ML20069L837

Forwards Draft Responses to Reactor Sys Branch Draft SER Re TMI Items II.K.3.13,II.K.3.15,II.K.3.16,II.K.3.18,II.K.3.21, II.K.3.22,II.K.3.24,II.K.3.25,II.K.3.30,II.K.3.31 & II.K.3.45.Response Will Be Incorporated Into FSAR Revs

ML20069L837
Person / Time
Site:	Limerick
Issue date:	04/27/1983
From:	Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Schwencer A Office of Nuclear Reactor Regulation
References
TASK-2.K.3.11, TASK-2.K.3.13, TASK-2.K.3.15, TASK-2.K.3.18, TASK-2.K.3.21, TASK-2.K.3.22, TASK-2.K.3.24, TASK-2.K.3.25, TASK-2.K.3.30, TASK-2.K.3.31, TASK-2.K.3.45, TASK-TM NUDOCS 8305020454
Download: ML20069L837 (13)

Text

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PHILADELPHIA ELECTRIC COMPANY 2301 M ARKET STREET P.O. BOX 8699 PHILADELPHI A. PA.19101

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EUGENE J. BR ADLEY assoctava esmaRAk COUNSEL DON ALD BLANKEN April 27, 1983 RUDOLPH A. CHILLEMI E. C. MIR M H A LL T. H, M AHER CORN ELL PAUL AUERBACH Assesvany esman AL counssk EDW A RD J. CU LLE N, J R.

THOM AS H. MILLER, JR.

GRENE A. McKENN A Assistant cousess6 Mr. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing U. S. Nuclear Regulatory Ccnmission Washington, D.C. 20555

Subject:

Iamerick Generating Station, Units 1 and 2 Response to Reactor Systens Branch Draft Safety Evaluation Report (DSER) on TMI Itens

Reference:

A. Schwencer to E. G. Bauer, Jr. letter dated March 11, 1983 File: GOVT l-1 (NPC)

Dear Mr. Schwencer:

The attached documents are draft responses to the Reactor Systems Branch DSER, Itan 2, relating to TMI iters. Ihe response to II.K.3.18, which was subnitted in the March,1983 revision of the FSAR, is being revised and will be formally incorporated into the FSAR revision scheduled for May, 1983. The responses to II.K.3.30 and II.K.3.31 will be forrally incorporated into the FSAR revision scheduled for April, 1983. The responses to II.K.3.13, II.K.3.15, II.K.3.16, II.K.3.22, II.K.3.24, II.K.3.25, and II.K.3.45 will be fonrally incorporated into the FSAR revision scheduled for May,1983. Please note that the response to II.K.3.21 has been subnitted via FSAR revision in March,1983.

Sincerely, 0 n 7,

Eu ene . Er y b IN/gra/Z-9 cc: See Attached Service List 8305020454 830427 PDR ADOCK 05000352 E pm

cc: Judge Lawrence Brenner (w/ enclosure)

Judge Richard F. Cole ' (w/ enclosure)

Judge Peter A. Morris (w/ enclosure)

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

Ann P. Hodgdon (w/ enclosure)

Mr. Frank R. Romano (w/ enclosure)

Mr. Robert L. Anthony (w/ enclosure)

Mr. Marvin I. IEwis (w/ enclosure)

Judith A. Dorsey, Esq. (w/ enclosure)

Charles W. Elliott, Esq. (w/ enclosure)

Mr. Alan J. Nogee (w/ enclosure)

Thmas Y. Au, Esq. (w/ enclosure)

Mr. Thmas Gerusky (w/ enclosure)

Director, Pennsylvania Emergency Managenent Agency (w/ enclosure)

' Mr. Steven P. Hershey (w/ enclosure)

James M. Neill, Esq. (w/ enclosure)

Donald S. Bronstein, Esq. (w/ enclosure)

Mr. Joseph H. hhite, III (w/ enclosure)

Walter W. Cohen, Esq. (w/ enclosure)

Robert J. Sugannan, Esq. (w/ enclosure)

Rodney D. Johnson (w/ enclosure)

Atomic Safety and Licensing Appeal Board (w/ enclosure)

Atanic Safety and Licensing Board Panel (w/ enclosure)

Docket and Service Section (w/ enclosure) q

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. Response to II.K.3.13 Analysis perfonned by the BWR Owners' Group (NEDO-24951) has concluded that changing the initiation setpoint of the HPCI/PCIC is unwarranted *.

'Ihe same NEDO report did recamend a modification to the RCIC circuitry to permit auto-restart of RCIC on low level after a high level trip.

Therefore, modifications to the BCIC circuitry are currently underway to delete the high water level turbine trip and to apply this signal to

- .the auto-close circuit of the steam supply valve. This will provide autmatic operation of the PCIC systen to trip at high water level and auto restart at low water level. This will be impletented prior to fuel load.

• The NBC has accepted this generic report.

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Response to II.K.3.15

'Ihe HPCI/BCIC steam line isolation logic is currently being nodified to aMress the spurious isolation of these systems due to the pressure spike which acumgunies their start-up. The modification consists of adding a time delay to the high flow trip logic of HPCI/BCIC. This will prevent the instantaneous pressure spike from causing a systen isolation. This will be implanented prior to fuel load.

'Ihis design change was submitted by the BWR Owners' Group and has been accepted generically by the NBC.

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Response to II.K.3.16 PECO endorses the BWR Owners' Group generic response to Its II.K.3.16 for Limerick. This response is described in NEDO-24951, "BWR Owners' Group NUREG-0737 Inplanentation: Analysec and Positions Subnitted to the USNRC," June, 1981. The following reccmnendations frm NEDO-24951 will be inplanented at Limerick in order to reduce the challanges to relief valves by approximately an order of magnitude:

(1)- Im Water level Isolation Setpoint (Reference Section 6.3.1.1.1 of NEDO-24951) . The RPV water level isolation setpoint for MSIV closure is being lowered frm level 2 to level 1 as part of the A'IKS modifications for Limerick.

(2) Im-Im Set Relief or Equivalent Manual Actions (Reference Section 6.3.1.3.1 of NEDO-24951) . This recantnendation assures that follcwing the initial pressurization the pressure will be relieved by one valve alone, and the remaining safety relief valve will not experience any subsequent actuation. At Iamerick this will be accamplished manually as described in "BWR Emergency Procedure Guidelines," Revision 1 (prepublication form subnitted January 31, 1981).

(3) Reduce MSIV Testing Frequency (Reference Section 6.3.1.4.4 of NEDO-24951) . A number of isolation events occur when the MSIV closure tests are being conducted. Reducing the MSIV testing frequency would result in a reduction in the number of isolation events. Appropriate reductions will be made to the frequency of testing for the Iamerick MSIV's.

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1 Response to II.K.3.18 The BWR Oaners' Group has subnitted a report in NEDO-24951 to the NRC in which they propose five options to address this concern. Limerick will take steps to make the required modifications to the ADS logic

-when the NRC rules on the acceptability of the proposed options. This Itodification will be impleented during the first refueling cutage, in accordance with NUREG-0737 Implenentation Requirenents.

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Response to II.K.3.21 Philadelphia Electric Cornpany endorses the BhR Owners' Group position l-for Item II.K.3.21 for Lunerick. This position was forwarded to the NBC by letter frun D. B. Walters (BWBOG) to D. G. Eisenhut (NFC) dated December 29, 1980. The conclusion of this position is that autanation i of the restart of the LPCI and CS will result in a net decrease in safety because of the cauplexity of the logic required. Ingic rnodifications to the LPCI and CS systaus are therefore not warranted at Limerick.

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Response to II.K.3.22 bbdifications are nw underway to modify the PCIC systs suction valve logic to autmatically switch suction frm the Condensate Storage Tank to the Suppression Pool on low Condensate Storage Tank level. This will be inplemented prior to fuel load.

Response to II.K.3.24 At Iamerick, the HPCI and RCIC cmpartment unit coolers are powered by onsite mergency power and therefore continue to be available during a loss of offsite power. The unit coolers are described in Section 9.4.2.2.

The mergency service water pumps which provide flow to the coolers are also powered frm onsite mergency power. Adequate space cooling is therefore assured during a loss of offsite power. There are no other supporting systes that require offsite power such that operation of the HPCI and RCIC systems would be impaired should offsite power be lost.

.The current Limerick design is therefore acceptable.

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Response to II.K.3.25 At Limerick, two systes are available for cooling the recirculation punp seals: The reactor enclosure cooling water (RECW) syst s and the recirculation pump seal purge systs.

Recirculation pump vendor test data has shown that if either one of these seal cooling systes is operating, seal tmperatures will remain within acceptable limits and excessive seal deterioration is not expected to occur.

The primary cooling for the recirculation pump seals is provided by the RECW system which cools the reactor water that flows to the lower seal cavity. After a loss of offsite power, the RECW pumps will be powered by onsite mergency power and will restart autmatically. The service water system, which normally provides cooling water to the RECW heat exchangers, will not be available, but cooling water to the heat eachangers can be provided via manual realignment of the Emergency Service Water (ESW) systs. If the RECW pumps do not restart or are unavailable for sme other reason, the ESW can be manunlly routed directly to the recirculation pump seals for cooling by way of the RECW piping.

Backup cooling is provided by the recirculation pump seal purge syst s which injects cool water from the Control Rod Drive (CRD) systs into the lower seal cavity. The CRD pumps are powered frm the emergency diesels and can be manually restarted once onsite power is available. Hence, the CRD pumps provide an alternate method that is available for seal cooling during a loss of offsite power.

Even in the rmote case where neither cooling source is reestablished and gross seal degradation occurs, the General Electric analysis (NEDO-24951) performed under the direction of the BWR Owners' Group and which is applicable to LimerirA has shown that the maximum coolant loss would be limited to 70 gpn per pump. This loss is small enough to be empensated for by nonnal or mergency reacter water level controls.

. Instrumentation for various parameters, including seal cavity pressure, seal staging and drain flows, drywell equipnent drain sump pump flow and drywell floor drain sump pump flow, is available to the operator to indicate potential seal failure. In addition, gross seal failure may lead to changes in drywell pressure, tmperature, or radioactivity, all of which are monitored and recorded in the control rom.

i It is therefore concluded that a total loss of recirculation pump seal cooling is not a probl e at Limerick and modifications are not necessary.

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, Response to II.K.3.30 I

'Ihe respcmse to the NFC small break model concerns was provided at a -

. neeting between the NBC and E on June 18, 1981. Informaticn provided at this meeting showed that, based on the small break test results and sensitivity studies, the existing GE small break IOCA model already satisfies the concerns of NUREG-0626 and is in cmpliance with 10CFR50, Appenduc K. Therefore, the GE model is acceptable relative to the concerns of Item II.K.3.30, and no model changes need be made to satisfy this item.

i' Documentation of the information provided at the June 18, 1981 meeting was provided via letter frm R. H. Buchholz (GE) to D. G. Eisenhut (NBC) dated June 26, 1981.

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,. 1 response to II.K.3.31 The small-break IOCA calculations included in the Iamerick IOCA analysis are discussed in Section 6.3.3.7 and Table 6.3-5. The references listed in Section 6.3.6 describe the currently approved Appendix K nethodology used to perform these calculations. Ccurpliance with 10CFR50.46 has previously been established for that methodology. As stated in the June 26, 1981 letter frm R. H. Buchholz (GE) to D. G. Eisenhut (NBC), no model changes are needed to satisfy NUREn-0737, Item II. K.3.31; therefore, there is no need to revise the calculations presented in Section 6.3.3.7.

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.- l Besponse to II.K.3.45 The Applicant endorses the BhR Owners' Group position on Iten II.K.3.45 for Limerick. This position is presented in NEDO-24951. "BhR Owners' Group NUREG-0737 Implementation: Analysis and Positions Subnitted to the USNRC," June 1981 and is sunmarized below.

An evaluation of alternate axle of depressurization other than full l actuation of the ADS was made by the Bh2 Owners' Group with regard to the effect of such reduced depressurization rates on core cooling and vessel integrity.

Depressurization by full ADS actuation constitutes a depressurization frcm about 1050 psig to 180 psig in approximately 3.3 minutes. The alternate modes of depressurization that were evaluated considered vessel depressurization over the same range (1050 psig to 180 psig) within two different time periods (6-10 minutes and 15-20 minutes) .

'Ihe cases mnsidered show that no appreciable improvenent can be gained by a slower depressurization based on core cooling considerations.

Since a full ADS blowdown is well within the design basis of the reactor pressure vessel and ADS is properly designed to minimize the threat to core moling, no change in the depressurization rate is necessary, and no modifications to Limerick are needed for this 'IMI iten.

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