ML20199C924

Forwards Addl Info as Requested by NRC in 981218 RAI & Addl Info Re Original Submittal of Proposed TS Number NPF-10/15-491.TS Involved RCS Temp Reduction & Volumetric Min Flow Rate

ML20199C924
Person / Time
Site:	San Onofre
Issue date:	01/13/1999
From:	Scherer A SOUTHERN CALIFORNIA EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-MA2238, TAC-MA2239, NUDOCS 9901190196
Download: ML20199C924 (12)

Text

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y soum,_,om EDISON m.,o sa_

ML,,m An UJISON /V7LRNA710ML* Company January 13, 1999 4

U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Gentlemen:

Subject:

Docket Nos. 50-361 and 50-362 l Response to the NRC Request for Additional Information to Support Proposed Technical Specification Number NPF-10/15-491, Reactor Coolant System (RCS) Temperature Reduction and Volumetric Minimum Flow Rate (TAC Nos. MA2238 and MA2239)

San Onofre Nuclear Generating Station Units 2 and 3

References:

1) Letter from James W. Clifford (NRC) to Harold B. Ray (SCE),

dated December 18, 1998,

Subject:

Request for Additional InformationonChangetoTca[ReactorCoolantSystem(RCS) cold leg temperature (T,,u)] Reduction and RCS Flow Measurement Technical Specification (TAC Nos. MA2238 and MA2239) San Onofre Nuclear Generating Station Units 2 and 3 '

2) Letter from Dwight E. Nunn (SCE) to the Document Control Desk (NRC),datedJune 19, 1998,

Subject:

Proposed Technical SpecificationNumberNPF-10/15-491, RCS Temperature Reduction and Volumetric Minimum Flow Rate, San Onofre Nuclear Generating Station, Units 2 and 3 This letter provides additional information as requested by the U. S. NRC in reference 1 concerning the proposed reduction in Reactor Coolant System (RCS) cold-leg temperature (T,,u) and the change in RCS flow measurement at San Onofre Units 2 and 3. The Southern California Edison Company (SCE) responses to the U. S. NRC's questions are provided as Enclosure 1 to this letter.

Also, SCE has reviewed the original submittal, reference 2, and determined

! that additional information was needed in Table H.1. Therefore, additional i information to support the conclusions provided on page H-12 of Table H.1 is provided in Enclosure 2.

9901190196 990113 ~~

PDR ADOCK 05000361 P

PDR. ,

(- P. O. Ikw 128 San Clemente, CA 92674-0128 QD f 7t

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i 949-368-7501 fax 949-368-7575 I

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l l Document Control Desk BACKGROUND By reference 2 SCE submitted Amendment Appiication Numbers 179 to Operating I License NPF-10 and 165 to Operating License NPF-15. These amendment applications requested the following changes:

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1) A reduction in the minimum primary Reactor Coolant System (RCS) cold leg temperature (T,,u) from 544 F to 535"F between the 70%

.and 100% rated thermal power levels.

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2) A conversion of the specified RCS minimum flow rate from a " Mass" l (i.e., Ibm /hr) to a " Volumetric" (gpm) flow basis.

3). Elimination of the maximum RCS flow rate limit.

If you have any questions or would like additional information on this subject, please feel free to contact me or Jack Rainsberry at (949) 368-7420.

l Sincerely, l

Enclosures I

cc: E. W. Merschoff, Regional Administrator, NRC Region IV l J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 J. W. Clifford, NRC Project Manager, San Onofre Units 2 and 3 l

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t ENCLOSURE 1 l-l  ;

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, The Southern California Edison Company (SCE) ,

I T,.u Program i Response to NRC Questions j i

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Enclosure 1 Page 1 of 7 l

Question 1:

Provide a comparison of key parameters (e.g., reactor coolant system (RCS) pressure, T.,,,,,,, Tc .io, T,,,, steam pressure, and steam generator (SG) outlet temperature) between the proposed operation with the T,,i, reduction in your l

submittal and the design basis analysis.

Response to Question 1:

l Parameter' Original Proposed Delta Design Operation-Pressurizer 2250 psia 2250 psia 0 psia ,

Pressure RCS T-ave 582*F 568*F - 14"F RCS T-hot 611*F 596*F - 15"F RCS T-cold 553*F 540*F - 13*F Steam 900 psia 800 psia - 100 psia Generator )

Pressure Steam 532*F 518*F - 14*F l Generator Outlet Temperature Note: All values are based on full power operation.

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Enclosure 1 Page 2 of 7 Question 2:

Provide a summary of the evaluations (including analytical methodology, assumptions, and maximum stress and fatigue usage factors) for the effects of T,,w reduction on the structural and pressure boundary integrity of the reactor vessel and internals, RCS piping, control rod drive mechanisms and

. housing, pressurizer, surge line (stratification), pressurizer spray nozzles, SGs, reactor coolant pumps, and pressurizer power-operated valves and safety valves. Identify changes in maximum stress and fatigue usage factors (at critical locations) from your evaluation.

Response to Question 2:

SUMMARY

OF STRUCTURAL EVALUATIONS FOR THE EFFEfTS OF T-COLD REDUCTION l

OBJECTIVE The objective of the structural evaluations was to determine the impact of T-cold reduction on the structural and pressure boundary integrity of the SONGS Units 2 and 3 Reactor Coolant System (RCS) components and supports. It should be noted that these evaluations have no bearing on the Bases for LC0 3.4.1, but have been performed to update the SONGS design basis to be consistent with intended future operation.

BACKGROUND ,

Southern California Edison (SCE) has completed engineering evaluations which justify a reduction in T-cold of 13 'F from the original SONGS Units 2 and 3 design value of 553 'F at full power operating conditions. This change in T-cold and the corresponding changes in other plant operating conditions modify the original design basis assumptions for the Normal Operating (N0P) conditions of the components. This change in N0P conditions was examined for its effect on the original calculations of expected stress and fatigue usage factors.

The decrease in T-cold can also affect component loading under accident conditior.s. The lower temperature increases the fluid density and thus increases the loading on the components during postulated pipe breaks. The er@inal plant design basis considered various size breaks at several i locations in the Main Coolant Loop (MCL) piping. Since the evaluation of the original design basis, additional analyses were performed to demonstrate that

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Enclosure 1 l Page 3 of 7 l

l a rupture of the RCS main loop piping would be preceded by detectable leaks, L

rather than resulting in sudden catastrophic failure. This methodology is l defined as Leak-Before-Break (LBB), and SCE has been authorized by the NRC to

! implement the LBB methodology at San Onofre Units 2 ard 1. (

Reference:

Letter from M. B. Fields (NRC) to H. B. Ray (SCE) dated April 11, 1996;

Subject:

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" Application of Leak-Before-Break Technology to Reactor Coolant System piping j at San Onofre Nuclear Generating Station, Units 2 and 3 (TAC Nos. M92949 and  ;

i M92950).") I Lower fluid temperatures tend to increase the calculated loading on the components associated with pipe break accident conditions (LOCA). The LBB methodology permits replacement of the severe MCL breaks with breaks in the branch lines, the largest of which are the shutdown cooling, safety injection, and surge lines. The controlling secondary side breaks (feedwater and main steam' lines) remain unchanged. Therefore, the structural evaluations performed for the T-cold Reduction Project take into account the combination of decreased loads due to LBB methodology and the effects of implementing T-cold reduction. For the faulted loads, use of LBB offsets the detrimental calculated effects of reduced T-cold.

ANALYTICAL METHODOLOGY The following lists the approach taken to evaluate the structural effects of reduced operating temperatures:

For the normal operating condition:

• Evaluate the effect on defined plant transients

• Reviewand/orquantifytheeffectontheaffectedRCScomponents

• Review and/or quantify the effect on reactor vessel internals e Review the effect on fuel assemblies for the faulted load condition:

• Examine the differences in hydraulic loading due to the lower

, temperature of both large MCL and branch line pipe (BLP) breaks.

• Evaluate the effect on overall primary system response with parametric

hydraulic loading conditions -

o RCS motions e RCS support loads

Enclosure 1 j Page 4 of 7 i e Piping loads e Reactor vessel internals loads i-

• - Quantify and evaluate the effect on all RCS components

! ~ SUMARY'0F EVALUATIONS

1. Norwal Oneratina Loads Evaluations were performed'to demonstrate the impact of T-cold on the plant i transients defined for the RCS components including: l l-

• Reactor Vessel

• Reactor Vessel Internals

• Reactor' Coolant Piping and Attached Nozzles I e Pressurizer

• Pressurizer Spray Nozzles l -e Surge Line (stratification) e Steam Generators  !

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• Reactor Coolant Pumps e Fuel Assemblies The. evaluations reviewed the design basis specifications and analyses for j these components and assessed and documented the changes which may occur l- because of T-cold reduction. 1t was demonstrated that T-cold reduction does l not have an impact on the component thermal transients or resulting thermal l stresses and fatigue for the primary side components. The evaluations also l included the effect of T-cold reduction on the primary side attachment noz.zles (shutdown cooling, safety injection, surge, drain, spray, charging, and i letdown lines).

! Because the pressurizer conditions will not change with reduced T-cold, there L is no effect on the structural integrity of the pressurizer safety valves.

! SONGS Units 2 and 3 are not fitted with pressurizer power-operated relief valves.

l A few steam generator components required further evaluation to justify that L the effect on the fatigue usage factors were acceptable. The results of these I

evaluations are described in section (3) below. Secondary side attachment j

nozzles were evaluated as part of the steam generator component evaluations.

-2 Faulted Load Evaluations l The implementation of the LBB methodology changed the loss-of-coolant accident

.(LOCA) load portions of the faulted load calculations. It was found that the l

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LOCA loads due to a Branch Line Pipe (BLP) break are significantly lower than those associated with a MCL break. The actual effects of the revised loads were examined in the evaluations as described by the following.

ihe impact of BLP breaks in conjunction with T-cold reduction was compared to the motions and loads from the existing design basis of large MCL breaks.

l This comparison included the effects on RCS motions, component support loads, I

and the corresponding loads for reactor vessel internals including fuel. The

, results show that the RCS response to large break LOCA bound those due to BLP i breaks with T-cold reduction implemented.  !

The effects of revised loads were reconciled for all components and documented j in appropriate stress calculations and design reports. It was determined that  ;

l the faulted stress levels for all components meet the applicable ASME Code l criteria. The affected components are:

e Reactor Vessel l e Reactor Vessel Internals e Reactor Coolant Piping and Attached Nozzles e Control Element Drive Mechanisms and Housings l

• Reactor Head Lift Rig I e Pressurizer

e Surge Line

• Steam Generators e Reactor Coolant Pumps e Fuel Assemblies .

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3. Imooct of T-cold Redubtton on Steam Generators The steam generator components were evaluated for a primary cold leg temperature of 540 'F, hot leg temperature of 596 'F, and secondary side i pressure of 794 psia. A review of existing calculations identified that the l following areas required additional analytical evaluations for determining the l effect of T-cold reduction:

ITEM COMPONENT LOADING CONCERN 1 Tubes Increase Primary to Secondary AP RG 1.121 Tube Degradation l

2 Tubes Fluid hydraulics due to higher Flow Induced l

density fluid Vibration 3 Tubesheet and Increase Primary to Secondary by Fatigue Adjacent Structure l l

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l Enclosure 1 l Page 6 of 7 l '4 Secondary Shell Increased stress amplitude due Fatigue to reduced secondary pressure 5 Feedwater Nozzle Increased stress amplitude due Fatigue to reduced secondary pressure )

l The analysis considered the decrease in secondary side pressure, the fluid l flow changes on tube vibrations, and the changes in certain operating transients that have an effect on the components' fatigue evaluations.

The lower operating secondary pressures and associated peak stresses for some i transient conditions caused the stress amplitude and the total calculated I fatigue usage factors to increase in steam generators. This is the result of i a conservative pairing of maximum and minimum peak stresses due to transients l considered in the original design. The evaluations performed and summarized in the following table demonstrated that the effect of T-cold reduction on the secondary side components is acceptable.

COMPONENT / TYPE OF EVALUATION CONCLUSION

= Tubes / Increased AP caused by lower

• RG 1.121 acceptable degradation secondary pressure used in the RG is 61.2% using min. code tensile ]

1.121 evaluation strength = 80 ksi.  !

= Tubes / Degradation

• 64.4% allowable degradation calculated with min. actual tensile strength from mill test reports = 87 ksi. This compares to original 64% allowable.

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Enclosure 1 l Page 7 of 7 l COMPONENT / TYPE OF EVALUATION CONCLUSION l

Tubes / Fluid Hydraulics No significant effect on tube  ;

l vibration stability.

• Limiting vertical section:

stability ratio = 0.75 < 1.0; increased from 0.67.

• Limiting U-bend section: i stability ratio = 0.71 < 1.0; decreased from 0.74.

Tubesheet and Adjacent Structures / U = fatigue usage factor:  ;

Fatigue

. Tubesheet top surface . U = 0.703 < 1.0; increased from 0.44.

. Secondary shell at top of tubesheet a U = 0.606 < 1.0; increased from interface 0.061.

Bottom head at bottom of tubesheet . U = 0.262 < 1.0; decreased from interface 0.274.

SecondaryShell/ Fatigue a U = 0.785 < 1.0 with pressure stresses; increased from 0.71.

• U = 0.851 < 1.0 with pressure and thermal stresses; increased from 0.71. i Feedwater Nozzle / Fatigue a U = 0.881 < 1.0 with pressure l stresses; increased from 0.652. l

• U = 0.965 < 1.0 with pressure l and thermal stresses; increased  !

from 0.652.

CONCLUSIONS The maximum stress and fatigue usage factors at critical locations of all primary system components and supports (after implementing T-cold reduction and eliminating MCL breaks from consideration) were found to be conservatively bounded. by the values given in the original design except for steam generators. After implementing T-cold reduction, some transients defined in the original fatigue evaluations for the steam generators have larger stress

amplitudes which result in higher calculated fatigue usage factors. In all l cases, stresses at critical locations meet ASME code criteria.

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ENCLOSURE 2 San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 T,,u Program Revised Paragraph Page H-14 of Table H.1

Enclosure 2 Page 1 of 1 Summar.y During a review of our original submittal, the Letter from Dwight E. Nunn (SCE) to the Document Control Desk (NRC), dated June 19, 1998,

Subject:

Proposed Technical Specification Number NPF-10/15-491, RCS Temperature Reduction and Volumetric Minimum Flow Rate, San Onofre Nuclear Generating Station, Units 2 and 3, a need was identified to provide additional information to support Table H.1.

Specifically, in Table H.1 for the Increase in Reactor Coolant Inventory events (UFSAR Section 15.5), the impact of the Tcold and flow Technical Specification change did not address the identified criteria of "No liquid flow though the [ primary safety valves] PSVs for the peak pressure case." For this criteria, the Inadvertent Operation of the ECCS during Power Operation with and without single failure continue to remain bounded by the Chemical and Volume Control System (CVCS) Malfunction with and without single failure, respectively. Therefore, the discussion for Sections 15.5.1.2 and 15.5.2.2 in Table H-1 continue to remain valid for the peak pressure and the no PSV liquid flow criteria. For the CVCS malfunction events, Section H-2 should be revised as follows:

Revision 15.5.1.1 Chemical and Volume Control System Malfunction 15.5.2.1 Chemical and Volume Control System Malfunction with Single Failure Evaluation of Reduction in Tg for Peak RCS Pressure and Liauid Release Throuah the PSVs For the peak reactor coolant system (RCS) pressure criterion, a sensitivity analysis had been performed to determine the impact of Tcold on the event. It was determined that the impact of reducing the Tcold from 560*F to 542*F resulted in a peak RCS pressure increase from 2592 psia to 2600 psia. Based on this sensitivity analysis and linear extrapolation, peak pressure would increase by less than 10 psia (adding a factor of 2 for conservatism) as a result of lowering Tcold to 533*F. The limiting analysis (i.e., CVCS Malfunction + Single Failure (SF)) was evaluated at 542*F and resulted in a peak RCS pressure of 2629 psia. Hence, the acceptance criterion for peak RCS pressure is not challei.ged as a result of reducing the minimum Technical Specification Tcold'from 544*F to 535'F. For the limiting peak pressure case for both the CVCS malfunction and the CVCS malfunction with single failure events, the analysis showed that no liquid was released through the PSVs as a-result of this reduction in Ta.