19 actions; and to provide more accurate data for Safety Issue: The requirement of separation Safety Issue: The currently available precipitation washout and plume rise models licensing evaluations in which the models are Clearly, there are an extraordinary number of open safety-related issues, some bearing directly on the Seabrook analysis. We fail to understand how the utility and the NRC staff can provide sufficient assurances that the Seabrook containment is virtually "fail-safe" in light of these considerable uncertainties remaining in understanding and modeling basic phenomena. OTHER COMMENTS ON POSSIBILITIES OF A SEVERE REACTOR ACCIDENT In light of the efforts by PSNH and NRC staff to demonstrate that the Seabrook containment is virtually "fail-safe" and that a miniscule probability exists of an accident with significant off-site consequences beyond 2 miles, Subcommittee staff felt it useful to review several knowledgeable sources' comments regarding severe reactor accidents. We found that even a cursory review suggests that the path being pursued by PSNH and the NRC staff is ill-considered, and that many acknowledged experts including some within the Commission itself have offered comments that reflect poorly on the effort to reduce the size of the EPZ. We provide a few examples below. 11 ... On the need for planning for serious reactor accidents, from the report of the Kemeny Commission: We must not assume that an accident of this or greater seriousness cannot happen again, even if the changes we recommend are made. Therefore, in addition to doing everything to prevent such accidents, we must be fully prepared to minimize the potential impact of such an accident on public health and safety, should one occur in the future." Report of the President's Commission on the Accident at Three Mile Island, p. 15. - On the question of the accuracy of probabilistic risk assessments estimations of core meltdown probabilities: "I believe that the recent Davis-Besse event illustrates that, in the real world, system and component reliabilities can degrade below those we and the industry routinely assume in estimating core melt frequencies. Our regulatory process should require margins against such degradations and also to reflect the uncertainties in our PRA estimates." Harold Denton, NRC Director of Nuclear Reactor Regulation, in memorandum dated June 27, 1985, to William J. Dircks. "There is a On the possibility of severe reactor accidents: distinct possibility of one or more additional severe reactor accidents, beyond the one at Three Mile Island, in the remaining life of the plants now in operation or under construction, unless the estimated accident frequency declines sharply with modifications, or has been significantly overestimated in current PRAS and actuarial inferences." NU REG-1070, "NRC Policy on Future Reactor Designs: Decisions on Severe Accident Issues in Nuclear Power Plant Regulations," August 1984, p. 108. On the possibility of an accident in the U.S. as severe as the Chernobyl accident: "...given the present level of safety being achieved by the operating nuclear power plants in this country, we can expect to see a core meltdown accident within the next 20 years, and it is possible that such an accident could result in off-site releases of radiation which are as large as, or larger than, the releases estimated to have occurred at Chernobyl. NRC Commissioner James Asselstine, in testimony before this Subcommittee, May 22, 1986. Even these few examples suggest that both the NRC staff and PSNH are proceeding under a dubious assumption when they attempt to demonstrate an invincible technical basis premised on containment strength and severely limited accident scenarios in arguing that the Seabrook plant should be considered essentially immune to severe accidents. PRIOR NRC COMMENTS ON CONTAINMENT PERFORMANCE In recent times the Subcommittee has had numerous exchanges with the NRC regarding emergency planning, containment analysis, and source terms. Several statements by the NRC have either directly contradicted the positions now being taken by NRC staff and the PL&G consultants or have emphasized that the present state of the art is characterized by substantial uncertainties in analytical techniques. It appears evident the NRC staff has ignored such qualifications in its attempt to assist the utility to justify a reduction in the size of the EPZ. The utility, its consultants, and PL&G have argued that the containment failure pressure is so high as to make a severe accident implausible. However, in a May, 1986 exchange between the Subcommittee and the NRC, the NRC clearly stated that uncertainties in containment analysis were too great to permit a precise estimate of containment failures. The exchange reads as follows: QUESTION 5. What degree of confidence does the NRC have in the ability of different containment buildings to prevent a major release of radiation during a core meltdown? For each type of containment building, what is the estimated probability of containment failure given a meltdown and state precisely what uncertainty bounds are assigned to this estimate and how it was calculated? ANSWER. At present, the NRC staff cannot temperatures and pressures associated with a combustion or release of large quantities of The exchange in Question 6 reiterates these inherent uncertainties. QUESTION 6: In a supplement to the record of ANSWER. The potential containment failure 2. 1. Early failures directly to the atmosphere; 4. 5. The exact failure modes and causes, and their quantitative information on the types of For PWR's with large, dry" and "subatmospheric" containment designs [like The above comments demonstrate that there are no ready answers to questions regarding containment failure. If the NRC is unable to supply any hard numbers and is reduced to citing analytical uncertainties and studies in progess, it is difficult to understand how in three short months PL&G could accomplish an analysis sufficiently well-grounded to warrant a change in emergency planning regulations for the Seabrook plant. POTENTIAL TECHNICAL PROBLEMS WITH THE UTILITY'S CLAIMS The utility's consultant has asserted that the pressure at which the containment would fail is significantly higher than for most other similar large dry containment pressurized water reactors. They attribute this greater strength to several factors, including a larger than average containment volume and more steel used in the construction of the containment. The Subcommittee staff questions whether PL&G has failed to consider adequately a number of significant phenomena and possible reactor accident scenarios, several of which could possibly result in outcomes very different from those cited by PL&G. A few examples of these technical questions may be summarized briefly as follows. (a). The conclusions of the PL&G report depend crucially on the assumption that the containment failure pressure is roughly 225 psig. This value is much higher than the estimated containment failure pressures for other plants, and may not be adequately supported. In addition, containment failures generally may be assumed to occur at points of structural discontinuity |