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No. 96-97 FOR IMMEDIATE RELEASE
(Tuesday, July 9, 1996)
NOTE TO EDITORS:
The Nuclear Regulatory Commission has received the attached report
from its Advisory Committee on Reactor Safeguards. The report, in the
form of a letter, provides comments on the NRC's severe accident research
program.
In addition, the NRC's executive director for operations has
received a letter report on a draft Regulatory Guide, DG-1047, "Standard
Format and Content for Applications to Renew Nuclear Power Plant
Operating Licenses."
#
Attachments:
As stated
. June 28, 1996
The Honorable Shirley Ann Jackson
Chairman
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT: SEVERE ACCIDENT RESEARCH
During the 432nd meeting of the Advisory Committee on Reactor Safeguards,
April 11-13, 1996, we completed our review of the status of the NRC
severe accident research program and severe accident codes. Our
Subcommittee on Severe Accidents held meetings on these matters on March
1 and April 8, 1996. During this review, we had the benefit of
discussions with representatives of the NRC staff and of the documents
referenced.
Conclusions and Recommendations
1. Severe accident research provides information essential to the
development of risk-informed regulation.
2. Severe accident research provides the basis for evaluating severe
accident management strategies.
3. The NRC nuclear safety research program budget continues to
decline, and various research efforts are being reduced or
eliminated. Periodic analysis should be performed to assure that
the remaining severe accident research efforts are focused on
topics that have the greatest impact on risk and the associated
uncertainties. Criteria should be developed for determining when
programs have met their objectives.
4. Results of the severe accident research have shown that there is no
threat of prompt containment failure posed by direct containment
heating (DCH) in Westinghouse large dry containments, alpha-mode
steam explosions, and Mark I liner melt-through. Research should
continue to:
. determine the impact of DCH on other containment types,
. develop codes to better model the hydrogen stratification and
detonation,
. determine the impact of ex-vessel steam explosions on the BWR
containments,
. understand the phenomenological aspects associated with
molten debris coolability,
. determine the impact of fuel coolant interaction on lower
head failure, and
. determine the threats posed to steam generator tubes by the
natural circulation induced by the core degradation
processes.
5. Quantification of uncertainties is essential to risk-informed
regulation. The NUREG-1150 effort contributed significantly to the
method for quantification of uncertainties. Additional effort is
needed to improve understanding and quantification of
phenomenological uncertainties and their impact on Level 2 PRA
results. We plan to provide more specific recommendations in this
area in the future, as needed.
6. The assurance of the availability of specialized experts to advise
the Commission is sometimes a tacit motivation for planning
research programs. We believe that such assurance is prudent and
should be explicitly recognized as a criterion in the funding of
research.
Discussion
We believe it is important that the staff periodically perform top down
assessments of research to assure that the work supports top level
objectives, to review priorities, and to identify research efforts that
have reached maturity and perhaps should be discontinued. In our view,
severe accident research should have the following top-level objectives:
support assessments of severe accident risk from operating plants,
provide a technical basis for reviewing accident management
procedures,
support the development of risk-informed regulation, and
provide a technical basis for evaluating advanced plant designs and
operational features.
Better Level 2 PRAs are needed to reduce the uncertainties associated
with the assessment of the risk to public health and safety. Severe
accident research provides the bases for improving Level 2 PRAs, many of
which have used unnecessarily simplistic models for severe accident
behavior. Severe accident research is needed to reduce the presently
large uncertainties in risk assessment results that are inimical to
making sound regulatory decisions.
The processes that lead to early failure of containment are of particular
importance to risk. Among such processes are DCH, fuel coolant
interactions, alpha-mode steam explosions, hydrogen detonations, direct
contact of core debris with containment structures, and steam generator
tube ruptures. Additional assessment of DCH is needed for CE, B&W, and
ice condenser containments, and for BWRs. Although it appears that large
dry containments and containments with igniters can accommodate hydrogen
combustion without failing, we believe that stratification and the
potential for local detonation needs additional investigation.
The extent to which debris can be cooled can be pivotal in determining
the likelihood of containment liner failure and long-term containment
basemat melt-through. Viable criteria for coolability of molten debris
either in-vessel or ex-vessel have not yet been developed.
A possible disadvantage of successful in-vessel debris cooling is the
potential failure of the reactor coolant system or steam generator tubes
caused by overheating from the convection of hot gases. Steam generator
tube ruptures that might occur as a consequence of, or coincident with, a
severe accident would provide a direct path for radionuclide release from
the reactor core to the environment. The NRC and industry are addressing
this issue, but we believe additional thermal hydraulic and radionuclide
transport code development will be required for resolution. The present
NRC codes are not capable of assessing this situation.
Currently, significant information in the severe accident area is being
developed in international cooperative programs. While we fully support
the bilateral agreements and the Cooperative Severe Accident Research
Program (CSARP), it is important for NRC that its domestic contractors
maintain capability in this area. Staff and contractors who are
knowledgeable of the physics and technology of severe accident phenomena
will be needed to resolve complex issues in this area, to enhance the
regulatory process, and to provide technical support in the event of a
real accident.
Dr. Dana A. Powers did not participate in the Committee's deliberation
regarding this matter.
Sincerely,
/s/
T. S. Kress
Chairman, ACRS
References:
1. Report dated August 18, 1992, from David A. Ward, Chairman, ACRS,
to Ivan Selin, Chairman, NRC, Subject: Severe Accident Research
Program Plan
2. U. S. Nuclear Regulatory Commission, SECY-95-004, dated January 4,
1995, from James M. Taylor, Executive Director for Operations, NRC,
for the Commissioners, Subject: Status of Implementation Plan for
Closure of Severe Accident Issues, Status of the Individual Plant
Examinations and Status of Severe Accident Research
3. U. S. Nuclear Regulatory Commission, NUREG/CR-6109, "The
Probability of Containment Failure by Direct Containment Heating in
Surry," May 1995
4. Nuclear Energy Institute, NEI 91-04, Revision 1, "Severe Accident
Issue Closure Guidelines," December 1994
5. Report (undated) by F. Cheung and K. Haddad, Pennsylvania State
University, Subject: Steady-State Observations and Theoretical
Modeling of Critical Heat Flux Phenomena on a Downward Facing
Hemispherical Surface
6. Sandia National Laboratories Letter Report, "Scaling and Design
Report for Lower Head Failure Experiments," May 1995
7. Secretary-General of the OECD Report, Senior Group of Experts on
Severe Accident Management (SESAM), "Severe Accident Management
Implementation," October 1995
8. Secretary-General of the OECD Draft Report, "Nuclear Safety
Research in OECD Countries, Areas of Agreement, Areas For Further
Action, Increasing Need For Collaboration," November 1995
9. Proceedings of the Specialist Meeting On Severe Accident Management
Implementation, held at Niantic, Connecticut, on June 12-14, 1995,
by the Committee on the Safety of Nuclear Installations, OECD
Nuclear Energy Agency
June 18, 1996
Mr. James M. Taylor
Executive Director for Operations
U. S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Mr. Taylor:
SUBJECT: DRAFT REGULATORY GUIDE DG-1047, "STANDARD FORMAT AND CONTENT
FOR APPLICATIONS TO RENEW NUCLEAR POWER PLANT OPERATING
LICENSES"
During the 432nd meeting of the Advisory Committee on Reactor Safeguards,
June 12-14, 1996, we discussed the subject draft Regulatory Guide with
representatives of the NRC staff and the Nuclear Energy Institute. We
also had the benefit of the documents referenced.
We have no objection to the staff proposal to issue the draft Regulatory
Guide for public comment. We plan to review the proposed final version
of this Guide after reconciliation of the public comments.
Dr. William J. Shack did not participate in the Committee's deliberations
regarding this matter.
Sincerely,
/s/
T. S. Kress
Chairman, ACRS
References:
1. U.S. Nuclear Regulatory Commission, Draft Regulatory Guide DG-1047,
"Standard Format and Content for Applications to Renew Nuclear
Power Plant Operating Licenses," transmitted by memorandum dated
April 18, 1996, from Scott F. Newberry, Office of Nuclear Reactor
Regulation, NRC, to John T. Larkins, ACRS
2. Nuclear Energy Institute, NEI 95-10 (Revision 0), "Industry
Guideline for Implementing the Requirements of 10 CFR Part 54ÄThe
License Renewal Rule," March 1996
3. U.S. Nuclear Regulatory Commission, SECY-96-059 dated March 18,
1996, from James M. Taylor, Executive Director for Operations, NRC,
for the Commissioners, "Activities Associated with the
Implementation of 10 CFR Part 54"
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