ML20072N789

Suppl to 781230 Application for Amend to Licenses NPF-4 & NPF-7 Allowing Operation at RCS Average Temp of 587.8 F. Info Reflects Results of Containment LOCA Reanalysis Performed by S&W

ML20072N789
Person / Time
Site:	North Anna
Issue date:	07/11/1983
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Clark R, Harold Denton Office of Nuclear Reactor Regulation
References
726B, NUDOCS 8307180097
Download: ML20072N789 (16)

Text

{{#Wiki_filter:s VINGINIA ELucTnIC AND Pownn COMPANY HicnwoNn,VIROINIA 20261 W.L. STEWART vice p......., July 11, 1983 NET &.Aw Op. 4 Trow. Mr. Harold R. Denton, Director Serial No.: 726B Office of-Nuclear Reactor Regulation HBR:cdk:0554C Attn: Mr. Robert A. Clark, Chief Docket Nos.: 50-338 Operating Reactors Branch No. 3 50-339 Division of Licensing License Nos.: NPF-4 U. S. Nuclear Regulatory Commission NPF-7 Hashington, D.C. 20555 Gentlemen: SUPPLEMENT TO AMENDMENT TO OPERATING LICENSES NPF-4 AND NPF-7 NORTH ANNA POWER STATION UNIT NOS. 1 AND 2 REACTOR COOLANT AVERAGE TEMPERATURE OF 587.8"F In our letter dated December 30, 1982 (Serial No. 726) Vepco requested an amendment to operating Licenses NPF-4 and NPF-7 to allow operation of the North Anna Unit Nos. I and 2 at a reactor coolant system average temperature of 587.8aF. Our letter dated April 25, 1983 (Serial No. 726A) provided supplemental information in the form of responses to NRC questions on the containment LOCA analyses performed in support of the uprating. The purpose of this letter is to provide revisions to the information supplied in the above references, to reflect the results of a recent containment LOCA reanalysis performed by Stone & Hebster Engineering Corporation (SWEC) for operation at a Tavg of 587.8"F. provides revisions to the Balance of Plant Licensing Summary forwarded to the NRC as Enclosure 2 to our December 30, 1982 letter. Attachment 2 provides revisions to the responses to Containment System Branch questions, forwarded to the NRC in our April 25, 1983 letter. Attachment 2 also discusses the changes in input parameters for the revised containment LOCA analysis by SHEC. Should you have any further questions, please contact us at your earliest convenience. Very truly yours, l. l H. L. Stewart Attachments i (1) North Anna Units 1 and 2, 587.8"F Reactor Coolant System, Stone & Webster / BOP Safety Evaluation Summary (Revision 1) hp [W i 9 F307180097 830711 L PDR ADOCK 05000338 p. PDR

d 4 VsmotutA EtzcTasc Awo Powra Cowpawir To Mr. Harold R. Denton Page 2 (2) Response to Containment Systems Branch For North Anna Operation at an RCS Average Temperature of 587.8"F (Revision 1) cc: Mr. James P. O'Reilly Mr. M. B. Shy:nlock Regional Administrator NRC Resident Inspector Region II North Anna Power Station Mr. Charlie Price Department of Health 109 Governor Street Richmond, Va. 23219 I +- >, g ,m - +. .f%s 3

J h l 7 ATTACHMENT 1 Revision 1 (pages 4 and 7) North Anna Units 1 & 2 587.8*F Reactor Coolant System Stone & Webster /B0P Safety Evaluation Sumary i r 1 i l i cdk/0554C/4

0-C. ACCIDD ANALYSIS AND EhTIRONMENTAL QUALIFICATIONS 1. Containment Loss of Coolant Accident (LOCA) An analysis has been performed to investigate the effect of the 7.5'T uprate on containment integrity and Net Positive Suction Head Available (NPSHA) for the Rccirculation Spray (RS) and Low Head Safety Injection (LHSI) pumps. The following are the results: 1. Containment Integrity Table I shows the effect on containment peak pressure, subatmospheric peak pressure, and depressurization time due to the uprate. The pressures increase slightly, but are still within the North Anna acceptance criteria and con-tainment design pressure rating. 2. NPSHA for RS and LESI Pumps Table 2 shows that there is no decrease in NPSHA for the inside RS pump, the outside RS pump, and the LHSI pump h7SHA in the recirculation mode. The NPSHA values are within the Rev NPSH requirements at design flow. 2. Containment Main Stear Line Break (MSLB) Analysis The basis of the steam line break analysis is the full guillotine main steam line break at the no-load (hot shutdown) condition. The no-load Reactor Coolant system and Steam Generator tempera-ture and pressure remain unchanged subsequent to the uprate. Although there are additional energy releases at power operating conditions due to increased (uprated) Steam Generator pressure and temperature, the no-load condition remains the limiting case. Therefore, the Main Steam Line Break post-accident conditions remain as previously analyzed. O M 4 , -, + w ,~, --e w%

Table 1 LOCA Containment Integrity Analysis Results 7.5*F UFSAR Uprate Required Containment Peak Pressure (psig) 40.3 40.9 <45 Rev Containment Depressurization Time (sec) 3310* 3340 < 3600 Containment Subatmospheric Peak (psig) .11 .08 (0 Table 2 LOCA NPSH Analysis Results 7.5'F Required UFSAR Uprate NPSH** Inside Recire. Spray Pumps (ft) 11.5 11.6 9.4 Outside Recire. Spray Pumps (ft) 16.6 17.0 11.0 Rev Low Head Safety Injection Pumps (ft) 15.5 15.9 13.4

• UFSAR value in Table 6.2-53.
  • At design flow.

Rev 7

E i ATTACHMENT 2 Revised Response (Revision 1) ] To Containment Systems Branch For North Anna Operation at an RCS Average Temperature of 587.8*F 4 d l ~ 4 6 r 1 1 cdk/0554C/5 e -- + O y g

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QUESTION 1: Provide containment LOCA reanalysis mass and energy release table for the pump suction DER corresponding to operation with a TAVG of 587.8"F.

RESPONSE

Tables 1-1 thru l-3, provide the revised mass and energy release data for the following limiting containment LOCA analysis cases presented in the Balance of Plant-Safety Evaluation for the 7.5"F uprate (Reference 2, Enclosure 1). Containment 7.5*F Uprate Mass /Enerav Calculation Value Data Peak Pressure 40.9 Psig Table 1-1 Subatmospheric Peak .08 Psig Table 1-2 Depressurization Time 3340 Seconds Table 1-3 It should be noted that the attached mass and energy release data was generated by LOCTIC to maximize peak containment pressures and depressurization times. This data has no direct relation to the mass and energy release data presented in Table 5 of the NSSS Safety Evaluation (Reference 1, Enclosure 1). The data in that table was generated as part of Westinghouse's LOCA-ECCS analysis and was calculated using assumptions which maximize clad temperature. QUESTION 2: Address the differences in analysis and input between the LOCA containment reanalysis as specified in Question 1 and the previous analysis in the FSAR.

RESPONSE

The containment LOCA analysis results presented in the UFSAR and in both l Revision 0 (Reference 1, Enclosure 2) and Revision 1 (Reference 2, Enclosure (

1) of the Balance of Plant (B0P) Safety Evaluation for the 7.5'F uprating, l

were generated using Stone & Webster's LOCTIC Code. However, for the 7.5 F uprating calculations, the current version of the LOCTIC Code was used. Calculations performed by Stone & Webster have shown that the different versions have negligible effect on the containment analysis. Although there were no differences in the containment model parameters (such as containment free volume, temperature and heat sinks, etc.) used in the UFSAR and 7.5*F (Revisions 0 and 1) uprating analyses, there were - differences in several parameters used in the generation of LOCA mass and energy and release rates. These differences are discussed below. l l l I' cdk/0554C/6

I UPRATED-RELATFD PARAMETER DIFFERENCES A. Steam Generator (SG) se:condary fluid temperature For the UFSAR analyses, an initial steam generator secondary fluid temperature of 525.2*F was used. For the uprating analyses (both Rev. O and 1), an uprated value of 535.8"F was used. II NON-UPRATE RELATED PARAMETERS DIFFERENCES The following nonuprate-related parameter differences included in the 7.5 F uprate analyses have been found to have a minimal impact on the containment LOCA analysis results. A. SG Secondary Liquid Volume For the UFSAR analyses, an initial SG secondary liquid mass of 89,600 lbm per SG was used. Stone & Webster studied the impact of applying a Westinghouse-recommended 210% margin to the SG Secondary 11guld mass. It was found that reducing the mass 10% gave the most conservative results. Therefore, a mass of 80640 lbm was used in the 7.5'F uprate analyses (Rev. O and 1). B. Accumulator Flow Rate For the UFSAR analyses, accumulator flow rate was input as a function of time. This function had been based on a bounding LOCA analysis performed by Westinghouse. For the 7.5 F uprate analyses Thisavoidedamassimbalancepresentwhenusing[ Gev. O and 1), LOCTIC g2nerated accumulator flow rates based on system conditions. the forcing function flow input. C. Reactor Vessel Volume Below Top of Core For the UFSAR calculations, a value of 2448 ft3 was used for the volume of water below the top of the core. For the 7.5'F uprate l analyses (Rev. O and 1), a more correct value of 1947 ft3 was used. I D. Core Heat Transfer Coefficient During Reflood For the existing UFSAR calculations, a core reflood heat transfer l coefficient of 1000 BTU /hr-ft2 *F was used in the analysis runs to calculate the 1st and 2nd containment peaks, and a RESAR 3 Curve (Reference attached Figure 2-1) was used in the analysis runs to calculate Depressurization Time, the 3rd containment peak and LHSI NPSH. For the Rev. O uprate analyses the RESAR 3 curve was used for all cases. For the Rev. 1 uprate analyses, a conservative value of 1000 BTU /hr-ft2-F was used for all cases. E. Outside Recirculation Spray (ORS) Flow Rate. For the UFSAR analyses and the Rev. O uprate analyses, an Outside Recirculation Spray (ORS) flow of 3700 gpm was used in the analysis runs to calculate 3rd peak, depressurization time and the NPSHA for L cdk/0554C/7

the Inside Recirculation Spray (IRS) pumps the ORS pumps and Low Head Safety Injection (LHSI) pumps. For the Rev. ) aprate analyses, the system flow rate of 3640 gpm based on the as-built system was used. The effect on the results was insignificant. F. ORS Cooler Overall Heat Transfer Coefficient The overall heat transfer coefficient changes as a result of the change in ORS pump flow rate, resulting in a slightly different value for the quantity UA. For the UFSAR and Rev. O uprate analyses a UA value of 3.67 x 10' BTU /hr *F per cooler was used with the ORS flow of 3700 gpm. For the Rev. 1 uprate analyses a UA value of 3.65 x 10' BTU /hr *F per cooler was used with the revised ORS flow of 3640 gpm. G. Initial Core Reflood Velocity The reflood velocity which determines the end of reflood (EOR) time, is taken from Westinghouse supplied data for the cold leg break cases. Westinghouse has confl.'med that existing reflood data for use in containment analyses remains bounding for the uprated conditions, thus the E0R time is unchanged. However, for Rev. O uprate Analyses, the E0R time did change since a LOCTIC calculated accumulator flow rate had been used in lieu of inputting accumulator flow rates (as a function of time) which had been based on the bounding LOCA analysis performed by Westinghouse. For the Rev. I uprate analyses, the E0R time was made to be consistent with the UFSAR EOR time, by appropriately adjusting the initial reflood velocity. The impact of this change on containment pressure is negligible. The change was made to ensure consistency between the E0R times for the UFSAR and uprate (Rev. 1) analyses. H. Peak Condensing Heat Transfer Coefficients For Heat Sinks and Times of Their Occurrence For the UFSAR and Rev. O uprate analyses, a peak heat transfer coefficient of 201 BTU /hr *F-ft2 (at 25 seconds) was used for the calculation of 3rd peak, depressurization time and LHSI NPSH. For the Rev. I uprating analyses, a more appropriate value of 258 BTU /hr *F-ft2 (at 16.1 seconds) was used. This change has an insignificant impact on the results since the long term effects of these cases are not sensitive to the peak transfer coefficient which occurs early in the accident. I. Quench Spray (QS) and Recirculation Spray (RS) Efficiency For the calculation of 2nd peak, the UFSAR and Rev. I uprate analyses used a spray efficiency of.9, while the Rev. O uprate analysis had incorrectly used a nonconservative value of 1.0. l J. NPSHR for the Inside (IRS) and Outside (ORS) Recirculation Spray Pumps and the LHSI Pumps i cdk/0554C/8

The TABLE below shows the NPSHR used in the calculation of the NPSHA for the IRS, ORS and LHSI pumps. The Rev. I uprate analyses reflected the correct NPSHR values This had no effect on the NPSH analyses since adequate margin exists between the NPSHA and NPSHR. Case UFSAR Uprate Rev. O Uprate Rev. 1 IRS NPSHR (ft) 9.4 8.8 9.4 ORS NPSHR (ft) 11.0 9.7 11.0 LHSI NPSHR (ft) 12.8 13.4 13.4 The combined effect of the above changes produces small differences from the UFSAR results, as confirmed by the data in response to Question 1.

References:

1. Letter from H. L. Stewart (Vepco) to Harold R. Denton (NRC), dated December 30, 1982, Serial No. 726, " Amendment to Operating Licenses NPF-4 and NPF-7, North Anna Power Station Unit Nos. 1 and 2, Proposed Technical Specification Change". 2. Letter from H. L. Stewart (Vepto) to Harold R. Denton (NRC), Serial No. 726B, " Supplement to Amendment to Operating Licenses NPF-4 and NPF-7, North Anna Nuclear Power Station Unit Nos. I and 2, Reactor Coolant Average Temperature of 587.8"F". cdk/0554C/9 ._J

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