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Safety Issue: Large uncertainties exist in the in-vessel fission-product release rates currently used in Source Term analysis with the Source Term Code Package (STCP). These rates are taken from early, old, relatively coarse, out-of-pile data at atmospheric pressure. The relatively volatile fission products not released from the fuel in-vessel and then removed by plate out in the reactor vessel are immediately released ex-vessel by melt-concrete interactions. In cases of early containment failure these fission products can become the largest part of the release to the containment.

Safety Issue: One of the safety issues is whether or not, in the absence of a hydrogen control system, given hydrogen-steam mixtures in large dry containments, would such mixtures autoignite and burn slowly destroying containment seals or, alternatively, burn rapidly and damage containment equipment and structures. The information would allow NRC to evaluate the need for changes in equipment qualification and hydrogen regulations. Another issue not resolved involves the quantitative loads to containment walls and electrical and mechanical safety equipment from local hydrogen detonations and the probability of wall and equipment damage."

Safety Issue: Significant uncertainties exist
in the safety analysis codes ability to
calculate reactor response to

feedwater-line/steamline breaks. These
uncertainties affect NRC's ability to evaluate
such effects as: 1) thermal shock to the
reactor vessel under pressure in the event of
a steamline break; and 2) coolant system
overpressurization during feedwater line

breaks.

Safety Issue: To confirm the adequacy of data obtained from plant instrumentation and the availability of electrical equipment under severe accident states which provide the basis for operational and emergency preparedness

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actions; and to provide more accurate data for
deterministic and probabilistic calculations
as they pertain to severe accident states.

Safety Issue: The requirement of separation
of the redundant trains of safe shutdown
equipment prescribed by 10CFR50, App. R cannot
be met in many control rooms. The rule
requires in such cases that an alternative or
dedicated shutdown capability be provided.
Whether in the event of a credible control
room fire the operator will have time enough
and the physical ability to transfer control
of the reactor to the alternative or dedicated
shutdown panel is at issue. There is also the
question of how much of the control room
equipment will survive and whether and when
control room operation can be resumed.

Safety Issue: The currently available
real-time atmospheric dispersion,

precipitation washout and plume rise models
for dose projections in emergency response
situations have not been adequately validated.
Accordingly, there is the potential for an
inappropriate, unreliable or inaccurate model
being used in an emergency response situation
and [sic] which would produce incorrect dose
projections....Risk assessments and other

licensing evaluations in which the models are
used have large uncertainties.

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

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Bevere reactor accidents. We found that even a cursory review suggests that the path being pursued by PSNH and the NRC staff is 111-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. provide a few examples below.

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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.

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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.

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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.

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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
specify with a high degree of preciseness, the
conditional probability (and uncertainty
bounds to be assigned) of containment failure
with a major release of radiation....There are
very low probability severe accident
conditions under which a containment may be
unable to prevent a major release. Although
containment structures are conservatively
designed to withstand the substantial

temperatures and pressures associated with a
major pipe rupture...they are not designed to
withstand the additional challenges that might
be associated with a complete core melt. Such
challenges include phenomena such as increased
pressures from an uncontrolled hydrogen
combustion or release of large quantities of
noncondensible gases from core-concrete
interactions....

The exchange in Question 6 reiterates these inherent uncertainties.

QUESTION 6: In a supplement to the record of
the Subcommittee's April 17, 1985 hearing, the
NRC staff wrote: "Analysis shows that there
are some kinds of accident sequences that
could cause failure of any containment
design...." For each type of containment
please enumerate each mode of potential
containment failure and the conditions that
can lead to each mode of failure. What is the
relative likelihood of different modes of
containment failure?

ANSWER. The potential containment failure
modes for severe accidents in all U.S. LWR
designs can be generally classified into six
groups. The definition of these groups
depends on the timing of the failure (relative
to core melting and major releases of
radioactive material into the containment) and
the failure location. These groups are:

1. Early failures directly to the atmosphere;
2. Early failures into other plant buildings;
3. Late failures directly to the atmosphere;
Late failures into other plant buildings;
Late failures into the ground;
6. No containment failure.

4.

5.

The exact failure modes and causes, and their
relative likelihoods, vary considerably among
plant types and even among plants of similar
containment design.... The NRC staff and
supporting contractors are presently engaged
in a major reassessment of the risks of
current commercial reactors. More

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