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U. S. NUCLEAR REGULATORY COMMISSION |
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No. 96-52 FOR IMMEDIATE RELEASE
(Friday, March 22, 1996)
NOTE TO EDITORS:
The Nuclear Regulatory Commission has received two
attached reports from its Advisory Committee on Reactor
Safeguards (ACRS). The reports, in the form of letters,
provide comments on:
--Recent probabilistic risk assessments performed by
Brookhaven National Laboratory on fires and certain fire
barrier issues; and
--Use of individual plant examinations in the regulatory
process.
In addition, the NRC's executive director for operations
received two ACRS reports. They provide comments on:
--An NRC program assessing the adequacy of a computer
code for simulating the behavior of the Westinghouse Electric
AP600 advanced pressurized water reactor design; and
--Resolution of generic safety issue 78, "Monitoring of
Fatigue Transient Limits for the Reactor Coolant System."
#
Attachments:
As stated
.
March 15, 1996
The Honorable Shirley Ann Jackson
Chairman
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT: REVIEW OF RECENT FIRE PROBABILISTIC RISK
ASSESSMENT
REPORTS BY BROOKHAVEN NATIONAL LABORATORY AND
CERTAIN FIRE BARRIER ISSUES
During the 429th meeting of the Advisory Committee on Reactor
Safeguards, March 7-9, 1996, we reviewed scoping fire
probabilistic risk assessments (PRAs) performed by Brookhaven
National Laboratory (BNL). We had the benefit of discussions
with representatives of the staff, BNL, and the National
Institute of Standards and Technology (NIST). Our Subcommittee
on Fire Protection discussed this matter during a meeting on
February 29, 1996. We also had the benefit of the documents
referenced.
At your request, we reviewed both the PRA model that evaluated
the strategy of using self-induced station blackout (SISBO) to
mitigate the consequences of a fire in the control room or
cable spreading room and the PRA-based scoping analysis of
degraded fire barriers. We also discussed the development of
alternate time-temperature curves for qualification of fire
barriers and the status of other fire protection issues.
To comply with Appendix R requirements, eight units have
procedures that require initiating a station blackout (SBO)
condition. An additional fifteen units have procedures for
dealing with fires in critical areas that could result in an
SBO. The PRA by BNL evaluated the effects of different schemes
for managing the electrical systems in the plant when a fire in
the control room has required use of the alternate shutdown
panel.
The study focused on the effectiveness of the procedures used
to mitigate the fire and did not address the probabilistic
treatment of fires. The scope of the study did not include a
number of issues that could affect the conclusions. For
example, the BNL study addressed neither the effects of fire
and smoke on human actions nor the possible damage to sensitive
electronic control and safety instrumentation. The study is
weak in the areas of modeling human actions for the manual
shutdown and restart of electrical equipment after an SBO
condition. Because of the limitations of the analysis and the
failure to quantify uncertainties, no substantive conclusions
can be drawn from this scoping study. The limitations of the
analysis should be addressed in Phase 2 of this study. A
meaningful uncertainty analysis should also be performed.
In the analysis of degraded fire barriers, BNL developed
core-damage frequencies for fire scenarios involving failures of
fire protection features such as cable tray fire barriers,
automatic detection and suppression systems, and fire barrier
penetrations. The PRA model did not examine degrees of fire
barrier degradation.
The analysis was based on event tree/fault tree models.
Although this is a step in the right direction, the analysis
does not use the best available methods for modeling fire
propagation, detection, and suppression. It does not model the
fundamental competition between the time to damage and the time
to detection/suppression. Most current fire PRAs have adopted
the competing processes model.
We also discussed the program proposed to the staff by NIST to
develop alternate time-temperature curves for nuclear power
plant fire barrier qualification. The program includes
development of models, ASTM E119-type full-scale furnace tests,
and test methods to simulate barrier response. We question the
need for this program. We have been told that alternate
time-temperature curves have been produced by the insurance
industry. Furthermore, a large number of fire models exist, some
of
which are being evaluated by the Department of Energy.
Although the need for new models is not clear, more validation
of these models with experimental data is needed. Some data
exist (NUREG/CR-6017). Comparisons with fire model simulations
show that the results are very sensitive to input parameters
that are not always well known.
The staff summarized the progress of licensee actions to
correct deficiencies associated with Thermo-Lag fire barriers.
The program appears to be meeting its objectives.
Sincerely,
/s/
T. S. Kress
Chairman
References:
1. Brookhaven National Laboratory, Draft Technical Letter
Report, FIN L-2629, "Risk Evaluation of the Response of
PWRs to Severe Fires in Critical Locations," May 30, 1995
(Draft Predecisional)
2. Brookhaven National Laboratory, Technical Evaluation
Report, FIN L-1311, "A Risk-Based Approach for Evaluation
of Fire Mitigation Features in Nuclear Power Plants,"
November 21, 1995 (Draft Predecisional)
3. U. S. Nuclear Regulatory Commission, NUREG/CR-6017 and
SAND93-0528, "Fire Modeling of the Heiss Dampf Reaktor
Containment," September 1995
.
March 8, 1996
The Honorable Shirley Ann Jackson
Chairman
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT: USE OF INDIVIDUAL PLANT EXAMINATIONS IN THE
REGULATORY PROCESS
During the 428th and 429th meetings of the Advisory Committee
on Reactor Safeguards, February 8-10 and March 7-9, 1996,
respectively, we discussed the Individual Plant Examination
(IPE) review process and findings with the NRC staff. Our
Subcommittee on IPEs also met with the staff and its contrac-
tors on January 26, 1996, to review this matter. We also had
the benefit of the documents referenced. This report is in
response to the December 27, 1995 Staff Requirements Memorandum
(SRM).
In the SRM, the Commission requested "the ACRS views on the
extent to which the current spectrum of IPEs can be used in the
regulatory process." We interpret this request as referring to
potential regulatory uses of the IPEs that were not delineated
in Generic Letter 88-20, "Individual Plant Examination for
Severe Accident Vulnerabilities." This report includes
comments on both the Generic Letter goals and the Commission
request.
Goals of Generic Letter 88-20
The purpose of the IPE program, as stated in Generic Letter
88-20, was for each licensee:
(1) to develop an appreciation of severe accident
behavior
(2) to understand the most likely severe accident
sequences that could occur at its plant
(3) to gain a more quantitative understanding of
the overall probabilities of core damage and
fission product releases
(4) to reduce, if necessary, the overall probabilities of
core damage and fission product releases by modifying, where
appropriate, hardware and procedures that would help prevent or
mitigate severe accidents.
We note that the IPEs were to be limited to the examination of
internal initiating events and internal floods with the reactor
at power and that individual and societal risks were not to be
estimated. Other programs deal with external events and
shutdown risk.
The IPE program has been successful at most utilities in
meeting goal (1) and, to a lesser extent, goals (2) and (3) of
the Generic Letter. Goal (4) of the Generic Letter also
appears to have been achieved. We were told that most
licensees discovered weaknesses and took corrective actions. In
addition, this program has been beneficial in educating a
broader segment of the NRC staff about the issues related to
these goals.
We were told by the staff that all licensees submitted a
Level-1 probabilistic risk assessment (PRA). Most licensees also
submitted a Level-2 PRA, although some addressed Level-2
phenomena in a rudimentary manner. The methods and data
sources used by different licensees varied widely. In some
cases, the choices appeared to be arbitrary. Some licensees
chose to include common-cause failures only for major
components, while others chose to ignore them completely.
It is difficult to determine the extent to which the
variability in IPE results for similar classes of plants is due
to
actual plant differences or to modeling assumptions. Although
some of the causes for this variability may be immediately
apparent, others are not. The latter include assumptions made
about success criteria, the assumed dependencies between
operator actions, and the level of decomposition in fault-tree
analyses. (We note that the fault trees were not requested as
part of the IPE submittals.)
An example of a potentially significant impact of modeling
differences is the range of core-damage frequencies (CDFs) for
BWR 3/4s that the staff has compiled. This range is from about
10-7 to about 10-4 per reactor-year. Although the staff has
stated that such differences are primarily due to plant
differences, this range of results seems unrealistic given the
similarity among BWR 3/4s.
Use of IPEs in the Regulatory Process
As discussed above, the quality and consistency of the IPEs
vary and the impact of assumptions and analytical models is
difficult to assess. On a case-by-case basis, however,
additional and extended use of these IPEs is possible. As
specific regulatory issues arise, the PRA Standard Review Plan
now being developed by the staff can serve as a template for
judging the quality and acceptability of the individual plant
PRA for the proposed application.
As the agency moves toward risk-informed regulation, there will
be an increasing need for full-scope PRAs that incorporate fire
risk, external events, other modes of operation, and
site-specific consequences. When requests for risk-informed
regulatory action arise, the NRC staff should make it clear
that a relevant PRA should be used.
To achieve these goals, especially consistency, some degree of
standardization will be required. Standardizing PRA models and
methods has been a controversial subject. Proponents argue
that it would create a basis for comparison of PRA results,
while opponents fear that it would inhibit methodological
developments. We recommend that IPEs be reviewed to identify
acceptable and unacceptable assumptions and/or models.
Codification of assumptions and models ought not inhibit the
continued development of PRA methods. These activities would
be a significant first step toward addressing the Commission's
statement in the SRM dated June 16, 1995, "that more meaningful
plant-to-plant or scenario-to-scenario comparisons based on
risk could be achieved if PRAs were done on a more
standardized, replicable basis."
We believe that the NRC could make additional use of the
present IPEs (except those that the staff has found to use
unacceptable methods or models) for a limited number of
applications (e.g., regulatory analyses and prioritization of
generic issues).
The staff stated that the CDFs for several PWRs are greater
than 10-4 per reactor-year. Several BWRs have CDFs that are
very close to 10-4 per reactor-year and the conditional
containment failure probabilities for BWR Mark I containments
range from about 0.02 to about 0.6. Although the PRAs have
limitations as discussed above, these numbers suggest that an
investigation would be warranted to reassess their validity and
to verify that the very low numbers reported by some other
plants reflect actual plant differences.
Our conclusion is that the IPE program has met successfully the
objectives of Generic Letter 88-20. This program has developed
a risk awareness, both in the utilities and the NRC, that will
contribute significantly to efforts to establish a
risk-informed and performance-oriented regulatory system. The
plant-specific .IPEs are an extremely valuable asset that should not be
permitted to languish unimproved and unused.
Sincerely,
/s/
T. S. Kress
Chairman
References:
1. Staff Requirements Memorandum dated June 16, 1995, from
Andrew L. Bates, Acting Secretary, NRC, to the File
regarding Meeting with ACRS on June 8, 1995
2. Staff Requirements Memorandum dated December 27, 1995,
from John C. Hoyle, Secretary, NRC, to John T. Larkins,
ACRS regarding Meeting with ACRS on December 8, 1995
3. Generic Letter 88-20, dated November 23, 1988, to All
Licensees Holding Operating Licenses and Construction
Permits for Nuclear Power Reactor Facilities, Subject:
Individual Plant Examination for Severe Accident
Vulnerabilities - 10 CFR .50.54(f)
.
March 19, 1996
Mr. James M. Taylor
Executive Director for Operations
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Mr. Taylor
SUBJECT: NRC STAFF PROGRAM ON THE ADEQUACY ASSESSMENT OF
THE RELAP5/MOD3 CODE FOR SIMULATION OF AP600
PASSIVE PLANT BEHAVIOR
During the 429th meeting of the Advisory Committee on
Reactor Safeguards, March 7-9, 1996, we reviewed the program
being conducted by the Office of Nuclear Regulatory Research
(RES) to assess the adequacy of the RELAP5/MOD3 code for
simulating the behavior of the Westinghouse AP600 passive
plant design. During this review, we had the benefit of
discussions with representatives and consultants of the NRC
staff and the Idaho National Engineering Laboratory (INEL).
Our Subcommittee on Thermal Hydraulic Phenomena held a
meeting on this matter on February 22-23, 1996. We also had
the benefit of the referenced documents.
We have been asked to comment on the approach and methodology
for demonstrating the adequacy of the RELAP5/MOD3 code
to calculate AP600 passive plant behavior in support of the
design certification review. We believe that the overall
approach and methodology being employed by RES for this
assessment is acceptable. Most of the necessary elements
are in place. A substantial amount of work remains,
however, and we believe that the schedule for successful
completion cannot be met.
Our comments and recommendations relative to this review,
primarily based on oral presentations, are:
. Since we last reviewed this program in 1994,
significant improvements have been made. The most significant
has been the increased emphasis on the code improvement
program. Other changes that have led to excellent
results include the involvement of outside technical
expertise, via the Thermal Hydraulic Expert Consultants
group and the direct involvement of RES technical
personnel in the research activities. Particularly
noteworthy accomplishments include the analysis of
water hammer, the treatment of flow oscillations
observed in the tests during injection from the
In-containment Refueling-Water Storage Tank and the
evaluation and explanation of strong thermal
stratification in the ROSA cold leg.
. RES should perform a more robust and complete top-down
system scaling analysis for ROSA, SPES, and OSU. An
entire transient should be evaluated to quantify the
effects of various distortions in the three facilities
and to demonstrate that the experimental database is
sufficient to validate the code. Any additional
distortions or anomalies identified should be added to
the list of distortions compiled by RES in late-1994,
and that remain to be addressed. The scaling effort
should be integrated with the Phenomena Identification
and Ranking Table.
. The thermal stratification that was seen in ROSA tests
for a one-inch cold-leg break was initially identified
potentially important safety issue for the AP600.
It has now been shown to be just a manifestation of
scale distortion in the ROSA facility. This
demonstrates the need to identify and explain anomalous
behavior.
. The thermal stratification in the Core Makeup Tank
(CMT) observed in the tests needs to be studied. Its
effects on core inventory have to be understood because
neither RELAP5/MOD3 nor the Westinghouse computer codes
can, at present, reliably predict thermal
stratification.
. The screening study for water hammer in the AP600
design addressed an important safety issue. The study
allows an analysis of the potential for such events and
provides a method for estimating the resulting loads in
susceptible areas. We recommend that this study be
published soon as a separate report.
. The documentation provided for our review did not, by
itself, furnish an adequate basis upon which we could
logically endorse the process. The documentation
provided to the Thermal Hydraulic Phenomena
Subcommittee in advance of the February 22-23, 1996 meeting was
inconsistent and contained results declared incorrect
by RES during the meeting. Furthermore, the
RELAP5/MOD3 Code Manual published in August 1995 was
not provided to us in time to support our review.
. RELAP5 is still undergoing significant and rapid
modifications. A calculation has not yet been
performed with a version of the code that contains all
the planned changes. Numerous calculations will need
to be performed to mature the code and validate it
using data obtained from various separate effects and
integral facilities tests.
Overall, the approach and methodology for qualifying
RELAP5/MOD3 for AP600 simulation appear to be adequate.
However, two possible "show stoppers" remain: 1) simulation
of the CMT thermal stratification and 2) simulation of
long-term cooling, which is still an issue. Serious
consideration should be given to addressing these obstacles.
.Dr. George Apostolakis did not participate in the Committee's
deliberations of this matter.
Sincerely,
/s/
T. S. Kress
Chairman
References:
1. Memorandum dated January 22, 1996 from M. W. Hodges,
Office of Nuclear Regulatory Research, NRC, to J.
Larkins, Advisory Committee on Reactor Safeguards, NRC,
transmitting:
- Volume 2 of 10 volumes of adequacy demonstration
reports, "Adequacy Assessment Overview"
- Idaho National Engineering Laboratory draft
report prepared for U.S. Nuclear Regulatory
Commission, "Adequacy Evaluation of RELAP5/MOD3
for Simulating AP600 Small Break Loss-of-Coolant
Accidents, Volume 2: Horizontal Integrated
Analysis of the AP600 1-Inch Diameter Cold Leg
Break," November 1995, with Appendices A-K
(Proprietary)
2. Idaho National Engineering Laboratory, draft report
prepared for U.S. Nuclear Regulatory Commission,
"Top-Down Scaling Analysis Methodology for AP600 Integral
Tests," January 1996
3. Letter report dated April 12, 1995, to James M. Taylor,
Executive Director for Operations, NRC, from T. S.
Kress, Chairman, Advisory Committee on Reactor
Safeguards, Subject: NRC Test and Analysis Program in
Support of AP600 Advanced Light Water Passive Plant
Design Review
4. Letter dated May 8, 1995, from James M. Taylor,
Executive Director for Operations, NRC, to T. S. Kress,
Chairman, Advisory Committee on Reactor Safeguards,
Subject: Staff Response to ACRS Letter Dated April 12,
1995, on NRC Test and Analysis Program in Support of
AP600 Advanced Light Water Passive Plant Design Reviews
.
March 14, 1996
Mr. James M. Taylor
Executive Director for Operations
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Mr. Taylor:
SUBJECT: RESOLUTION OF GENERIC SAFETY ISSUE 78, "MONITORING
OF FATIGUE TRANSIENT LIMITS FOR THE REACTOR
COOLANT SYSTEM"
During the 429th meeting of the Advisory Committee on
Reactor Safeguards, March 7-9, 1996, we completed our
deliberations on the resolution of the subject Generic
Safety Issue that we started during our 424th meeting,
September 7-8, 1995. We had the benefit of discussions with
representatives of the NRC staff and of the documents
referenced.
This Generic Safety Issue was originally developed to deter-
mine whether licensees need to perform transient monitoring
to ensure compliance with requirements concerning fatigue
failure. The transient monitoring concern was subsumed in
the Fatigue Action Plan, which was reported as complete in
SECY-95-245, "Completion of the Fatigue Action Plan."
The current scope of the Generic Safety Issue is focused on
the evaluation of risk from fatigue failure. The staff
completed a study that demonstrated that the risk from
fatigue failure of the primary coolant pressure boundary
components is very small. The analyses used in the study
were based on the assumption that the probability of crack
initiation by fatigue in a component subject to cyclic loads
and the probability of crack propagation through the wall
are independent. The product of these probabilities was
used to calculate the change in core-damage frequency caused
by fatigue failure of a component.
The analyses, as presented to us by the staff to demonstrate
its conclusion, lacked sufficient detail to be convincing.
Additional discussions with the staff demonstrated that more
complete analyses using the PRAISE code have led to the same
conclusion. The PRAISE analyses of the failure probability
of primary system piping assumed that a distribution of
cracks existed in a component and calculated the probabilities of
crack propagation through the wall and failure.
Parametric studies using the PRAISE code showed that the
calculated probabilities of failure are small, even when
very conservative loads and flaw-size distributions are
assumed. The staff provided a careful quantification of
uncertainty of fatigue crack initiation. We recommend such
consideration of uncertainties in any future analyses
regardless of the technical approach adopted.
We believe that the staff's conclusion concerning the risk
significance of fatigue failure of reactor components is
correct. Thus, we agree that this Generic Safety Issue is
resolved.
Dr. Shack did not participate in the Committee's deliberations
regarding this matter.
Sincerely,
/s/
T. S. Kress
Chairman
References:
1. Memorandum dated August 18, 1995, from Charles Serpan,
Jr., NRC Office of Nuclear Regulatory Research, to John
T. Larkins, ACRS Executive Director, Subject: Proposed
Resolution of Generic Safety Issue 78, "Monitoring of
Fatigue Transient Limits for the Reactor Coolant
System"
2. SECY-95-245 dated September 25, 1995, from James M.
Taylor, Executive Director for Operations, to the
Commissioners, Subject: Completion of the Fatigue
Action Plan
3. Memorandum dated October 27, 1995, from Jeff Keisler
and Omesh Chopra, Argonne National Laboratory, to Craig
Hrabal, NRC Office of Nuclear Regulatory Research,
Subject: Uncertainty Estimates for the Probability of
Fatigue Crack Initiation in Reactor Components,
NUREG/CR-6335, ANL-95/15
4. U. S. Nuclear Regulatory Commission, NUREG/CR-6237,
"Statistical Analysis of Fatigue Strain-Life Data for
Carbon and Low-Alloy Steels," August 1994
5. U. S. Nuclear Regulatory Commission, NUREG/CR-6335,
"Fatigue Strain-Life Behavior of Carbon and Low-Alloy
Steels, Austenitic Stainless Steels, and Alloy 600 in
LWR Environments," June 1995
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