ML20094J378

Forwards Responses to 840611 Request for Addl Info Re SER Confirmatory Issues 3,5 & 6 & SER Open Issue 29.Related Draft FSAR & Design Assessment Rept Page Changes Also Encl

ML20094J378
Person / Time
Site:	Limerick
Issue date:	08/08/1984
From:	Kemper J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Schwencer A Office of Nuclear Reactor Regulation
Shared Package
ML20094J383	List: ML20094J378 ML20094J420 ML20094J449 ML20094J464
References
OL, NUDOCS 8408140303
Download: ML20094J378 (27)

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l cc: Judge Lawrence Brenner (w/ enclosure-1)

-Judge Richard F.LCole (w/ enclosure 1)-

. Troy B. Conner,,Jr., Esq. (w/ enclosure 1)

Ann P. Hodgdon,'Esq. (w/ enclosure 1)'

Mr. Frank R. Romano' (w/ enclosure 1)

Mr. Robert L. Anthony- . (w/ enclosure 1)

Charles.W. Elliot, Esq.- (w/ enclosure 1)

Zori G. Fe'rkin, Esq. (w/ enclosure 1)

Mr. Thomas Gerusky (w/ enclosure 1) i Director,-Penna. Emergency  : (w/ enclosure 1)-

l Management Agency I Angus R. Love, Esq. (w/ enclosure-1)

David Wersan, Esq. (w/ enclosure 1)

- Robert J. Sugartmn, Esq. (w/ enclosure 1)-

Spence W. Perry,-Esq. (w/ enclosure 1)

Jay M. Gutferrez, Esq. . (w/ enclosure 1) i Atomic Safety & Licensing - (w/ enclosure 1) -

.Appea1 Board Atomic Safety & Licensing (w/ enclosure 1)

Board Panel' Docket & Service Section. (w/ enclosure s )

. Martha W. Bush,Esq. (w/ enclosure 1) -!

Mr. James Wiggins (w/ enclosure 1) '

Mr. Timothy R. S. Cartpbell (w/ enclosure 1).

Ms. Phy1IIs Zitzer . (w/ enc 1osure 1) '

Judge Peter A.. Morris (w/ enclosure 1)

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Responses to >RC EB Questions 1, 2, 3, and 4, from NRC letter dated June'11, 1984, including 8 Related Draft FSAR and DAR Page Changes.

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1. SER Confirmatory Issue # 3 - Startup Test Specification for 80P Piping The response to MEB SER Question 210.58 concerning the startup test specification for B0P. piping as contained in FSAR, Revision 28 stated that interim test specifications governing the scope of startup testing of B0P piping have been prepared and will be made available to the NRC for review when requested. Provide the staff a copy of the interim test specifications.

Response

The following test specifications are provided as Attachment 2:

a) Specification 8031-P-362, Specification for Test Requirements for Hot Deflection Testing of ASME Section III Nuclear Class 1, 2, 3 l and ANSI B31.1 Bechtel Piping for the Limerick Generating Station, Units 1 and 2.

b) Specification 8031-P-363 Specification for Requirements for Steady State Vibration Testing of ASME Section III Nuclear Class 1, 2, 3 and ANSI B31.1 Bechtel Piping for the Limerick Generating Station, Units 1 and 2.

c) Specification 8031-P-364, Specification for Test Requirements for Dynamic Transient Testing of ASME Section III Nuclear Clasc 1, 2, 3 and ANSI B31.1 Bechtel Piping for the Limerick Generating Station, Units 1 and 2.

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2. SER Confinnatory Issue # 5 - Suppression Pool Hydrodynamic Load Reconciliation The response to MEB Question 210.69, suppression pool hydrodynamic load reconciliation, as contained in FSAR Revision 27 stated that Section 3.9.has been changed to provide the New Loads Adequacy Evaluation. Based on a review of the information provided in FSAR Section 3.9, Revision 27 and the Design Assessment (DAR) of Limerick, we have determined that the following areas are incomplete.

a. Provide additional infonnation to clearly define the scope of the suppression pool hydrodynamic load reconciliation program for Limerick. Specifically, clarify the statement in Section 7.2.1.10 of DAR, Revision 5 that "as described in Section 7.1.5, all seismic Category I BOP piping systems located inside the containment, reacter enclosure and control structure are analyzed for seismic and hydrodynamic loads " and.the statement in Section 7.2.1.11 of DAR, Revision 8 that, "all seismic Category I B0P equipment i,s re-assessed for hydrodynamic and seismic loads (Section 7.1.7)." Sections 7.1.5 and 7.1.7 only address the design assessment methodology and do not clearly define the scope of the design assessment program as to whether all of the BOP piping components, equipment and their supports have been included in the design assessment.

With respect to NSSS, Section 7.2.1.12 of DAR, Revision 5 stated that NSSS piping and safety-related equipment have been assessed for hydrodynamic and seismic loads. It is not clear whether all of the NSSS piping components, equipment and their supports have been included in the design assessment. It is the staff's position that all safety-related B0P and NSSS piping components, equipment and their supports affected by the hydrodynamic load, both inside and outside containment have to be re-assessed in the hydrodynamic load reconciliation program. Provide a corrmi tment to comply with this position. Indicate the methods employed for the design re-assessment program such as actual reanalysis or spectra comparison.

b. Provide additional information to clearly identify the status and the results of the design re-assessment for suppression pool hydrodynamic loads. Specifically, identify whether changes in design such as additional supports, modification of existfng supports or any other plant modifications are required as a result of the suppression pool hydrodynamic load reconcilation and provide a cormitment and schedule of completion of design changes for til the affected -

safety-related piping components, equipment and their supports for both BOP and NSSS. Currently, FSAR Section 3.9, Revision  ;

27 and Sections 7.2.1.11 and 7.2.1.12 of DAR, Revision '8 do not '

contain this information and Section 7.2.1.10 of DAR Revision 5 does not address the status of implementation of d ,ign changes.

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Response

'a) The text in DAR Sections 7.2.1.10, 7.2.1.11, and 7.2.1.12 wi11 be revised to clarify the scope of the hydrodynamic ' load reconcilation program. Specifically, these sections will reflect

.that all safety-related piping, components, equipment, and their supports have been included in the design assessment. DAR Sections 7.1.5, 7.1.6 and 7.1.7 describe the design assessment methodology used in the BOP and NSSS evaluation.

i b) The assessment of all safety-related BOP and NSSS piping, components, equipment, and their supports has been completed.

DAR Sections 7.2.1.10, 7.2.1.11, and 7.2.1.12 and FSAR Section 3.9 will be revised to reflect that all structural modifications l necessitated by reconciliation of the suppression pool hydrodynamic l

loads have been completed. A swmary of those modifications is given below:

o Addition of the downcomer bracing system as discussed in DAR Section 7.1.2.2.

o Addition or modification of pipe supports as required to accm modate the hydrodynamic loads.

o Modification of safety-related equipment and associated supports where required to sustain the additional hydrodynamic loads, e.g., additional bracing was provided

for all the safety-related motor control centers.

l The above referenced FSAR and DAR draft page changes are attached.

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DRAET LGS DAR Pressure time histories in the wetwell are used to investigate the reactor enclosure and control structure response to SRV and LOCA loads. Maximum time history force responses and broadened I response spectra curves are approximately used to assess the adequacy of associated structural. components. The assessment methodology of the reactor enclosure and control structure is presented in Section 7.1.1.2.-

l The mode shapes, modal frequencies, and hydrodynamic response spectra of the reactor enclosure and control structure are presented ,in Appendix B.

The results of the structural assessment are summarized in Appendix E. -

2.2.2 CONTAINMENT SUBMERGED STRUCTURES ASSESSMENT

SUMMARY

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Load cpabinations:for the downcomer bracing and suppression chamber columns are presented in -Table 5.3-1. Load combinations for the downcomers are presented in Table 5.5-1. The hydrodynamic design assessment methodology for the downcomers, bracing, and columns is presented in Sections 7.1.2 and 7.1.4.

The results of the analysis.are presented in Appendix D. (

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~, __ The suppression pool liner plate loa Q re combined in accordance with Table 5.2-1. Results from the analysis. indicate that.no- .

structural modification is required (see Sections $.7 1,.L and' 7.2.1.5). -

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2.2.3 BOP PIPING SYSTEMS ~ ASSESSMENT

SUMMARY

g/c,rc Containment and reactor enclosure BOP piping systems ;r L _ ' . .; -

analyzed by the methods presented in Section 7.1.5. The load combinations for piping are described in Table 5.6-1. The results of the analysis are presented in Appendix F.

2.2.4 NSSS ASSESSMENT

SUMMARY

2.2.4.1 Introduction General Electric Company performed a design assessment of Limerick Unit 1 to demonstrate that the NSSS piping and safety- (

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related equipment have sufficient capability to accommodate combinations of seismic and hydrodynamic loadings. The scope of the evaluation included the reactor pressure vessel (P.PV), RPV internals and associated equipment, main steam and recirculation piping, and GE-supplied floor mounted equipment, pipe mounted equipment, and control and instrumen gtati equipment -

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$nd odl W065 deb $Qf&, i The methodologies described in Section 7.1.6 were used to perform the evaluation. Load combinations and acceptance criteria listed in Table 5-7.1 were used for the evaluation of ASME Class 1, 2 and 3 piping, equipment, and supports.

2.2.4.2 Desian Assessment Results The results of the assessment have demonstrated that the NSSS piping and safety-related equipment have sufficient capability to accommodate combinations of seismic and hydrodynamic loadings for the normal, upset, emergency and faulted conditions.

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Detailed results of the NSSS piping and major safety-related equipment evaluations are given in FSAR Sections 3.9 and 3.10. l

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2.2.5 BOP EQUIPMENT ASSESSMENT

SUMMARY

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Safety related BOP equipment in the contatnment, reactor ~

enclosure,' and control structure are assessed by the methods contained in Section 7.1.7. Loads are combined as shown in Table 5.8-1.

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2.2.6 ELECTRICAL RACEWAY SYSTEM ASSESSMENT

SUMMARY

The electrical raceway system located in the containment, reactor enclosure, and control structure is assessed for load

-combinations in accordance with Table 5.9-1. The assessment *~

methodology and analysis results are presented in Chapter 7.

2.2.7 HVAC DUCT SYSTEM ASSESSMENT

SUMMARY

' The HVAC duct system located in the containment, reactor enclosure, and control structure is assessed for load

'! combinations in accordance with Table 5.10-1. The assessment methodology and analysis results are presented in Chapter 7.

2.2-3 Rev. 8, 04/84

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r LGS DAR iA The refueling head and flange.were found to have no stresses

-exceeding the specified allowable limits. I The leaktightness of the flanged joint is investigated for the combined jet forces.

effect of temperature, pressure, seismic, SRV, LOCA and Vertical separation at the flange faces is prevented by providing sufficient bolt preload to offset uplift due to the applied loads. Similarly, relative horizontal movement between the flange faces is prevented by the bolt preload induced frictional forces. A preload of 157K per bolt is required to maintain leaktightness at the flange joints.

7.2.1.9.2 Suppression Chamber Access Hatch, CRD Removal Hatch, and Equipment Hatch For these components, CBI's analysis indicated that there are no stresses in excess of the specified allowable limits when considering the additional hydrodynamic loading.

7.2.1.9.3 Equipment Hatch-Personnel Airlock The equipment hatch with personnel airlock has been assessed for hydrodynamic and seismic loads. Modifications to some cap screws I

of the attachment brackets are required to accommodate the additional hydrodynamic loading. The equipment hatch with personnel airlock and all related components are within the specified allowable limits.

7.2.1.10 s 0n N ypeks BOP Pipino and MSRV Systems Marcins As described in Section 7.1.5, all Seismic Category I BOP piping 4

y;t;;; located inside the containment, reactor enclosure, and control oads.

structurgsens analyzed for seismic and hydrodynamic Table 5.6-1. Ine loads from the analyses are combined as described in supports -~- Additional supports and modification of existing

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hydrodynamic and setsmic loads for some piping systems,e-5 tresses and stress margins for selected BOP piping systems are summarized in Appendix F.

The stress reports for the evaluation of the BOP piping will be available for NRC review.

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All Seismic Category 1 BOP equipment #- -

____d for hydrodynamic and seismic loads (Section 7.1.7) via the Limerick Seismic Qualification Review Team (SORT) program. or each piece of BOP equipment, a five-page SORT summary form has een prepared documenting the re-evaluation of the equipment. =

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Gul NIE su.PPC uNSSSJpjgtg3 W ety-relafte quipmentp ave been assessed for nydrodynamic and sefsmic loads Detailed results o Af-evaluation are given in FSAR Sections 3.9 and 3.10.F In addition, i

General Electric Co. has prepared Seismic Qualification Reevaluation (SOR) Program forms,.NSSS Loads Adequacy Evaluation (NLAE) Program Summary reports, and design stress reports to document the assessment of seismic and hydrodynamic loads on NSSS piping one safety-related equipment . These forms and reports

will be available for NRC review. ~

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• W 7.2.2 ACCELERATION RESPONSE SPECTRA

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l 7.2.2.1 Containment Structure The method of analysis and load description for the acceleration response spectrum generation are outlined in l

Section 7.1.1.1.1.6.1. From a review of the acceleration response spectra curves for the containment structure, the maximum spectral accelerations are tabulated for 1 percent damping of critical. For SRV and LOCA loads, the maximum l

spectral accelerations are presented in Table 7.2-1.

The hydrodynamic acceleration response spectra of the containment structure are presented in Appendix A.2.

7.2.2.2 Reactor Enclosure and Control Structure The method of analysis and load applications for the computation of the hydrodynamic acceleration response spectrum in the reactor enclosure and the control structure are described in Section 7.1.1.2. The response spectra of the reactor enclosure and the control structure are shown in Appendix B.

7.2-7 Rev. 8, 04/04

r Mo a d' Ih D 0"' N LGS FSAR There are no open discharge pressure relieving devices with limited runs of discharge piping mounted on ASME Code Class I, 2, and 3 systems.

b. Closed Discharge ,

A closed discharge system is characterized by piping between the valve and a tank, or some other terminal end. Under steady-state conditions, there are no net unbalanced forces. The. initial transient response and resulting stresses are determined by using either a time-history computer solution, or a conservative equivalent static solution. In calculating initial l transient forces, pressure and momentum terms are l included. Water slug effects are also considered.

Time-history dynamic analysis is performed for the discharge piping and its supports. The effect of the loading on the header is also considered. The design

load combinations for a given transient are shown in i Table 3.9-11, and the design criteria and stress limits are shown in Tables 3.9-12 and 3.9-16.

3.9.3.4 Component Supports Furnished with the NSSS 3.9.3.4.1 Piping l

l Hangers are designed in accordance with ANSI B31.7. In general, the load combinations for the various operating conditions correspond to those used to design the supported pipe. Design transient cyclic data are not applicable to hangers because no fatigue evaluation is necessary to meet the code requirements.

All hangers are designed, fabricated, and assembled so that they cannot become disengaged by the movement of the supported pipe or equipment after they are installed. The design load on hangers

'is the load caused by dead weight. The hangers are calibrated :: ,

ensure that they support the design load at both their not and

, cold load settings. Hangers provide a specified down travel and up travel in excess of the specified thermal movement.

Snubbers are not supplied by GE; however, required load capacity and snubber location for NSSS piping systems are determined by GE as a part of the NSSS piping system design and analysis scope.

Rev. 27, 12/83 3,9-82

LGS FSAR  !

I The entire piping system, including valves and~the suspension system between anchor points, is mathematically modeled for complete structural analysis. In the mathematical model, the snubbers are modeled as springs with a given stiffness depending on the snubber size. The analysis determines the forces and moments. acting on each component and the forces acting on the snubbers due to all dynamic loading conditions defined in the piping design specification. The design load on snubbers includes those loads caused by seismic forces (OBE and SSE),

system anchor movements, and reaction forces caused by relief valve discharge, turbine stop valve closure, and other hydrod namic forces (SRV, LOCA, AP).

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The snubber location and loading direction are decided by estimation so that the stresses in the piping system have acceptable values. The snubber locations and direction are refined by performing the computer analysis on the piping system as described above.

The spring constant required by the suspension design specification for a given load capacity snubber is compared against the spring constant used in the piping system model. If

(. the spring constants are not in agreement, they are brought into agreement, and the-system analysis is redone to confirm.the snubber loads.

i If the stiffness of the backup structure for the snubber is not large compared to that of the snubbers, the reduced effective snubber stiffness (spring constant) is used in the_ analysis to account for backup structure flexibility.

Snubber design is discussed in Section 3.9.3.5.2~ .

l 3.9.3.4.2~ NSSS Floor-mounted Equipment (Pumps,' Heat Exchangers, and RCIC and HPCI Turbines)

The ECCS pumps, RCIC and SLC pumps, RHR heat exchanger, and RCIC and HPCI turbines are analyzed to verify the adequacy of their support structure under various plant operating conditions. In all cases, the stress loads in the critical support areas are within'ASME Code allowables.M The loading conditions, stress criteria, and allowable and calculated stresses in the critical support areas are summarized in Tables 3.9-6(1), (m), (n), (o),

(q), (r),. (t), and (ae).

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3.9-83 Rev. 27, 12/83

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3. SER Confirmatory Issue #6 - Pressure Isolation Valves Leak Testing The Surveillance Requirement pertaining to leak testing of pressure isolation valves (PIVs) presented in Section 4.4.3.2.2 of Limerick Draft Technical Specification is not complete. In addition to the two requirements- currently identified in Limerick draft . Technical Specification Section 4.4.3.2.2, the' staff requires the PIVs to be leak tested (a) prior to entering the Hot Shutdown whenever the plant has been in Cold Shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 months and (b) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action ~or flow through the valve. Provide additional information to assure that the Limerick plant has the following plant features: (1) full closure of PIV's is verified in the control ~ room by direct monitoring position indicators, (2) inadvertent opening of PIV's is prevented by. interlocks which require the prinary system pressure to be below subsystem design pressure prior to openings, and (3) gross intersystem leakages into the -low-pressure core spray, residual heat removal / low-pressure coolant injection, and residual heat removal / shutdown cooling return and suction lines would be detected by high-pressure alarms and increases in the suppression pool level. With these plant features in p, lace, the PIV's are controlled and verified continuously rather than at the intervals specified in (a) and (b) above and then, the exception for relief from the surveillance requirements (a) and (b) could be accepted.

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Response

The Limerick Generating Station Technical Specifications (Section 4.4.3.2, as modified during the NRC meetings, held June 11-15, 1984) and the Limerick Pump an? Valve Inservice Testing Program Plan require that Reactor Coolant System Pressure Isolation Valves (RCS-PIV) be

-leak tested:

a) At least once-per 18 months, and b) Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.

The additional surveillance requirements (a) and (b) listed in the question above are not required because Limerick has the following features:

1) All RCS-PIV's listed in Tech. Spec. Table 3.4.3.2-1 have position indication in the control room.

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2) All low pressure piping systems isolated by the RCS-PIV's listed in Tech. Spec. Table 3.4.3.2-1 are protected by interlocks which require the reactor coolant system pressure to be below the low l

pressure system design pressure before a direct path may be achieved to the reactor. These interlocks are described along with all safety related high pressure / low pressure system

- interlocks in FSAR Section 7.6.1.2.

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3) Any pressure increase caused by leakage past the Core Spray RCS-PlV's listed in Tech. Spec. Table 3.4.3.2-1 will be sensed and alarmed in the control room when the set point listed in the table is exceeded. After the first refueling outage, any pressure increase caused by leakage past~ the RHR system RCS-PIV's in Tech. Spec.

Table 3.4.3.2-1 will be sensed and alarmed in the control room as above. Before the first refueling outage, the RHR pump discharge line pressure will be observed and recorded once per shif t--from indicators in the auxiliary-equipment room to inform the operators

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'4 SER Open Issue #29 - Stiff Pipe Clamps For all safety-related piping in the NSSS and 80P scope, identify all locations where stiff pipe clamps are used TEef. IE Information Notice, No. 83-80, Use of Specialized Stiff Pipe Clamps). Indicate whether or not stiff clamps are located at or near welds on elbows.

For those stiff clamps located at or near welds on elbows, provide information.to assure that the effects of the clamp-induced pipe loadings have been adequately considered in the Limerick piping .

design and show that the calculated piping stresses for these situations are within applicable code allowables. The information on E-System pipe clamps for the core spray line and feedwater line provided in the letter from J. Kemper to R. Purple dated May 4,

, 1983 is acceptable. In addition, for such clamps, we will require I a commitment to ensure post-installation control of the clamp preload.

Response

A list of E-System clamps installed on BOP and NSSS piping is attached.

This list identifies hanger numbers, part numbers and clamp locations for each piping system. Stress evaluations to consider clamp induced stresses for E-System clamps located at or near elbow welds have been j completed. These stress evaluations were performed for BOP piping.

The evaluation results showed that piping stresses are within the applicable code allowables. These results concur with investigations by both General Electric Company and Bechtel Corporation which indicated that " stiff" pipe clamps do not cause stresses or fatigue levels higher l than the governing stresses or fatigue levels in these piping systems.

Preload requirements for the E-System clamp installation are controlled by specification 8031-P-143-30-7. This specification is also used to control post installation preload of the E-System clamps.

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l Attactment 2 /

Response to NRC MEB OuestIon 1 from NRC Letter Dated June 11, 1984 Including the following specifciations:

Specification 8031-P-362, Specification for Test Requirmients for Hot Deflection Testing of ASE Section III Nuclear Class 1, 2, 3 and ANSI B31.1 Bechtel Piping for the Limerick Generating Station, Units 1 and 2.

Specifiction 8031-P-363, Specification for Requirements for Steady State Vibration Testing of ASE Section III Nuclear Class 1, 2, 3 and ANSI B31.1 Bechtel Piping for the Limerick Generating Station, tkif ts 1 and 2.

Specification 8031-P-364, Specification for Test Requirements for Dynamic Transient Testing of ASE Section III Nuclear Class 1, 2, 3 and ANSI B31.1 Bechtel Piping for the Limerick Generating Station, Units 1 and 2.

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