NUREG-1150

Last updated April 26, 2019

NUREG-1150 "Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants", published December 1990 by the Nuclear Regulatory Commission (NRC) is a follow-up to the WASH-1400 and CRAC-II safety studies that employs the methodology of plant-specific Probabilistic Risk Assessment (PRA). The research team, led by Denwood Ross, Joseph Murphy, and Mark Cunningham, concluded that the current generation of nuclear power plants exceeded NRC safety goals.

Nuclear Regulatory Commission United States government agency

The 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 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.

WASH-1400, 'The Reactor Safety Study', was a report produced in 1975 for the Nuclear Regulatory Commission by a committee of specialists under Professor Norman Rasmussen. It "generated a storm of criticism in the years following its release". In the years immediately after its release, WASH-1400 was followed by a number of reports that either peer reviewed its methodology or offered their own judgments about probabilities and consequences of various events at commercial reactors. In at least a few instances, some offered critiques of the study's assumptions, methodology, calculations, peer review procedures, and objectivity. A succession of reports, including NUREG-1150, the State-of-the-Art Reactor Consequence Analyses and others, have carried-on the tradition of PRA and its application to commercial power plants.

CRAC-II is both a computer code and the 1982 report of the simulation results performed by Sandia National Laboratories for the Nuclear Regulatory Commission. The report is sometimes referred to as the CRAC-II report because it is the computer program used in the calculations, but the report is also known as the 1982 Sandia Siting Study or as NUREG/CR-2239. The computer program MACCS2 has since replaced CRAC-II for consequences of radioactive release.

"This study was a significant turning point in the use of risk-based concepts in the regulatory process and enabled the NRC to greatly improve its methods for assessing containment performance after core damage and accident progression." However significant, and sometimes unrealistic, conservatisms were applied in this study and it is (as of 2006 [update] ) being replaced with a new state-of-the-art study entitled State-of-the-Art Reactor Consequence Analyses(see below).

The State-of-the-Art Reactor Consequence Analyses (SOARCA) is a study of nuclear power plant safety conducted by the Nuclear Regulatory Commission. The purpose of the SOARCA is assessment of possible impact on population caused by major radiation accidents that might occur at NPPs. This new study updates older studies with the latest state-of-the-art computer models and incorporates new plant safety and security enhancements.

Results

Results of NUREG-1150 (page 12-3):

Average probability of an individual early fatality per reactor per year:

NRC Safety Goal: 5 x 10⁻⁷
Typical pressurized water reactor (PWR): 2 x 10⁻⁸
Typical boiling water reactor (BWR): 5 x 10⁻¹¹

Pressurized water reactors (PWRs) constitute the large majority of the world's nuclear power plants and are one of three types of light water reactor (LWR), the other types being boiling water reactors (BWRs) and supercritical water reactors (SCWRs). In a PWR, the primary coolant (water) is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms. The heated water then flows to a steam generator where it transfers its thermal energy to a secondary system where steam is generated and flows to turbines which, in turn, spin an electric generator. In contrast to a boiling water reactor, pressure in the primary coolant loop prevents the water from boiling within the reactor. All LWRs use ordinary water as both coolant and neutron moderator.

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 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 water system, which turns to steam and drives the turbine. The BWR was developed by the Argonne National Laboratory and General Electric (GE) in the mid-1950s. The main present manufacturer is GE Hitachi Nuclear Energy, which specializes in the design and construction of this type of reactor.

Average probability of an individual latent cancer death per reactor per year:

NRC Safety Goal: 2 x 10⁻⁶
Typical PWR: 2 x 10⁻⁹
Typical BWR: 4 x 10⁻¹⁰

Using the data on pages 3–5, 3-7, 4-5 and 4-7 the probability of some U.S. plant having core damage is about 30% over 20 years - this number doesn't include containment failure, which is conservatively estimated at 8% for PWRs (page 3-13, weighting by the probabilities at the bottom) and 84% for BWRs (page 4-14, same technique). Assuming that the 104 current-design (2005) U.S. plants are similar to the two "typical" plants, the chance of a major release of radiation is under 8% every 20 years.

The typical BWR was the Peach Bottom plant and the typical PWR was the Surry plant.

Peach Bottom Atomic Power Station, a nuclear power plant, is located 50 miles (80 km) southeast of Harrisburg in Peach Bottom Township, York County, Pennsylvania, on the Susquehanna River three miles north of the Maryland border.

Parts of NUREG-1150 were compiled by Sandia National Laboratories, which continues to do such research.

The Sandia National Laboratories (SNL), managed and operated by the National Technology and Engineering Solutions of Sandia, is one of three National Nuclear Security Administration research and development laboratories. In December 2016, it was announced that National Technology and Engineering Solutions of Sandia, under the direction of Honeywell International, will take over the management of Sandia National Laboratories starting on May 1, 2017.

NUREG-1420 contains the Kouts' Committee peer review of NUREG-1150.

NRC disclaimer of CRAC-II and NUREG-1150

The NRC, which initially conducted the NUREG-1150 study, has issued the following statement:

The U.S. Nuclear Regulatory Commission has devoted considerable research resources, both in the past and currently, to evaluating accidents and the possible public consequences of severe reactor accidents. The NRC's most recent studies have confirmed that early research into the topic led to extremely conservative consequence analyses that generate invalid results for attempting to quantify the possible effects of very unlikely severe accidents. According to the NRC, these previous studies did not reflect current plant design, operation, accident management strategies or security enhancements. They often used unnecessarily conservative estimates or assumptions concerning possible damage to the reactor core, the possible radioactive contamination that could be released, and possible failures of the reactor vessel and containment buildings. These previous studies also failed to realistically model the effect of emergency preparedness. The NRC staff is currently pursuing a new, state-of-the-art assessment of possible severe accidents and their consequences.

External links

Related Research Articles

Fault tree analysis (FTA) is a top-down, deductive failure analysis in which an undesired state of a system is analyzed using Boolean logic to combine a series of lower-level events. This analysis method is mainly used in the fields of safety engineering and reliability engineering to understand how systems can fail, to identify the best ways to reduce risk or to determine event rates of a safety accident or a particular system level (functional) failure. FTA is used in the aerospace, nuclear power, chemical and process, pharmaceutical, petrochemical and other high-hazard industries; but is also used in fields as diverse as risk factor identification relating to social service system failure. FTA is also used in software engineering for debugging purposes and is closely related to cause-elimination technique used to detect bugs.

Hope Creek Nuclear Generating Station is a thermal nuclear power plant located in Lower Alloways Creek Township, in Salem County, New Jersey, United States, on the same site as the two-unit Salem Nuclear Power Plant. The plant is owned and operated by PSEG Nuclear LLC. It has one unit, a boiling water reactor (BWR) manufactured by GE. The complex was designed for two units, but the second unit was cancelled in 1981. It has a generating capacity of 1,268 MWe. The plant came online on July 25, 1986, licensed to operate until 2026. In 2009, PSEG applied for a 20-year license renewal, which it received in 2011. With its combined output of 3,572 megawatts, the Salem-Hope Creek complex is the largest nuclear generating facility in the Eastern United States and the second largest nationwide.

Perry Nuclear Generating Station nuclear power plant

The Perry Nuclear Power Plant is located on a 1,100-acre (450 ha) site on Lake Erie, 40 miles (65 km) northeast of Cleveland in North Perry, Ohio, US. The nuclear power plant is owned by First Energy Nuclear Operating Corporation.

Surry Power Station is a nuclear power plant located in Surry County in southeastern Virginia, in the South Atlantic United States. The power station lies on an 840-acre (340 ha) site adjacent to the James River across from Jamestown, slightly upriver from Smithfield and Newport News. Surry is operated by Dominion Generation and owned by Dominion Resources, Inc.

Probabilistic risk assessment (PRA) is a systematic and comprehensive methodology to evaluate risks associated with a complex engineered technological entity or the effects of stressors on the environment for example.

A containment building, in its most common usage, is a reinforced steel or lead structure enclosing a nuclear reactor. It is designed, in any emergency, to contain the escape of radioactive steam or gas to a maximum pressure in the range of 275 to 550 kPa. The containment is the fourth and final barrier to radioactive release, the first being the fuel ceramic itself, the second being the metal fuel cladding tubes, the third being the reactor vessel and coolant system.

Advanced boiling water reactor type of generation III nuclear reactor

The advanced boiling water reactor (ABWR) is a Generation III boiling water reactor. The ABWR is currently offered by GE Hitachi Nuclear Energy (GEH) and Toshiba. The ABWR generates electrical power by using steam to power a turbine connected to a generator; the steam is boiled from water using heat generated by fission reactions within nuclear fuel. Kashiwazaki-Kariwa unit 6 is considered the first Generation III reactor in the world.

The Price-Anderson Nuclear Industries Indemnity Act is a United States federal law, first passed in 1957 and since renewed several times, which governs liability-related issues for all non-military nuclear facilities constructed in the United States before 2026. The main purpose of the Act is to partially compensate the nuclear industry against liability claims arising from nuclear incidents while still ensuring compensation coverage for the general public. The Act establishes a no fault insurance-type system in which the first approximately $12.6 billion is industry-funded as described in the Act. Any claims above the $12.6 billion would be covered by a Congressional mandate to retroactively increase nuclear utility liability or would be covered by the federal government. At the time of the Act's passing, it was considered necessary as an incentive for the private production of nuclear power — this was because electric utilities viewed the available liability coverage as inadequate.

The AP1000 is a nuclear power plant designed and sold by Westinghouse Electric Company. The plant is a pressurized water reactor with improved use of passive nuclear safety. The first AP1000 began operations in China at Sanmen Nuclear Power Station, where Unit 1 became the first AP1000 to achieve criticality in June 2018.

WASH-740 was a report published by the U.S. Atomic Energy Commission (USAEC) in 1957. This report, called "Theoretical Possibilities and Consequences of Major Accidents in Large Nuclear Power Plants", estimated maximum possible damage from a meltdown with no containment building at a large nuclear reactor.

The Economic Simplified Boiling Water Reactor (ESBWR) is a passively safe generation III+ reactor design derived from its predecessor, the Simplified Boiling Water Reactor (SBWR) and from the Advanced Boiling Water Reactor (ABWR). All are designs by GE Hitachi Nuclear Energy (GEH), and are based on previous Boiling Water Reactor designs.

Nuclear safety in the United States is governed by federal regulations issued by the Nuclear Regulatory Commission (NRC). The NRC regulates all nuclear plants and materials in the United States except for nuclear plants and materials controlled by the U.S. government, as well those powering naval vessels.

The McGuire Nuclear Station is a nuclear power plant located about 17 miles (27 km) northwest of Charlotte, North Carolina, on the state's largest lake, Lake Norman. It is a 32,500-acre (13,200 ha) lake created in 1963 by Duke Power for the Cowans Ford Hydroelectric Station. The McGuire units use the lake's water for cooling.

Core damage frequency (CDF) is a term used in probabilistic risk assessment (PRA) that indicates the likelihood of an accident that would cause severe damage to a nuclear fuel in nuclear reactor core. Core (severe) damage accidents are considered extremely serious because severe damage to the fuel in the core prevents adequate heat removal or even safe shutdown, which can lead to a nuclear meltdown. Some sources on CDF consider core (severe) damage and core meltdown to be the same thing, and different methods of measurement are used between industries and nations, so the primary value of the CDF number is in managing the risk of core accidents within a system and not necessarily to provide large-scale statistics.

General Electric's BWR product line of Boiling Water Reactors represents the designs of a large percentage of the commercial fission reactors around the world.

References

Direct correspondence with the NRC via Scott Burnell ( srb3@nrc.gov )

This page is based on this Wikipedia article
Text is available under the CC BY-SA 4.0 license; additional terms may apply.
Images, videos and audio are available under their respective licenses.