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136-421: There is no definitive origin for the term. United States Nuclear Regulatory Commission historian Tom Wellock notes that scram is English-language slang for leaving quickly and urgently, and he cites this as the original and most likely accurate basis for the use of scram in the technical context. Scram is sometimes cited as being an acronym for safety control rod axe man or safety cut rope axe man . This

272-720: A FOIA request. NRC conducts audits and training inspections, observes the National Nuclear Accrediting Board meetings, and nominates some members. The 1980 Kemeny Commission's report after the Three Mile Island accident recommended that the nuclear energy industry "set and police its own standards of excellence". The nuclear industry founded the Institute of Nuclear Power Operations (INPO) within 9 months to establish personnel training and qualification. The industry through INPO created

408-449: A LBLOCA (large-break loss-of-coolant accident – a massive pipe rupture leading to catastrophic loss of coolant pressure within the reactor, considered the most threatening "design basis accident" in probabilistic risk assessment and nuclear safety and security ), which is anticipated to lead to the temporary exposure of the core; this core drying-out event is termed core "uncovery", for the core loses its heat-removing cover of coolant, in

544-545: A boiling water reactor would be feasible for use in energy production. He found that it was, after subjecting his reactors to quite strenuous tests, proving the safety principles of the BWR. Following this series of tests, GE got involved and collaborated with Argonne National Laboratory to bring this technology to market. Larger-scale tests were conducted through the late 1950s/early/mid-1960s that only partially used directly generated (primary) nuclear boiler system steam to feed

680-512: A defence in depth philosophy, which is a design philosophy that is integrated throughout construction and commissioning . A BWR is similar to a pressurized water reactor (PWR) in that the reactor will continue to produce heat even after the fission reactions have stopped, which could make a core damage incident possible. This heat is produced by the radioactive decay of fission products and materials that have been activated by neutron absorption . BWRs contain multiple safety systems for cooling

816-622: A 1952 U.S. Atomic Energy Commission (AEC) report by Fermi, the AEC declassified information on the Chicago Pile. The report includes a section written by Wilson's team shortly after the Chicago Pile achieved a self-sustaining chain reaction on December 2, 1942. It includes a wiring schematic of the rod control circuitry with a clearly labeled "SCRAM" line (see image on the right and pages 37 and 48). The Russian name, AZ-5 ( АЗ-5 , in Cyrillic ),

952-514: A BWR and PWR is that in a BWR, the reactor core heats water, which turns to steam and then drives a steam turbine. In a PWR, the reactor core heats water, which does not boil. This hot water then exchanges heat with a lower pressure system, which turns water into steam that drives the turbine. The BWR was developed by the Argonne National Laboratory and General Electric (GE) in the mid-1950s. The main present manufacturer

1088-409: A BWR core is substantiated by a calculation that proves that 99.9% of fuel rods in a BWR core will not enter the transition to film boiling during normal operation or anticipated operational occurrences. Since the BWR is boiling water, and steam does not transfer heat as well as liquid water, MFLCPR typically occurs at the top of a fuel assembly, where steam volume is the highest. FLLHGR (FDLRX, MFLPD)

1224-482: A BWR tends to balance the savings due to the simpler design and greater thermal efficiency of a BWR when compared with a PWR. Most of the radioactivity in the water is very short-lived (mostly N-16, with a 7-second half-life ), so the turbine hall can be entered soon after the reactor is shut down. BWR steam turbines employ a high-pressure turbine designed to handle saturated steam, and multiple low-pressure turbines. The high-pressure turbine receives steam directly from

1360-524: A BWR: MFLCPR, FLLHGR, and APLHGR must be kept less than 1.0 during normal operation; administrative controls are in place to assure some margin of error and margin of safety to these licensed limits. Typical computer simulations divide the reactor core into 24–25 axial planes ; relevant quantities (margins, burnup, power, void history) are tracked for each "node" in the reactor core (764 fuel assemblies x 25 nodes/assembly = 19100 nodal calculations/quantity). Specifically, MFLCPR represents how close

1496-490: A PWR, where the turbine steam demand is set manually by the operators, in a BWR, the turbine valves will modulate to maintain reactor pressure at a setpoint. Under this control mode, the turbine output will automatically follow reactor power changes. When the turbine is offline or trips, the main steam bypass/dump valves will open to direct steam directly to the condenser. These bypass valves will automatically or manually modulate as necessary to maintain reactor pressure and control

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1632-551: A city block but would not have presented an immediate health hazard. Twelve years into NRC operations, a 1987 congressional report entitled "NRC Coziness with Industry" concluded, that the NRC "has not maintained an arms length regulatory posture with the commercial nuclear power industry ... [and] has, in some critical areas, abdicated its role as a regulator altogether". To cite three examples: A 1986 Congressional report found that NRC staff had provided valuable technical assistance to

1768-458: A decrease in neutron multiplication , and thus shutting down the reactor without use of the control rods. In the PWR, these neutron absorbing solutions are stored in pressurized tanks (called accumulators) that are attached to the primary coolant system via valves. A varying level of neutron absorbent is kept within the primary coolant at all times, and is increased using the accumulators in the event of

1904-430: A failure of all of the control rods to insert, which will promptly bring the reactor below the shutdown margin. In the BWR, soluble neutron absorbers are found within the standby liquid control system , which uses redundant battery-operated injection pumps, or, in the latest models, high pressure nitrogen gas to inject the neutron absorber solution into the reactor vessel against any pressure within. Because they may delay

2040-632: A high power output (1350 MWe per reactor), and a significantly lowered probability of core damage. Most significantly, the ABWR was a completely standardized design, that could be made for series production. The ABWR was approved by the United States Nuclear Regulatory Commission for production as a standardized design in the early 1990s. Subsequently, numerous ABWRs were built in Japan. One development spurred by

2176-742: A library, which also contains online document collections. In 1984 it started an electronic repository called ADAMS, the Agencywide Documents Access and Management System, for its public inspection reports, correspondence, and other technical documents written by NRC staff, contractors, and licensees. It was upgraded in October 2010 and is now web-based. Of documents from 1980 to 1999 only some have abstracts and/or full text; most are citations. Documents from before 1980 are available in paper or microfiche formats. Copies of these older documents or classified documents can be applied for with

2312-402: A model of the fuel assembly but power it with resistive heaters. These mock fuel assemblies are put into a test stand where data points are taken at specific powers, flows, pressures. Experimental data is conservatively applied to BWR fuel to ensure that the transition to film boiling does not occur during normal or transient operation. Typical SLMCPR/MCPRSL (Safety Limit MCPR) licensing limit for

2448-399: A more homogeneous distribution of the power: in the upper side the density of the water is lower due to vapour formation, making the neutron moderation less efficient and the fission probability lower. In normal operation, the control rods are only used to keep a homogeneous power distribution in the reactor and to compensate for the consumption of the fuel, while the power is controlled through

2584-445: A reactor that has not had a constant power history, the exact percentage is determined by the concentrations and half-lives of the individual fission products in the core at the time of the scram. The power produced by decay heat decreases as the fission products decay, but it is large enough that failure to remove decay heat may cause the reactor core temperature to rise to dangerous levels and has caused nuclear accidents , including

2720-410: A safety-related contingency developed. For example, if the reactor got too hot, it would trigger a system that would release soluble neutron absorbers (generally a solution of borated materials, or a solution of borax ), or materials that greatly hamper a chain reaction by absorbing neutrons, into the reactor core. The tank containing the soluble neutron absorbers would be located above the reactor, and

2856-417: A sales tool to help push American technology to foreign governments, when "lobbying for the purchase of equipment made by Westinghouse Electric Company and other domestic manufacturers". This gives the appearance of a regulator which is acting in a commercial capacity, "raising concerns about a potential conflict of interest ". San Clemente Green, an environmental group opposed to the continued operation of

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2992-468: A scram to insert the control rods, as it is the most reliable method of completely inserting the control rods, and prevents the possibility of accidentally withdrawing them during or after the shutdown. Most neutrons in a reactor are prompt neutrons ; that is, neutrons produced directly by a fission reaction. These neutrons move at a high velocity , so they are likely to escape into the moderator before being captured . On average, it takes about 13 μs for

3128-402: A series of notched positions with fixed intervals between these positions. Due to the limitations of the manual control system, it is possible while starting-up that the core can be placed into a condition where movement of a single control rod can cause a large nonlinear reactivity change, which could heat fuel elements to the point they fail (melt, ignite, weaken, etc.). As a result, GE developed

3264-415: A set of rules in 1977 called BPWS (Banked Position Withdrawal Sequence) which help minimize the effect of any single control rod movement and prevent fuel damage in the case of a control rod drop accident. BPWS separates control rods into four groups, A1, A2, B1, and B2. Then, either all of the A control rods or B control rods are pulled full out in a defined sequence to create a " checkerboard " pattern. Next,

3400-514: A single core-damaging event during their 100-year lifetimes. Earlier designs of the BWR, the BWR/4, had core damage probabilities as high as 1×10 core-damage events per reactor-year. This extraordinarily low CDP for the ESBWR far exceeds the other large LWRs on the market. Reactor start up ( criticality ) is achieved by withdrawing control rods from the core to raise core reactivity to a level where it

3536-468: A small group of engineers accidentally increased the reactor power level on an experimental reactor to such an extent that the water quickly boiled. This shut down the reactor, indicating the useful self-moderating property in emergency circumstances. In particular, Samuel Untermyer II , a researcher at Argonne National Laboratory , proposed and oversaw a series of experiments: the BORAX experiments —to see if

3672-465: A tortuous path where the water droplets are slowed and directed out into the downcomer or annulus region. The "dry" steam then exits the RPV through four main steam lines and goes to the turbine. Reactor power is controlled via two methods: by inserting or withdrawing control rods (control blades) and by changing the water flow through the reactor core . Positioning (withdrawing or inserting) control rods

3808-451: Is GE Hitachi Nuclear Energy , which specializes in the design and construction of this type of reactor. A boiling water reactor uses demineralized water as a coolant and neutron moderator . Heat is produced by nuclear fission in the reactor core, and this causes the cooling water to boil, producing steam. The steam is directly used to drive a turbine , after which it is cooled in a condenser and converted back to liquid water. This water

3944-450: Is $ 1,059.5 million, with 3,895.9 full-time equivalents (FTE), 90 percent of which is recovered by fees. This is an increase of $ 3.6 million, including 65.1 FTE, compared to FY 2014. NRC headquarters offices are located in unincorporated North Bethesda, Maryland (although the mailing address for two of the three main buildings in the complex list the city as Rockville, MD ), and there are four regional offices. The NRC territory

4080-401: Is a limit on fuel rod power in the reactor core. For new fuel, this limit is typically around 13 kW/ft (43 kW/m) of fuel rod. This limit ensures that the centerline temperature of the fuel pellets in the rods will not exceed the melting point of the fuel material ( uranium / gadolinium oxides) in the event of the worst possible plant transient/scram anticipated to occur. To illustrate

4216-505: Is an abbreviation for аварийная защита 5-й категории ( avariynaya zashhchita 5-y kategorii ), which translates to "emergency protection of the 5th category" in English. In any reactor, a scram is achieved by inserting large amounts of negative reactivity mass into the midst of the fissile material, to immediately terminate the fission reaction. In light-water reactors , this is achieved by inserting neutron-absorbing control rods into

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4352-483: Is broken down into four geographical regions; until the late 1990s, there was a Region V office in Walnut Creek, California which was absorbed into Region IV, and Region V was dissolved. In these four regions NRC oversees the operation of US nuclear reactors , namely 94 power-producing reactors, and 31 non-power-producing, or research and test reactors. Oversight is done on several levels. For example: NRC has

4488-619: Is closely related to the National Academy for Nuclear Training, not a government body, and referred to as independent by INPO, the Nuclear Energy Institute, and nuclear utilities. but not by the NRC, all of whom are represented on the board. The 1982 Nuclear Waste Policy Act directed NRC in Section 306 to issue regulations or "other appropriate regulatory guidance" on training of nuclear plant personnel. Since

4624-464: Is designated by the president to be the chairman and official spokesperson of the commission. The chairman is the principal executive officer of the NRC, who exercise all of the executive and administrative functions of the commission. The current chairman is Christopher T. Hanson . President Biden designated Hanson as chairman of the NRC effective January 20, 2021. The current commissioners as of September 24, 2024: President Biden has nominated

4760-503: Is done via cranes and under water. For this reason the spent fuel storage pools are above the reactor in typical installations. They are shielded by water several times their height, and stored in rigid arrays in which their geometry is controlled to avoid criticality. In the Fukushima Daiichi nuclear disaster this became problematic because water was lost (as it was heated by the spent fuel) from one or more spent fuel pools and

4896-421: Is easily varied by simply increasing or decreasing the forced recirculation flow through the recirculation pumps. The two-phase fluid (water and steam) above the core enters the riser area, which is the upper region contained inside of the shroud. The height of this region may be increased to increase the thermal natural recirculation pumping head. At the top of the riser area is the moisture separator. By swirling

5032-447: Is evident that the nuclear chain reaction is self-sustaining. This is known as "going critical". Control rod withdrawal is performed slowly, as to carefully monitor core conditions as the reactor approaches criticality. When the reactor is observed to become slightly super-critical, that is, reactor power is increasing on its own, the reactor is declared critical. Rod motion is performed using rod drive control systems. Newer BWRs such as

5168-569: Is in place to ensure that the highest powered fuel rod will not melt if its power was rapidly increased following a pressurization transient. Abiding by the LHGR limit precludes melting of fuel in a pressurization transient. APLHGR, being an average of the Linear Heat Generation Rate (LHGR), a measure of the decay heat present in the fuel bundles, is a margin of safety associated with the potential for fuel failure to occur during

5304-423: Is known as the advanced boiling water reactor (ABWR). The ABWR was developed in the late 1980s and early 1990s, and has been further improved to the present day. The ABWR incorporates advanced technologies in the design, including computer control, plant automation, control rod removal, motion, and insertion, in-core pumping, and nuclear safety to deliver improvements over the original series of production BWRs, with

5440-433: Is licensed to operate, the fuel vendor/licensee simulate events with computer models. Their approach is to simulate worst case events when the reactor is in its most vulnerable state. APLHGR is commonly pronounced as "Apple Hugger" in the industry. PCIOMR is a set of rules and limits to prevent cladding damage due to pellet-clad interaction. During the first nuclear heatup, nuclear fuel pellets can crack. The jagged edges of

5576-408: Is monitored with an empirical correlation that is formulated by vendors of BWR fuel (GE, Westinghouse, AREVA-NP). The vendors have test rigs where they simulate nuclear heat with resistive heating and determine experimentally what conditions of coolant flow, fuel assembly power, and reactor pressure will be in/out of the transition boiling region for a particular fuel design. In essence, the vendors make

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5712-453: Is required that the decay heat stored in the fuel assemblies at any one time does not overwhelm the ECCS. As such, the measure of decay heat generation known as LHGR was developed by GE's engineers, and from this measure, APLHGR is derived. APLHGR is monitored to ensure that the reactor is not operated at an average power level that would defeat the primary containment systems. When a refueled core

5848-435: Is saturated with a steam quality of about 15%. Typical core flow may be 45,000,000 kg/h (100,000,000 lb/h) with 6,500,000 kg/h (14,500,000 lb/h) steam flow. However, core-average void fraction is a significantly higher fraction (~40%). These sort of values may be found in each plant's publicly available Technical Specifications, Final Safety Analysis Report, or Core Operating Limits Report. The heating from

5984-408: Is switched to a "Three-Element" control mode, where the controller looks at the current water level in the reactor, as well as the amount of water going in and the amount of steam leaving the reactor. By using the water injection and steam flow rates, the feed water control system can rapidly anticipate water level deviations and respond to maintain water level within a few inches of set point. If one of

6120-492: Is terminated by the automatic insertion of the control rods. So, when the reactor is isolated from the turbine rapidly, pressure in the vessel rises rapidly, which collapses the water vapor, which causes a power excursion which is terminated by the Reactor Protection System. If a fuel pin was operating at 13.0 kW/ft prior to the transient, the void collapse would cause its power to rise. The FLLHGR limit

6256-452: Is the normal method for controlling power when starting up a BWR. As control rods are withdrawn, neutron absorption decreases in the control material and increases in the fuel, so reactor power increases. As control rods are inserted, neutron absorption increases in the control material and decreases in the fuel, so reactor power decreases. Differently from the PWR, in a BWR the control rods ( boron carbide plates) are inserted from below to give

6392-429: Is then returned to the reactor core, completing the loop. The cooling water is maintained at about 75 atm (7.6 MPa , 1000–1100 psi ) so that it boils in the core at about 285 °C (550 °F). In comparison, there is no significant boiling allowed in a pressurized water reactor (PWR) because of the high pressure maintained in its primary loop—approximately 158 atm (16 MPa, 2300 psi). The core damage frequency of

6528-585: Is to regulate the nation's civilian use of byproduct, source, and special nuclear materials to ensure adequate protection of public health and safety, to promote the common defense and security, and to protect the environment. The NRC's regulatory mission covers three main areas : The NRC is headed by five commissioners appointed by the president of the United States and confirmed by the United States Senate for five-year terms. One of them

6664-554: The ABWR and ESBWR as well as all German and Swedish BWRs use the Fine Motion Control Rod Drive system, which allows multiple rods to be controlled with very smooth motions. This allows a reactor operator to evenly increase the core's reactivity until the reactor is critical. Older BWR designs use a manual control system, which is usually limited to controlling one or four control rods at a time, and only through

6800-565: The Emergency Core Cooling System . The ECCS is designed to rapidly flood the reactor pressure vessel, spray water on the core itself, and sufficiently cool the reactor fuel in this event. However, like any system, the ECCS has limits, in this case, to its cooling capacity, and there is a possibility that fuel could be designed that produces so much decay heat that the ECCS would be overwhelmed and could not cool it down successfully. So as to prevent this from happening, it

6936-777: The San Onofre Nuclear Plant , said in 2011 that instead of being a watchdog, the NRC too often rules in favor of nuclear plant operators. In 2011, the Tōhoku earthquake and tsunami led to unprecedented damage and flooding of the Fukushima Daiichi Nuclear Power Plant . The subsequent loss of offsite power and flooding of onsite emergency diesel generators led to loss of coolant and subsequent Nuclear meltdown of three reactor cores. The Fukushima Daiichi nuclear disaster led to an uncontrolled release of radioactive contamination, and forced

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7072-768: The United States . However, the case for widespread nuclear plant construction was eroded due to abundant natural gas supplies. Many license applications for proposed new reactors were suspended or cancelled. These will not be the cheapest energy options available, therefore not an attractive investment. In 2013, four reactors were permanently closed: San Onofre 2 and 3 in California, Crystal River 3 in Florida, and Kewaunee in Wisconsin. Vermont Yankee , in Vernon,

7208-592: The United States Atomic Energy Commission . Its functions include overseeing reactor safety and security, administering reactor licensing and renewal, licensing radioactive materials , radionuclide safety, and managing the storage, security, recycling, and disposal of spent fuel . Prior to 1975 the Atomic Energy Commission was in charge of matters regarding radionuclides. The AEC was dissolved, because it

7344-572: The 'National Academy for Nuclear Training Program' either as early as 1980 or in September 1985 per the International Atomic Energy Agency . INPO refers to NANT as "our National Academy for Nuclear Training" on its website. NANT integrates and standardizes the training programs of INPO and US nuclear energy companies, offers training scholarships and interacts with the 'National Nuclear Accrediting Board'. This Board

7480-498: The 'Operator Requalification Rule' 59 FR 5938, Feb. 9, 1994, allowing each nuclear power plant company to conduct the operator licensing renewal examination every six years, eliminating the requirement of NRC-administered written requalification examination. In 1999, NRC issued a final rule on operator initial licensing examination, that allows companies to prepare, proctor, and grade their own operator initial licensing examinations. Facilities can "upon written request" continue to have

7616-481: The 1979 Three Mile Island accident in Pennsylvania, the NRC has often been too timid in ensuring that America's commercial reactors are operated safely: Nuclear power regulation is a textbook example of the problem of "regulatory capture" — in which an industry gains control of an agency meant to regulate it. Regulatory capture can be countered only by vigorous public scrutiny and Congressional oversight, but in

7752-668: The 32 years since Three Mile Island, interest in nuclear regulation has declined precipitously. An article in the Bulletin of the Atomic Scientists stated that many forms of NRC regulatory failure exist, including regulations ignored by the common consent of NRC and industry: A worker (named George Galatis ) at the Millstone Nuclear Power Plant in Connecticut kept warning management, that

7888-561: The AZ-5 shutdown system was initiated after a core overheat. RBMK reactors were subsequently either retrofitted to account for the flaw, or decommissioned. Not all of the heat in a nuclear reactor is generated by the chain reaction that a scram is designed to stop. For a reactor that is scrammed after holding a constant power level for an extended period (greater than 100 hrs), about 7% of the steady-state power will remain after initial shutdown due to fission product decay that cannot be stopped. For

8024-670: The Allegations Program, Office of Investigations, Office of Nuclear Security and Incident Response, Region I, Region II, Region III, Region IV, Office of the Chief Information Officer, Office of Administration, Office of the Chief Human Capital Officer, and Office of Small Business and Civil Rights. Of these operations offices, NRC's major program components are the first two offices mentioned above. NRC's proposed FY 2015 budget

8160-511: The EISs and found significant flaws, included failure to consider significant issues of concern. It also found that the NRC management had significantly underestimated the risk and consequences posed by a severe reactor accident such as a full-scale nuclear meltdown. NRC management asserted, without scientific evidence, that the risk of such accidents were so "Small" that the impacts could be dismissed and therefore no analysis of human and environmental

8296-571: The Fukushima disaster, the NRC prepared a report in 2011 to examine the risk that dam failures posed on the nation's fleet of nuclear reactors. A redacted version of NRC's report on dam failures was posted on the NRC website on March 6. The original, un-redacted version was leaked to the public. The un-redacted version which was leaked to the public highlights the threat that flooding poses to nuclear power plants located near large dams and substantiates claims that NRC management has intentionally misled

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8432-799: The General Counsel, Office of International Programs, Office of Public Affairs, Office of the Secretary, Office of the Chief Financial Officer, Office of the Executive Director for Operations). Christopher T. Hanson is the chairman of the NRC. There are 14 Executive Director for Operations offices: Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation , Office of Nuclear Regulatory Research, Office of Enforcement, which investigates reports by nuclear power whistleblowers , specifically

8568-520: The Japanese Government to evacuate approximately 100,000 citizens. Gregory Jaczko was chairman of the NRC when the 2011 Fukushima disaster occurred in Japan. Jaczko looked for lessons for the US, and strengthened security regulations for nuclear power plants . For example, he supported the requirement that new plants be able to withstand an aircraft crash. On February 9, 2012, Jaczko cast

8704-530: The NRC about the pipes and the NRC also did not know they existed. In March 2011, the Union of Concerned Scientists released a study critical of the NRC's 2010 performance as a regulator. The UCS said that over the years, it had found the NRC's enforcement of safety rules has not been "timely, consistent, or effective" and it cited 14 "near-misses" at U.S. plants in 2010 alone. In April 2011, Reuters reported that diplomatic cables showed NRC sometimes being used as

8840-588: The NRC approved a 20-year extension for the license of Vermont Yankee Nuclear Power Plant , although the Vermont state legislature voted overwhelmingly to deny an extension. The plant had been found to be leaking radioactive materials through a network of underground pipes, which Entergy had denied under oath even existed. At a hearing in 2009 Tony Klein, chairman of the Vermont House Natural Resources and Energy Committee had asked

8976-470: The NRC mailed the license to the West Virginia postal box. Upon receipt of the license, GAO officials were able to easily modify its stipulations and remove a limit on the amount of radioactive material they could buy. A spokesman for the NRC said that the agency considered the radioactive devices a "lower-level threat"; a bomb built with the materials could have contaminated an area about the length of

9112-494: The Nuclear Regulatory Commission is "on the defensive to prove it is doing its job of ensuring safety". In October 2011, Jaczko described "a tension between wanting to move in a timely manner on regulatory questions, and not wanting to go too fast". In 2011 Edward J. Markey , Democrat of Massachusetts, criticized the NRC's response to the Fukushima Daiichi nuclear disaster and the decision-making on

9248-544: The SCRAM [Safety Control Rod Axe Man] story until many years after the fact. Then one day one of my fellows who had been on Zinn's construction crew called me Mr. Scram. I asked him, "How come?" And then the story. Leona Marshall Libby , who was present that day at the Chicago Pile, recalled that the term was coined by Volney Wilson who led the team that designed the control rod circuitry: The safety rods were coated with cadmium foil, and this metal absorbed so many neutrons that

9384-434: The US's first research effort in nuclear power being devoted to the PWR, which was highly suited for naval vessels (submarines, especially), as space was at a premium, and PWRs could be made compact and high-power enough to fit into such vessels. But other researchers wanted to investigate whether the supposed instability caused by boiling water in a reactor core would really cause instability. During early reactor development,

9520-572: The United States and elsewhere. In March 2007, undercover investigators from the Government Accountability Office set up a false company and obtained a license from the Nuclear Regulatory Commission that would have allowed them to buy the radioactive materials needed for a dirty bomb. According to the GAO report, NRC officials did not visit the company or attempt to personally interview its executives. Instead, within 28 days,

9656-573: The United States. The origins and development of NRC regulatory processes and policies are explained in five volumes of history published by the University of California Press . These are: The NRC has produced a booklet, A Short History of Nuclear Regulation 1946–2009 , which outlines key issues in NRC history. Thomas Wellock , a former academic, is the NRC historian. Before joining the NRC, Wellock wrote Critical Masses: Opposition to Nuclear Power in California, 1958–1978 . The NRC's mission

9792-499: The absorption solution, once the system was triggered, would flow into the core through force of gravity, and bring the reaction to a near-complete stop. Another example was the Isolation Condenser system , which relied on the principle of hot water/steam rising to bring hot coolant into large heat exchangers located above the reactor in very deep tanks of water, thus accomplishing residual heat removal. Yet another example

9928-436: The case of a BWR, light water. If the core is uncovered for too long, fuel failure can occur; for the purpose of design, fuel failure is assumed to occur when the temperature of the uncovered fuel reaches a critical temperature (1100 °C, 2200 °F). BWR designs incorporate failsafe protection systems to rapidly cool and make safe the uncovered fuel prior to it reaching this temperature; these failsafe systems are known as

10064-438: The chain reaction was stopped. Volney Wilson called these "scram" rods. He said that the pile had "scrammed," the rods had "scrammed" into the pile. Other witnesses that day agreed with Libby's crediting "scram" to Wilson. Wellock wrote that Warren Nyer, a student who worked on assembling the pile, also attributed the word to Wilson: "The word arose in a discussion Dr. Wilson, who was head of the instrumentation and controls group,

10200-471: The cladding remains intact for the life of the rod. The BWR concept was developed slightly later than the PWR concept. Development of the BWR started in the early 1950s, and was a collaboration between General Electric (GE) and several US national laboratories. Research into nuclear power in the US was led by the three military services. The Navy, seeing the possibility of turning submarines into full-time underwater vehicles, and ships that could steam around

10336-407: The control rods upon any interruption of the electric current. In both the PWR and the BWR there are secondary systems (and often even tertiary systems) that will insert control rods in the event that primary rapid insertion does not promptly and fully actuate. Liquid neutron absorbers ( neutron poisons ) are also used in rapid shutdown systems for heavy and light water reactors. Following a scram, if

10472-477: The core after emergency shut down. The reactor fuel rods are occasionally replaced by moving them from the reactor pressure vessel to the spent fuel pool. A typical fuel cycle lasts 18–24 months, with about one third of fuel assemblies being replaced during a refueling outage. The remaining fuel assemblies are shuffled to new core locations to maximize the efficiency and power produced in the next fuel cycle. Because they are hot both radioactively and thermally, this

10608-454: The core creates a thermal head that assists the recirculation pumps in recirculating the water inside of the RPV. A BWR can be designed with no recirculation pumps and rely entirely on the thermal head to recirculate the water inside of the RPV. The forced recirculation head from the recirculation pumps is very useful in controlling power, however, and allows achieving higher power levels that would not otherwise be possible. The thermal power level

10744-402: The core, although the mechanism by which rods are inserted depends on the type of reactor. In pressurized water reactors the control rods are held above a reactor's core by electric motors against both their own weight and a powerful spring. A scram is designed to release the control rods from those motors and allows their weight and the spring to drive them into the reactor core, rapidly halting

10880-408: The downcomer or annulus region, which is separated from the core by a tall shroud. The water then goes through either jet pumps or internal recirculation pumps that provide additional pumping power (hydraulic head). The water now makes a 180-degree turn and moves up through the lower core plate into the nuclear core, where the fuel elements heat the water. Water exiting the fuel channels at the top guide

11016-400: The earthquake could have altered the geometry. The fact that the fuel rods' cladding is a zirconium alloy was also problematic since this element can react with steam at temperatures above 1,500 K (1,230 °C) to produce hydrogen, which can ignite with oxygen in the air. Normally the fuel rods are kept sufficiently cool in the reactor and spent fuel pools that this is not a concern, and

11152-532: The event of a major safety contingency for at least 48 hours following the safety contingency; thence, it would only require periodic refilling of cooling water tanks located completely outside of the reactor, isolated from the cooling system, and designed to remove reactor waste heat through evaporation. The simplified boiling water reactor was submitted to the United States Nuclear Regulatory Commission , however, it

11288-428: The event of a transient requiring the quenching of steam), as well as the drywell, the elimination of the heat exchanger, the steam dryer, the distinctive general layout of the reactor building, and the standardization of reactor control and safety systems. The first, General Electric ( GE ), series of production BWRs evolved through 6 iterative design phases, each termed BWR/1 through BWR/6. (BWR/4s, BWR/5s, and BWR/6s are

11424-417: The examinations prepared and administered by NRC staff, but if a company volunteers to prepare the examination, NRC continues to approve and administer it. Since 2000 meetings between NRC and applicants or licensees have been open to the public. Between 2007 and 2009, 13 companies applied to the Nuclear Regulatory Commission for construction and operating licenses to build 25 new nuclear power reactors in

11560-449: The feedwater heaters enters the reactor pressure vessel (RPV) through nozzles high on the vessel, well above the top of the nuclear fuel assemblies (these nuclear fuel assemblies constitute the "core") but below the water level. The feedwater enters into the downcomer or annulus region and combines with water exiting the moisture separators. The feedwater subcools the saturated water from the moisture separators. This water now flows down

11696-480: The first chain reaction was Norman Hilberry . In a letter to Raymond Murray (January 21, 1981), Hilberry wrote: When I showed up on the balcony on that December 2, 1942 afternoon, I was ushered to the balcony rail, handed a well sharpened fireman's axe and told, "If the safety rods fail to operate, cut that manila rope ." The safety rods, needless to say, worked, the rope was not cut... I don't believe I have ever felt quite as foolish as I did then. ...I did not get

11832-524: The five-member NRC had become "captive of the industries that it regulates". Numerous different observers have criticized the NRC as an example of regulatory capture The NRC has been accused of having conflicting roles as regulator and "salesman" in a 2011 Reuters article, doing an inadequate job by the Union of Concerned Scientists , and the agency approval process has been called a "rubber stamp". Frank N. von Hippel wrote in March 2011, that despite

11968-660: The following to fill a seat on the commission. They await Senate confirmation. The NRC consists of the commission on the one hand and offices of the executive director for Operations on the other. The commission is divided into two committees (Advisory Committee on Reactor Safeguards and Advisory Committee on the Medical Uses of Isotopes) and one Board, the Atomic Safety and Licensing Board Panel, as well as eight commission staff offices (Office of Commission Appellate Adjudication, Office of Congressional Affairs, Office of

12104-442: The fuel, and reactor power increases. As flow of water through the core is decreased, steam voids remain longer in the core, the amount of liquid water in the core decreases, neutron moderation decreases, fewer neutrons are slowed enough to be absorbed by the fuel, and reactor power decreases. Thus the BWR has a negative void coefficient . Reactor pressure in a BWR is controlled by the main turbine or main steam bypass valves. Unlike

12240-477: The future. This will reduce demand for replacement new builds. Byrne and Hoffman wrote in 1996, that since the 1980s the NRC has generally favored the interests of nuclear industry, and been unduly responsive to industry concerns, while failing to pursue tough regulation. The NRC has often sought to hamper or deny public access to the regulatory process, and created new barriers to public participation. Barack Obama , when running for president in 2007, said that

12376-405: The heated surface to increase drastically to once again reach equilibrium heat transfer with the cooling fluid. In other words, steam semi-insulates the heated surface and surface temperature rises to allow heat to get to the cooling fluid (through convection and radiative heat transfer). Nuclear fuel could be damaged by film boiling; this would cause the fuel cladding to overheat and fail. MFLCPR

12512-531: The industry's own Institute for Nuclear Power Operations (INPO), an organization formed by utilities in response to the Three Mile Island Accident. One example involves the license renewal program that NRC initiated to extend the operating licenses for the nation's fleet of commercial nuclear reactors. Environmental impact statements (EIS) were prepared for each reactor to extend the operational period from 40 to 60 years. One study examined

12648-405: The inner walls of the fuel cladding which are resistant to perforation due to pellet-clad interactions, and the second is a set of rules created under PCIOMR. The PCIOMR rules require initial "conditioning" of new fuel. This means, for the first nuclear heatup of each fuel element, that local bundle power must be ramped very slowly to prevent cracking of the fuel pellets and limit the differences in

12784-400: The leading fuel bundle is to "dry-out" (or "departure from nucleate boiling" for a PWR). Transition boiling is the unstable transient region where nucleate boiling tends toward film boiling . A water drop dancing on a hot frying pan is an example of film boiling. During film boiling a volume of insulating vapor separates the heated surface from the cooling fluid; this causes the temperature of

12920-524: The lone dissenting vote on plans to build the first new nuclear power plant in more than 30 years when the NRC voted 4–1 to allow Atlanta-based Southern Co to build and operate two new nuclear power reactors at its existing Vogtle Electric Generating Plant in Georgia. He cited safety concerns stemming from Japan's 2011 Fukushima nuclear disaster , saying "I cannot support issuing this license as if Fukushima never happened". In July 2011, Mark Cooper said that

13056-571: The most common types in service today.) The vast majority of BWRs in service throughout the world belong to one of these design phases. Containment variants were constructed using either concrete or steel for the Primary Containment, Drywell and Wetwell in various combinations. Apart from the GE designs there were others by ABB (Asea-Atom), MITSU, Toshiba and KWU (Kraftwerk Union). See List of boiling water reactors . A newer design of BWR

13192-486: The neutrons to be slowed by the moderator enough to facilitate a sustained reaction, which allows the insertion of neutron absorbers to affect the reactor quickly. As a result, once the reactor has been scrammed, the reactor power will drop significantly almost instantaneously. A small fraction (about 0.65%) of neutrons in a typical power reactor comes from the radioactive decay of a fission product. These delayed neutrons , which are emitted at lower velocities, will limit

13328-513: The nuclear accidents at Three Mile Island and Fukushima I . Nuclear Regulatory Commission The United States Nuclear Regulatory Commission ( NRC ) is an independent agency of the United States government tasked with protecting public health and safety related to nuclear energy. Established by the Energy Reorganization Act of 1974 , the NRC began operations on January 19, 1975, as one of two successor agencies to

13464-492: The nuclear industry already had developed training and accreditation, NRC issued a policy statement in 1985, endorsing the INPO program. NRC has a memorandum of agreement with INPO and "monitors INPO activities by observing accreditation team visits and the monthly NNAB meetings". In 1993, NRC endorsed the industry's approach to training that had been used for nearly a decade through its 'Training Rule'. In February 1994, NRC passed

13600-423: The nuclear reaction by absorbing liberated neutrons. Another design uses electromagnets to hold the rods suspended, with any cut to the electric current resulting in an immediate and automatic control rod insertion. In boiling water reactors , the control rods are inserted up from underneath the reactor vessel. In this case a hydraulic control unit with a pressurized storage tank provides the force to rapidly insert

13736-469: The opposing group (B or A) is pulled in a defined sequence to positions 02, then 04, 08, 16, and finally full out (48). By following a BPWS compliant start-up sequence, the manual control system can be used to evenly and safely raise the entire core to critical, and prevent any fuel rods from exceeding 280 cal/gm energy release during any postulated event which could potentially damage the fuel. Several calculated/measured quantities are tracked while operating

13872-404: The pellet can rub and interact with the inner cladding wall. During power increases in the fuel pellet, the ceramic fuel material expands faster than the fuel cladding, and the jagged edges of the fuel pellet begin to press into the cladding, potentially causing a perforation. To prevent this from occurring, two corrective actions were taken. The first is the inclusion of a thin barrier layer against

14008-477: The petition asks the NRC to halt proceedings to approve the standardized AP1000 and Economic Simplified Boiling Water Reactor designs. The petitioners asked the NRC to supplement its own investigation by establishing an independent commission comparable to that set up in the wake of the less severe 1979 Three Mile Island accident . The petitioners included Public Citizen , Southern Alliance for Clean Energy , and San Luis Obispo Mothers for Peace . Following

14144-1453: The proposed Westinghouse AP1000 reactor design. In 2011, a total of 45 groups and individuals from across the nation formally asked the NRC to suspend all licensing and other activities at 21 proposed nuclear reactor projects in 15 states until the NRC completed a thorough post- Fukushima nuclear disaster examination: The petition seeks suspension of six existing reactor license renewal decisions ( Columbia Generating Station , WA Davis–Besse Nuclear Power Station , OH, Diablo Canyon Power Plant , CA, Indian Point Energy Center , NY, Pilgrim Nuclear Generating Station , MA, and Seabrook Station Nuclear Power Plant , NH); 13 new reactor combined construction permit and operating license decisions ( Bellefonte Nuclear Generating Station Units 3 and 4, AL, Bell Bend, Callaway Nuclear Generating Station , MO, Calvert Cliffs Nuclear Generating Station , MD, Comanche Peak Nuclear Power Plant , TX, Enrico Fermi Nuclear Generating Station , MI, Levy County Nuclear Power Plant , FL North Anna Nuclear Generating Station , VA, Shearon Harris Nuclear Power Plant , NC, South Texas Nuclear Generating Station , TX, Turkey Point Nuclear Generating Station , FL, Alvin W. Vogtle Electric Generating Plant , GA, and William States Lee III Nuclear Generating Station , SC);a construction permit decision (Bellefonte Units 1 and 2); and an operating license decision ( Watts Bar Nuclear Generating Station , TN). In addition,

14280-417: The public for years about the severity of the flooding. Boiling water reactor A boiling water reactor ( BWR ) is a type of light water nuclear reactor used for the generation of electrical power. It is the second most common type of electricity-generating nuclear reactor after the pressurized water reactor (PWR), which is also a type of light water nuclear reactor. The main difference between

14416-479: The pumps could be repaired during the next refueling outage. Instead, the designers of the simplified boiling water reactor used thermal analysis to design the reactor core such that natural circulation (cold water falls, hot water rises) would bring water to the center of the core to be boiled. The ultimate result of the passive safety features of the SBWR would be a reactor that would not require human intervention in

14552-468: The rate at which a nuclear reactor will shut down. Due to flaws in its original control rod design, scramming an RBMK reactor could raise reactivity to dangerous levels before lowering it. This was noticed when it caused a power surge at the startup of Ignalina Nuclear Power Plant Unit number 1, in 1983. On April 26, 1986, the Chernobyl disaster happened due to a fatally flawed shutdown system, after

14688-510: The rates of thermal expansion of the fuel. PCIOMR rules also limit the maximum local power change (in kW/ft*hr), prevent pulling control rods below the tips of adjacent control rods, and require control rod sequences to be analyzed against core modelling software to prevent pellet-clad interactions. PCIOMR analysis look at local power peaks and xenon transients which could be caused by control rod position changes or rapid power changes to ensure that local power rates never exceed maximum ratings. For

14824-519: The reactor (or section(s) thereof) are not below the shutdown margin (that is, they could return to a critical state due to insertion of positive reactivity from cooling, poison decay, or other uncontrolled conditions), the operators can inject solutions containing neutron poisons directly into the reactor coolant. Neutron poison solutions are water-based solutions that contain chemicals that absorb neutrons, such as common household borax , sodium polyborate , boric acid , or gadolinium nitrate , causing

14960-472: The reactor core passes through steam separators and dryer plates above the core and then directly to the turbine , which is part of the reactor circuit. Because the water around the core of a reactor is always contaminated with traces of radionuclides due to neutron capture from the water, the turbine must be shielded during normal operation, and radiological protection must be provided during maintenance. The increased cost related to operation and maintenance of

15096-456: The reactor was estimated to be between 10 and 10 (i.e., one core damage accident per every 10,000 to 10,000,000 reactor years). Steam exiting the turbine flows into condensers located underneath the low-pressure turbines, where the steam is cooled and returned to the liquid state (condensate). The condensate is then pumped through feedwater heaters that raise its temperature using extraction steam from various turbine stages. Feedwater from

15232-401: The reactor's heatup and cooldown rates while steaming is still in progress. Reactor water level is controlled by the main feedwater system. From about 0.5% power to 100% power, feedwater will automatically control the water level in the reactor. At low power conditions, the feedwater controller acts as a simple PID control by watching reactor water level. At high power conditions, the controller

15368-414: The reactor. The high-pressure turbine exhaust passes through a steam reheater which superheats the steam to over 400 degrees F (204.4 degrees celcius) for the low-pressure turbines to use. The exhaust of the low-pressure turbines is sent to the main condenser. The steam reheaters take some of the turbine's steam and use it as a heating source to reheat what comes out of the high-pressure turbine exhaust. While

15504-562: The reheaters take steam away from the turbine, the net result is that the reheaters improve the thermodynamic efficiency of the plant. A modern BWR fuel assembly comprises 74 to 100 fuel rods , and there are up to approximately 800 assemblies in a reactor core , holding up to approximately 140 short tons of low-enriched uranium . The number of fuel assemblies in a specific reactor is based on considerations of desired reactor power output, reactor core size and reactor power density. A modern reactor has many safety systems that are designed with

15640-479: The response of LHGR in transient imagine the rapid closure of the valves that admit steam to the turbines at full power. This causes the immediate cessation of steam flow and an immediate rise in BWR pressure. This rise in pressure effectively subcools the reactor coolant instantaneously; the voids (vapor) collapse into solid water. When the voids collapse in the reactor, the fission reaction is encouraged (more thermal neutrons); power increases drastically (120%) until it

15776-417: The restart of a reactor, these systems are only used to shut down the reactor if control rod insertion fails. This concern is especially significant in a BWR, where injection of liquid boron would cause precipitation of solid boron compounds on fuel cladding, which would prevent the reactor from restarting until the boron deposits were removed. In most reactor designs, the routine shutdown procedure also uses

15912-649: The resulting design to a larger size of 1,600  MWe (4,500 MWth). This Economic Simplified Boiling Water Reactor (ESBWR) design was submitted to the US Nuclear Regulatory Commission for approval in April 2005, and design certification was granted by the NRC in September 2014. Reportedly, this design has been advertised as having a core damage probability of only 3×10 core damage events per reactor-year. That is, there would need to be 3 million ESBWRs operating before one would expect

16048-487: The risk posed to the nation by approximately two orders of magnitude (i.e., the true risk is about 100 greater than NRC represented). These findings were corroborated in a final report prepared by a special Washington State Legislature Nuclear Power Task Force, titled, "Doesn't NRC Address Consequences of Severe Accidents in EISs for re-licensing?" In Vermont, the day before the 2011 Tōhoku earthquake and tsunami that damaged Japan's Fukushima Daiichi Nuclear Power Plant ,

16184-464: The so-called "100% rod line", power may be varied from approximately 30% to 100% of rated power by changing the reactor recirculation system flow by varying the speed of the recirculation pumps or modulating flow control valves. As flow of water through the core is increased, steam bubbles ("voids") are more quickly removed from the core, the amount of liquid water in the core increases, neutron moderation increases, more neutrons are slowed to be absorbed by

16320-695: The spent fuel rods were being put too quickly into the spent storage pool and that the number of rods in the pool exceeded specifications. Management ignored him, so he went directly to the NRC, which eventually admitted that it knew of both of the forbidden practices, which happened at many plants, but chose to ignore them. The whistleblower was fired and blacklisted. Terrorist attacks such as those executed by al-Qaeda on New York City and Washington, D.C. , on September 11, 2001 , and in London on July 7, 2005 , have prompted fears that extremist groups might use radioactive dirty bombs in further attacks in

16456-440: The success of the ABWR in Japan is that General Electric's nuclear energy division merged with Hitachi Corporation's nuclear energy division, forming GE Hitachi Nuclear Energy , which is now the major worldwide developer of the BWR design. Parallel to the development of the ABWR, General Electric also developed a different concept, known as the simplified boiling water reactor (SBWR). This smaller 600 megawatt electrical reactor

16592-461: The turbine and incorporated heat exchangers for the generation of secondary steam to drive separate parts of the turbines. The literature does not indicate why this was the case, but it was eliminated on production models of the BWR. The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR: the torus (used to quench steam in

16728-515: The two feedwater pumps fails during operation, the feedwater system will command the recirculation system to rapidly reduce core flow, effectively reducing reactor power from 100% to 50% in a few seconds. At this power level a single feedwater pump can maintain the core water level. If all feedwater is lost, the reactor will scram and the Emergency Core Cooling System is used to restore reactor water level. Steam produced in

16864-418: The two-phase flow in cyclone separators, the steam is separated and rises upwards towards the steam dryer while the water remains behind and flows horizontally out into the downcomer or annulus region. In the downcomer or annulus region, it combines with the feedwater flow and the cycle repeats. The saturated steam that rises above the separator is dried by a chevron dryer structure. The "wet" steam goes through

17000-440: The utility seeking an operating license for the controversial Seabrook plant . In the late 1980s, the NRC 'created a policy' of non-enforcement by asserting its discretion not to enforce license conditions; between September 1989 and 1994, the 'NRC has either waived or chosen not to enforce regulations at nuclear power reactors over 340 times'. Finally, critics charge that the NRC has ceded important aspects of regulatory authority to

17136-478: The water flow (see below). Some early BWRs and the proposed ESBWR (Economic Simplified BWR made by General Electric Hitachi) designs use only natural circulation with control rod positioning to control power from zero to 100% because they do not have reactor recirculation systems. Changing (increasing or decreasing) the flow of water through the core is the normal and convenient method for controlling power from approximately 30% to 100% reactor power. When operating on

17272-546: The world without refueling, sent their man in engineering, Captain Hyman Rickover to run their nuclear power program. Rickover decided on the PWR route for the Navy, as the early researchers in the field of nuclear power feared that the direct production of steam within a reactor would cause instability, while they knew that the use of pressurized water would definitively work as a means of heat transfer. This concern led to

17408-459: Was even performed. Such a conclusion is scientifically indefensible given the experience of the Three Mile Island , Chernobyl , and Fukushima accidents. Another finding was that NRC had concealed the risk posed to the public at large by disregarding one of the most important EIS requirements, mandating that cumulative impacts be assessed (40 Code of Federal Regulations §1508.7). By disregarding this basic requirement, NRC effectively misrepresented

17544-493: Was having with several members of his group," Nyer wrote. "The group had decided to have a big button to push to drive in both the control rods and the safety rod. What to label it? 'What do we do after we punch the button?,' someone asked. 'Scram out of here!,' Wilson said. Bill Overbeck, another member of that group said, 'OK I'll label it SCRAM.'" The earliest references to "scram" among the Chicago Pile team were also associated with Wilson's shutdown circuitry and not Hilberry. In

17680-407: Was notable for its incorporation—for the first time ever in a light water reactor —of " passive safety " design principles. The concept of passive safety means that the reactor, rather than requiring the intervention of active systems, such as emergency injection pumps, to keep the reactor within safety margins, was instead designed to return to a safe state solely through operation of natural forces if

17816-599: Was perceived as unduly favoring the industry it was charged with regulating. The NRC was formed as an independent commission to oversee nuclear energy matters, oversight of nuclear medicine , and nuclear safety and security . The U.S. AEC became the Energy Research and Development Administration (ERDA) in 1975, responsible for development and oversight of nuclear weapons . Research and promotion of civil uses of radioactive materials, such as for nuclear non-destructive testing , nuclear medicine, and nuclear power ,

17952-469: Was shut down on December 29, 2014. New York state eventually closed Indian Point Energy Center , in Buchanan, 30 miles from New York City, on April 30, 2021. In 2019 the NRC approved a second 20-year license extension for Turkey Point units 3 and 4, the first time NRC had extended licenses to 80 years total lifetime. Similar extensions for about 20 reactors are planned or intended, with more expected in

18088-1087: Was split into the Office of Nuclear Energy, Science & Technology within ERDA by the same act. In 1977, ERDA became the United States Department of Energy (DOE). In 2000, the National Nuclear Security Administration was created as a subcomponent of DOE, responsible for nuclear weapons. Following the Fukushima nuclear disaster in 2011, the NRC developed a guidance strategy known as "Diverse and Flexible Coping Strategies (FLEX)" which requires licensee nuclear power plants to account for beyond-design-basis external events (seismic, flooding, high-winds, etc.) that are most impactful to reactor safety through loss of power and loss of ultimate heat sink. FLEX Strategies have been implemented at all operating nuclear power plants in

18224-484: Was supposedly coined by Enrico Fermi when he oversaw the construction of the world's first nuclear reactor . The core , which was built under the spectator seating at the University of Chicago's Stagg Field , had an actual control rod tied to a rope with a man with an axe standing next to it; cutting the rope would mean the rods would fall by gravity into the reactor core, shutting the reactor down. The axe man at

18360-439: Was the omission of recirculation pumps within the core; these pumps were used in other BWR designs to keep cooling water moving; they were expensive, hard to reach to repair, and could occasionally fail; so as to improve reliability, the ABWR incorporated no less than 10 of these recirculation pumps, so that even if several failed, a sufficient number would remain serviceable so that an unscheduled shutdown would not be necessary, and

18496-399: Was withdrawn prior to approval; still, the concept remained intriguing to General Electric's designers, and served as the basis of future developments. During a period beginning in the late 1990s, GE engineers proposed to combine the features of the advanced boiling water reactor design with the distinctive safety features of the simplified boiling water reactor design, along with scaling up

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