BSEP-97-0085, Forwards ECCS Evaluation Models Which Summarizes Results & Shows Absolute Value of Changes & Errors Approx 23 F for GE7 Fuel

Forwards ECCS Evaluation Models Which Summarizes Results & Shows Absolute Value of Changes & Errors Approx 23 F for GE7 Fuel

ML20136H144
Person / Time
Site:	Brunswick
Issue date:	03/10/1997
From:	Jury K CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BSEP-97-0085, BSEP-97-85, NUDOCS 9703190063
Download: ML20136H144 (11)

Text

_ . . . .

i . CP&L t

Carolina Power & Ught Company P,0. Box 10429 Southport, NC 28461-0429 l March 10,1997 f

SERIAL: BSEP 97-0085

[ 10 CFR 50.46(a)(3)(ii) l U. S. Nuclear Regulatory Commission i ATTENTION: Document Control Desk l Washington, DC 20555 l

BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 ECCS EVALUATION MODELS l

Gentlemen:

l The Code of Federal Regulations (CFR), Title 10, paragraph 50.46(a)(3)(i) requires the licensee estimate the impact of changes and errors in Emergency Core Cooling System (ECCS) evaluation models or in the application of these models. Paragraph 10 CFR 50.46(a)(3)(ii) l specifies reporting requirements based on the sum of the absolute value of these changes and l _ errors in calculated peak cladding temperature (PCT). If the absolute sum of the changes or l errors is significant (i.e., exceeds 50 F), a 30-day report is required. If not, an annual report is l required summarizing the effect on the limiting ECCS analysis.

Since the 1988 revision to 10 CFR 50.46, General Electric (GE) has been compiling all changes and errors in their approved SAFER /GESTR ECCS evaluation method and providing that information to the NRC and Carolina Power & Light (CP&L) Company. It has been the position of GE and CP&L that this generic report satisfied the requirements of 10 CFR 50.46(a)(3)(i) and (ii) for the Brunswick Steam Electric Plant, Unit Nos.1 and 2.

l l The NRC staff has recently informed GE of the need for each licensee to file a report for each i f

plant. The report may use applicable GE " generic" reports, plus any plant-unique changes or errors. This letter is intended to fulfill this reporting requirement.

CP&L has evaluated the impact of generic and plant-unique changes and errors using References 1 through 15. Enclosure 1 summarizes the results and shows the absolute value of fI the changes and errors is approximately 23 F for GE7 fuel (10 F for GE13 fuel). This value represents the impact on the Brunswick Steam Electric Plant of PCT changes and errors which have occurred since 1988. No changes in plant design or operation are required as a result of this letter.

9703190063 970310 E PDR ADOCK 05000324 P PDR l

h,k .

NRC Document Control Desk BSEP 97-0085 / Page 2 Please refer any questions regarding this submittal to Mr. Mark Turkal, Supervisor - Licensing at (910) 457-3066.

Sincerely, Keith R. Jury Manager- Regulatory Affairs Brunswick Nuclear Plant WRM/wrm Enclosures

1. ECCS Analysis Change / Error Summary

2. List of Regulatory Commitments i

t

I l

! l l ,

l l

i ' 1 NR'C Document Control Desk l l

BSEP 97-0085 / Page 3 l l

1 pc (with enclosures)-

U. S. Nuclear Regulatory Commission ATTN.: Mr. Luis A. Reyes, Regional Administrator ,

101 Marietta Street, N.W., Suite 2900 )

Atlanta, GA 30323-0199  ;

l Mr. C. A. Patterson NRC Senior Resident Inspector- Brunswick Steam Electric Plant, Units 1 and 2 U.S. Nuclear Regulatory Commission

-l l ATTN.: Mr. David C. Trimble, Jr. (Mail Step OWFN 14H22) j l 11555 Rockville Pike '

Rockviile, MD 20852-2738 The Honorable J. A. Sanford Chairman - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 -

i 1

l l

I-i i

l l

.... - -. _.- -- . . . . . ..-. - ~ _ _ ,.... - _ . . - ... - ... - . . . . - . . - . .- . _ _ _

ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62

. ECCS EVALUATION MODELS L

1 h

l BRUNSWICK SAFER /GESTR-LOCA ECCS ANALYSIS CHANGE / ERROR

SUMMARY

l t

I i

i i

l l

l l

l i

t i

, ._ _ -. . - . - . _ _ . _ . _ . _ . _ . _ . . . - - . . . . _ _ _ _ . _ . . _ _ - _ . . _ . . . . _ . . _ . _ _ _ _ _ . . _ . m m.m l

l i

1

, BSEP SAFER /GESTR LOCA ECCS ANALYSIS CHANGE / ERROR

SUMMARY

l Change Notre or Penod Change or Error Desenption GE BWR Eshmated BSEP Cumulatue is cumulative or Error Document Covered Estimated PCT BSEP PCT BSEP PCT i Nohce Date PCT Impact change. change l

Impact and greater than Eshmated PCT 50*F?-

l Base (09M8) N/A Change to SAFER /GESTR-LOCA Methodology N/A Approx. Approx. Yes LOCA (NEDE 23785, SAFER 02) -660*F 660*F Analysis Calculated results: Switch to Switch to NEDC Unit 1 Unit 2 Best Estimate Best 31624P Licensing Basis PCT 1533*F (GE7)1537'F (GE7) LOCA Estimate Rev 0 Peak Local Oxidahon 0299% (GE7) 0.306% (GE7) Method LOCA Metal-WaterReaction 0.046% (GE7)0.036% (GE7) Method I NEDC (02S 0) N/A This revision was to correct cbcumentation errors (the N/A N/A N/A No l 31624P reported value of an input parameter) found in Reset by Rev 1 Revision 0. However, Revision 1 contained U11533*F GE7 06/01/89

! pubhcation errors and was never dstributed for use. U21537aF GE7 SER NO CALCULATIONS WERE MADE.

NFM 06/13/90 10/17/88- Codng enors in SAFER - energy balance in unheated +20*F to Unknown, 20*F No 023-90 06/1 % 0 bottom node, modeling of steam quenching by the 100*F assume +20*F LPCI system, calculabon of the bypass void profile.

U11553*F GE7

! U21557'F GE7 l NEDC N/A Corrected publication errors in Revision 0. NO N/A No (07/90) 20*F No i 31624P~ CALCULATIONS WERE MADE.

I Rev 2 U11553*F GE7 l U21557'F GE7 l

I NFM 03/12/91 06/13/90 No errors or changes GE nohfied NRC that they are N/A No 20*F No

025-91 03/12/91 in the process of adapting SAFER /GESTR L methodology to accommodate GE11. U11553*F GE7 U21557*F GE7 i

?

! NFM 06/26 S 2 03/12/91- No changes or errors. Sensitivity to changes in less than N/A 20*F No 058-92 06/26 S 2 computer operating system or small changes in input 150*F parameters (jet pump loss coefficients). This does U11553*F GE7 not apply to BSEP because BSEP analyses have not U21557'F GE7 been run on a new computing system.

l F.

1 i

?

1

, 1 BSEP SAFERIGESTR LOCA ECCS ANALYSIS CHANGE / ERROR

SUMMARY

l Change Notice or Period Change or Error Desenpbon - GE BWR Estimated BSEP Cumulative is cumulative 1 or Error Document covered Estimated PCT BSEP PCT BSEP PCT l

Notice Date PCT Impact change. change Impact and greater than Estimated PCT 50*F7 j l

NFM 06/30 S 3 06/26 S 2 Two moor codng errors, improper upper plenum flow iS*F Unknown, 25*F No  !

, 09493 06/30 S 3 initiahzation and sgn error in a pressure drop assume +5*F l l calculation for the top of the hot channel  :

U11558*F GE7 Noted that model sensilmty to small input parameter U21562*F GE7

changes could be greater than that orginally reported j

l. in NFM 058-92. Vanaten could be greater than +/- 1 l 50*F for some BWR/4 plants with LPCI injectxm into I the lower plenum (total variabon of less than 85*F for i most cases but with one case showing a range of 102*F).

Also noted that the identified sensitivity is related to l

the explicit numencal treatment in SAFER combined with rapid and simultaneous vanations of multiple .

parameters. Sensitivity will be limited through better  !

control of time steps in the computations.

].

The second and third paragraphs above do not apply to BSEP because BSEP analyses have not been l rerun LPCs. Sept.1993 N/A GE Document GE NE-20845-0393, DRF-A00 N/A increase of 13*F 13*F No 00S 05316.

• Low Pressure Core Spray Out-of4ervice for Report BSEP Uruts 1 and 2? U11546*F GE7 Cumulative U21550*F GE7 change Section 3.2 establishes a basehne between the reduced

! current version of SAFER (SAFER 04) and the onginal from 25'F to BSEP heensing basis verson of SAFER (SAFER 02). 13*F based SAFER 04 includes several modehng improvements on revised relative to SAFER 02. estimate of prior effects This basehne companson also includes several ECCS obtained parameter relaxations relative to the hcensing basis from the (L3=LL1 changed from 519.5 to 517 and L1=LL3 LPCS-00S changed form 3695 to 3580). These water level Report

l. setpomt changes impact the PCT by only l approximately 1*F and therefore this basehne calculation gives a very good indcation of the effect of modeling changes and errors on the licensing basis PCTs.

The results indicate that, with the updated version of SAFER, the licensing basis PCT for BSEP2 GE7 fuel increases from 1537'F to 1550*F. (Approximately 1*F of this increase is due to the reduced water level setpoints assumed)

NFM 0741S4 06/30 S 3- No changes or errors. N/A N/A 13*F No 088-94 0741S4 i' U11546*F GE7 U21550*F GE7 1

i NFM 06/24S 5 0741S4- No changes or errors N/A N/A 13*F No

! 087 95 06/24S 5 4 U11546*F GE7 U21550*F GE7 i

I i

r

. . - - ._ . -._ . . . _ ~ _ _ . . . _ . . _ _ _ _ _ _ _ . _ _ _ _ . _ . - - ~ - .- . -

l BSEP SAFER /GESTR LOCA ECCS ANALYSIS CHANGE / ERROR

SUMMARY

Change Notee or Penod Change or Error Descrption GE BWR Estimated BSEP Cumulatwe is cumulatwo or Error Document Covered Eshmated PCT BSEP PCT BSEP PCT Notes Date PCT Impact change. change Impact and greater than Estimated PCT 50*F7 1

NFM 12/15/95 074)1/94- Revises NFM 087 95 Pointed out that m March of 10*F Unknown, 23*F No 278-95 12/15/95 1995 a domeste utshty asked about the impact of the assume 10*F RPV bottom head drain (BHD) on LOCA evaluabons GE stated that the BHD impact is believed to be less U11556*F GE7 than 10*F. U21560*F GE7 NEDC (01/96) PFA BSEP GE13 LOCA analysis (NEDC-23785, N/A Unit 1 beensing 23*F GE7 No 31624P SAFER 04) basrs PCT 10*F GE13 Sup.3 increased by 2*F Rev. O Relatwo to the base analysis, this ana?ysis was done relatwo to onginal with a revised version of the SAFER model GE7 calculabons but no increase i Calculated results: relatwe to the l GE13 current eshmated

! Licensing Basis PCT 1535*F GE 7 PCT.

Peak localOxidabon 0.27 % 027%

Metal-Water Reachon 0.036 % 0036% (GE7) 10*F increase for GE13 due to the BHD issue Summary of BSEP ECCS Results as of B1C11

and B2C12 U11556*F GE7 Unit 1 Unit 2 Licensing Basis PCT 1535*F GE13 1537'F GE7 U11545*F GE13

, Peak Local Oxidation 0.299%GE7 0.306%GE7 U21545*F GE13 l MetatWater Reachon 0.046% GE7 0036% GE7 l

Note that this analysis dd not attempt to model BHD effects Therefore, thePCTimpactdesenbedmNFM 278-95 applies to the GE 13 analysis.

NFM 02/20/96 12/15/95- Revises NFM 278-95. Just adds a drawing to more N/A N/A No

' 23*F GE7 020 96 02/20/96 eleaily descrbe the BHD concern No change in 10*F GE13 eshmated PCT impact. U11556*F GE7 U21560*F GE7 l

U11545'F GE13 U21545*F GE13 NFM 06/28/96 02/02/96 No changes or errors but some apphcations used the less than tFA 23*F GE7 No 088-96 06/28/96 wrong number of fuel rods for GE9/10/11/12/13 30*F 10*F GE13

, (designs containing large water rods) U11556*F GE7 i U21560*F GE7 l It has been confirmed that this nobce does not apply to BSEP LOCA evaluabons. U11545'F GE13 U21545'F GE13 I

I

i  !

BRUNSWICK STEAM ELECTRIC PLANT l PCT CHANGE

SUMMARY

l. (SEPTEMBER 1988 THROUGH JUNE 1996 j t

Unit and Fuel Type Greatest PCT Current Estimated Change from Reported in a LOCA PCT Reported PCT ,

l Document Submitted to NRC j I

BSEP1 GE7* 1533 F 1556 F +23 F l

BSEP1 GE13 1535 F 1545 F +10 F r

! BSEP2 GE7* 1537 F 1560 F +23 F

j. BSEP2 GE13 ' 1535 F 1545 F +10P 1

GE7 bounds GE8, GE9, and GE10.

i l'

i t

1 i

+

?

1 4 f 5

3- j N

i CHANGE AND ERROR REPORT

REFERENCES:

1) Letter, R. C. Mitchell to Director of Nuclear Reactor Regulation, " Reporting of Changes i and Errors in ECCS Evaluation Models," June 13,1990 (MFN 023-90). ]

1

2) Letter, P. W. Marriot to Office of Nuclear Reactor Regulation, " Reporting of Changes and l Errors in ECCS Evaluation Models," March 12,1991 (MFN 025-91).

l

3) Letter, S. J. Stark to Office of Nuclear Reactor Regulation, " Reporting of Changes and 1 Errors in ECCS Evaluation Models " June 26,1992 (MFN 058-92).

4) Letter, R. C. Mitchell to Office of Nuclear Reactor Regulation, " Reporting of Changes and Errors in ECCS Evaluation Models," June 30,1993 (MFN 090-93).

5) Letter, R. C. Mitchell to Office of Nuclear Reactor Regulation, " Reporting of Changes and l Errors in ECCS Evaluation Models," July 1,1994 (MFN 088-94).

l

6) Letter, J. F. Klapproth to Document Control Desk (R.C. Jones, Jr.), " Reporting of Changes and Errors in ECCS Evaluation Models," June 24,1995 (MFN 087-95). i

7) Letter, R. J. Reda to Document Control Desk (R.C. Jones, Jr.), " Reporting of Changes and Errors in ECCS Evaluation Models," December 15,1995 (MFN 278-95).

1

8) Letter, R. J. Reda to Document Control Desk (R.C. Jones, Jr.), " Reporting of Changes and Errors in ECCS Evaluation Models," February 20,1996 (MFN 020-96).

9) Letter, R. J. Reda to Document Control Desk (R.C. Jones, Jr.), " Reporting of Changes and Errors in ECCS Evaluation Models," June 28,1996 (MFN 088-96).

BSEP LOCA ANALYSIS REFERENCES

10) NEDC-31624P, " Brunswick Steam Electric Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis (Revision 0)," September 1988.

11) Letter, E. G. Tourigny to L. W. Eury, " SAFER /GESTR-LOCA Analysis, Brunswick Steam Electric Plant, Units 1 and 2 (TAC Nos. 72854/72855)," June 1,1989 (NRC-89-401).

Transmitted NRC SER for NEDC-31624P, Revision 0.

l 12) NEDC-31624P, " Brunswick Steam Electric Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis (Revision 2)," July 1990.

13) Letter, Ngoc B. Le to L. W. Eury, " Revision of SAFER /GESTR-LOCA Analysis - Brunswick Steam Electric Plant, Units 1 and 2 (TAC Nos. 77585 and 77586)," January 10,1991

! (NRC-91-018). Transmitted NRC SER for NEDC-31624P, Revision 2.

14) NEDC-31624P, " Brunswick Steam Electric Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis: Application to GE13 Fuel," January 1996.

Transmitted to NRC with B2C12 COLR. NRC SER not required since this is an application of an approved methodology, not a methodology change.

l l

l l

.)

15) ' GE-NE-208-05-0393 (DRF-A00-05316), " Low Pressure Core Spray Out-Of-Service for BSEP Units 1 and 2," September 1993. (This report has not been sent to the NRC. It contains a comparison between SAFER 02 and SAFER 04 results.]

l l

l i

l l

ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 ECCS EVALUATION MODELS l

LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Carolina Power & Ught (CP&L)

Company in this document. Any other actions discussed in the submittal represent intended or planned actions by CP&L. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager - Regulatory Affairs at the Brunswick Steam l Electric Plant of any questions regarding this document or any associated regulatory commitments.

1 l

Commitment Committed date or outage l 1. None. N/A t

{

i l

l l

f