Report to the Congress from the Presidential Commission on Catastrophic Nuclear Accidents


Appendix B
The Nature of Severe Nuclear Accidents

I.   Introduction

    In October 1975, the NRC issued the final results of a 3-year study of the risks from postulated accidents during the operation of nuclear power reactors of the type used in the United States. This study had been initiated in 1972 by the Atomic Energy Commission. This report. known as The Reactor Safety Study (RSS) or by its document number, WASH 1400, was the first comprehensive study that attempted to quantify a variety of health risks and economic risks associated with power reactor accidents. Since that time, about 40 reactors have been analyzed using the same general methodology as WASH 1400 but with considerably improved computer codes and experience data. The most recent and the most detailed of these studies has been the effort undertaken by the NRC to analyze five different reactors using the very latest methodology and experience data available. It is interesting that two of the five reactors analyzed are the same two analyzed in WASH 1400. In June 1989, the second draft of this work, “Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,” or NUREG 1150, was issued for public comment. All the quantitative results referred to below come from these two reports.

    There is widely held belief among reactor safety analysts that the risks of severe nuclear accidents are small. This conclusion rests in part upon probabilistic analysis of the type discussed above. This appendix describes the sources of the risk and its magnitude as determined in these two studies.

II.  The Source of Risk

    During full power operation, a nuclear power reactor generates a large amount of radioactivity. Most of this radioactivity consists of fission products, resulting from the fission process, which are produced inside the reactor fuel. The fuel is uranium dioxide, a ceramic material that melts at about 5,000F. The fuel effectively contains the radioactive fission products unless it is heated to the melting point. At temperatures in this range, essentially all the gaseous forms of radioactivity will be released from the fuel. In addition, some of the more volatile forms of the solid fission products may be released as fine aerosols. If either of these forms were to be released into the atmosphere, they could be spread by the prevailing winds.

    Design requirements for U.S. plants mandate that plants have systems to contain any radioactivity accidentally released from the fuel. The main system for accomplishing this is the containment building, an air-tight structure that surrounds the reactor. In addition, all reactors have a system for removing aerosols from the containment atmosphere. in many reactors this system consists of a water spray that can create the equivalent of a heavy rainstorm inside the containment. Boiling water reactors (BWRs) accomplish this function by passing any released gases through a pool of water.

    The principal goal of the reactor safety philosophy is to prevent the accidental release of radioactivity. This is accomplished by designing a system in which the chance of overheating the fuel accidentally is low. As a back-up, systems are added that prevent the release of radioactivity to the atmosphere even if it were released from the fuel. Despite these efforts one can always postulate ways in which these systems might fail to prevent the accidental release of radioactivity. It is the task of probabilistic safety assessment (PSA) to identify how this might happen, to determine how likely it is to happen, and finally to determine the health effects and economic impacts of the radioactive releases upon the public.

III.  Probabilistic Safety Assessment

    All PSAs begin by developing the causes and likelihood of heating the fuel to its melting point due to either external (earthquakes, floods, tornadoes, etc.) or internal causes. This analysis involves developing a logical relationship between the failures of plant components and operators and the failure of system safety function. The result of this analysis is an estimate of the probability of accidentally melting the fuel, a condition often called “core melt.” Of the plants analyzed thus far, most have an estimated likelihood of core melt of between I in 10,000 and I in 100,000 per plant year. The core melt frequency for the Surry and the Peach Bottom reactors from WASH 1400 and NUREG 1150 are given in Figures B.1 and B.2, taken from NUREG 1150. It is interesting to note that despite all the progress we have made in PSA techniques, the answers are quite close and, in fact, the uncertainty ranges overlap.

    The second step in a PSA analysis is to determine the type and amount of radioactivity that might be released in the different accidents identified. Figures B.3 and BA compare the amounts of radioactivity calculated to be released for several similar accidents for Surry and Peach Bottom. These fractions of the various types of radioactivity released are called the “source terms” for the accident. As can be seen in these figures, WASH 1400 in most cases had significantly larger source terms than NUREG 1150. The lower values of NUREG 1150 are the result of new information gained from major research in the United States, Japan, and Western Europe. These experiments and the measurements at Three Mile Island confirm that the values used in WASH 1400 were too high.

    The final step in a PSA is to calculate the effects on public health and the economic impact of any radioactivity released in the accident. Sophisticated computer models have been developed to do this calculation. These models require input of the source terms, the population density around the site, and weather data for recent years from the plant site. The code then calculates thousands of cases to generate curves that give the magnitude of given risks versus their probabilities.

    The result of the WASH 1400 calculation for early fatalities is given in Figure B.5. Figure B.6 gives the curves for the latent cancer fatalities at both plants. The pressurized water reactor (PWR) curve is for the Surry Plant and the BVVR curve is for the Peach Bottom plant. The dotted lines show the results of the NUREG 1150 calculation for each plant. Such curves are generated for early injuries as well as for latent effects including cancer, thyroid nodules and genetic effects. Note that the curve in Figure 13.5 gives the frequency in units per reactor year for events of a given size or greater. For example, using the average curve, we could say that once in 10 million years of plant operation we would expect an accident that produced 100 or more early fatalities. Note that the curves in Figure B.5 have a wide range of consequences from quite small at high frequencies to quite large at very low frequencies. Curves of this shape are typical of all accidents where a number of factors affect the magnitude of the event. Thus, highway accidents in which 1 or 2 are killed will be much more frequent that those in which, say, 5 or 10 are killed.

IV.  Catastrophic Accidents

    This Commission was charged with studying how the United States should cope with a catastrophic nuclear accident. Clearly this refers to accidents of low probability near the high consequence end of the scale. These extreme accidents come about only if the various factors affecting the magnitude of the consequences all are in nearly their worst states. Thus, for example, the core must melt, then the containment must fail above ground level, the wind must be blowing toward an area of relatively high population, inversion conditions must prevail, and civil protection efforts must fail to be effective. This is what is required to produce one of the largest consequence events on the curves. As you can see from Figure B.5, such events have a probability of about I in I billion per reactor year. This very low frequency is the result of the approximate probability of the following five factors:

PROBABILITY OF CORE MELT

1 in 10,000 per plant year

PROBABILITY OF CONTAINMENT FAILURE

1 In 100

PROBABILITY OF BAD WIND DIRECTION

1 In 10

PROBABILITY OF INVERSION CONDITION

1 In 10

PROBABILITY OF EVACUATION FAILURE

1 In 10

The product of these possibilities is 1 in 1 billion

    In the United States, we are currently operating about 100 reactors, so the frequency of an extreme event is about 100 times I/billion or about I in 10 million per year. Thus an extreme event would be expected about once every 10 million years on the average. Of course, this assumes nothing will be done to make plants safer in the 10 million years. Clearly, such a calculation is not very meaningful. A more meaningful question might be, “What is the probability of an extreme event in the remaining lifetime of today's plants?” The remaining lifetime is about 30 years per plant or a total of (30)(100) = 3,000 plant years. Multiplying 3,000 times I in a billion gives about I in 3.3 million.

    Curves of the type shown in Figures B.5 and 9.6 are calculated for each of the major consequences. The magnitude of the three latent effects versus frequency of occurrence are summarized in Table B.1.* The normal incidence of these effects are given in the bottom row. Note the first column gives the frequency for an industry of 100 plants. A comparison with NUREG 1150 shows that, for the latent effects, NUREG 1150 gets results that are about a factor of three lowerthan WASH 1400. This appears to be mainly due to the lower source term developed in the NUREG 1150 study.











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