ML18022A900
| ML18022A900 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/23/1992 |
| From: | Mccarthy D CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLS-92-199, NUDOCS 9207300175 | |
| Download: ML18022A900 (8) | |
Text
ACCELERATED DISTRIBUTION DEMONSTPA.TION SYSTEM REGULAT INFORMATION DISTRIBUTION STEM (RIDS)
I ACCESSION NBR:9207300175 DOC.DATE: 92/07/23 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH.NAI.".E AUTHOR AFFILIATION SCCA'RTHY,D.C.
Carolina Power
& Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
"Annual Rept of ECCS Evaluation Model Changes,"
covering June 1991 to May 1992.Marginal utilization tables, indicating effects of permanent assessment of peak clad temp margin of large
& small break LOCA encl.
R DI'STRIBUTION CODE:
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TITLE: OR Submittal: General Distribution NOTES:Application for permit renewal filed.
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D NOTE TO ALL"RIDS" RECIPIENTS:
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SME Carolina Power 8 Light Company IJUL 33 ]99P.
SERIAL:
NLS-92-199 10 CFR 50.46 United States Nuclear Regulatory Commission ATTENTIONs Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL CHANGES Gentlemen:
The purpose of this letter is to provide the annual report pursuant to 10 CFR 50.46(a)(3)(ii) for the Shearon Harris Nuclear Power Plant (SHNPP) regarding the estimated effect of changes or errors in Emergency Core Cooling System (ECCS) evaluation models or in the application of the models.
This report covers the period of June 1991 through May 1992.
There have been no changes to the Westinghouse ECCS Evaluation Model during this reporting period.
However, there are supplements to the analyses of record which were implemented during the previous reporting period, June 1990 through May 1991, which continue to be applicable.
These supplements resulted in permanent assessment of Peak Clad Temperature (PCT) margin as reported to the NRC by Carolina Power 6 Light Company (CPGL) on July 26, 1991.
The enclosures to this letter provide the Margin Utilization Tables which show the effects of permanent assessment of PCT margin for various issues for the Large and Small Break Loss of Coolant Accident (LOCA) analyses of record.
The information provided is identical to that provided in the July 26, 1991 letter, with one exception.
Subsequent investigation has resulted in an adjustment to the value of peak clad temperature for Small Break LOCA (SBLOCA) due to a revision to the Rod Internal Pressure Assumption from 10'F to 33.3'F.
As a point of clarification, the July 26, 1991 report showed 30'F for the SBLOCA Rod Internal Pressure Assumption.
This item consisted of two components, SBLOCA Cladding Creep Model (20'F) and SBLOCA Rod Internal Pressure Assumption (10'F).
In this annual report, these items are listed separately as Items D.3 and D.4, and only the Rod Internal Pressure value has changed.
The Margin Utilization Tables also include a listing of current open issues which are being investigated, but for which no conclusions have yet been reached.
Should the resolution of these issues impact the 10 CFR 50.46 ECCS evaluation, then CPt L will report them in the next annual report.
- Likewise, if any of the issues result in significant change in the calculated
- PCT, as defined by 10 CFR 50.46, they will be reported to the NRC in accordance with the 30-day reporting requirement.
<90036 411 Fayettevitte Street o P. O. Box 1551 o Raleigh, N. C. 27602 9207300i75 92072%
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NLS-92-199 Page 2
As stated in a February 19, 1992 letter, OPED plans to reanalyze both the large and small break LOCA events prior to Cycle 6 for SHNPP, to coincide with a change in nuclear fuel vendors.
Accordingly, these analyses of record will be approved and in place to support Refueling Outage No. 5, currently scheduled to begin in March 1994.
Questions regarding this matter may be referred to Mr. R.
W. Prunty at (919) 546-7318.
Yours very truly, D. C. McCarthy Manager Nuclear Licensing Section SDC/sdc
Enclosures:
CC ~
Mr. S.
D. Ebneter Mr. N. B. Le Mr. J.
E. Tedrow
REPORTXNG OF 10CFR50.46 MARGXN UTXLXZATXON LARGE-BREAK LOCA Plant Name:
UtilityName:
Shearon Harris Unit 1 (CQL)
Carolina Power
& Light'ompany A.
Anal sis of Record Vantage 5 Reload Transition Safety Report, February 1989 Evaluation Model: 1981 With BASH, FQT 2.45, FAH 1.65 SGTP 6%, Other: Vantage 5
Fuel Margins Used: Transition Core Penalty PCT 2105.2'F APCT
+ 50'F B.
D.
Prior LOCA Model Assessments
- 1989 (Permanent Assessment of PCT Margin - Letter ¹: N/A)
Prior LOCA Model Assessments
- 1990 (Permanent Assessment of PCT Margin - Letter ¹: CQL-91-011)
Prior LOCA Model Assessments
- 1991 (Permanent Assessment of PCT Margin - Letter ¹: CQL-91-035)
APCT
+
O' APCT
+
O'
- 1. Fuel Rod Initial Condition Inconsistency 2.
LBLOCA Burst and Blockage Assumption 3.
Steam Generator Tube Seismic/LOCA Assumption~
APCT
+ 10'F APCT
+
O' APCT
+
O' E.
10CFR50.59 Safet Evaluations Permanent Assessment of PCT Mar in F.
Issue:
SI Delay Time Increase Letter:
CQL-90-573 and 90CP*-G-0080 Current Permanent PCT APCT
+
4'F PCT 2169.2'F G.
Current LOCA Model Issues No PCT Assessments Made'till Under Investi ation
- l. IFBA Fuel (see reference 3)
- 2. Containment Initial Temperature (see reference 1) 3.
WREFLOOD Discrepancies (see reference 7) 4.
Steam Generator Tube Collapse Under Combined LOCA/Seismic Loads (partial through-wall cracking)
(see reference 8) 5.
LOCTA Coding Errors (see reference 3) 6.
IMP Data Base Errors (see reference 5)
- 7. Hot Leg Recirculation Switchover (see reference 6) 8.
Core Average Zirc-Water Reaction (see reference 4)
- 9. Beginning of Life Rod Internal Pressure Uncertainties (see reference 2)
~ 1.8 percent steam generator tube plugging margin was assessed to account for possible steam generator tube collapse in the event of a seismic event coincident with a LOCA.
Note that this issue accounted for only full through-wall cracking.
(1708HNP)
REPORTING OF 10CFR50.46 MARGIN UTILIZATION SMALL-BREAK LOCA Plant Name:
Utility Name:
Shearon Harris Unit 1 (CQL)
Carolina Power 6 Light Company B.
D.
Anal sis of Record Vantage 5 Reload Transition Safety Report, February 1989 Evaluation Model:
- NOTRUMP, FQT 2.50, FAH - 1.65 SGTP - 6%, Other:
Vantage 5
Prior LOCA Model Assessments
- 1989 (Permanent Assessment of PCT Margin - Letter ¹: N/A)
Prior LOCA Model Assessments
- 1990 (Permanent Assessment of PCT Margin - Letter ¹: CQL-91-011)
Prior LOCA Model Assessments
- 1991 (Permanent Assessment of PCT Margin - Letter ¹: CQL-91-035)
PCT 1779.8'F APCT
+
O' BPCT
+
O'
- 1. Fuel Rod Initial Condition Inconsistency 2.
NOTRUMP Solution Convergence Reliability 3.
SBLOCA Cladding Creep Model 4.
SBLOCA Rod Internal Pressure Assumptioni BPCT
+ 37'F APCT
+
O' APCT
+ 204F APCT -
+ 33.3'F E.
10CFR50.59 Safet Evaluations Permanent Assessment of PCT Mar in
- 1. Issue:
AFV Enthalpy Delay Time Increase Letter:
CQL-90-535 APCT
+ 36'F
- 2. Issue:
ECCS Flow Shortfall Evaluation Letter:
90CP*-G-0080 and CQL-90-573 Current Permanent PCT APCT
+ 75'F PCT 1981.1'F Current LOCA Model Issues No PCT Assessments Made Still Under Investi ation
- 1. Beginning of Life Rod Internal Pressure Uncertainty Impact on Safety Analysis (see reference 2) 2.
IFBA Fuel (see reference 3) 3.
LOCTA Coding Methodology Issues (see reference 3)
- 4. Main Feedwater Isolation (see reference 4) 5.
IMP Data Base Errors (see reference 5)
- 6. Hot Leg Recirculation Switchover (see reference 6) 7.
Core Average Zirc-Water Reaction (see reference 4) 8.
SBLOCA Burst and Blockage (see reference 4) i This penalty was increased to account for the low backfill pressure IFBA fuel.
(1708HNP)
REFERENCES 1.
ET-NRC-92-3699, "Results of Technical Evaluation of Containment Initial Temperature Assumptions for Large-Break Loss-of-Coolant Accident,"
June 1,
1992 from N. J. Liparulo (W) to NRC.
2.
ET-NRC-92-3695, "Interim Report of a Deviation or Failure to Comply Pursuant to 10CFR21.21(a)(2)," April 30, 1992 from N. J. Liparulo (W) to NRC.
3.
ET-NRC-92-3718, "Interim Report of a Deviation or Failure'to Comply Pursuant to 10CFR21.21(a)(2)," July 1, 1992 from N. J. Liparulo (W) to NRC.
ET-NRC-91-3647, "Interim Report of a Deviation or Failure to Comply Pursuant to 10CFR21.21(a)(2),"
December 20, 1991 from S.
R. Tritch (W) to NRC.
5.
ET-NRC-92-3655, "Interim Report of a Deviation or Failure to Comply Pursuant to 10CFR21.21(a)(2),"
January 21, 1992 from S.
R. Tritch (W) to NRC.
6.
ET-NRC-92-3712, "Interim Report of a Deviation or Failure to Comply Pursuant to 10CFR21.21(a)(2),"
June 23, 1992 from N. J. Liparulo (W) to NRC.
7.
92CP*-G-0078, "Shearon Harris Unit 1 - Cycle 5 Preliminary Expanded Text Reload Safety Evaluation (RSE) Report," June 25, 1992 from Beth Pearson McAtee (W) to Thomas Dresser, CP&L.
8.
CQL-92-025, "Shearon Harris Unit 1 - Steam Generator Tube Deformation and Potential Secondary to Primary Leakage,"
June 3,
1992 from G. J.
Murray (W) to J.
F. Nevill, CP&L.
(1708HNP)