Safety Research: 1955-1978

Reactor Safety: 1953-1978

It is important to note that the following studies were initiated before TMI and Chernobyl ever occurred. It is no coincidence that Chernobyl did not happen in the U.S. Reactor safety has been the motivation behind the design of safety and containment systems ever since the first power reactors were built in the last 1950's. Large-scale accidents were assumed to be possible and therefore the siting and design criteria for plants should reflect this fact.

Reactor safety research began with the first nuclear reactors. The following is a summary of the more important tests and studies that were performed in this time period.

BORAX: 1953

One of the first major tests was performed at the Idaho National Engineering Laboratory. The BORAX test demonstrated that when the reactor core loses water, the reactor automatically shuts itself down. This phenomena is observed because water is both the coolant and the moderator in the reactor therefore when the water is lost the amount of thermal neutrons present decreases. During the test, the BORAX 1 reactor was deliberately destroyed through large reactivity insertion. BORAX was the first test that showed that a light-water moderated reactor could not explode like a bomb because the expulsion of cooling water during the excursion would stop the fission reaction before it reached an explosive rate.

The BORAX tests were performed at atmospheric pressure. As all power reactors operate at high pressure, a similar series of tests were performed at high pressure. In 1955, a subsequent series of tests named SPERT (Special Power Excursion Reactor Test) confirmed that a nuclear explosion could not occur in power reactors. The SPERT reactors (there were 3 in all) were pressurized and were subjected to large step and ramp reactivity excursions to assess the kinetic behavior of reactors. These studies provided a tremendous amount of data on reactor stability, feedback effects, and transient behavior.

WASH-740: 1955

This study was an effort to use the available mathematical tools to estimate the costs to public health cost of a severe nuclear power plant accident. The estimates were based on the assumptions that: 1) the reactor would have no containment, and 2) the fuel would vaporize into small micron-size pellets that would enhance their dispersal. Among other items: this report concluded that the chance of a major nuclear accident in a large (500MW) plant is between 1 in 100000 and 1 in 1 billion, and that the likelihood of a member of the public being killed by such an accident is 1 in 50 million.

In a related experiment, approximately 1 pound of irradiated fuel was melted in order to estimate release fractions of radioisotopes (Creek and Parker). In the Containment Series Experiments, tests were performed to assess the impact of a steam (non-nuclear) explosion caused by a pipe break. Designs of coolant loop pipes were based upon these results.

KEWB: 1956

The Kinetic Experiments on Water Boilers (KEWB) project began in 1954 with the objective of obtaining and evaluating data on the dynamic behavior of aqueous homogeneous reactors. The core of this reactor was a solution of uranyl sulfate and had a power level of 50 kW. Although the tests were successful and much information was obtained, homogeneous reactors never became commercially viable.

Fuel Meltdown Tests: 1956-1959

At Oak Ridge National Laboratory (Tennessee) and elsewhere, the Atomic Energy Commission contracted varous private companies to conduct fuel meltdown tests. Meltdown studies with zirconium-uranium assemblies in a steam atmosphere were conducted by MSA Research corporation to provide data on fission product release, melt characteristics, and hydrogen formation. Some of the major results were:

  • the major fission products released are xenon and iodine with as much as 60% and 25% escpaing, respectively,
  • actual fission product release was greatly reduced by condensation in the reactor vessel,
  • core meltdown in a steam atmosphere results in considerable production of hydrogen gas, but that the possibility of a hydrogen explosion is small due, in part, to the lack of oxygen.

    In another test at INEL in 1958, a series of meltdown studies with field releases of fission products were conducted to determine release quantities, airborne activity, and ground deposition of fission products.

    In the TREAT experiments (Transient Reactor Test Facility) of 1959, a special reactorwas built to study nuclear fuel melting in thermal reactors. Although primarily suited to fast-reactor studies, the facility was also used for general meltdown research. FLECHT: 1970 - 1985

    Performed by GE and Westinghouse, to assess the ability of emergency core cooling systems to cool the fuel pins and cladding during emergency conditions. It was demonstrated that the ECCS is designed to provide sufficient coolant to the core under most emergency conditions.

    Power Burst Test: 1972

    Built at the Idaho National Engineering Laboroatory (INEL), this facility was designed for the purpose of testing fuel and cladding behavior under normal and abnormal conditions to determine the margins of safety. The fuel and cladding were tested to determine failure and melting points for the materials and conditions previously unidentified which would impair fuel integrity.

    BWR Blowdown/ECC Program (BD/ECC): 1975-1981

    This program used the TLTA to investigate the effect of BWR/6 scaling and an 8x8 simulated fuel bundle on the LOCA blowdown response. As the name suggests, tests were also run to simulate complete LBLOCA transients, including the effect of Emergency Core Coolant injection. Two small break LOCA tests were also conducted, as well as a core uncovery under slow loss of coolant (boiloff) transient.

    Reactor Safety Study: An Assessment of Accident Risk in U.S. Commercial Nuclear Power Plants (WASH-1400): 1975

    The Reactor Safety Study was the first comprehensive evaluation of the risk due to all accidents that were thought at the time to be possible in U.S. light water nuclear power plants. The advances in probabilistic risk assessment (PRA) that grew out of this study (also known as the Rassmussen Report) are being applied more and more to this day. The goals of the WASH-1400 study were to:

  • perform a realistic assessment of risk to the public from reactor accidents,
  • develop methodological approaches for performing the assessments and to understand their limitations,
  • provide and independent check on the effectiveness of reactor safety practice of industry and government and identify areas for future safety research.

    A detailed description of the Reactor Safety Study is beyond the scope of this document (please refer to the references), however, the major conclusions can be summarized as follows:

  • The most likely core meltdown accident has modest consequences to the public.
  • Reactor accidents have consequences which are no larger, and often much smaller, than those to which the population is already exposed.
  • The frequency of reactor accidents is smaller than that of most other accidents which have similar consequences.

    The results are applicable to a 100-LWR population (The U.S. has about 110 reactors in operation) and cannot be applied to the RBMK design.

    The WASH-1400 study was subject to an extensive critical review by the American Physical Society, the U.S. Encironemntal Protection Agency, the Electric Power Research Institute, the Uniuon of Concerned Scientists (a group opposed to nuclear power), and teh Risk Assessment Review Group (RARG). The RARG group relied much on the results and testimony of the other three critiques. The NRC charged RARG with the task of conducting an independent review that would clarify the achievements and limitations and recommend the proper uses of the PRA methodology in the regulatory process. The RARG report (also known as the Lewis report) concluded that the WASH-1400 study was overall a 'conscientious and honest effort", a substantial advance for quantitative analysis of reactor safety, and a sound methodology that should be used by the NRC.

    The Lewis report was critical of the Executive Summary of the WASH-1400 report for not adequately indicating the full extent of the consequences of, and the uncertainties in the probabilities for, reactor accidents. The Lewis group maintained that the uncertainty in the probabilities was higher than that suggested in the Summary, but that the probabilities themselves were accurate. For this reason, the NRC withdrew endorsement of the Summary although it did not repudiate the study itself. The numerical results are NOT regarded as optimistic and most likley are quite pessimistic. The response of the industry has been to improve the PRA techniques and reduce the uncertainty.

    One important finding of the study was that transients, small LOCAs, and human errors can make important contributions to overall risk. Both the LEWIS and EPRI reviews regarded this as being inadequately refleted in NRC research and regulatory policies. Since then, more emphasis has been assigned to addressing these factors.

    Loss-of-Fluid Test (LOFT): 1978 - 1980

    Also built at INEL, this facility was originally designed to be a single test to melt a PWR core and to study, among other things, fission-product behavior in the vessel, containment building, and the atmosphere.

    Prior to commencement of the LOFT tests in 1978, a series of blowdown tests were performed on the facility to test the various safety features. The SEMISCALE series began in 1963 and involved heated loops but no fuel, and took many years to complete.

    The development of larger LWR's required more information on the performance of Emergency Core Cooling Systems (ECCS) and other engineered safety systems, thus the LOFT program was expanded. The facility was a 50MW PWR core with scaled coolant loops and extensive instrumentation. To conduct small and large break loss-of-coolant accidents (LOCAs), one of the coolant loops was designed with passive pump and steam generator components plus a supression vessel to hold the effluent and to simulate containment building back pressure.

    The LOFT facility was used to conduct the first large break LOCA, as well as a number of small break LOCAs, operational transient events, and a final severe fuel damage test. A total of 36 experiments were conducted under the auspices of the NRC and the Organization for Economic Cooperation and Development (OECD). Some of the major conclusions of these tests included:

  • core thermal response in large break LOCAs is much less severe than initially projected,
  • ECCS design is effective in core protection over the range of LOCA break sizes
  • both two-phase natural circulation and primary system feed-and-bleed areeeffective in removing core decay heat
  • in anticipated-transient-without-SCRAM (ATWS) events, the core goes sub-critical and pressure-relief capacity is adequate.

    The LOFT results were also used extensively to valaidate computer codes used for predicting accident behavior.

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