S1G
The S1G intermediate neutron flux beryllium sodium cooled reactor was built by the General Electric (GE) Company, hence the G designation. The Kesselring Site in New York was constructed in the mid-1950's to allow the full scale testing of the SIR (Submarine Intermediate Reactor) prototype plant which was being designed by the Knolls Atomic Power Laboratory for the Navy.
This was the first liquid metal (sodium) cooled reactor developed for use on a submarine. The design promised a more compact reactor with greater thermal efficiency and higher power density than the Nautilus’ PWR, while delivering superheated steam to drive the turbines. Fuel was UO2 clad in stainless steel, with beryllium as a moderator and reflector. Fuel enrichment was 90% [versus 97% in the 21st Century].
The 1st core was designed for an operating life of 900 equivalent full power hours Sodium coolant was circulated by electromagnetic (EM) pumps, with flow regulated by changing voltage. Steam generator tubes, including superheater tubes, were double-walled. Primary sodium coolant flowed inside the tubes. Secondary water / steam flowed outside the tubes. The space between the tube double walls was filled potassium-sodium alloy to transfer heat from the primary to the secondary system and provide a barrier against direct sodium-water contact.
General Electric operated the Knolls Atomic Power Laboratory for the AEC for the purpose of designing Naval power plants; the SIR project stemmed from a redirection of what had originally been a plan for a civilian liquid metal cooled breeder reactor, whose progress and funding became uncertain. According to testimony by Admiral Rickover (given some years later in 1957), the project was changed over to being a Navy project exclusively in order to ensure that funding continued.
Development of the Kesselring Site began in 1948 with Government acquisition of the 3,900 acres of land. The Site was known as the West Milton Site until 1968 when it was renamed The Kenneth A. Kesselring Site in honor of a former Knolls Atomic Power Laboratory (KAPL) General Manager. The first power plant at the Site went into operation in 1955. This was the liquid sodium cooled Naval prototype for the second nuclear powered submarine, the original SEAWOLF.
The containment built for the S1G originally and is the largest spherical nuclear power plant containment vessel constructed to date. This sphere was 225 feet in diameter, and was thus larger than those built for Dresden 1, Garigliano (Italy), Yankee Rowe or any other.
The unique job of anchoring the 3,850 ton ball was completed by the Rust Engineering Company of Pittsburgh. Built over its foundation and supported by a ring of steel columns, the sphere posed the engineering problem of how to secure it in place without developing stresses that would break or distort its 1-inch skin. The engineering company solved it by alternately pouring a three-foot layer of limestone inside one day, and a three-foot layer of concrete outside the next, until the huge ball was secured. The SIR, later known as S1G reactor first achieved criticality on March 20, 1955; the plant generated its first useful power on May 18, 1955. In order to obtain the maximum press benefit from the startup of the plant, GE and the AEC staged a press event and on July 18, 1955 actually used the plant's generating equipment to sell power back to the commercial provider, thus claiming the first commercial sale of power generated by a nuclear plant at that time.
The neutron spectrum was intermediate in energy. It used UO2 fuel clad in stainless steel with Be used as a moderator and a reflector. The maximum temperature in the fuel could reach 1,700 +/- 300 °F with a maximum sheath temperature of 900 °F, with a cycle time of 900 hours or 900 / 24 = 37.5 days.
Because of its compact core, Admiral Hyman G. Rickover, the father of the US nuclear navy, had a sodium-cooled reactor built for the second US nuclear submarine. Seawolf was launched on July 21, 1955, and conducted sea trials in January 1957. After acceptance, Seawolf operated as an active unit of the Atlantic Fleet and in 1958 made a record-breaking submerged run of two months, traveling more than 13,000 miles submerged, producing air and water for her crew the entire time. Seawolf operated more than two years and steamed 71,000 miles on her sodium-cooled reactor.
After sea trials in 1957, however, Rickover had the reactor replaced by a pressurized water reactor. Duperheater bypass problems causing mediocre performance and as a result of a sodium fire. The steam turbines had their blades replaced to use saturated rather than superheated steam. The reactor was housed in a containment vessel designed to contain a sodium fire.
In 1958 the Navy had her refitted with a pressurized water reactor similar to the one in Nautilus, and that design is still the standard today. His summary of his experience with the sodium-cooled reactor pretty aptly characterized the problems that have been subsequently experienced in attempts to commercialize sodium-cooled breeder reactors. These reactors are “expensive to build, complex to operate, susceptible to prolonged shutdown as a result of even minor malfunctions, and difficult and time-consuming to repair.”
Because of low vapor pressure, an excellent performance as a heat transfer medium, good stability against radioactive rays and of particularly small slowing-down power, liquid sodium was a popular candidate for use in naval reactors. Sodium, the reactor coolant, is an alkali metal with a melting point of 208°F. It is a strong reducing agent that is capable of removing oxide films normally present on materials like stainless steels. It has several physical and nuclear properties wlrch make it attractive as a reactor coolant. Specifically, it has a low capture cross section for fast neutrons, a high thermal conductivity making it a good heat transfer fluid, and a high normal boiling point (1630°F), which allows reactor operation at a low pressure.
Undesirable properties include its chemical reactivity with water/steam and the fact that it burns spontaneously in air at temperatures above 300-400°F. Fires could represent the dominant risk contributor, especially given the unique characteristics of metal fires such as very high temperatures and fire suppression challenges.
Sodium cooled nuclear rectors include a primary continuously circulating loop of liquid sodium which is sequentially pumped through a reactor core to and from an intermediate heat exchanger. The reactors also include a secondary sodium loop. This loop obtains heat energy at the intermediate heat exchanger in the reactor and transmits heat energy to a steam generator for the generation of power.
The cycle of sodium through a sodium primary loop or a secondary loop typically divides into a "hot leg" and a "cold leg." Taking the case of the primary loop in a sodium cooled reactor, the so-called "hot leg" of the sodium cycle begins at the reactor core and continues until heat is extracted at an intermediate heat exchanger. The cold leg begins at the outlet of the intermediate heat exchanger and includes passage through the pump of this invention and ends at the reactor core inlet. The cycle of the primary loop sodium interior of the reactor endlessly repeats in circulation through the loop.
The "hot leg" and "cold leg" terms are relative. Typically, the cold leg of each loop is a temperature in excess of 600° F.; the hot leg is even more extreme --temperatures in the range of 1200° F. are common. The sodium in the sodium primary loop is radioactive, with the activity originating from activated -corrosion and fission products and from neutron activation of the sodiuin. Temperatures in the secondary loop are 50-100°F lower than in the primary system and the sodiuin is not radioactive.
In such facilities, it is necessary to detect the leakage of the liquid sodium from essential parts promptly and securely, which demands the development of a detecting apparatus of moderate cost. Conventionally, there are typically employed three methods for detecting the leakage of liquid sodium.
The first known method is to use the so-called electrode type detector consisting of a ceramic-insulated tube accomodating a metal wire, while the second known method utilizes a smoke detector sensitive to the smoke of sodium oxide which results from the leakage, and the third known method is to utilize a hydrogen detector sensitive to hydrogen gas which is produced as a resultant of the reaction of the sodium and atmospheric water vapor.
These known measures have been found, however, to be inconvenient or unsatisfactory for the reasons discussed below. Namely, the first method is not capable of immediate detection of the leak, because the elctrode-type detector is sensitive only to a relatively large quantity of sodium leakage, and also is likely to respond to electrically conductive substances other than sodium. In addition, the detector of this type cannot be placed at any desired location due to the structure thereof. In the second method, the detector tends to by erroneously operated or actuated by smoke from sources other than sodium oxide. The smoke detector is often invalid because only a part of the leaked sodium passes through the heat insulator surrounding the leaking point to the atmosphere to generate the smoke. The third method requires an expensive detector and thus is not practical when a large number of detectors in numerous places is necessary.
In the operation of liquid sodium cooled nuclear reactors, it may be necessary to shut down the fission reaction of the fuel to deal with emergencies or carry out routine maintenance services. Reactor shut down is attained by inserting neutron absorbing control rods into the core of fissionable fuel to deprive the fuel of the needed fission producing neutrons. However decay of the fuel in the shut down reactor continues to produce heat in significant amounts which must be dissipated from the reactor unit.
The heat capacity of the liquid metal coolant and adjacent structure aid in dissipating the residual heat. However, the structural materials of the nuclear reactor may not be capable of safely withstanding prolonged high temperatures. For example the concrete of the walls of the typical housing silo may splay and crack when subjected to high temperatures. Accordingly, auxiliary cooling systems are commonly utilized to safely remove heat from the nuclear reactor structure during shut down.
Water cooled reactors operate at or near the boiling point of water. Any significant rise in temperature results in the generation of steam and increased pressure. By contrast, sodium or sodium-potassium has an extremely high boiling point, in the range of 1800 degrees Fahrenheit at one atmosphere pressure. The normal operating temperature of the reactor is in the range of about 900 degrees Fahrenheit. Because of the high boiling point of the liquid metal, the pressure problems associated with water cooled reactors and the steam generated thereby are eliminated. The heat capacity of the liquid metal permits the sodium or sodium-potassium to be heated several hundred degrees Fahrenheit without danger of materials failure in the reactor.
Reactor startup is often the most risky time, particularly when the startup is of a new reactor design. Great care is taken to start a new reactor in a systematic fashion. The core is loaded and tested at zero power. Primary and secondary flow systems are tested at zero power. Finally, the full system is operated at increasing power levels to ensure it is working properly. A component may not have been tested at 100% operating conditions until these first full-power tests at startup. Even if a component has been properly designed, manufacturing defects could result in sodium leaks (the so-called “infant mortality” failures). Welds, particularly welds of dissimilar metals, represent a potential failure point at startup, and great care is taken to ensure their integrity prior to operation.
As the reactor is operated, the sodium in the primary system will become radioactive. Sodium-24 (t1/2=15 hours) is generated from naturally-occurring sodium-23 capturing neutrons in the core region. The level of radiation in the sodium will approach a constant level after a few days of operation. This induced radioactivity introduces additional complications with released sodium. The location of a primary loop sodium leak may be difficult or impossible for personnel to reach due to the high levels of radiation present. If sodium burns, large quantities of oxide aerosols are released. Radioactive sodium in these emissions poses a significant health hazard to plant workers and fire-fighting personnel. In addition, if the sodium vapor is released to the environment, it could represent a health threat to the general public. Perhaps more importantly, it will certainly create a public outcry and could prevent the restart of the reactor.
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