ML20148N271

Submits Revised Commitment Re Crediting Containment Pressure in ECCS NPSH LOCA Analysis.Will Change Seawater Inlet Temp Administrative Limit from 65 F to 68 F

ML20148N271
Person / Time
Site:	Pilgrim
Issue date:	06/20/1997
From:	Olivier L BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BECO-2.97.063, NUDOCS 9706260304
Download: ML20148N271 (14)

Text

~ _ , . ..

, \

_e. I

' g 10 CFR 50.90

) ,

so,aan esiaan

Pilgrirn Nuclear Power Station Rocky Hill Road j Plyrnouth, Massachusetts 02360

( L J. Olivier June 20, 1997 Vice President Nuclear operations j and Station Director BECo Ltr. 2.97.063 i

i U.S. Nuclear Regulatory Commission i

Attention: Document Control Desk Washington, DC 20555 i

Docket No. 50-293 i License No. DPR-35 k

SUBJECT:

Revised Commitment Conceming Crediting Containment Pressure in ECCS NPSH LOCA Analysis '

References:

1) Boston Edison Letter (BECo letter No. 97.004) to NRC dated January 20,1997, entitled " Request for Review"

2) Boston Edison Letter (BECo letter No. 97.008) to NRC dated January 30,1997, entitled "Significant Hazards Evaluation for Pilgrim Nuclear Power Station's Net Positive Suction Head Analyses"

3) Boston Edison Letter (BECo letter No. 97.023) to NRC dated February 27,1997, entitled " Supplemental Submittal on Pilgrim Station NPSH Analysis"

4) Boston Edison Letter (BECo letter No. 97.042) to NRC dated April 11,1997, entitled

" Revised Request for License Amendment to Credit Containment Pressure in ECCS )

NPSH LOCA Analyses" /

Boston Edison Letter (BECo Letter No. 97.035) dated May 14,1997, " Response to 5)

Request for Additional Information /

By this letter, Boston Edison Company changes a commitment made in our letter dated April 11, 1997, conceming seawater inlet temperature as it relates to emergency core cooling system 8/

(ECCS) net positive suction head (NPSH). We are changing the seawater inlet temperature

. administrative limit from 65*F to 68 F. ,

l Backaround By letter dated April 11,1997, BECo informed the NRC of an evaluation of emergency core cooling system (ECCS) net positive suction head (NPSH) at Pilgrim using the larger ECCS strainers installed during refueling outage 11. This analysis concluded containment pressure was not needed to maintain sufficient NPSH using a 65 F heat sink temperature and current design basis values for debris volume.

9706260304 970620 11lllll 11ll111111lll llIGl Illllisillis 'Iulllll Hillllllllllfulillllll PDR ADOCK 05000293 ' ' ' '

P PDR

_yO-

in the April 11,1997, letter we also committed to enter the loss of containment cooling limiting l condition of operation (LCO) whenever seawaterinlet temperature exceeded 65 F.

This commitment was an interim compensatory measure until the NRC reviewed and approved our January 20,1997, request for a license amendment that would credit containment pressure l and raise the seawater inlet temperature assumed (to 75 F)in Pilgrim's ECCS NPSH analysis. I Our April 11,1997,65 F commitment was predicated on receiving NRC approval of the January 1 20,1997, request prior to Pilgrim's seawater inlet temperature rising to the 65 F limit.

Temperatures greater than 65 F occur sporadically during the summer months generally as a i result of short duration environmental influences (e.g., tidal action, diurnal effects and wind i direction). However, approval of our amendment has not arrived, and inlet temperature is  !

approaching 65 F at a faster rate than originally anticipated.

1 While 65 F is not specifically mentioned in Pilgrim's Technical Specifications, exceeding it under our April 11,1997, ccmmitment requires Pilgrim to be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This shutdown would be required because the NRC has determined an unreviewed safety question (USQ] exists when credit for containment pressure is required in ECCS NPSH analysos.

Resolution of the USQ for Pilgrim through the license amendment process (10CFR50.90) was the l subject of our January 20,1997, letter to the NRC.

New Commitment We are changing our April 17,1997, commitment. An administrative limit will require entering the loss of containment cooling limiting condition for operation whenever seawater inlet temperature l' exceeds 68 F.

Reason for Reauest and Justification for Chanae

. Analysis has demonstrated that Pilgrim can operate safely with a 75 F seawater inlet temperature with credit taken for containment pressure. (See the above referenced correspondence associated with our requested license amendment). Discussions with the NRC concerning our license amendment request indicate approval is imminent but will not likely be received prior to seawater inlet temperature exceeding 65 F.

Analysis also indicates Pilgrim can operate safely with a 68 F seawater inlet temperature without taking credit for containment pressure. Therefore, the containment pressure, which is central to ine USQ, is not an issue at or below 68 F, and plant operation in the USQ regime does not occur at or below 68 F. The 50.59 evaluation documenting this conclusion is attached.

Summary We submitted an amendment request concerning ECCS NPSH that addresses the seawater inlet temperature issue in Reference 1. The commitment change made in this letter will be used until the NRC completes its review and approval of the amendment request during periods when seawater inlet temperature does not exceed 68 F. Station procedures will be revised to allow interim operation using 68 F as the absolute vs. indicated temperature requiring entry into the containment cooling LCO.

2

~

l 4 . -

Should you have further questions concerning this issue, please contact P. M. Kahler at (508)

'830-7939.

• o L. J. Olivier {

l PMK/dmc/ECCS1

Attachment:

Pilgrim 10CFR50.59 Review for Operation with a Seawater Inlet Temperature of 68 F. .

I cc: Mr. Alan B. Wang, Project Manager l

Project Directorate I-3 i Office Of Nuclear Reactor Regulation 1 Mail Stop: OWF 14B2 )

1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region i 475 Allendale Road King of Prussia, PA 19406 Senior Resident inspector Pilgrim Nuclear Power Station 1 l

Mr. Robert Hallisey )

Radiation Control Program Center for Communicable Diseases Mass Dept. of Public Health 305 South Street Jamaica Plain, MA 02130 Mr. Jim Muckerheide 400 Worcester Road j P.O. Box 1496 Framingham, MA 01701 1 Mr. Peter Laporte, Director Mass. Energy Management Agency 400 Worcester Road P.O. Box 1496 Framingham, MA 01701-0313 3

-+-.a  ; -.

S;fity Evrluition No. 3108+ l

^

SAFETY EVALUATION )

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

Calc M-664 RBCCW l Calc M-662 RHR P.D. Harizi NESG Mech. Eng. Calc M-734 Core Spray D:scription of Proposed change, test, or experiment: Interim evaluation of Containment Heat R*moval systems with an Ultimate Heat Sink (UHS) temperature of 68cF by extrapolation of oriainal FSAR desian basis LOCA analysis with a 65 F UHS. i l

SAFEW EVALUATION CONCLUSIONS:

Yes No

1. @ 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 S 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. O E 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 E May the proposed activity create the possibility of an accident of a different type than any previously evaluated in the Final Safety Analysis Report'i

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

BASIS FOR SAFEW EVALUATION CONCLUSIONS:

This interim evaluation demonstrates that the oriainal licensina basis assumptions are conservatively met with the new ECCS suction strainers based on the oriainal L OCA analysis usina a 65'F UHS and extrapolatina to a 68*F UHS and current desian basis values for debris loadina. See Attachment 1.

Safety Performed Evaluation by [AO ')dvuld Date 6-/9-97 y i NOP83ES Pev. 8 Exhibn i Sheet 1 of 4

Snfaty Evtluttinn No.

• Lot

. SAFETY EVALUATION PILGRIM NUCLEAR POWER STATION A. APPROVAL Comments:

1 l

L^/ An w. e 6 Mr7 N/A 4

%ipline Divihon Mgr./Da'te ' Supporting Discipline Division MgrJDate i

B. REVIEW / APPROVAL 0 comments: .

i

/gd j -Ce J L. p.og. es, qg d

[ S&SA DiEsion Mgr./ D5te  % p.

NOTES: 1) Items (14) and (15) are not required for Safety Evaluation prepared by the Hant Department.

2) The independent technical review of Plant Department Safety Evaluations is documented in item C below.

C. ORC REVIEW O This proposed change involves an unreviewed safety question and a request for authorization of this change must be filed with the NRC prior to implementation.

7 This proposed change does not involve an unreviewed safety question.

ORC Chairperson Date d!84!92 ORC Meeting Number 97- N cc:

NOP83E5 Rev. 8 Exhibit 1 Sheet 2 of 4

- - . . - - . - . . - = - . . - . - - . . . - . - - , , .- - -.

< SIfaty Ev1lutti::n

~

No. St 04-D. FSAR Review Sheet

List FSAR text, diagrams, and indices affected by this change and corresponding FSAR revision.

f i' Preliminary revision to the affected FSAR

' Affected FSAR Section Section is shown on:

NONE NOTE:

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 aspart ofthefinal response to NRC Bulletin 96-03 and/or as part ofan updated accident analysis for higher heat sink temperatures.

i l

l l

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

Prepared by: DM Date: 6-/f47

' Approved by: Nk Date: 4-7.-r7

\

FINAL FSAR REVISION - Prepared in accordance with NOP83E4 following operational I

turnover of related systems, structures, or components for use at PNPS. j l

NOP83E5 Rev. 8 Exhibit 1 Sheet 3 of 4

Siftty Evtlu:. tion No. 'A OLf E. SAFETY EVALUATION WORK SHEET '

A. System / Component Failure and Consequence Analyses.

System / Component Failure Modes Effects of Failure Comments 1 RHR and Core Spray LOCA pipe break jet Debris accumulation on Effect of debris was evaluated

impingement torus suction strainers for and the Increased suction head a

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

j See Attachments.

t 2 Containment Heat RBCCW cooling Equipment may not All affected systems were Removal Systems water at higher than achieve rated heat evaluated as summarized in original equipment removal and/or EQ Calculation M-664 for 4

rating temperatures conditions may be suppression pool temperatures up l may occur. exceeded. to 178F and RBCCW

temperatures of 98F.

3 General Reference Material Review FSAR CALCULATIONS REGULATORY GUIDES SECTION PNPS TECH. DESIGN SPECS. PROC. STANDARDS CODES SPECS 4.8.5.1 Section 3.2.H Calculation M-664 Rev. O Reg. Guide 1.82 Rev.1 6.4.3 Section 3.5.A & B Calculation M-734 Rev. O NRC Bulletin 96-03 14.5 Section 3.7.A Calculation M-662 Rev. E2 Section 4.7.A.2 GE Report GE-NE-B13-01805-11 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 Id Date 6-/f- U

]

l l

NOP83E5 Rev. 8 j Exhibit 1 Sheet 4 of 4 )

. 1

, SE- Mof Attachm::nt 1 Paga 1 of 6 Safety Evaluation - Attachment 1 l

A. Description of Chance

)

This Safety Evaluation provides an interim analysis of Containment Heat Removal systems and RHR and Core Spray pump Net Positive Suction Head (NPSH) conditions following a Design Basis Loss of Coolant Accident (DBA-LOCA) with an Ultimate Heat Sink (UHS) temperature of 68 F. This interim evaluation is based on the current design  ;

basis analysis for LOCA-generated debris, new RHR and Core Spray pump suction l strainers, and an extrapolation of the original FSAR DBA-LOCA analysis with a 65 F heat sink.

)

B. Purpose of the Chance .

l 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 l Spray pump suction strainers were replaced with large capacity stacked disk strainers as i part of the response to NRC Bulletin 96-03 [Ref. 2). To support Pilgrim operation at UHS I temperature above 65 F, it is necessary to produce this interim Safety Evaluation that is -

based on the new strainer debris capacity and an extrapolation of the original FSAR i DBA-LOCA analysis based on a 65'F heat sink temperatura. - The postulated LOCA- l generated debris is the current 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 i debris without affecting pump performance using an NPSH margin that is very i conservatively based on zero containment positive pressure following a DBA-LOCA. l l

Proper operation of the Containment Heat Removal systems also requires that the peak Suppression Pool and UHS temperatures following a design basis LOCA be within established limits. This evaluation includes or references analyses of increased UHS ,

temperatures on the Containment Heat Removal systems. l 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:

Containment Heat Removal Systems Residual Heat Removal (RHR) System Core Spray System

2. Indirectly Affected:

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

- - - _ _ .__ _. . _..__.-__.__.__._.-.____m m

, SE- U 0 Attachment 1 Pago 2 of 6

~

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

The following documents are referred to by [Ref. #] in this SE:

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

[2] NRC Bulletin 96-03 " Potential Plugging of Emergency Core Cooling Suction  !

Strainers by Debris in Boiling-Water Reactors".

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

[4] BECo Calculation M-862 Rev. E2 "RHR and Core Spray Pump NPSH and Suction j Pressure Drop". 1

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

l

[6] GE Report GE-NE-B13-01805-11 " Effects of Fiberglass Insulation Debris on Pilgrim ECCS Pump Performance" January 1996, SUDDS/RF # 96-02.

[7] BECo Calculation M-664 Rev. O " Containment Heat Removal".

[8] SE-2983 " Evaluate results of a design verification for safety related cooling systems using the FSAR site maximum SSW temperature of 75F'."  !

[9] BECo Specification E-536 Rev,6 " Environmental Parameters for Use in the Environmental Qualification of Electrical Equipment per 10CFR50.49".

[10] BECo Calculation M-665 Rev. O " Equipment Area Cooler Performance".

[11] BECo Calculation N-127 Rev.1 " Secondary Containment Heatup Analysis".

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  ;

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 removalin the Suppression 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 Suppresolon Pool Both the LPCI and LPCS are required to mitigate the consequences of the vanous postulated LOCA and Steam Line Break (SLB) accidents by providing emergency core cooling and containment cooling via the RHR operating modes.

~

The Containment Heat Removal systems include RHR, RBCCW, and SSW. These systems control the environment within primary and secondary conteinment following the loss of coolant accident. Together, these systems must maintain all safety-related equipment within established operating limits and within the established Environmental Qualification (EQ) limits.

3 ,

SE- 310S Attachm:nt 1 Pags 3 of 6 E. Effect on Functions As a consequence of a LOCA or SLB, the NUKON insulation in the vicinity 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 i

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 j 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.

i The peak Suppression Pool and UHS temperatures following a design basis LOCd are

! the primary considerations for operation of the Containment Heat Removal systems.

} The heat transfer from containment, the RBCCW cooling water temperature, and l resulting environmental conditions in primary and secondary containment are dependent l on the Suppression Pool temperature and the UHS temperature.

F. Analysis of Effect on Functions

The maximum Suppression Pool temperature for which containment positive l pressurization is not required for NPSH analysis is 169 F as shown in Attachment 2.

3 This represents a three degree increase from the original FSAR value of 166'F. Based j on the three degree increase in peak Suppression Pool temperature, the UHS may be j increased by three degrees to 68*F by simple extrapolation.

1 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 j evaluation are based on the current design basis analysis for LOCA-generated debris, new suction strainer debris capacity, and an extrapolation of the original FSAR DBA-l LOCA analysis with a 65 F heat sink as described above.

The calculated 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 1 minimum available margin for LOCA debris for the limiting Core Spray pump is 0.03 ft

{ assuming zero positive pressure following a DBA-LOCA with a peak Supprr ssion Pool temperature of 169 F per Attachment 2. Therefore, there is adequate NPSH margin to

] accommodate the postulated debris loading without affecting pump performance.

The heat transfer achieved by the RHR and RBCCW systems with a peak Suppression 4

Pool temperature of 169 F and UHS of 68 F is equal to the hect transfer assumed in the j original FSAR analysis with a peak Suppression Pool temperature of 166*F and UHS of i 65'F and as confirmed in Calculation M-664 (Ref. 7]

I The current Environmental Qualification (EQ) profiles for equipment are based on containment temperature profiles that are referenced in SE-2983 (Ref. 8] for design

(

4 i

• SE- Mo4 Attachmtnt 1 Pago 4 of 6 1 l

basis accident analysis with a 75'F UHS and are now included in the EQ Specification l ,

)

E-536 [Ref. 9] that is the controlling document for the EQ program. SE-2983 is currently i

^

under review by the NRC and there are issues relating to the decay heat inputs used in  !

the analysis and the use of containment positive pressure for NPSH analysis. However, i this current evaluation is based on a 68'F UHS and does not require consideration of t

containment positive pressure, it can also be concluded that the containment EQ i profiles currently used by Pilgrim are bounding for a 68 F UHS even if a decay heat input

{ were used that is 10% higher than referenced in SE-2983.

- It is also necessary to evaluate the acceptability of the peak' RBCCW cold-side loop temperature for the safety-related equipment, other than the RHR heat exchanger ,

i which is cooled by RBCCW under emergency conditions. These components include the following:

1

1. RHR, HPCI, RCIC area coolers. 1

2. RHR pump seal coolers. 1

3. Core Spray pump thrust bearing cooler.

1 The peak RBCCP We loop temperature evaluated in [Ref. 7] is 98 F when the i Suppression Poolis a. S'F. The RHR, HPCI, and RCIC area coolers are analyzed at various water inlet temperatures in [Ref.10]. The design water inlet temperature for rating area cooler performance is 85 F. Water inlet temperatures up to 100 F are used j'

in the [Ref.10] performance analyses. The area cooler performance at or above the 98'F water inlet condition was used to calculate K-factors that are included in the secondary containment accident thermal analysis [Ref.11]. The effect of RBCCW peak cold-side loop temperature on area cooler performance is thereby included in the j analysis that determines the post-accident area temperatures in secondary containment. .

, i The RHR pump seal coolers are evaluated in [Ref. 7] as fixed heat loads with a maximum 98'F inlet temperature. The temperature rise in the cooler is only 10.1'F for an outlet temperature of 108.1*F. The pump's mechanical seal purging arrangement consists of a tap on the pump discharge from which water passes through a cyclone separator with the clean discharge passing through the seal cooler and then injecting into the seal chamber for flushing and cooling. The seal cooler is a small shell & tube heat exchanger that uses RBCCW water on the shell side to cool the seal purge water.

4 The RHR pump requires a seal cooler because it must operate in the shutdown cooling mode where the water temperature can exceed 300*F. For the safety-related modes of operation (LPCI and ECCS), the water is drawn from the Suppression Pool with a maximum temperature of 178'F. The RHR mechanical sealis a John Crane Model 881 which is a commonly used rotating-spring O-ring pusher design, equipped with Viton O-rings. For this type of seal, cooling is required to hold the seal chamber below the temperature at which water will flash ct the seal faces (212'F). The seal itself can withstand up to 300*F and is thereforo not the limiting factor. The seal will operate satisfactorily at temperatures approa(hing 212"F, therefore the calculated cooling water temperatures are well within the acceptable range. It is concluded that although effective cooling is provided, the accident modes do not require the seal cooling function for proper seal operation. That is the reason the Core Spray pumps, with similar seals (Durametallic Type PTO) do not have seal coolers.

_ __ - . _, ,_. _ , _, - - ~ - _

. SE- MOY Attechment 1 Pcgo 5 of 6

• The Core Spray pump thrust bearing coolers are evaluated in [Ref. 7) as fixed heat loads with a maximum 98'F inlet temperature. The temperature rise in the cooler is only 7.6 F for an outlet temperature of 105.6*F. The thrust bearings are a duplex set of angular contact ball bearing in an oil bath in which the cooling coils are also immersed. Ball bearings with oil lubrication are typically considered to operate in the range of 120 to 160'F with an upper limit of 180 F, therefore the calculated cooling water temperatures are well within the acceptable range. The motor outline drawing specifies a maximum ,

cooling water inlet temperature of 165'F for short periods.

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 l basis, and NPSH margin is very conservatively based on zero containment positive 1 pressure following a DBA-LOCA, there is no unreviewed safety question involved for _i plant operation that remains within the defined limits of a 68'F heat sink.  !

1. - Q: May the proposed activity increase the "vaability of occurrence of an accident previously evaluated in the Fine Safety Analysis Report?

A: No, there are no new accident 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 performance or operation of the Containment Heat Removal systems.

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. The components of the Containment Heat Removal systems remain within their established operating limits.

1

4. Q: May the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the Final Safety 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. j i

I I

i

]

. _ _ __. _ _ ~ _ _ _ _ _ _ . _ . _ - _ _ _ _ _ _ _ __ . _ . _ _ _

, SE- Hof Attachment 1 a

Pcg3 6 of 6

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.

7. Q: Does the proposed activity reduce the margin of safety as defined in the i

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 l increase in suction head loss from the postulated debris accumulation is  ;

within the margin for NPSH available to the ECCS pumps. Since the NPSH -l

available at the pump suction exceeds the NPSH required, the pump will achieve its rated performance. Therefore, there is no effect on ECCS pump performance and no change in the margin of safety as determined by the ,

accident analyses. The bases for the Technical Specification requirements 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 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 as adequate when NPSHA is simply greater than NPSHn (corrected for air Ingestion when appropriate).

The components of the Containment Heat Removal systems, including RHR, RBCCW, and SSW, remain within their established operating limits with a peak Suppression Pool temperature of 169 F with a 68*F UHS.

Considerations regarding Environmental Qualification of equipment are also bounded by the current EQ specification for affected equipment.

SE-3104 Attachment 2 -

Page 1 of 1 Table Sa - NPSH & Max Suction dP (Table 4) (Eq8) (Ref S&6) (Eq9) (Eq 10)

Available Max Suction Suction Suction Margin for dP Torus Spec Elev Head Wetwell Availabl6 LOCA Measured Temp Vapor Volume Head Loss Press Available Required Margin Debris @ IST Tm Press Pvp Vsp hz hst Pc NPSHA NPSHa NPSHu NPSMu Conditions (F) (psia) (ft'/lbm) (ft) (ft) (psig) (ft) (ft) (ft) (ft) (psi)

Core Spray Pumps A & B @ 4400 GPM:

169.00 5.8629 0.016440 12.50 2.38 0.000 31.03 29.00 2.03 0.03 1.82 0.500 32.21 29.00 3.21 1.21 1.82 3.635 39.64 29.00 10.64 8.64 1.82 5.600 44.29 29.00 15.29 13.29 1.82 RHR Pumps A & D @ 5100 GPM:

169.00 5.8629 0.016440 12.50 2.62 0.000 30.79 23.00 7.79 5.79 2.06 0.500 31.97 23.00 8.97 6.97 2.06 -

3.635 39.40 23.00 16.40 14.40 2.06 5.600 44.05 23.00 21.05 19.05 2.06 RHR Pumps B & C @ 5100 GPM:

169.00 5.8629 0.016440 12.50 2.02 0.000 31.39 23.00 8.39 6.39 1.83 0.500 32.57 23.00 9.57 7.57 1.83 3.835 40.00 23.00 17.00 15.00 1.83 5.600 44.65 23.00 21.65 19.65 1.83 Note: [1] Equations and tables referenced here are contained in BECo calcultir.n M662.

[2] This Table is a modified version of Table 5 from BECo calculation M662, Rev. E2 where the torus temperature in column one is changed from 166F to 169F and the vapor pressure and specific volume changed accordingly per the steam tables. The available margin for LOCA debris is reduced from 0.96 feet to 0.03 feet which is greater than the current debris load of 0.01 feet per BECo calculation M734.

[3] Two feet is subtracted from the total available margin to account for Inservice Testing (IST) measurement variations.

The remaining margin is available for LOCA debris.

- _ _ _ _ ___________ ____ _ _________ _________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _