NUREG/CR-5465, Discusses Evaluation of PRA & Fire Hazards Analysis,Per Util 880512 & 0104 Submittals.Documentation & Data Not as Complete as Normally Found in well-documented Pra. NUREG/CR-5465 & Staff List Encl
| ML20029C404 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/21/1991 |
| From: | Le N Office of Nuclear Reactor Regulation |
| To: | Eury L CAROLINA POWER & LIGHT CO. |
| References | |
| RTR-NUREG-CR-5465 TAC-67123, TAC-67124, TAC-68295, TAC-68296, NUDOCS 9103270375 | |
| Download: ML20029C404 (7) | |
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UNITED STATES
.o NUCLEAR REGULATORY COMMISSION WASHINo TON, D. C. 20666 j
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%...4 March 21, 1991 j
M Docket Nos.
50-325 and 50-324 gs Mr. Lynn W. Eury e stf#
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Executive Vice President Power Supply Carolina Power & Light Company r-Post Office Box 1551 r
Raleigh, North Carolina 27602
Dear Mr. Eury:
SUBJECT:
EVALUATION OF PROBABillSTIC RISK ASSESSMENT AND FIRE HAZARDS ANALYSIS - BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND2(TACN05.68295,68296,67123,AND67124)
By letter dated May 12, 1988, Carolina Power & Light Company (CP&L) provided a copy of the Brunswick Steam Electric Plant, Units 1 and 2 (BSEP), Probabilistic Risk Assessment (PRA), and by letter dated January 4,1988. CP&L provided a copy of the BSEP Probabilistic Fire Analysis (PFA). The Nuclear Regulatory Commission (NRC) staff has undertaken a limited-scope review of both submittals and has contracted with the Idaho National Engineering Laboratory (INEL) to perform the review.
As part of the review process, the NRC staff and INEL staff performed a plant walk-down on August 25 and 26,1988, to familiarize themselves with major safety systems, components and structures important to the risk review.
The staff also conducted a meetin 25, 1988, and requested additional information-(RAI) g with your staff on Auguston selected accident scenario containment venting design details, and other operating procedures that were not documented in the BSEP PRA.
You provided a response to the above staff RAI on October 25, 1988.
The INEL review of the above-submitted BSEP PRA and PFA has been completed and the results documented in NUREG/CR-5465, which was published in November
'1989. A copy of this HUREG is enclosed (see Enclosure 1). A copy of this NUREG was also provided to your staff for information in early 1990.
The 1NEL review concluded that the PRA (level-1, internal and external-initiators) and the PFA were, generally, well performed, comprehensive and competent risk assessments. The methodologies used for the PRA are generally consistent-with current practices, and the analysis of the contributors to core damage at BSEP was evaluated in a systematic fashion. The PRA provides adequate.information regarding contributors to the core cooling and decay heat removal system failures and failure probability estimates. However, the documentation and the bases of assumptions and the data were not as complete as normally found in a well-documented PRA.
l NRC FILE CENTER COPY lEE'9281RggBlay h
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i Mr. Lynn W. Eury 2
The staff has reviewed the INEL's final evaluation report documented in the NUREG/CR and agrees with its conclusions and recommendations.
In Enclosure 2, the: staff lists additional areas of the BSEP PRA and PFA, beyond those mentioned in NUREG/CR-5465, that could be upgraded to improve the analyses.
These areas and those in the NUREG need to be addressed for the staff to be able to establish the usefulness of the analyses as a licensing tool for plant modifications, improvement, and overall plant safety enhancement.
In conclusion the staff finds the BSEP PRA and PFA to be a fair evaluation of-the most likely paths to core damage at the plant. The staff also recognizes that many plant modifications are in process or have recently been made that could have an effect on the PRA and on INEL's review.
This concludes the staff efforts on TAC Nos. 68295/68296, and 67123/67124 Sincerely, c
Orignal Signed By:
Ngoc B. Le, Project Manager Project Directorate 11 1 l
Division of Reactor Projects. 1/11 Office of Nuclear Reactor Regulation
Enclosures:
1.
Staff List i'
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See next page DISTRIBUTION See attached page DFC
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Mr. L. W. Eury Brur,swick Steam Electric Plant Carolina Pows e & Light Company Units 1 and 2 Cc:
Mr.' Russell B. Starkey, Jr.
Mr. H. A. Cole Vice President Special Deputy Attorney General Brunswick Nuclear Project State of North Carolina P. O. Box 10429 F. O. Box 629 Southport, North Carolina 28461 Raleigh, North Caroline 21602 Mr.R.E. Jones $LightCompany General Counsel Mr. Robert P. Gruber Carolina Pcwer.
Executive Director a
P; 0. Box 1551 Public Staff - NCUC Raleigh,' North Carolina' 27602 P. O. Box 29520
~ Ms. frankie Rabon Board of Commissior,ers P. O. Box 249
. Bolivia, North Carolina 28422 Resident Inspector-U. S. Nuclear Regulatory Commission Star Route 1 P. 0. Box 208
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Southport, North Carolina 28461
~ Regional Administrator, Region 11 U. S. Nuclear Regulatory Commission
~
.101 Marietta Street, Suite 2900
- Atlanta, Georgia 30323 Mr. Dayne H.. Brown Director. _
DivisionofRadiatIonProtection N; C. Department of Environmental.
Commerce and Natural Resources.
P. 0. Box 27687.
- Raleigh, North Carolina 27611-76*7 Mr. J.:L.' Harness Plant General Manager -
Brunswick Steam Electric Plant
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- P.'O. Box 10429
- Scuthport, North Carolina 28461 o
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____._-._--------------n EGO-2579 i
i Review of tae Brunswick i
Steam Electric Plant Proba3iListic Risx Assessment i
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Prepared by M.B. Sattison, P.R. Davis, D.G. Satterwhite W.E. Gilmore, R.E. Gregg Idaho National En glnerring Laboratory EG&G Idaho, Inc.
Prepared for U.S. Nuclear Regulatoiy Commission
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ENCLOSURE 2 AREAS OF RECOMMENDED IMPROVEMENT IN THE BRUNSWICK PRA 1.
Evaluate the fa61ure to close of equipment drain valves, drywell floor drains, and check valves in the Brunswick PRA flooding analysis.
Provide plant-specific data for equipment imputant to safety, such as ECCS and RHR pumos, batteries, electrical bussus, breakers, the srrvice water system, and the component cooling water system.
Plant-specific frequencies should be calculated for initiating events that the units have experienced, such as loss of feedwater and turbin i trip ' Where safety systems are known to have had availability /relhbility problems in the past (e.g., some ECCS valves), t?w licensee should use plant-specific data.
3.
The PRA should address the effect of battery room HVAC failures (particularly the effect on the battery-electrolytt specific gravity needed to maintain guaranteed capacity).
The PRA should address the need for battery room cooling following a loss of AC power event.
The PRA should clarify whether the room heatup calculations (Appendix A.2) include the effect of the heat source from postulated pipe breaks in the HPCI or RCIC steam lines.
Clarify how dependent failures from steam line breaks in the HPCI or RCIC systems outside containment were evaluated in the PRA.
4 Sinm the RHR system is very important in maintaining adequate lorig-term containment heat removal, the licensee should provide additional justification that the KlR pump seals do not need any cooling in the long term following a MSIV closure event or a small LOCA.
5.
As explained in detail in NUREG/CR-5465, the estimation of human error rates in the Brunswick PRA is not well documented.
The hman error rate estimates appear to be lower by a factor of 2 to 10 than other mom up-to-date and better docukented human error studies.
Based on 6n analysis of the sensitivity of core damage frequency estimates to changes in human error rates, the licensee, at a minimum, should reexamine its human error estimates for events XI (failure to inhibit ADS during an ATWS). LH (failure to control reactor water level subsequent to a controlled depressurization event following an ATWS), and MS sfailure tu scram manually given failure of electrical portions of the scram system).
6..
The Brunswick fire PRA does not provide sufficient information to determine the effec'. of postulated systems interactions that could result from consequential failures (e.g., loss of ventilation, loss of an ac bus, hot gas pr0pagation through partial fire barriers). As an example, the staff review of the fire PRA indicates that the service water pump house may not have adequate fire barriers betwren the service water pumps. The staff is unable to determine whether a postulated fire in this *~'a is or is not a safety concern.
7.
Based on previous BWR PRAs, the staff would expect to find the following sequences to be potentially important core damage or risk contributors.
The following sequences are not well documented or not documented at all in the Brunswick PRA and should be addressed, a.
Sequences with Rx trip in combination with a break in the tail pipe 4
of a safety / relief valve and a subsequent failure to provide vapor suppression, q
b.
Sequences involving failures of tne drywell coolert, c.
Sequences involvin; failures of the reactor water level measurement system in combination with failures in the de power system, d.
Sequences involving late failures of the operator to initiate manual scram following an inadvertent open relief valve event, e.
Sequences involving a stuck-open relief valve event immediately following a reactor isolation event with a subsequent failure to scram.
-Peto.:1 pal Contributors: G. Kelly
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E. Chelliah 4
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DIST BUTION W$
NRC & Local PDRs S. Varga 14 E4 s
G. Lainas 14 H3 E. Adensam P. Anderson N. Le OGC E. Jordan MNbB 3701 G, Kelly 10 E4
'. Chellish NLN 316
.. Beckner 10 E4-ACRS(10)
Brunswick File l
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