ML20213D485

Forwards SER Input as Result of 801014-17 Site Review, Identifying Concerns Re Equipment Seismic Qualification Program

ML20213D485
Person / Time
Site:	Summer
Issue date:	12/24/1980
From:	Noonan V Office of Nuclear Reactor Regulation
To:	Tedesco R Office of Nuclear Reactor Regulation
References
CON-WNP-0322, CON-WNP-322 NUDOCS 8101270832
Download: ML20213D485 (8)

Text

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s DEC 2 41980 j

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_9 Docket No. 50-395

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MEMORANDUM FOR: Robert L. Tedesco, Assistant Director o

for Licensing 2l2 m

Division of Licensing o

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Vincent S. Iloonan Assistaat Director

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Materials & Qualification Engineering Divirion of Engineering i'

SUBJECT:

SAFETY EVALUATION REPORT, EQUIPMENT SEISMIC QUALIFICATION FOR V.S. SUMMER NUCLEAR STATION

Reference:

1.) Letter from J.P. Knight to R.C. DeYoung, dated August 26, 1976.

2.) Letter from T.C. Nichols, Jr. to R.R. Denton, dated December 2,1980.

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Plant Name:

V.C. Summer Unit 1 Applicant: South Carolina Electric & Gas Company RSSS: Westinohouse Docket No.: 50-395 Licensing Stage: Operattng License Responsible Branch: Licensing Branch 2 l

Project Manager:

W. Kane Requested Completion Date: N/A l

Review Status: Continuing i

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l The Seismic Qualification Review Team consisting of engineers from the l

Equipment Qualification Branch and the Brookhaven National Laboratory l

has conducted a plant site review of V.C. Summer Unit 1 safety related mechanical and electrical equipment of both NSSS and BOP on October *14-17, 1980. Although this plant was docketed prior to October 27, 1972, the 80P equipment seismic qualification program was aimed at full compliance with the current criteria of Standard Review Plan (SRP) l Sections 3.9.2 and 3.10, instead of IEEE 344-1971. This was confirmed i

by our review. The NSSS electrical equipment for Stamer Station, however, was originally qualified to IEEE 344-1971. This equipment was procured on l

CONTACT:

A. Lee NRR/DE/EQB

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b RRC Form 318 A (4 79) NRCM 02040

Robert L. Tedesco a

qualification demonstrsimilar basis to that whi

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(See References 1 and 2). n test programch was qualified under th NSSS electrical equip atio equipm In ment seismic qualificatiorder to furtherapproved b review.ent, together includes threeA total of 25 pieces with that of NSSS mecha i seismic confim the adequacy ofstaff in Aug on cal, program, certain HSSS e additional pieces of equipment was n

Our in the considered forare included in the p review,put to the Safety Evaluation R is selected at the acceptability of enclosed.

conclusion ofreviewwhich site As is cant to provprogram with respectthe applicant's NSSS noted, we identified someeport (SER), as a the visit.

to SRP 3.9.2 and 3 10 In addition', ide concernsof the SQRT End 30P equipment s i additional information t reports we of nine (requested as to the criteria e

review.

requestedsmic qualification the applicant to pro iof equipment sele applicant's infomationWe will re) port the r 9

o clarify an. We pieces the d resolve these the appli-trip report, in a v de the test and anal concerns.

esults of such confimatory re clarifying the above con or a follow ysis supplement up confirmatory to this SER.

view cerns as identified in and review of our WAB M S1 cedByi 8

Vincent S. Hoonan

Enclosure:

DivisionMaterials & Qualific, Assistant Dire As stated cc: R. Vollmer of Engineeringations Engineering D. Eisenhut Z. Rosztoczy R. Bosnak A. Schwencer C. Hofmayer P. Chen W. Kane A. Lee occk. 02($

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Robart L. Tedesco a similar basis to that which was qualified under the Westinghouse seismic qualification demonstration test program approved by the staff in August 1976 (See References 1 and 2).

In order to further confirm the adequacy of the fiSSS electrical equipment seismic qualification program, certain HSSS electrical equipment, together with that of NSSS mechanical, are included in the plant site review. A total of 25 pieces of equipment was considered for review which includes three additional pieces selected at the conclusion of the visit.

Our input to the Safety Evaluation Report (SER), as a result of the SQRT review, is enclosed. As is noted, we identified some concerns as to the i

acceptability of the applicant's NSSS and B0P equipment seismic qualification program with respect to SRP 3.9.2 and 3.10 criteria. We requested the appli-cant to provide additional infomation to clarify and resolve these concerns.

In additio'n, we requested the applicant to provide the test and analysis reports of nine (9) pieces of equipment selected for a follow up confirmatory review. We will report the results of such confirmetory review and review of the applicant's information clarifying the above concerns as identified in our trip report, in a supplement to this SER.

grasinal Sisned BY3 Vincent S. Noonan, Assistant Director Materials & Qualifications Engineering Division of Engineering

Enclosure:

l As stated cc: R. Vollmer D. Eisenhut Z. Rosztoczy R. Bosnak A. Schwencer C. Hofmayer P. Chen W. Kane A. Lee l

l OFFICE k.,

SURNAME k.. Alee; M.

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/80 NRC FORM 318 (9-76) NRCM 0240 D U.S. GOVERNMENT PRINTING OFFICE: 1979 289-369

DEC 2 41980 Robert L. Tedesco a similar basis to that which was qualified under the Westinghouse seismic qualification demonstration test program approved by the staff in August 1976 (See References 1 and 2).

In order to further confirm the adequacy of the NSSS electrical equipment seismic qualification program, certain NSSS electrical equipment, together with that of NSSS mechanical, are included in the plant site review. A total of 25 pieces of equipment was considered for review which includes three additional pieces selected at the conclusion of the visit.

Our input to the Safety Evaluation Report (SER), as a result of the SQRT review, is enclosed. As is noted, we identified some concerns as to the acceptability of the applicant's NSSS and BOP equipment seismic qualification program with respect to SRP 3.9.2 and 3.10 criteria. We requested the appli-cant to provide additional information to clarify and resolve the:;e concerns.

In addition, we requested the applicant to provide the test and analysis reports of nine (9) pieces of equipment selected for a follow up confirmatory review. We will report the results of such confinnatory review and review of the applicant's information clarifying the above concerns as identified in our trip report, in a supplement to this SER.

O Mate b SI ona ic tant Director

& Qualifications Engineering Division of Engineering

Enclosure:

As stated cc: R. Vollmer D. Eisenhut Z. Rosztoczy l

R. Bosnak l

A. Schwencer C. Hofmayer i

P. Chen W. Kane l

A. Lee i

Equipment Qualification Branch Input for V.C. Summer Safety Evaluation Report 3.10 Seismic Qualification of Category I Mechanical and Electrical Equipment Mechanical and electrical (includes instrumentation, control and electrical) equipment and components required to perform a safety function are designed to meet Seismic Category I design criteria. Seismic requirements established by the seismic system analysis have been incorporated into equipment speci-fications to assure that the equipment purchased or designed meets seismic requirements equal to or in excess of the requirements for Seismic Cate-gory I equipment and components; either by appropriate analysis or by qualification testing or a combination of analysis and testing.

The applicant has implemented a seismic qualification program for seismic Category I mechanical and electrical equipment and the associated supports for that equipment to provide assurance that such equipment can be expected to function properly, and that structural integrity of the equipment and its supports will not be impaired during the excitation and vibratory forces imposed by the safe shutdown earthquake and under the conditions of post earthquake operation. The above qualification program was implemented while Standard Review Plan (SRP) Sections 3.9.2 and 3.10 were being published and therefore was aimed at full compliance with these SRP Sections. SRP 3.9.2 and 3.10 specify criteria which when conformed with satisfies the applicable portions of GDC 2 of Appendix A to 10 CFR 50. SRP 3.10 references Regulatory Guide 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants," and IEEE Standard 344-1975, "IEEE Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Gerferating Stations."

The principle change from earlier criteria is to require consideration of equipment multi-mode response and biaxial coupling effects.

2-Our Seismic Qualification Review Team (SQRT) performed a review at the plant site on October 14-17, 1980 to determine the extent to which the qualification of the equipment, as installed in V.C. Summer, meets cur-rent licensing criteria as described in SRP Sections 3.9.2 and 3.10.

During this review we evaluated a representative sample of twenty-two pieces of Seismic Category I mechanical and electrical equipment, both in NSSS and B0P. Among the eighteen B0P equipment selected, a review of the qualification of the Reactor Building Cooling Unit Damper Actuator and the Radiation Monitoring Control Panel had not been completed by the applicant's architect engineer, Gilbert / Commonwealth Associates, and therefore qualification reports were not available for review.

In addi-tion, the qualification documents for Main Steam Isolation Valve, although having been approved by Gilbert /Comonwealth Associates, was not available for review during the visit, but were provided to the SQRT at the conclu-sion of the visit. The complete documents for 480 Volt Substations were only briefly reviewed during the visit, and were also provided to SQRT for further review. The Hydrogen Analyzer Panels had not been delivered to the plant and complete information was not available during the plant site visit. Among the four NSSS equipment selected, the appropriate quali-fication documents for the Post Accident Monitors (PAM) Indicators were not available for review.

In addition to the six outstanding qualification reports identified above, the SQRT, at the conclusion of the visit, requested the applicant to pro-vide the test and analysis reports for three additional selected pieces of equipment, encompassing both the B0P and NSSS category, to be included

. in a followup confirmatory review. These equipment are Diesel Generator and Associated Equipment (Electrical and Air Starting Controls), Accumulator Tanks, and Electrical Containment Penetrations and Miscellaneous Connectors.

Our review of the available equipment qualification when compared with current criteria of. SRP Sections 3.9.2 and 3.10 identified the need to clarify the details of the qualification for some items of equipment. For example:

1) the design of the supports for the Battery Charger need to be clarified since they were bolted to the test table, but are welded to the floor in the plant; 2) on the Charging Pump, some small pipes are loosely supported and clarification of their safety significance is needed; and

3) in all the safety related valves reviewed, the justification of the "g" levels used for qualification needs to be documented and verified with the as-built piping analysis results. The details of our review and the con-cerns identified in both the B0P and NSSS equipment qualification are des-cribed in the report of our October 14-17 trip to the plant.

In order to complete our review we have requested the applicant to provide the following information:

1)

Identify all equipment still to be qualified and provide documenta-tion to demonstrate the completion of the qualification program. Pro-vide SQRT " Qualification Summary of Equipment" forms for this equip-ment and update the forms provided for the site visit.

2) Review and revise, as necessary, FSAR tables in Chapter 3 to include updated information of all safety related systems and components.

. 3) Provide a copy of the revised SQRT Tables which include a list of equipment and the summary of the qualification program.

4) For all safety related valves describe the design procedure used to demonstrate that the accelerations used in the valve qualifica-tion equal or exceed the accelerations obtained in the final as-built piping analysis. Provide specific information for the valves reviewed by SQRT.

5) Provide qualification reports for the four pieces of equipment not available during the visit and the three additional pieces of equip-ment selected by the staff at the conclusion of the visit.

6) Provide confirmation that Westinghouse's generic response spectra for equipment qualification envelop the corresponding plant specific required response spectra.

7) Clarify details as discussed in our trip report concerning the quali-fication of Component Cooling Water Pump and Motor, Turbine Appur-tenances for Turbine Driven Emergency Feedwater Pump, Charging Pump, RHR Pumps, Battery Chargers, Control Valves, and Pressure and Dif-ferential Pressure Transmitters.

Based on the results of the review to date, we conclude that an appropriate equipment qualification program has been defined for the seismic Category I mechanical and electrical equipment which will provide adequate assurance that such equipment will function properly during and after the excitation from vibratory forces imposed by the Safe Shutdown Earthquake. We are

. j continuing our review of the implementation of the program and will report on our conclusions, including our evaluation of the additional information requested of the applicant as discussed above, in a supplement to this report.

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k UNITED STATES NUCLEAR REGULATORY COMMISSION j

WASHINGTON, D. C. 20555 i

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December 30, 1980 Task Action Plan A-39 Docket Nos.: 50-353, 50-352/353, 50-367, 50-373/374, 50-387/388, 50-410, 50-322, 50-397 i

MD10RANDUM FOR: Karl Kniel, Chief Generic Issues Branch Division of Safety Technology FROM:

T. M. Su, A-39 Task Manager Generic Issues Branch, DST APPLICANT:

Members of the Mark II Owners Group

SUBJECT:

MEETING WITH REPRESENTATIVES OF MARK II OWNERS GROUP AND GENERAL ELECTRIC COMPANY TO DISCUSS THE SUPPRESSION POOL TEMPERATURE LIMITS Backaround The BWR containments are equipped with a number of safety / relief valves (SRV) to release mass and energy from the primary system for overpressure protection.

Inadvertant actuation of a SRV will also result in primary system blowdown. Steam released from the primary system is discharged to the suppression pool where it is condensed. Extended blowdown, however, may heat the suppression pool to the point, where the degree of subcooling of suppression pool water is insufficient to condense the steam smoothly. Severe vibratory loads may result. The current practice is to limit the maximum suppression pool temperature below this potential threshold temperature. This limit was established at a local 'emperature of 200*F. The applicants, however, believe that there is eviuence from test data to relax this limit. The purpose of this meeting was to discuss these test data and the justification for the assumptions used to calculate the suppression pool temperature response.

An attendance list and a enoy of the meeting handouts are enclosed.

Summary 1.

Temoerature Limits Mr. J. Post of General Electric Company presented the data for SRV condensation tests performed at Kraftwerk Union (KWU) facilities in Germany and CNEN Laboratory in Italy. Although these data had been presented to the NRC staff and their consultants k

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S Karl Kniel December 30, 1980 new approach to analyze the data by using subcooling as the key parameter has been made. Results of this new analysis show that the quencher device is capable of stable steam condensation up to 212*F.

Thedegreeofsubcoolingforthesetesgsrangesfrom8'Fto20*Fwith mass flux ranging from 2 to 160 lbm/ft -sec.

The applicants, therefore, concluded that stable condensation can be achieved by maintaining the suppression pool temperature subcooling greater than 8*F at the quencher discharge. On this basis, the applicants proposed the suppression pool temperature limits as follows:

2 a.

For mass flux greater than 140 lbm/ft -sec, the local pool temperature should not exceed 200*F; 2

b.

For mass flux below 140 lbm/ft -sec, the local pool temperature limit can be increased to 212*F.

l The applicants stated that these limits are justified by the test data and the geometries of Mark II suppression pools, which provide a minimum of 16*F subcooling.

The staff expressed a concern regarding the applicability of CNEN tests, which were performed in subscaled facilities. The adequacy of scaling, instrumentation setup and data interpretation has to be reviewed and evaluated before any conclusion ca'n be made.

Thestaffalsoexpressedaconcernonthevibragoryloadsexhibited in the KWU tests for mass flux below 140 lbm/ft -sec. These loads are relatively low in magnitude but exhibit high frequencies in l

comparison with initial air clearing loads. GE indicated that they l

will evaluate the data to detemine these loads.

2.

Justification for Assumptions The key assumptions used in the calculation of suppression pool temperature response include time allowed the operator to scram the reactor, availability of RHR system operating in pool cooling mode, and main condenser as a heat sink for primary system blowdown. The applicants discussed the basis for each assumption.

C. Graves of RSB expressed a concern on the potential of cutting off the RHR pool cooling mode in response to high containment pressure as a result of suppression pool heat-up. He also questioned the methodology

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used to calculate the mass and energy release through the SRV.

3 Karl Kniel December 30, 1980 The applicants also presented the results of a sensitivity study on each assumption. A five minute delay in manual scram time could increase the peak pool temperature by as much as 12*F. Most of the assumptions show insignificant effect on peak pool temperature.

In response to the requirement by R. Frahm of RSB during the meeting held on April 10, 1980, the applicants have investigated the suppression pool temperature response to FSAR Chapter 15 events.

GE stated that results of this study show that the Chapter 15 events yield lower peak pool temperature than that calculated on the basis of the current design cases.

3.

Conclusion T. Su of the NRC staff provided the staff's comments on the presentation which are summarized below.

The staff and its consultants believe that it is appropriate to evaluate the quencher performance of steam condensation on the basis of degree of subcooling. Since the CNEN data are based on subscaled tests, the applicability of this data base has to be determined. In order to expedite the review of pool temperature limit, we believe that the cut-off point of mass flux should be based on the KWU data, which have been reviewed extensively in the past.

We require that the applicants provide justification to demonstrate that the containment structure, piping and equipment can accommodate the oscillatory loads resulting from steam condensation for the mass flux below the cut-off point.

With respect to the calculation of pool temperature response, we will require the applicants to provide a detailed description of the assumptions and methods used to calculate the mass and energy l

released to the suppression pool. This includes an identification l

of all computer programs and a description of the data transferred to initiate any succeeding calculations. The description should also include the availability of each system related to the events of SRV actuations such as RHR in pool cooling mode (e.g., automatic cut-off with high containment pressure) and transient feedwater flow following feedwater pump trip. The applicants were informed that they should contact C. Graves (942-9404) of RSB to obtain the detailed requirements for the resolution of this issue.

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Karl Kniel December 30, 1980 The applicants indicated that the additional information required to resolve the issues stated above vill be made available for the staff's review by the end of March, i:181. The staff indicated, however, that an early submittal N.s of this infonnation is needed to meet the current schedule of cort.htion date, i.e. February,1981.

The completion schedule for this ta c has been slipped for several months as a result of requests maC Sy the Mark II owners group to relax the temperature limit. Further sl:.. age of the schedule may jeopardize the availability of staff r 2sourcet to review this task.

j?4 T. M. Su, A-39 Task Manager Generic Issues Branch Division of Safety Technology We

LIST OF ATTENDEES T. M. Su, HRC/ DST /GIB C. Economos, BNL/NRC A. Sonin, MIT (for BNL)

T. Lee, NRC/RSR J. E. Metcalf, S&W D. Desmarais, S&W H. R. Johnson, EBASCO W. M. Davis, GE L. Schell, Penn. Power & Light Co.

O. A. Nossardi, Bechtel (S.F.)

L. Steinert, GE R. W. Riley, Cin. Gas & Elec. Co.

P. T. Mairose, WPPSS G. E. Gottfried, S&L R. Ralph, Commonwealth Edison J. Post, GE C. Graves, NRC/DSI, RSB C. Lin, BNL/NRC R. F. McClelland, GE H. C. Urang, GE P. Norian,'NRC/ DST /GIB w

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DECEf1BER 1980

. APRIL 1980 MEETING C0ilCERNS:

1)

SENSITIVITY OF THE ANALYSIS TO VARIATIONS I. THE PARAMETER (ASSUMPTION)

VALUES

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2)

COMPARISON OF FSAR CH.' 15 TO LTPT TRANSIENTS

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NON-MECHANISTIC ASSUMPTIONS Q

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DECEMBER 1980 APPENDICIES TO." WHITE PAPER

1)

SENSITIVITY OF THE AllALYSIS TO VARIATIONS IN THE PARAMETER VALUES 2)

COMPARISON OF FSAR CH, 15 TO LTPT TRAllSIENTS 3)

ASSUMPTION BASIS l

" WHITE PAPER" REVISION 1)

NON-MECHANISTICASSUMPTIONS s.

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f DECEMBER 1980 CHANGES TO THE " WHITE PAPER" ASSUMPTI0tlS OLD NEW 1)

NON-MEDIANISTIC TREATMENT OF MAKEUP VIA 1)

MECHANISTIC TREATMENT OF FEEDWATER FEEDWATER SYSTEM ADDITION A)

TURBINE-DRIVE FEEDPUMPS B)

MOTOR-DRIVEN FEEDPUMPS 2) 0FFSITE POWER Ui1AVAILABLE EXCEPT CASE 1A 2) 0FFSITE POWER AVAILABLE ALL CASES NO INITIATIbH OF HPCSMIPCI OR RCIC HPCS/HPCI INITIATED, NO RCIC.

3) 3)

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DECEMBER 1980 f

NEW ASSUMPTIONS TO THE

" WHITE PAPER" 1.

FEEDWATER ADDITION TO THE REACTOR PRESSURE VESSEL (RPV).

A.

PLANTS UTILIZING TURBINE-DRIVEN FEEDPUMPS.

I)

UPON MAIN STEAM ISOLATION VALVE (MSIV)

CLOSURE, THE TURBINE-DRIVEN FEEDPUMPS SUPPLY FEEDWATER UNTIL THE DISCHARGE HEAD FALLS BELOW THE REACTOR PRESSURE.

II)

FOR CASES WHERE MAIN STEAM ISOLATION HAS OCCURRED, THE CONDENSATE (BOOSTER)

PUMP (S) SUPPLIES FEEDWATER TO THE RPV WHEN THE REACTOR PRESSURE FALLS BELOW THE CONDENSATE (BOOSTER) PUMP DISCHARGE HEAD.

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DECEMBER 1980 NEW ASSUMPTIONS TO THE

" WHITE PAPER" i

1.

FEEDWATER ADDITION TO THE REACTOR PRESSURE VESSEL (RPV).

j B.

PLANTS UTILIZING MOTOR-DRIVEN FEEDPUMPS I)

FEEDPUMPS SUPPLY FEEDWATER TO THE RPV UNTIL THE FEEDPUMPS TRIP DN AN AUTOMA-TIC SIGNAL (E.G., VESSEL HIGH WATER LEVEL TRIP).

II)

AFTER THE FEEDPUMPS HAVE TRIPPED, THE CONDENSATE (BOOSTER) PUMP (S) SUPPLIES

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FEEDWATER TO THE RPV WHEN THE REACTOR PRESSURE FALLS BELOW THE CONDENSATE (BOOSTER) PUMP DISCHARGE HEAD.

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D.ECEMBER 1980

" WHITE P PER" 1.'

FEEDWATER ADDITI0il TO THE REACTOR PRESSURE VESSEL (RPVI.'

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FEEDWATERWILLBESUPPLIEDTOTHERPVSAS D~ SCRIBED AB0VES UNTIL THE EllTHALPY OF THE FEEDWATER iS LESS THAN OR EQUAL TO THE ENTHALPY OF THE SUPPRESSION POOL WATER.'

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DECEMBER 1980 j

NEW ASSUMPTIONS TO THE

" WHITE PAPER" 1.

HPCI/HPCS INJECTION OF SUPPRESSION P0OL WATER 9

INTO THE RPV IS DETERMINED BY THE AUTOMATIC START AND STOP SIG1ALS GENERATED (E.G., LOW-LOW RPV LEVEL STARTS THE HPCI/HPCS INJECTION, HIGH RPV LEVEL STOPS HPCI/HPCS INJECTION).

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WHITE PAPER SUPPL!ME!!TAL INF0PJ'iAT10f}

e ASSUMPTI0tiBASIS 9

SEriSITIVITYSTUDIES e

FSAR CHAPTEP,15 CotiPARISON e

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JS POST /12/80 1.

ASSUMPTION BASIS e

KEY ASSUMPTIONS INCLUDED EQUIPMENT AVAILIBILITY OPERATOR ACTIONS e

ASSUMPTION AND BASIS 1)

MSIV CLOSURE 2)

RHR POOL COOLING 3)

MANUAL DEPRESSURIZATION 4)

SUPPRESSION POOL WATER 5)

FEEDWATER 6) 0FFSITE POWER 7)

MANUAL SCRAM 8)

SCRAM ON HIGH DRYWELL PRESSURE 9)

MAIN CONDENSER l

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1, MSIV CLOSURE ASSUMPT10N:

e 3.5 SECOMDS AFTER ISOLATION SIGNAL IF ISOLATION EXPECTED', ASSUME SIGNAL AT TIME = 0'.'

e BASIS:

MINIMIZE STEAM OUT REACTOR IS CONSERVATIVE {

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0.5 SEC TYPICAL INSTRUMENT DELAY o

TESTED CLOSURE IS 5 TO 10 SECONDS o

3 SECOND LINEAR CLOSURE IS CONSERVATIVE e

SIGNAL AT TIME = 0 IS CONSERVATIVE' e

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JS POST /12/80 3.

22.

RHR POOL COOLING ASSUMPTION:

e POOL C00LIMG OM 10 MINUTES AFTER TS1 BASISr o

POSITIVE ALARM AT TS1 e

STAtlDARD OPERATOR RESPONSE IN NORMAL AND EMERGENCY PROCEDURES o

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JS POST /12/80 4.

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V 3.

MANUAL DEPRESSURIZAT!0N j

l ASSUMPTION:

o MANUAL DEPRESSURIZATION OF REACTOR BEGUN AT TS4 (120 F).

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e EXCEPTION IF EVENT IS ALREADY DEPRESSURIZATING REACTOR AT A SUFFICIENT RATE e

TERMINATED WHEN INITIATE SHUTDOWN COOLING BASIS:

e.

OPERATOR TRAINED TO RESPOND BY DEPRESSURIZING TO UTILIZE HEAT SINKS e

NORMAL AND EMERGENCY PROCEDURES CALL FOR DEPRESSURIZATION AT TS4 o

SIGNIFICANT TIME LAPSE SINCE PREVIOUS ALARMS e

REACTOR SYSTEM UNDER CONTROL FOLLOWING TRANSIENT EVENT e

RHR SHUTDOWN COOLING IS SUFFICIENT TO REMOVE LONG TERM DECAY HEAT t

JS POST /12/80 5.

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4.

SUPPRESSION POOL WATER ASSUMPTION:

o WATER MASS IN PEDESTAL NOT INCLUDED AS HEAT SINK BASIS:

o MIXING BETWEEN PEDESTAL WATER AND MAIN POOL IS NEGLECTED o

MIllIMIZING POOL VOLUME IS CONSERVATIVE F

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FEEDWATER ASSUMPTION:

e TURBINE DRIVEN PUMPS CONTINUE TO INJECT DURING C0ASTDOWN e

MOTOR DRIVEN PUMPS CONTINUE TO INJECT

-e AT LOW REACTOR PRESSURE CONDENSATE (BOOSTER) PUMPS INJECT e

CONTINUED AS LONG AS FEEDWATER ENTHALPY GREATER THAN POOL ENTHALPY BASIS:

o CONSERVATIVE TO INCLUDE FEEDWATER ADDITION FOR CONTAINMENT

RESPONSE

e MECHANISTIC TREATMENT FOLLOWING ISOLATION e

RECOGNIZE OPERATOR CONFIDENCE IN FEEDWATER SYSTEM e

CONSERVATIVE AS LONG AS FEEDWATER TEMPERATURE IS GREATER THAN POOL TEMPERATURE e

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JS POST /L2/80 7.

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6.

0FFSITE POWER ASSUMPTION:

o 0FFSITE POWER AVIAI.ABLE ALL CASES MSlS:

o CONSISTENT WITH CONTINUED FEEDWATER FLOW ASSUMPTION REQUIRED TO DRIVE CONDENSATE (BOOSTER) PUMPS AND MOTOR e

DRIVEN FEEDWATER PUMPS SINCE CONTINUED FW IS CONSERVATIVE, 0FFSITE POWER AVAILABLE e

IS CONSERVATIVE e

h JS POST /12/80 8.

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7. MANUAL SCRAM ASSUMPTION:

e FOR INADVERTENT SORV CASES e

MANUAL SCRAM AT TS3 (110 F)

.e OPERATOR TRANSFERS REACTOR MODE SWITCH FROM "RUN" TO " SHUTDOWN" BASIS.:

o SORV IS TRANSIENT REACTOR INTEGRITY NOT THREATENED NO AUTOMATIC SCRAM SIGNAL o

POSITIVE INDICATION OF SCRV o

ALARM AT TS1 AND AT TS3 e

OPERATOR TRAINED TO RESPOND TO SORV EVENT SCRAM REQUIRED BY TECHNICAL SPECIFICATIONS AND IN NORMAL e

AND EMERGENCY PROCEDURES OPERATING EXPERIENCE SHOWS SCRAM OCCURS WELL BEFORE TS3 o

ASSUMING. REACTOR AT FULL POWER UNTIL POOL REACHES TS3 e

IS CONSERVATIVE m

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8.

SCRAM'ON HIGH DRYWELL PRESSURE ASSUMPTION:

o FOR SBA EVENT e

SCRAM AND ISOLATION AT TIME = 0 MSlS.:

e HIGH DRYWELL PRESSURE SIGNAL GENERATED BEFORE VENTS ARE CLEARED AND POOL HEATUP BEGINS e

SCRAM IS AUTOMATIC ON HIGH DRYWELL PRESSURE e

SPURIOUS ISOLATION AT TIME = 0 MAXIMIZES ENERGY ADDITION TO THE POOL e

SBA ANALYSIS DOES NOT TAKE CREDIT FOR ENERGY HELD UP AND HEAT SINKS IN DRYWELL.

ALL BREAK FLOW / ENERGY HEATS UP POOL DIRECTLY 1

JS POST /12/80 10,

9. -MAIN CONDENSER ASSUMPTION':

o FOR SORV WITH SINGLE FAILURE OF LOSS OF ONE RHR MAlf! CONDENSER REMAINS AVAILABLE AS HEAT SINK o

BASIS:

o SORV IS A TRANSIENT o

fl0RMAL PLANT EQUIPMENT WILL REMAIN AVAILABLE o

MAIN CONDENSER IS KNOWN PREFERENTIAL HEAT SINK e

EVENT SEQUENCE WILL NOT GENERATE SIGNAL THAT WOULD MAKE MAIN CONDENSER UNAVAILABLE 1

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EVENT DESCRIPTION e

REACTOR AT FULL POWER, POOL JUST BELOW TS1 e

SRV SPURIOUSLY OPENS AND STICKS OPEN e

PLANT AUTOMATIC CONTROLS RESPOND TO STABILIZE REACTOR j

SYSTEM e

OPERATOR ATTEMPTS TO CLOSE SORV, THEN SCRAMS AND PLACES RHR IN POOL COOLING e

DECAY IN REACTOR STEAM GENERATION FOLLOWING SCRAM TCV CLOSE TURBINE BYPASS DOES NOT OPEN AUTOMATICALLY a

REACTOR WATER LEVEL MAINTAINED BY FEEDWATER OPERATO,R OPENS TURBINE BYPASS VALVES TO USE MAIN CONDENSER e

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SENSITIVITY STUDIES e

KEY PARAMETERS AND ASSUMPTIONS e

FOR ZIMMER PLANT SPECIFIC CASES BASE CASE IN REV 0 WHITE PAPER l

e APPLICABILITY TYPICAL OF ALL MARK II PLANTS REV 1 WHITE PAPER ASSUMPTIONS HAVE MINOR IMPACT e

SINGLE PARAMETERS VARIED 4LT FROM BASE CASE GIVEN NEGATIVElif IS LOWER POOL TEMP POSITIVEllT IS HIGHER P0OL TEMP 5

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TABLE D-1 SUM 4ARY OF THE PEAK POOL TEMPERATURE SENSITIVITY STUDIES CHANGE IN SUPPRESSION PLANT POOL PEAK TEMPERATURE EVENT PARAHETER VARIATIONS

('F) 3A Service Water 95'F BASE Temperature 85*F

- 1. 3 75'F

-2.6 65"F

-3.8 1A Manual Scram Time 5 minutes (when SASE Tpool = 110 F) l 3 minutes

-5.4 10 minutes

+11.7 1A Time at which Hain 20 minutes BASE Condenser is available 10 minutes

-13.9 30 minutes

+5.0 3A Time for Initiation of 10 minutes BASE RIIR Pool Cooling 20 minutes 40.2 1

30 minutes

+0.6 3B Time for Initiation of 20 minutes (when BASE Manual Depressurization Tpool = 120'F) 15 minutes 0.0 30 minutes

+0.1 3B Manual Depressurization 100*F/hr BASE Rate 200*F/hr

-2.4 h

125"F/hr

-0.7

f 75'F/hr

+0.1 l-3 3A Initial Pool Water Mass 5.83 x 10" Ibm /(LWL)

BASE

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5.97 x 10 lba/(HWL)

- 1. I

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FSAR CHAPTER'15 COMPARIS0N 4

e COMPARIS0N OF LONG TERM POOL TEMPERATURE EVENTS AND

.'FSAR CHAPTER lui EVENTS e

EXAMINE FROM TWO PERSPECTIVES WOULD FSAR CHAPTER 15 EVENT GIVE GREATER POOL HEATUP THAN LONG TERM POOL TEMPERATURE EVENT?

DOES LONG TERM POOL TEMPERATURE EVENT MAKE ANY ASSUMPTIONS THAT COMPROMISE OR DEGRADE FSAR CHAPTER 15 EVENTS?

e KEY RESULT

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LONG TERM POOL TEMPERATURE EVENTS HAVE HIGHER PEAK POOL TEMPERATURE THAN FSAR CHAPTER 15 EVENTS THAT HAVE STEAM DISCHARGE THROUGH SRVS 1

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FSAR CHAPTER -15~ COMPARISON PERFORMANCE OF COMPARISON:

e EVENTS CLASSIFIED SIMILAR TO SORY

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SIMILAR TO ISOLATION / SCRAM SIMILAR TO SBA NO POOL HEATUP N/A - FOR PWR'S ONLY o

EXAMINED CASE BY CASE ASSUMPTIONS RESULTS t

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a FSAR CHAPTER 15~ COMPARISON CHAPTER 15 EVENTS e.

FSAR EVENTS PREDOMINENTLY ISOLATION / SCRAM EVENTS e

CALCULATE SHORT TERM REACTOR RESPONSE UNTIL RPV STABILIZED e

MECHANISTIC REACTOR SYSTEM TREATMENT DETAILED REACTOR MODEL NO CONTAINMENT MODEL e

ANY SINGLE FAILURE CONSIDERED LEAVES TWO RHR FOR POOL COOLING p

6 m

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IR PORT /1? /90 10

I TABLE E-1 FSAR CHAPTER 15 EVENTS VERSUS LONG TERM POOL TEMPERATURE EVENTS SIMILAR LONG TERM POOL TEMPERATURE EVENTS SORV SMALL NO POOL AT ISOLATION /

BREAK' TEMPERATURE FSAR CHAPTER 15 EVENTS POWER SCRAM ACCIDENT INCREASE 15.1.1 Loss of Feedwater Heating X

2 Feedwater Controller Failure - Maximum Demand X

3 Pressure Regulator Failure - Open X

4 Inadvertent Safety Relief Valve Opening X

5 PWR Steam Piping Break N/A 6

Inadvertent RHR Shutdowp Cooling Operation X

15.2.1 Pressure Regulator Failure - Closed X

2 Generator Load Reject X

3 Turbin'e Trip X

4 MSIV Closures X

5 Loss of Condenser Vacuum X

6 Loss of AC Power X

7 Loss of Feedwater Flow X

8 Feedwater Line Break X

9 Failure of RHR Shutdown Cooling X

15.3.1 3

Recirculation Pump Trip X

2 Recirculation Flow Control Failure -

X Decreasing Flow I

3 Recirculation Pump Seizure X

4 Recirculation Pump Shaft Break X

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M TABLE E-1 (CONTINUED)

FSAR CHAPTER 15 EVENTS VERSUS LONG TERM POOL TEMPERATURE EVENTS

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1 SIMILAR LONG TERM POOL TEMPERATURE EVENTS l

SORY SMALL NO POOL l

AT ISOLATION /

BREAK TEMPERATURE l

FSAR CHAPTER 15 EVENTS POWER SCRAM ACCIDENT INCREASE l

15.4.1 Rod Withdrawal Error - Low Power X

2 Rod Withdrawal Error - At Power X

3 Control Rod Haloperation X

4 Abnormal Startup of Idle Recirculation Pump X

5 Recirculation Flow Control Failure with X

Increasing Flow

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Chemical and Volume Control System Halfunctions N/A 7

Hisplaced Bundle Accident X

8 Spectrum of Rod Ejection Assemblies N/A 9

Control Rod Drop Accident X

1 15.5.1 Inadvertent HPCI/HPCS Startup X

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2 Chemical Volume Control System Malfunction N/A I

15.6.1 Inadvertent Safety Relief Valve Opening X

2 Instrument Line Break X

3 Steam Generator Tube Failure N/A

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Steam System Piping Break Outside Containment X

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Loss-of-Coolant Accidents X

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Feedwater Line Break Outside Containment X

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I QUENCHER CONDENSATION PERFORMANCE t

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OBJECTIVE 1

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QUENCHER CONDENSATION PERFORMANCE OBJECTIVE SHOW BY TEST DATA THAT THE LOCAL TEMPERATURE LIMIT CAN BE RAISED AT LOW fMSS FLUX DUE TO THE SUBC00 LING AVAILABLE TO THE QUENCHER DEVICE 1

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QUENCHER CONDENSATION PERFORMANCE PROPOSED RANGE FOR STABLE CONDENSATION TLOCAL 200*F, G >140 LBM/FT -SEC 2

TLOCAL 212*F, G6 140 LBM/FT -SEC 2

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QUENCHER CONDENSATION PERFORMANCE CONCLUSION FROM DATA e

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MARK II QUENCHER SUBMERGENCE IS 13 To 23 FEET e

CURRENT LIMIT (200 F) dTsun g 2 8

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