, SRS nonreactor nuclear facilities maintain prefire plans prepared by the management and engineering staff of each facility with the assistance by the Criticality Safety Engineering Group, SRS fire safety engineers, and the Area Fire Department, as necessary. These plans help provide a framework for the successful combination of firefighting and criticality safety. The CSEG approves the prefire plans for each facility in which criticality safety is of concern.
The SRS maintains criticality alarm systems, or Nuclear Incident Monitors (NIMs). The primary purpose of NIM systems is to minimize, by means of quick detection and alarm, the acute dose received by personnel from a criticality (and potential recriticality) accident in areas where the cumulative absorbed dose in free air might exceed 12 rads. The secondary purpose of the NIM system is to notify people to stay clear of the evacuated area and to notify appropriate response teams.
Emergency procedures for criticality accidents are prepared for each SRS facility in which criticality safety controls are instituted or criticality alarm systems are installed. Such emergency plans are approved by the appropriate management and the cognizant Criticality Safety Engineering Group, and consistent with the Site Emergency Plan (WSRC 1994a).
E.8 Accident Descriptions
The larger facilities contain a variety of processes, equipment, and techniques used depending on the intended function. In determining the source terms for use in accident analysis, DOE examined the appropriate process or section of a facility for the specific material and adjusted the source term to correspond where necessary. The tables in Section E.3 list the "modes" or conditions to reflect the selection for that material. The following paragraphs explain the accident titles used in the tables in Section E.3.
Unpropagated fire – A fire that has localized impact and does not spread. It can be caused by ignition of flammable solvent, spontaneous burning of plutonium metal exposed to oxygen, or other causes. Radioactive particulates are dispersed in the immediate area of the fire and some might be released to the environment (e.g., during a filter fire). The fire lasts for a short period because the amount of combustible material is limited.
Inadvertent transfer – An unplanned transfer of a solution or liquid to an unintended location due to personnel error. The usual causes of such accidents are incorrectly installed piping connections or overflows from a vessel into a sump resulting from human errors.
Coil and tube failure – Some process vessels and tanks have internal coils for cooling or heating the stored solutions. The coils usually contain water or steam. The pressure inside the coils is normally higher than the pressure in the vessel. Should the coils leak or fail their internal pressure could be lost, resulting in radioactive solution entering the cooling water (or steam) system. If the leak is undetected, the contaminated water could be released through the system to the atmosphere without treatment.
Inadvertent criticality – These events are discussed in Section E.7.
Severe earthquake – An earthquake that would be expected every 5,000 years. The severity or magnitude is based on an assumed horizontal ground acceleration of 20 percent of the acceleration due to gravity. An earthquake of this magnitude could result in structural damage and a loss of confinement of nuclear materials.
Rupture storage container – Certain radioactive materials can cause a buildup of gases inside the container in which they are stored (e.g., metal can) if it contains organic materials (e.g., plastic bags). Other materials (e.g., plutonium metal) can oxidize and gain moisture if the container is not completely airtight. Eventually, the pressure buildup can cause the storage container to bulge or rupture. This could disperse the material in the area around the container and result in exposure of a worker performing routine surveillance.
Eructation - A thermal or chemical reaction causes material to spew from its container. This could be an energetic event resulting in localized contamination. For the materials discussed in this EIS, such events would occur inside the canyons and no workers would be directly affected.
Red-oil explosion - So named because the substance causing the explosion is a thick red liquid produced by the inadvertent addition of organics to a high nitrate solution. The event can be very energetic and can result in a sudden localized explosion. The radiological consequences would probably be confined to areas within the canyon facilities.
Tornado - A tornado exerts pressure due to high wind speed on the surfaces of a structure. The resulting damage could cause releases of stored materials within the structure or could disperse materials stored in pads.
Uncontrolled reaction - Adjustments are routinely made to solutions to produce a reaction under known controlled conditions. If an adjustment (e.g., adding acid) or a change in condition (e.g., heating the contents) produces unexpected or rapid reaction, that reaction is "uncontrolled." The energy from this type of reaction could cause radioactive solutions to overflow or erupt outside the tank in which they are stored.
Propagated fire - A fire that goes beyond the area of ignition. For the materials discussed in this EIS, a propagated fire does not self-extinguish. For example, it might spread from a glovebox into the surrounding room or other areas of the facility.
Basin overflow/draindown - An unplanned movement of water, either into the reactor basins (causing an overflow) or from the reactor basins (draindown), which results in a flow of the basin water to sumps or storm drains and into the Savannah River. Basin overflow would normally be caused by human error; basin draindown could be caused by a breach of the basin integrity due to an earthquake.
Hydrogen explosion - Hydrogen gas is generated by radiolysis when water is in a tank or can with nuclear materials. If a sufficient quantity of the atmosphere in the container is hydrogen, the gas can detonate or explode, rupturing the container and releasing nuclear material.
Energetic event - Exergetic events cause penetration of the primary confinement barrier and, if sufficiently energetic, can result in the bypass of a secondary barrier. Medium energetic events include a cabinet fire, an uncontrolled reaction, and criticality.
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