ML20042C314

Further Response to 820301 Interrogatories & Requests for Production of Documents Per 820319 Telcon.Certificate of Svc Encl

ML20042C314
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	03/26/1982
From:	Earley A HUNTON & WILLIAMS, LONG ISLAND LIGHTING CO.
To:	SHOREHAM OPPONENTS COALITION
Shared Package
ML20042C307	List: ML20042C306 ML20042C311 ML20042C314
References
ISSUANCES-OL, NUDOCS 8203310201
Download: ML20042C314 (10)

Text

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for the eleven items mentioned in the contention.

FSAR Section 7.1.does contain information about the regulatory classification of Shoreham's systems, but LILCO does not have a list of the sort requested by SOC. There are available, however, plan views of each elevation in the Reactor Building that show the loca-tion of all major equipment. Safety related equipment can be identified by an asterisk in the equipment

! identification number between the system designation and the specific equipment number (e.g., lEll*P-014C). i These plan views are figures 3.8.1-1 through 3.8.1-6 ,

i in the FSAR.

If this information is not adequate, LILCO will make available for SOC's review the relevant as-built drawings as ordered by the Board. The only safety related or class IE equipment located in the secondary containment are the isolation valves and sample tubing for the equipment associated with SOC 3(e), (f) and (g). It is these drawings that will be provided.

For the other eight items in the contention, LILCO either does not intend to install equipment or there is no safety related equipment in the secondary contain-ment.

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l SOC Contention 7.A(2)

6. Attached is a revision to a diagram included with LILCO's March 17, 1982 response.

SOC Contention 16

2. As noted in LILCO's March 17, 1982 answers, a response to NUREG-0630 was in the process of being completed.

Attached is a copy of that response, SNRC-679, dated March 15, 1982.

SOC Contention 19 LILCO was ordered to answer certain questions related to SOC Contention 19 by April 4, 1982.

Answers are being prepared and will be sent to SOC within the time allowed by the Board.

Respectfully submitted, LONG ISLAND LIGHTING COMPANY n ji W.~ Taylor R ey,'III Anthony F. arley, Jr. f[/ .

V Daniel 0. Flanagan Hunton & Williams P.O. Box 1535 Richmond, Virginia 23212 Dated: March 26, 1982

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SOC Contention 16, #2

' ,7 LONG ISLAND LIGHTING COM PANY j.'[h,}h SHOREHAM NUCLEAR POWER STATION P.O. BOX 618. NORTH COUNTRY RO AD e WADING RIVER. N.Y.11792 March 15, 1982 SNRC-679 Mr. Harold Denton, Director .

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Safety Evaluation Report - Licensing Condition No. 2 Fission Gas Release, Ballooning and Rupture Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

References:

1) NEDE 23785-1-P "GESTR-LOCA and SAFER Models for the Evaluation of the Loss of Coolant

- Accident," Revision 1, December 1981, Volumes 1 and 2

2) NEDE 23786-l'-P " Fuel Rod Prepressurization -

Amendment 1," May 1978

3) Letter from R. H. Bucholz (GE) to L. S.

Rubenstein (NRC), General Electric Fuel Clad Swelling and Rupture Model," May 15, 1981

4) Letter from J. F. Quirk (GE) to L. S.

Rubenstein (NRC), " General Electric Analytical Model for Calculation of Cladding Rupture Strain and Maximum Local Oxidation in LOCA Analysis," October 19, 1981

5) Letter from J. F. Quirk (GE) to L. S.

• Rubenstein (NRC), " General Electric Analytical .

Model for Calculat'.on of Local Oxidation in LOCA Analysis," Seitember 14, 1981

Dear Mr. Denton:

The Shoreham Nuclear Power Station - Unit 1 Safety-Evaluation Report (SER) , Supplement No. 1 states in Sections 4.2.3.2 and 4.2.3.3 that the Shoreham license will be conditioned to require ECCS reanalysis for second cycle and beyond utilizing models that (1) account for effects of high burnup fission gas release and prepressurized fuel, (2) accommodate the information in NUREG-0630, including its effects on local oxidation, and-(3) have been reviewed

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1 e l March 15, 1982 SNRC-679 Page 2 and approved by the NRC. LILCD has determined that this licensing condition is unwarranted, and that no further analysis is necessary.

The following provides our basis for this determination.

Effects of Fission Gas Release and Prepressurized Fuel An improved fission gas release (FGR) model was submitted to the NRC for review as part of General Electric's overall fuel perfor-mance code (Reference 1, Volume I) in December 1981. Concurrent with the submittal of the improved fuel performance model, GE submitted an improved ECCS evaluation model (Ref erence 1, Volume II) based on more realistic loss-of-coolant accident (LOCA) analysis methods and inputs. Use of the combined realistic ECCS evaluation model and improved fuel performance (FGR) model has been chown to provide large margins in calculated PCTs. The NRC approval of the models is expected by December, 1982. .

The ECCS Calculations for Shoreham were done using the current GE evaluation mcdel with the older fission gas release model. The calculated PCT was within the 10CFR50.46 limit of 22000F, Any calculations performed using the realistic ECCS model with the latest FGR correlation would predict PCTs substantially lower than the current analysis. Therefore, a specific reanalysis for the Shoreham plant using the latest FGR model is not necessary. A GE licensing topical report (Reference 2) showed that the use of pre-pressurized fuel in the BWR reduced the calculated PCT by 00F to 600F. Since the current Shoreham ECCS analysis predicts PCT values less than 22000F a reanalysis accounting for prepressurized fuel would only provide improved margin and is not justified.

Fuel Cladding Swelling and Rupture (Including Local Oxidation)

General Electric has performed several generic studies ',o address the NRC concern related to the fuel cladding swell and rupture model utilized in the current GE-BWR evaluation model for loss-of-coolant accident (LOCA) analyses. The results of those studies have been submitted to.the NRC (see References 3, 4, and 5) . Although their review is not yet complete, the NRC staff has agreed verbally with GE on the content of the report, and all of the key issues are believed resolved. ,

The submitted studies show that no changes to the current GE fuel cladding swell and rupture model are required to meet 10CFR50 Appendix K requirements for loss-of-coolant accidents. Key points from these studies are listed below:

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March 15, 1982 SNRC-679 Page 3

1. The GE model conservatively bounds 90% of all experimental data relevant to BWR conditions. This data base includes results from GE experiments, as well as from NUREG-0630 and other sources (see Reference 3) which were obtained under conditions prototypical of the BWR (i.e., cold shroud, slow heatup rates).

2. Sensitivity studies were run to determine effects of increased rupture strain on the peak cladding temperature (PCT). These studies were performed using a base c ase plant with a long reflood time and a short blowdown perlod which was bounding for all BWRs. The majority of the stidies were performed using prepressurized 8x8 fuel as used in Shoreham. ,

3. Several different sensitivity studies were performed to compare the effects of various bundle location configurations of high rupture strains to the results from the current model. The high strain cases were also compared to modified base cases to assure similar percentages of flow blockages.

4. All the studies show decreases in PCT (up to 400F) with the higher rupture strains. The reduction in PCT is due mainly ,

to the increased heat transfer area available at the higher strains for the ruptured rods.

5. Zircaloy oxidation heating has always been accounted for in' current GE model (see Reference 5) . In these sensitivity studies, it was shown that any temperature increase due to increased zircaloy oxidation heating for the case of higher strain was more than offset by the improved heat removal from the rods due to larger surface area.

6. In these sensitivity studies, it was shown that increasing i the maximum perforation strain had an insignificant effect on the calculated maximum local oxidation fraction, i.e., a .

greater than 50% increase in strain amounted to only a 5%

increase in the calculated local oxidation fraction. The reason for this small sensitivity is the decrease in cladding temperature as a result of the larger strain.

These studies submitted to the NRC justify continued use of the current General Electric cladding swelling and rupture model for BWR LOCA analysis. No changes to that model are necessary to meet the 10CFR50 Appendix K requirements.

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March 15, 1982 UNRC-679 Page 4 Models Reviewed and Approved by NRC The Shoreham ECCS calculations were done with General Electric's current evaluation model which has been reviewed and. approved by the NRC.

Reanalysis with more realistic GE models will only provide increased PCT margin for Shoreham. This reanalysis requirement is therefore unnecessary and should be removed from the licensing condition.

Please advise if you have any questions or require additional .

information.

Very truly yours, ongner epwa rw J. L. Smith Manager, Special Projects

Shoreham Nuclear Power Station RWG:mp cc: J. 'Higgins bcc: A.F. Earley R.A. Kubinak R.A. Loper M.H. Milligan A.E. Pedersen E.J. Youngling .

E.J. Brabazon .

J.T. Murphy D. Toner J.U. Valente

- W. Tunney P. Bohm '-

Eng. File /SR2..A21.010 '

Dist. List #14

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In the Matter of LONG ISLAND LIGHTING COMPANY (Shoreham Nuclear Power Station, Unit 1)

Docket No. 50-322 (OL)

CERTIFICATE OF SERVICE i

I hereby certify that copies of LILCO's RESPONSE TO SUFFOLK COUNTY INTERROGATORIES AND TO SUFFOLK COUNTY SECOND SET OF INTERROGATORIES, LILCO's RESPONSE TO SUFFOLK COUNTY's REQUEST FOR PRODUCTION OF DOCUMENTS and LILCO's FURTHER RESPONSE TO SOC's MARCH 1, 1982 INTERROGATORIES AND REQUESTS FOR PRODUCTION OF DOCUMENTS were served upon the following people by first-class mail, postage prepaid, on March 26, 1982, ,

except for the asterisked people, who were served by-hand or by Federal Express on March 26, 1982.

I Lawrence Brenner, Esq.* Atomic Safety and Licensing Administrative Judge Appeal Board Panel  ;

Atomic Safety and Licensing U.S. Nuclear Regulatory Board Panel Commission t U.S. Nuclear Regulatory Washington, D.C. 20555 ,

Commission '

l Washington, D.C. 20555 Atomic Safety and Licensing Board Panel Dr. Peter A Morris

• U.S. Nuclear Regulatory Administrative Judge Commission Atomic Safety and Licensing Washington, D.C. 20555 ,

Board Panel U.S. Nuclear Regulatory Bernard M. Bordenick, Esq.

Commission David A. Repka, Esq.

Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Dr. James H. Carpenter

• Washington, D.C. 20555 Administrative Judge Atomic Safety and Licensing David J. Gilmartin, Esq.

Board Panel Attn: Patricia A. Dempsey, Esq.

U.S. Nuclear Regulatory County Attorney Commission Suffolk Councy Department of Law Washington, D.C. 20555 Veterans Memorial Highway Hauppauge, New York 11787 Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555

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Herbert H. Brown, Esq.* Howard L. Blau, Esq.

Lawrence Coe Lanpher, Esq. 217 Newbridge Road Karla J. Letsche, Esq. Hicksville, New York 11801 4 Kirkpatrick, Lockhart, Hill, Christopher & Phillips Matthew J. Kelly, Esq.

8th Floor Staff Counsel, New York 1900 M Street, N.W. State Public Service Commission Washington, D.C. 20036 3 Rockefeller Plaza Albany, New York 12223 Mr. Mark W. Goldsmith

• Energy Research Group Mr. Jay Dunkleberger 400-1 Totten Pond Road New York State Energy Office Waltham, Massachusetts 02154 Agency Building 2 Empire State Plaza MHB Technical Associates

• Albany, New York 12223 1723 Hamilton Avenue Suite K j San Jose, California 95125 l Stephen B. Latham, Esq.*

Twomey, Latham & Shea 33 West Second Street P. O. Box 398

! Riverhead, New York 11901

Ralph Shapiro, Esq.

Cammer and Shapiro, P.C.

9 East 40th Street New York, New York 10016

\

Daniel O. Fla%naga$

Hunton & Williams 707 East Main Street P. O. Box 1535 Richmond, Virginia 23212 DATED: March 26, 1982 -

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