ML18101B284
| ML18101B284 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/21/1996 |
| From: | Olshan L NRC (Affiliation Not Assigned) |
| To: | Eliason L Public Service Enterprise Group |
| Shared Package | |
| ML18101B285 | List: |
| References | |
| GL-88-20, TAC-M74461, TAC-M74462, NUDOCS 9603260225 | |
| Download: ML18101B284 (9) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 Mr. Leon R. Eliason Chief Nuclear Officer & President-Nuclear Business Unit Public Service Electric and Gas Company Post Office Box 236 Hancocks Bridge, NJ 08038 March 21, 1996
SUBJECT:
INDIVIDUAL PLANT EXAMINATION (!PE) SUBMITTAL-INTERNAL EVENTS, SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 (TAC NOS. M74461 AND M74462)
Dear Mr. Eliason:
By submittal dated July 30, 1993, Public Service Electric and Gas Company (PSE&G) submitted the Salem IPE for internal events and internal flooding in response to Generic Letter 88-20, Supplement 1.
Enclosed is our Staff Evaluation Report (SER) (Enclosure 1) of this submittal. Also included with the SER are the contractors' Technical Evaluation Reports (TERs).
The review was performed which examined the IPE results for their "reasonableness" considering the design and operation of Salem.
The staff employed Science & Engineering Associates, Inc., Concord Associates, and Scientech Inc., to review the front-end analysis, human reliability analysis, and back-end analysis, respectively, of the IPE submittal. Their TERs*are provided as Enclosure 2, 3, and 4.
These contractor TERs were reviewed by the IPE "Senior Review Board" (SRB) as part of the RES quality assurance process.
The SRB is comprised of RES staff and consultants at Sandia and Brookhaven National Laboratories with PRA expertise.
The Salem IPE has estimated a core damage frequency (CDF) of 5.2E-5/reactor-year for Unit 1 and 5.5E-5/reactor year for Unit 2 from internally initiated events, including a contribution from internal flooding of about 7E-6/reactor year for each unit.
However, there are some differences between units in the relative contributions to the CDF from various accident types and initiating events.
For Unit 1, station blackout contributes 41%, transients 25%, loss of coolant accidents (LOCA) 15%, internal flooding 14%, anticipated accidents without scram (ATWS) 3%, and steam generator tube rupture (SGTR) 1%.
For Unit 2, transients contribute 36% to the CDF, station blackout 31%, LOCAs 16%, internal flooding 13%, ATWS 2%, and SGTR <1%.
The licensee indicated that these differences are mainly as a result of differences that existed in unit specific data related to initiating evehts, testing and maintenance.
The licensee stated that for a sequence (or event) to be considered a vulnerability, "... it had to pass the screening criteria.. [NUREG 1335] and contribute inordinately to the Salem core damage with respect to either (1) other Salem core damage sequences or events, or (2) in comparison to similar sequences or events for other nuclear power plants as determined from published probabilistic risk assessment results." Based on this definition, the licensee initially identified two vulnerabilities in the IPE submittal -
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L. Eli as on one from internal flooding (rupture of the demineralized water pipe i~ the relay room and the 64 foot switchgear room).and another from *procedures (which the licensee indicated was common to* Westinghouse pressurized,~atet reactors) addressing interfacing systems LOCA (ISLOCA).*
Howe-veY'~* *as discussed in the transmittal letter for the IPE submittal, further.analyse*s performed by the licensee identified additional floor drains.that were unacco~nted for in the flooding analysis and therefore the *flooding vuln.erabi*iity. identified in the submittal is considered, by the licensee, to*be: invalid.. The'license~
indicated that there was insuffi'dent time. to update.the'IPE submittal prior to submitting it. A procedural improvement~ was* identifi~d for. the. ISLOCA vulnerability and the licensee submitted a work request.~o the Westinghouse Owner's Group (WOG) to evaluate and revise the Emergency Response* Guidelines.
- In response to the* staff's request' fo.r additional.information, PSE&G indicated that in response to the above noted work.,request the WOG' s response to PSE&G indicated that a change was made to Revisi-0n lb,. 'of the Eme~gency Response Guidelines to correct this potential procedure deficiancy.
The licensee has implemented this improvement which the li~ense esti~ates reduces the CDF(5.2E-5/reactor-year) by 1% and the large early release frequency (4.0E-6/reactor-year) by 4%.
!~,
Based on our review, we conclude that PSE&G has met* the intent of Generic Letter 88-20, Supplement 1 and we consider TAC Nos. M74461 and M74462 complete.
Sincerely,
/S/
- Leonard N. Olshan, Project Manager Project Directo~ate 1-2 Division of Reactor Projects - I/II.
Office of Nuclear Reactor Regulation Docket Nos. 50-272/311
Enclosures:
- 1. Staff Evaluation Report
- 2.
Technical Evaluation Report (Front End)
- 3. Technical Evaluation Report (Back End)
- 4.
Human Reliability Analysis Report cc w/encl 1:
See next page DISTRIBUTION:
See next page OFFICE NAME DATE
one 'from internal flooding (rupture of the demineralized water pipe in the relay room and the 64 foot switchgear room) and another from procedures (which the licensee indicated was common to Westinghouse pressurized water reactors) addressing interfacing systems LOCA (ISLOCA).
However, as discussed in the transmittal letter for the IPE submittal, further analyses performed by the licensee identified additional floor drains that were unaccounted for in the flooding analysis and therefore the flooding vulnerability identified in the submittal is considered, by the licensee, to be invalid.
The licensee indicated that there was insufficient time to update the IPE submittal prior to submitting it. A procedural improvement was identified for the ISLOCA vulnerability and the licensee submitted a work request to the Westinghouse Owner's Group (WOG} to evaluate and revise the Emergency Response Guidelines.
In response to the staff's request for additional information, PSE&G indicated that in response to the above noted work request the WOG's response to PSE&G indicated that a change was made to Revision lb, of the Emergency Response Guidelines to correct this potential procedure deficiency.
The licensee has implemented this improvement which the license estimates reduces the CDF(5.2E-5/reactor-year) by 1% and the large early release frequency (4.0E-6/reactor-year) by 4%.
Based on our review, we conclude that PSE&G has met the intent of Generic Letter 88-20, Supplement 1 and we consider TAC Nos. M74461 and M74462 complete.
Docket Nos. 50-272/311 Sincerely,.
Leonard N. Olshan, Project Manager Project Directorate 1-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
I. Staff Evaluation Report
- 2.
Technical Evaluation Report (Front End)
- 3. Technical Evaluation Report
{Back End)
- 4.
Human Reliability Analysis Report cc w/encl 1:
See next page
Mr. Leon R. Eliason~
Public Service Electric & Gas Company cc:
Mark J. Wetterhahn, Esquire Winston & Strawn 1400 L Street NW Washington, DC 20005-3502 Richard Fryling, Jr., Esquire Law Department - Tower SE 80 Park Place Newark, NJ 07101 Mr. Clay Warren General Manager - Salem Operations Salem Generating Station P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Louis Storz Sr. Vice President - Nuclear Operations Nuclear Department P.O. Box 236 Hancocks Bridge, New Jersey 08038 Mr. Charles S. Marschall, Senior Resident Inspector Salem Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Han cocks Bridge, NJ 08038 '
Dr. Jill Lipoti, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415 Maryland Office of People's Counsel 6 St. Paul Street, 21st Floor Suite 2102 Baltimore, Maryland 21202 Ms. R. A. Kankus Joint Owner Affairs PECO Energy Company 965 Chesterbrook Blvd., 63C-5 Wayne, PA 19087 Mr. Elbert Simpson Salem Nucl. Generating Station, Units I and 2 Richard Hartung Electric Service Evaluation Board of Regulatory Commissioners 2 Gateway Center, Tenth Floor Newark, NJ 07102 Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Lower Alloways Creek Township c/o Mary 0. Henderson, Clerk Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Mr. Frank X. Thomson, Jr., Manager Licensing and Regulation Nuclear Department P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. David Wersan Assistant Consumer Advocate Office of Consumer Advocate 1425 Strawberry Square Harrisburg, PA 17120 Ms. P. J. Curham MGR. Joint Generation Department Atlantic Electric Company P.O. Box 1500 6801 Black Horse Pike Pleasantville, NJ 08232 Carl D. Schaefer External Operations - Nuclear Delmarva Power & Light Company P.O. Box 231 Wilmington, DE 19899 Public Service Commission of Maryland Engineering Division Chief Engineer 6 St. Paul Centre Baltimore, MD 21202-6806 Senior Vice President - Nuclear Engineering Nuclear Department P.O. Box 236 Hancocks Bridge, New Jersey 08038
LETTER TO LEON R. ELIASON. FROM LEONARD N. OLSHAN. DATED March 21~ 1996 DISTRIBUTION w/Enclosures 1, 2, 3, and 4 Docket File PUBLIC LOlshan ACRS DISTRIBUTION w/Enclosure 1 SVarga JZwolinski JStolz MO' Brien PDI-2 Reading File ERodrick, RES MDrouin, RES EKelly, RGN-1 EButcher RHernan OGC
SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2, INDIVIDUAL PLANT EXAMINATION STAFF EVALUATION REPORT ENCLOSURE l
I.
INTRODUCTION On July 30, 1993, the Public Service Electric and Gas Co.(PSE&G) submitted the Salem Generating Station (SGS) Unit 1 and 2 Individual Plant Examination (IPE) submittal in response to Generic Letter (GL) 88-20 and associated supplements.
On April 25,, 1995, the staff sent questions to the licensee requesting additional information.
The licensee responded in a letter dated August l, 1995.
A "Step l" review of the SGS IPE submittal was performed and invrilved the efforts of Science & Engineering Associates, Inc., Scientech, Inc./Energy Research, Inc., and Concord Associates in the front-end, back-end, and human reliability analysis (HRA), respectively. The Step 1 review focused on whether the licensee's method was capable of identifying vulnerabilities.
Therefore, the review considered (1) the completeness of the information and (2) the reasonableness of the results given the SGS design, operation, and history. A more detailed review, a "Step 2" review, was not performed for this IPE submittal. A summary of contractors' findings is provided below.
Details of the contractors' findings are in the technical evaluation reports (Enclosures 2, 3, and 4).
In accordance with GL 88-20, PSE&G proposed to resolve Unresolved Safety Issue (USI) A-45, "Shutdown Decay Heat Removal Requirements."
No other specific USls or generic safety issues were proposed for resolution as part of the SGS
. IPE.
I I.
EVALUATION SGS is a Westinghouse 4 loop pressurized water reactor (PWR) with a large dry containment.
The SGS IPE has estimated a core damage frequency (CDF) of 5.2E-5/reactor-year for Unit 1 and 5.SE-5/reactor-year for Unit 2 from internally initiated events, including a contribution from internal flooding of about 7E-6/reactor year for each unit. However, there were some differences between units in the relative contributions to CDF from various accident types and initiating events.
For Unit 1, station blackout contributes 41%, transients 25%, loss of coolant accidents (LOCA) 15%,
internal flooding 14%, anticipated transients without scram (ATWS) 3%, and steam generator tube rupture (SGTR) 1%.
For Unit 2, transients contribute 36%
to the CDF, station blackout 31%, LOCAs 16%, internal flooding 13%, ATWS 2%,
and steam generator tube rupture <1%.
The licensee indicated that these differences are mainly as a result of differences that existed in unit specific data related to initiating events, testing and maintenance.
Based on the licensee's IPE process used to search for decay heat removal (OHR) vulnerabilities, and review of SGS plant-specific features, the staff finds the licensee's OHR evaluation consistent with the intent of the USI A-45 (Decay Heat Removal Reliability) resolution.
However, it is noted that The licensee performed an HRA to document and quantify potential failures in human-system interactions and to quantify human-initiated recovery of failure events.
The licensee identified the following operator actions as some of the more important actions in the estimate of the CDF: recovery of offsite power, establishment of alternate ventilation after loss of ventilation to control and relay rooms, shutdown of the reactor from the remote shutdown panel, transfer of emergency core cooling systems from injection to recirculation, control of steam generator level and miscalibration of undervoltage sensors.
The licensee evaluated and quantified the results of the severe accident progression through the use of a containment event tree and considered uncertainties in containment response through the use of sensitivity analyses.
The licensee's back-end analysis appeared to have considered important severe accident phenomena.
Among the SGS Units 1 (and Unit 2) conditional containment failure probabilities: early containment failure is 4.8% (5.1%)
with overpressurization and alpha mode failure being the primary contributors; late containment failure is 37% (44%) with overpressurization and combustion.
of non-condensible gases generated by molten core concrete interactions being the primary contributors and bypass is 1% (1%) with SGTR and interfacing systems LOCA sequences the primary contributors.
The containment remains intact 56% (47%) of the time.
The licensee's response to containment performance improvement program recommendations is consistent with the intent of GL 88-20 and the associated Supplement 3.
Some insights and unique plant safety features identified at SGS by the licensee are:
- 1.
Given a LOSP, at least 2 of 3 emergency diesel generators must operate to provide AC power.
Each emergency diesel generator can provide sufficient power for only one service water pump and two service water pumps must operate to provide adequate cooling.
- 2.
An alternate source of power is available at SGS from an onsite gas turbine, if the emergency diesel generators fail.
- 3.
ECCS switchover from injection to recirculation is a manual operation.
- 4.
Motor driven auxiliary feedwater pumps are dependent on room cooling.
- 5.
Control air and chilled water systems are capable of being cross-tied between Units 1 and 2 at SGS and credit is taken in the analysis for this capability.
The licensee stated that for a sequence (or event) to be considered a vulnerability, "... it had to pass the screening criteria.. [NUREG 1335] and contribute inordinately to the SGS core damage with respect to either (1) other SGS core damage sequences or events, or (2) in comparison to similar sequences or events for other nuclear power plants as determined from.
published probabilistic risk assessment results." A sequence or event was considered to contribute inordinately if the CDF results were unusually high.
Based on this definition, the licensee initially identified two vulnerabilities in the IPE submittal.
One from internal flooding (rupture of the demineralized water pipe in the relay room and the 64 foot elevation, switchgear room) and another from procedures (which the licensee indicated was common to Westinghouse pressurized water reactors) addressing ISLOCA.
However, as discussed in the transmittal letter for the IPE submittal, further
\\1
'. ~
analyses performed by the licensee identified additional floor drains that were unaccounted for in the flooding analysis and therefore the flooding vulnerability identified in the submittal is considered, by the licensee, to be invalid.
The licensee indicated that there was insufficient time to update the IPE submittal prior to submitting it. A procedural improvement was identified for the ISLOCA vulnerability and the licensee submitted a work request to the Westinghouse Owner's Group to evaluate and revise the Emergency Response Guidelines.
In response to the staff's request for additional information, Public Service Electric and Gas Company (PSE&G) indicated that in response to the above noted work request the WOG's response to PSE&G indicated that a change was made to Revision lb, of the Emergency Response Guidelines to correct this potential procedure deficiency.
The licensee has implemented this improvement which the license estimates reduces the CDF by 1% (5.2E-5 to 5.lE-5/reactor-year) and the large early release frequency by 4% (4.0E-6 to 3.79E-6/reactor.year).
No other improvements wer.e identified. However, it was indicated in the submittal that the SGS Probabilistic Risk Assessment (PRA) is used as a primary tool for the assessment and prioritization of design discrepancies.
The licensee indicated. that more than 900 separate discrepancies h_ave been prioritized using the PRA through Apr.il 1993.. None of the discrepancies were identified in the submittal.
III.
CONCLUSION Based on the above findings, the staff notes that: (1) the licensee's IPE is complete with regard to the information requested by GL 88-20 (and associated guidan~e NUREG-1335), and (2) the IPE results are reasonable given the SGS design, operation, and history.
As a result, the staff concludes that the licensee's IPE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the SGS IPE has met the intent of GL 88-20.
It should be noted that the staff's review primarily focused on the licensee's ability to examine the SGS for severe accident vulnerabilities. Although certain aspects of the IPE were explored in more detail than others, the review is not intended to validate the accuracy of the licensee's detailed findings (or quantification estimates) that stemmed from the examination.
Therefore, this SER does not constitute NRC approval or endorsement of any IPE material for purposes other than those associated with meeting the intent of GL 88-20.
Date:
March 21, 1996