ML20137W233

Notifies NRC That Requested License Amend Is No Longer Needed to Support Plant Restart from Refueling Outage 11. SE & Calculation M-734,encl

ML20137W233
Person / Time
Site:	Pilgrim
Issue date:	04/11/1997
From:	Boulette E BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20137W238	List: ML20137W233 ML20137W290
References
BECO-2.97.042, NUDOCS 9704180061
Download: ML20137W233 (11)

Text

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Pilgnm Nuclear Power Station $

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' E. T. Boulette, PhD Senior Vice President - Nuclear :

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g BECo Ltr. 2.97 042  !

'U.S. Nuclear Regulatory Commission I

Attention: Document Control Desk'  ;

Washington, DC 20555 j

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Docket No. 50-293 l License No. DPR-35 I i

6 Revised Request for License Amendment  !

to Credit Containment Pressure in ECCS NPSH LOCA Analyses i

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- In BECo letter # 2.97.004, a license amendment for Pilgrim Nuclear Power Station was requested to allow credit for containment pressure in net positive suction head (NPSH) analyses for emergency core cooling system (ECCS) pumps.' That request requires the review and approval of our new containment heat removal analysis including the 75 F seawater temperature design change. The

letter also requested NRC review and approval of the license amendment in time to support Pilgrim Station restart from refueling outage #11 (RFO#11).

. However, some NRC questions on the new analysis remain unresolved, and it is unlikely these

questions will be resolved by the Pilgrim RFO#11 restart date. Therefore, as discussed with the i: NRC staff on March 31,1997, we have evaluated NPSH using the larger ECCS pump suction ,

strainers installed during RFO#11 and determined sufficient NPSH is available to meet pump NPSH j requirements based on no containment positive pressure following a DBA-LOCA. A 10CFR50.59 safety evaluation was prepared based on this calculation. The evaluation concludes the original j .

licensing basis assumptions are conservatively met with the larger ECCS suction strainers when /

limited to the original LOCA analysis based on a 65 F heat sink temperature and current design /  !

basis values for debris volume. - Based on the aforementioned calculation and safety evaluation )

(see attachments), restart of PNPS is justified without the need for the requested license }_j amendment.-  !

I Because the calculation and safety evaluation are based on a 65 F seawater inlet temperature, i

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BECo commits to incorporate an administrative limit prior to restart that requires entering the loss of I containment cooling limiting condition for operation (LCO) whenever seawater inlet temperature l exseeds 65'F.:

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. U. S. Nuclear Regulatory Commission Page 2 ,

l Also, please note the NRC deferred Pilgrim's final response to a related issue (strainer debris) in Bulletin 96-03 until the end of 1998 by NRC letter dated March 11,1997 (Mr. Patrick D. Milano (NRC] to Mr. E. Thomas Boulette [BECo]).

In summary, this letter notifies the NRC that the requested license amendment is no longer needed  !

to support restart. However, we still intend to resolve the remaining NRC questions so that the [

requested license amendment in letter #2.97.004 may be granted in time to support power '

operation at'seawaterinlet temperatures above 65 F. Generally, seawater temperatures greater than 65*F occur sporadically at Pilgrim Station during the summer months. '

17 you have any questions regarding this letter, please contact Mr. Jeffrey Keene at (508)830-7876  !

or P. M. Kahler at (508) 830-7939.

h E. T. Boulette, PhD ETB/PMK/avf/npsh2 Attachments: 1) Pilgrim Safety Evaluation #3088

2) Pilgrim Calculation M-734 cc: Mr. Alan B. Wang, Project Manager Project Directorate 1-3 I Office Of Nuclear Reactor Regulation l Mail Stop: OWF 1482 1 White Flint North 11555 Rockville Pike l Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road - )

King of Prussia, PA 19406

.I Senior Resident inspector j Pilgrim Nuclear Power Station l l

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l ATTACHMENT 1 TO BECO LTR. 2.97.042 PILGRIM SAFETY EVALUATION # 3088 l

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p $ OF 2O_. Safety Evtluttirn L~ -

No. M9R SAFETY EVALUATION

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PILGRIM NUCLEAR POWER STATION Document Initiator Dept. Division No. Calc. No. System Name  ;

RHR l P.D. Harizi NESG Mech. Eng. Calc M-734 Core Spray D:scription of Proposed change, test, or experiment: Interim evaluation of ECCS Pumo NPSH with new stacked disk suction strainers usina oriainal FSAR desian basis LOCA

- analysis with a 65'F heat sink.

SAFETY EVALUATION CONCLUSIONS:

Yes No

1. O @ May the proposed activity increase the probability of occurrence of an accident previously evaluated in the Final Safety Analysis Report?

2. O @ May the aroposed activity increase the consequences of an accident previous y evaluated in the Final Safety Analysis Report?

[ 3. O @ May the proposed activity increase the probability of occurrence of a malfunction of ec uipment important to safety previously evaluated in the Final Safety Ana ysis Report?

4. May the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report?

5. O @ May the proposed activity create the possibility of an accident of a different

type than any previously evaluated in the Final Safety Analysis Report?

6. O @ May the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the Final Safety Analysis Report?

7. O @ Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

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BASIS FOR SAFETY EVALUATION CONCLUSIONS:

This interim evaluation demonstrates that the oriainal licensina basis assumptions are conservatively mrt with the new ECCS suction strainers when limited to the oriainal LOCA analysis based on a 65 F h7at sink temperature and current desian basis values for debris loadina. See Attachment 1.

Safety Evaluation Performed by NONmO PD.HARIZi Date N-OB-97

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NOP83E5 Rev. 8 Exhibit 1 Sheet 1 of 4

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.o- SAFETY EVALUATION PILGRIM NUCLEAR POWER STATION A.' APPROVAL ,

Commentsi Alha_

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-Discipline Division dgr./Ddte' A/k Supporting Discipline Division Mgr./Date  :

i B; REVIEW / APPROVAL '

O Comments: _

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- NOTES:

1) ' Items (14) and (15) are not required for Safety Evaluation prepared by the Plant Department.

2) - The independent technical review of Plant Department Safety Evaluations is

documented in item C below.

C. ORC REVIEW

- . 1 0 This proposed change involves an unreviewed safety question and a request for

[ authonzation of this change must be filed with the NRC prior to implementation.

[ This proposed change does not involve an unreviewed safety question.  !

ORC Chairperson AA1~/ Date V//o 497 i o

ORC Meeting Number. (/ 47- SC - j j

cC' i NOP83E5 Rev. 8 Exhibit 1.

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w-. - .- ' L D. FSAR Rev!ew Sheet List FSAR text, diagrams, and indices affected by this change and corresponding FSAR revision.

Preliminary revision to the affected FSAR Affected FSAR Section Section is shown on:

NONE

. NOTE: i This SEprovides an interim evaluation based on limiting conditions.

The assumptions used are conservatively bounded by the current FSAR. .

The FSAR will be appropriately updated as part ofthefinal response ,

to NRC Bulletin 96-03 and/or as part ofan updated accident analysis  ;

for higher heat sink temperatures.

PRELIMINARY FSAR REVISION (to be completed at time of Safety Evaluation preparation).

Prepared by: . b' Date: M-08-97 Approved by: . Date: [f,['T/

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FINAL FSAR REVISION - Prepared in accordance with NOP83E4 following operational turnover of related systems, structures, or components for use at PNPS.  :

4 NOP83E5 Rev. 8 Exhibit 1 Sheet 3 of 4

," l FBI / b d-Ob lP I OF S;faty Evr_lustien L .__ __._ cN g,, gg g,R E. SAFETY EVALUATION WORK SHEET A. System / Component Failure and Consequence Analyses.

System / Component Failure Modes Effects of Failure Comments

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RHR and Core Spray LOCA pipe break jet 4 1 impingement ~

Debris accumulation on torus suction strainers for Effect of debris was evaluated )

s and the increased suction head i 4

destroys insulation RHR and Core Spray loss is within the margin for and transports debris pumps increases suction NPSH available to the ECCS l into torus. head losses. pumps.

See Attachments. ,

2 3

4 1,

General Reference Material Review e

i FSAR CALCULATIONS REGULATORY GUIDES SECTION PNPS TECH. DESIGN SPECS. PROC. STANDARDS CODES SPECS '

4.8.5.1 Section 3.2.H Calculation M-734 Rev. O Reg. Guide 1.82 Rev.1 6.4.3 Section 3.5.A & B Calculation M-662 Rev. E2 NRC Bulletin 96-03 l 14.5 Section 3.7.A GE Report GE-NE-B13-01805-11 Section 4.7.A.2 B. For the proposed hardware change, identify the failure modes that are likely for the components consistent with FSAR assumptions. For each failure mode, show the consequences to the system, structures, or related components. Especially show how the failure (s) affects the assigned safety basis (FSAR text for each system) or plant safety functions (FSAR Chapter 14 and Appendix G.)

Prepared by AD hd 'PD.HARl2l 9 Date 04-0 8-77 NOP83E5 Rev. 8 Exhibit 1 Sheet 4 of 4

,' . hFa t fd'M -6 _ SE-3088 Attachment 1 fn.._..

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._.~.._-e Page 1 of 4 Safety Evaluation - Attachment 1 A. Description of Chanae This Safety Evaluation provides an interim analysis of RHR and Core Spray pump i Net Positive Suction Head (NPSH) conditions following a Design Basis Loss of Coolant l Accident (DBA-LOCA). This interim evaluation is based on the current design basis )

analysis for LOCA-generated deois, new RHR and Core Spray pump suction strainers, j and the original FSAR DBA-LOCA analysis with a 65 F heat sink.

i B. Purpose of the Chanae I Replacement of the original drywell piping insulation in 1984 and the potential effect on ECCS pump NPSH was evaluated in SE-2971 [Ref.1). In RFO-11, the RHR and Core Spray pump suction strainers were replaced with large capacity stacked disk strainers as part of the response to NRC Bulletin 96-03 [Ref. 2). To support the Pilgrim restart from RFO-11, it is necessary to produce this interim Safety Evaluation that is based on the new strainer debris capacity and the original FSAR DBA-LOCA analysis based on a 65 F heat sink temperature. The postulated LOCA-generated debris is the c.urrent Pilgrim design basis value from an analysis performed in accordance with Regulatory Guide 1.82 Rev.1 [Ref. 3]. With these conditions, it is demonstrated that there is adequate NPSH margin to accommodate the postulated debris without affecting pump performance using an NPSH margin that is very conservatively based on zero containment positive pressure following a DBA-LOCA. This evaluation will remain applicable until the Pilgrim design basis analysis is upgraded in accordance with Regulatory Guide 1.82 Rev. 2 as part of the final resolution for NRC Bulletin 96-03 and/or is superseded by an updated accident analysis for higher heat sink temperatures.

C. Systems. Subsystems. Components Affected

1. Directly Affected:

Residual Heat Removal (RHR) System Core Spray System

2. Indirectly Affected:

Reactor Building Closed Cooling Water (RBCCW) System Salt Service Water (SSW) System

3. List drawings, FSAR, Tech. Spec., other documents:

The follovdng documents are referred to by (Ref. #) in this SE:

[1] SE-2971 " Replace all piping thermalinsulation in the drywell with Owens-Corning NUKON fiberglass blanket insulation", 25-MAR-96.

[2] NRC Bulletin 90-03 " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors".

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[3] Regulatory Guide 1.82 Rev.1, " Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident", U.S. Nuclear Regulatory Commission, November,1985.

[4] BECo Calculation M-662 Rev. E2 "RHR and Core Spray Pump NPSH and Suction Pressure Drop".

[5] BECo Calculation M-734 Rev. O "RHR and Core Spray Pump Suction Strainer Debris Head Loss NPSH Evaluation".

[6] GE Report GE-NE-B13-01805-11 " Effects of Fiberglass insulation Debris on  ;

. Pilgrim ECCS Pump Performance" January 1996, SUDDS/RF # 96-02.

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-[7] FSAR Section 14.5.3 " Loss of Coolant Accident".

D. Functions of Affected Systems / Components

-The Residual Heat Removal (RHR) and Core Spray (CS) Pumps are part of the Core )

Standby Cooling Systems (CSCS)(FSAR Section 6). The RHR Pumps provide low l pressure coolant injection (LPCI) to the reactor after depressurization either due to a Loss of Coolant Accident (LOCA) or by operation of the Automatic Depressurization System (ADS). The RHR Pumps also provide for decay heat removal in the Suppression i Pool Cooling and Containment Spray modes of operation (FSAR Section 4.8). The CS Pumps provide low pressure core spray (LPCS) flow to the vessel in a continuous recirculation mode from the suppression pool. Both the LPCI and LPCS are required to mitigate the consequences of the various postulated LOCA and Steam Line Break (SLB) accidents by providing emergency core cooling and containment cooling via the RHR operating modes.

E. Effect on Functions As a consequence of a LOCA or SLB, the NUKON insulation in the vicinlty of the break may be damaged or destroyed by the jet impingement forces. The fiberglass debris generated by the line break may then be transported from the drywell into the suppression pool. Insulation shreds and fibers in various forms may continue to transport through the suppression pool water and ultimately some portion may accumulate on the suction strainers of the operating ECCS pumps. The accumulated debris on the strainers would increase the head loss of the strainer and thereby decrease the Net Positive Suction Head (NPSH) available to the ECCS pumps. If a sufficient amount of debris accumulates on the strainer, the margin for NPSH available to the pump may be exceeded resulting in cavitation, reduced performance, and potential -

damage to the pump.

F. Analysis of Effect on Functions The effect of LOCA-generated debris on the NPSH available to the RHR and Core Spray pumps is evaluated in Calculation M-734 [Ref. 5). The assumptions used in this interim evaluation are based on the current design basis analysis for LOCA-generated debris, new suction strainer debris capacity, and the original FSAR DBA-LOCA analysis with a 05'F heat sink as described above in Section B.

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' The c alculated total volume of LOCA-generated debris from [Ref. 6]is 23 ft'. Applying the entire volume to one suction strainer with 2 RHR and 1 Core Spray pump operating at maximum flow, the head loss due solely to the debris is less than 0.01 ft. The minimum available margin for LOCA debris for the limiting Core Spray pump is greater than 2 feet based on only the containment pressure available prior to the accident (0.5 psig) and is equal to 0.9 feet assuming zero pusitive pressure following a DBA-LOCA [Ref. 4]. Therefore, there is adequate NPSH margin to accommodate the postulated debris loading without affecting pump performance.

The total margin for NPSH available as described in FSAR Section 14.5.3.1.3 greatly exceeds the value of 2 feet at the peak suppression pool temperaturo when the equilibrium pressure in containment is included in the NPSH calculation. This SE and Calculation M-734 [Ref. 5] do not include the contribution to the available NPSH margin from the equilibrium conditions that exist for the containment atmosphere. The FSAR method for evaluating NPSH is consistent with the original design basis calculations for NPSH margin as described in SE-2971 [Ref.1]. Therefore, for this interim evaluation, the limiting assumptions used are conservatively bounded by the original design basis and the FSAR.

G. Summary Since this evaluation is based on a DBA-LOCA analysis that is from the original FSAR, together with an evaluation of LOCA-generated debris that comprises the current design basis, and NPSH margin is very conservatively based on zero containment positive pressure following a DBA-LOCA, there is no unreviewed safety question involved for plant operation that remains within the defined limits of a 65 F heat sink.

1. Q: May the proposed activity increase the probability of occurrence of an ]

accident previously evaluated in the Final Safety Analysis Report?

A: No, there are no new axident initiators or changes to the existing assumptions for the probability of any event considered in the FSAR. '

2. Q: May the proposed activity increase the consequences of an accident  !

previously evaluated in the Final Safety Analysis Report?

A: No, there is no change to the consequences for postulated accidents since there is no change to the assumed RHR and Core Spray pump l performance.

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3. Q: May the proposed activity increase the probability of occurrence of a )

malfunction of equipment important to safety previously evaluated in the  ;

Final Safety Analysis Report?

A: No, there is adequate NPSH margin to accommodate the postulated debris without affecting pump performance.

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4. Q: May the proposed activity increase the consequences of a malfunction of ,

equipment important to safety previously evaluated in the Final Safety i Analysis Report?

A: No, there is no change in the equipment failure assumptions for the accident analysis. ,

5. Q: May the proposed activity create the possibility of an accident of a different type than any previously evaluated in the Final Safety Analysis Report?  ;

A: No, there is no changes or effect upon the events considered in the FSAR  !

accident analyses. -

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6. Q: May the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the Final Safety Analysis Report?

A: No, there is no change in the way that equipment failures are considered for accident analyses.

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7. Q: Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?

A: No, the potential effect from insulation debris accumulating on ECCS pump suction strainers has been evaluated [Ref. 5]. The conclusion is that the increase in suction head loss from the postulated debris accumulation is l within the margin for NPSH available to the ECCS pumps. Since the NPSH available at the pump suction exceeds the NPSH required, the pump will achieve its rated performance. Therefore, there is no effect on ECCS pump l' performance and no change in the margin of safety as determined by the accident analyses. The bases for the Technical Specification requirements 1 regarding Core Spray, LPCI, and Containment Cooling (Sections 3.5.A & B) do not prescribe NPSH criteria per se but it is an implicit assumption for the i pump performance criteria that adequate NPSH be provided. There is no requirement that a specific amount of excess NPSH margin be available after all postulated degradations have been included in the analysis. Furthermore,

[Ref. 3) explicitly defines a design c.s adequate when NPSHA SSlmp!y l greater than NPSHa (corrected for air ingestion when appropriate). ]

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