ML17261B032

Forwards Annual Rept of ECCS Model Revs as Applicable to Facility,Per 10CFR50.46.Mods to Model for Small Break LOCA Do Not Affect Calculated Peak Clad Temp

ML17261B032
Person / Time
Site:	Ginna
Issue date:	03/28/1990
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson Office of Nuclear Reactor Regulation
References
NUDOCS 9004130169
Download: ML17261B032 (11)

Text

ACCELERATED DISTjUBUTION DEMONSHRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9004130169 DOC.DATE: 90/03/28 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas 6 Electric Corp.

RECIP.NAME RECIPIENT AFFILIATXON JOHNSON,A.R. . Project Directorate I-3

SUBJECT:

Forwards annual rept of ECCS model revs as applicable to facility,per 10CFR50.46.

DISTRXBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR Submittal: General Distribution .S NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 RECIPIENT COPIES RECIPIENT COPIES A ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL N

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ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N Y. 14649.0001 TELEP NON E March 28, 1990 AREA COOK 7ld 546-2700 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Mr. Allen R. Johnson PWR Project Directorate I-3 Washington, D.C. 20555

Subject:

10CFR50.46 Annual Report ECCS Evaluation Model Revisions R.E. Ginna Nuclear Power Plant Docket No. 50-244

Reference:

1) NS-NRC-89-3463, "10CFR50.46 Annual Notification for 1989 of Modifications in the Westinghouse ECCS Evaluation Models", Letter from W.J. Johnson (Westinghouse) to T.E. Murley (NRC), Dated.

October 5, 1989.

2) NS-NRC-89-3464, "Correction of Errors and Modifications to the NOTRUMP Code in the Westinghouse Small Break LOCA ECCS Evaluation Model Which Are Potentially Significant", Letter from W.J. Johnson (Westinghouse) to T.E. Murley (NRC), Dated October 5, 1989.

Dear Mr. Johnson:

This letter provides the annual report of Emergency Core Cooling System (ECCS) model revisions as they apply to R.E.

Ginna. Zn References 1) and 2), Westinghouse Electric Corporation provided information regarding modifications to their ECCS evaluation models to NRC Staff. References 1) and

2) describe the generic effects of the model revisions for both large and small break Loss Of Coolant Accidents (LOCA).

The attachment to this letter provides information regarding the effects of the ECCS evaluation model modifications on the Ginna UFSAR Chapter 15.6.4 LOCA analysis.

Modifications to the model cause the large break LOCA Peak Clad Temperature (PCT) to increase by 2 F to 1889 F.

9004i30i69 90032S PDR ADOCK 05000244 R PDC i p'gget5/44'/

Modifications to the model for small break LOCA do not affect calculated. peak clad temperature.

Very truly yours, Robert C. Mecredy Division Manager Nuclear Pr'oduction RWEi088 Enclosures xc: Mr. Allen R. Johnson (Mail Stop 14D1)

Project Directorate I-3 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector

ATT 'O 10CFR50. 46 ANNUAL PORT Effect of Westinghouse ECCS Evaluation Model Modifications on the LOCA Analysis Results Found in Chapter 15.6.4 of the R.E. Ginna (RG&E) Nuclear Power Plant Updated, Final Safety Analysis Report and. WCAP-11609 Containing the Steam Generator Tube Plu in Re ort for Ginna Nuclear Power Station The October 17, 1988 revision to 10CFR50.46 required applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory'ommission (NRC) of errors and changes in the ECCS Evaluation Models on an annual basis, when the errors and changes are not. significant. Reference 1 defines a significant error or change as one which results in a calculated peak fuel cladding temperature different by more than 50'F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50'F.

In References 2 and 3, information regarding modifications to the Westinghouse large break and small break LOCA ECCS Evaluation Models was submitted to the NRC. The following presents an assessment of the effect of the modifications to the Westinghouse ECCS Evaluation Models on the Loss-Of-Coolant Accident (LOCA) analyses results found in WCAP-11609, "Steam Generator Tube Plugging (SGTP) Report for Ginna Nuclear Power Station", October 1987 and Chapter 15.6.4 of the R.E. Ginna (RG&E) Nuclear Power Plant Updated Final Safety Analysis Report.

LARGE BREAK LOCA - EVALUATION MODEL CHANGES The large break LOCA analysis for R.E. Ginna (RG&E) was examined to assess the effect of the applicable modifications to the Westinghouse large break LOCA ECCS Evaluation Model on Peak Cladding Temperature (PCT) results reported. in WCAP-11609. The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP) level of 12'-o.

The large break analysis was subsequently reanalyzed and licensed for 15% tube plugging and. the results were documented in WCAP-11609, "Steam Generator Tube Plugging (SGTP) Report for Ginna Nuclear Power Station", October 1987. The large break LOCA analysis results were calculated. using the 1981 version of the Westinghouse large break LOCA ECCS Evaluation Model which is documented in WCAP-9220-P-A (Reference 4). The current licensing basis analysis assumed .the following information important to the large break LOCA analysis NSSS power level 102-o of 1520 MWt Fuel Type ,14 X14 OFA Pellet Edge Configuration Chamfer Uniform Steam Generator Tube Plugging Level 15%

i Nuclear Peaking Factors of 2.32 for the Total Peaking Factor and 1.66 for the Enthalpy Rise Peaking Factor.

~~

r

~~or R.E. Ginna (R), the limiting break reslked from the double ended. guillotine rupture of the cold leg piping with a discharge coefficient of CD = 0.4. The calculated peak cladding temperature was 1887'F including the applicable penalties. The PCT value of 1887'F includes a 6'enalty to account for an upper plenum injection and. a core crossflow penalty'of 10'F.

The following modifications to the Westinghouse ECCS Evaluation Model were evaluated to determine if they would affect the current licensing basis -large break LOCA analysis results for R.E. Ginna for 15% SGTP found.'n WCAP-11609.

1981 ECCS Evaluation Model: (Not Max-SI Limited,)

In the 1981 version of the Westinghouse ECCS Evaluation Model, a modification was made to delay downcomer overfilling. The delay corresponds to backfilling of the intact cold legs. Data from tests simulating cold leg injection during the post-large break LOCA reflood phase which have adequate safety injection flow to condense all of the- available steam flow show a significant amount of subcooled. liquid to be present in the cold leg pipe test section. This situation corresponds to the so-called maximum safety injection scenario of ECCS Evaluation Model analyses.

The R.E. Ginna (RGGE) LOCA analysis performed with the Westinghouse 1981 large break LOCA ECCS Evaluation Model is not affected. by the WREFLOOD code modifications since the maximum safeguards safety injection flow assumption is not limiting.

1981 ECCS Evaluation Model.: (Two-Loop Plants)

In the 1981'version of the Westinghouse ECCS Evaluation Model, the pressurizer is modeled as being attached to the broken (faulted) loop in the SATAN-VI code for calculating large break blowdown behavior. Sensitivity studies were performed to determine if pressurizer location in the noding scheme was the most limiting this position. The results indicated that two-loop Westinghouse PWRs are sensitive'to the pressurizer nodal location and that in some cases modeling the pressurizer in the intact (non-faulted) loop resulted. in a slight increase in the calculated Peak Cladding Temperature (PCT). For a two loop plant, core cooling is provided by negative core flow and. the negative core flow period. lasts through most of the remaining blowdown period; The concern regarding pressurizer location relates to the negative core flow period which is crucial for core cooling in a two-loop plant. With the pressurizer on the broken loop, pressurizer flow is a large contributor to break flow (pump side), lessening the contribution from the upper plenum and leaving a large upper plenum inventory for negative core flow later in blowdown.

For Ginna, a penalty of 2'F due to modeling the pressurizer in the intact (non-faulted) loop was assessed to be used in tracking margin to the 10CFR50.46 limit on the current licensing basis analysis for 15-o SGTP contained. in WCAP-11609. The PCT for the limiting C =0.4 break is 1887'F with applicable penalties, and thus adding the 2'F increase brings the PCT, to 1889'F which is well below the 10CFR50.46 limit.

ks discussed abov modifications to the Nestgghouse large break LOCA ECCS Evaluation Model could affect the result by altering the

~ PCT.

A. Analysis Calculated Result 1887'F B. Modifications to Westinghouse ECCS Evaluation Model 2oF C. ECCS Evaluation Model Modifications Resultant PCT 1889 F SMALL BREAK LOCA - EVALUATION MODEL CHANGES The small break. LOCA analysis for R.E. Ginna (RG&E) was also examined to assess the effect of the applicable modifications to the Westinghouse ECCS Evaluation Models on Peak Cladding Temperature (PCT) results reported in Chapter 15.6.4.1 of the UFSAR. The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP) level of 12%. The small break LOCA analysis was subsequently evaluated and licensed for a SGTP level increase from 12: to 15%. The small break LOCA event was not reanalyzed because the results are not sensitive to a SGTP level increase from 12% to 15:. The evaluation was documented, and transmitted in WCAP-11609, "Steam Generator Tube Plugging (SGTP) Report for Ginna Nuclear Power Station", October 1987. The small break LOCA analysis results were calculated using the Westinghouse small break LOCA ECCS Evaluation Model documented in WCAP-8970 (Reference 12) which utilized the WFLASH computer code. For R.E. Ginna (RG&E), the limiting size small break resulted from a 6 inch equivalent diameter break in the cold leg. The calculated peak cladding temperature was 1092 F. The analysis assumed the following information important to the small break analyses:

NSSS power level 102% of 1520 MWt Fuel Type 14 Z14 OFA Pellet Edge Configuration Chamfer Uniform Steam Generator Tube Plugging Level 12-o Nuclear Peaking Factors of 2e32 for the Total Peaking Factor and 1.66 for the Enthalpy Rise Peaking Factor.

600 GPM Total Auziliary Feedwater Flow.

The following modification to the Westinghouse ECCS Evaluation Model were evaluated to determine if they would affect the current licensing basis small break LOCA analysis results for R.E. Ginna.

WFLASH ECCS Evaluation Model Following the accident at Three Mile Island Unit 2, additional attention was focused on the small break LOCA and. Westinghouse submitted a report, WCAP-9600 (Reference 5), to the Nuclear Regulatory Commission (NRC) detailing the performance of the Westinghouse small break LOCA Evaluation Model which utilized the WFLASH computer code. In NUREG-0611 (Reference 6), the NRC staff questioned the validity of certain models in the WFLASH, computer

~ t1 censees to justify contin V

code and required acceptance of the model.Section II.K.3.30 of NUREG-0737 (Reference 7), which clarified the NRC Post-TMI requirements regarding small break LOCA modeling, required that the licensees revise the small break LOCA ECCS models along the guidelines specified in NUREG-0611.

Following the issuance of NUREG-0737, Westinghouse and the Westinghouse Owners Group decided to develop the NOTRUMP (Reference

9) computer code for use in a new small break LOCA ECCS Evaluation Model (Reference 10). The NRC approved the use of NOTRUMP for small break LOCA ECCS analyses in May 1985. Since approval of the NOTRUMP small break LOCA ECCS Evaluation Model in 1985, the WFLASH computer code has not been maintained as. part of the Westinghouse ECCS Evaluation Model computer codes.

In section II.K.3.31 of NUREG-0737, the NRC required that each licensee submit a new small break LOCA analysis using an NRC approved small break LOCA Evaluation Model. which satisfied the requirements of NUREG-0737 section II.K.3.30. NRC Generic Letter 83-35 (Reference 8) relaxed the requirements of Item II.K.3.31 by allowing a more generic response and providing a basis for retention of the existing small, break LOCA analyses. Provided that the previously existing model results were demonstrated to be conservative with respect to the new small break LOCA model approved under the requirements of NUREG-0737 II.K.3.30 (NOTRUMP),

plant specific analyses using the new small break LOCA Evaluation Model would not be required. In WCAP-11145 (Reference 11),

Westinghouse and the Westinghouse Owners Group demonstrated that the result's obtained from calculations with WFLASH were conservative relative to those obtained. with NOTRUMP. Compliance with Item II.K.3.31 of NUREG-0737 could be completed by referencing WCAP-11145 and supplying some plant specific information.

Westinghouse, therefore, has not been modifying, investigating, or evaluating proposed changes to the WFLASH .small break LOCA ECCS Evaluation Model.

As discussed above, none of the modifications to the Westinghouse small break LOCA ECCS Evaluation Model would affect the small break LOCA analysis results by altering the PCT.

A. Analysis Calculated Result 1092'F B. Modifications to Westinghouse ECCS Evaluation Model + 0oF C. ECCS Evaluation Model Modifications Resultant PCT 1092'F CONCLUSION An evaluation of the effect of modifications to the Westinghouse ECCS Evaluation Model as reported in References 2 and 3 was performed for both the large break LOCA and small break LOCA analyses results found in WCAP-11609 and Chapter 15.6.4 of the R.E.

Ginna Nuclear Power Plant Updated Final Safety Analysis Report.

The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP) level of 12-o. The large break LOCA analysis was subsequently reanalyzed and licensed for an increase to 15: tube plugging. The small break LOCA analysis was evaluated and was not impacted by the increase

in SGTP level to . The results of the ana es and evaluations for the increase in. SGTP level to 15-o were documented in WCAP-11609, "Steam Generator Tube Plugging (SGTP) Report for Ginna.

Nuclear Power Station".

When the effects of the ECCS model changes were combined with the current plant analysis results, it was determined that compliance with the requirements of 10CFR50.46 would be maintained.

REFERENCES "Emergency Core Cooling Systems; Revisions to Acceptance Criteria", Federal Register, Vol. 53, No. 180, pp. 35996-36005, dated September, 16, 1988.

2. NS-NRC-89-3463, "10CFR50.46 Annual Notification for 1989 of Modifications in the Westinghouse ECCS Evaluation Models,"

Letter from W.J. Johnson (Westinghouse) to T.E. Murley (NRC),

dated. October 5, 1989.

3. NS-NRC-89-3464, "Correction of Errors.and Modifications to the NOTRUMP Code in the Westinghouse Small Break LOCA ECCS Evaluation Model Which Are Potentially Significant", Letter from W.J. Johnson (Westinghouse) to T.E. Murley (NRC), dated October 5, 1989.

4. WCAP-9220-P-A, Revision 1 (Proprietary), WCAP-9221-A, Revision 1 (Non-Proprietary), "Westinghouse ECCS Evaluation Model 1981 Version", 1981, Eicheldinger, C.

5. WCAP-10924-P-A (Proprietary), WCAP-12130-A (Non-Proprietary),

"Westinghouse Large Break LOCA Best Estimate Methodology",

Hochreiter, L.E., et.al., January 1987.

6. Report on Small Break Accidents for Westinghouse Nuclear Steam Supply System", WCAP-9601 (Non-Proprietary), June 1979, WCAP-9600 (Proprietary), June 1979.

7. "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants", NUREG-0611, January 1980.

8. "Clarification of TMI Action Plan Requirements", NUREG-0737, November 1980.

9. "Clarification of TMI Plan Item II.K.3.31, "NRC Generic Letter 83-85 from D.G. Eisenhut, November 2, 1983.

10. "NOTRUMP - A Nodal Transient Small Break and General Network Code", WCAP-10079-P-A (Proprietary), WCAP-10081-A (Non-

'Proprietary), Lee, N., et. al., August 1985.

"Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary), Lee, N., et. al., August 1985.

12. WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary),

"Westinghouse Emergency Core Cooling System Small Break October 1975 Model", April 1977.