LR-N970092, Provides Addl Info Re Proposed Resolution Approach for NRC Bulletin 96-003, Potential Plugging of ECCS Suction Strainers by Debris

Provides Addl Info Re Proposed Resolution Approach for NRC Bulletin 96-003, Potential Plugging of ECCS Suction Strainers by Debris

ML20148C852
Person / Time
Site:	Hope Creek
Issue date:	05/20/1997
From:	Eric Simpson Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
IEB-96-003, IEB-96-3, LR-N970092, NUDOCS 9705290189
Download: ML20148C852 (10)

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. s Public Service Doctric and Gas Company E. C.Simpson Public Service Electric and Gas Company P.O. Box 236. Hancocks Bridge, NJ 08038 609-339-1700 Sere Vee Freedent Nuctear Ergreng MAY 2 0 ES7 LR-N970092 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen

PROPOSED RESOLUTION APPROACH - NRC BULLETIN 96-03 i POTENTIAL PLUGGING OF ECCS SUCTION STRAINERS BY DEBRIS l HOPE CREEK GENERATING STATION 1 FACILITY OPERATING LICENSE NPF-57 l DOCKET NO. 50-354 l l

Public Service Electric & Gas Company (PSE&G) provided an interim response to NRC Bulletin 96-03 by letter dated November 4, 1996. ,1 This letter provides additional information regarding PSE&G's  !

proposed resolution to the issues raised in the Bulletin.

PSE&G plans to install large capacity passive strainers on the j Core Spray and Residual Heat Removal (RHR) system pump suction 1 lines to ensure long term operability of the emergency core cooling systems (ECCS). PSE&G is using the Boiling Water Reactor Owners Group (BWROG) Utility Resolution Guidance (URG) as the basis for sizing passive ECCS suction strainers.

I PSE&G had anticipated the NRC would approve the URG before completing modification scoping for installation of the new strainers during the Hope Creek fall 1997 refueling outage.

Without an NRC approved technical basis for our resolution approach, procurement and installation of the replacement strainers represents a significant financial risk. Therefore, in response to NRC Staff concerns regarding the resources required to review all possible combinations of options under the URG, PSE&G is providing the attached information to enable the NRC to g j perform a plant-specific review of the proposed resolution for i Hope Creek. g')3 i Due to the lead time required to obtain the replacement strainers, PSE&G is proceeding with detailed strainer design and i procurement. PSE&G is requesting NRC review and approval of the l proposed resolution approach by July 1, 1997.

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. . MAY 2 01997 Document Control Desk LR-N970092 PSE&G proposes to meet with the NRC staff at its earliest convenience to review the Hope Creek resolution approach.

Should there be any questions concerning this submittal, please do not hesitate to contact us.

Sincerely, l

Attachment I Affidavit l I

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l 95-4933 I

e gg 201997 Document Control Desk LR-N970092 C Mr. H. J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. D. Jaffe, Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. R. Summers USNRC Senior Resident Inspector (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625 I

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95-4933

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i MAY 2 01997 Document Control Desk LR-N970092 PRD/ i i

BC Senior Vice President - Nuclear Operations (XO4)

Senior Vice President - Nuclear Engineering (N19)

General Manager - Hope Creek Operations (H07)

Director - Quality Assurance / Nuclear Safety Review (X01)

Director - Plant-Engineering & Proj ects (N25)

Director - Hope Creek and Component Engineering (X07)

Operations Manager - Hope Creek (H01)

Manager - System Engineering - HC (H18) l ~HC Design Engineering Manager (N51)

Plant Engineering Manager (N32)

Manager - Nuclear Safety Review (N38)

Manager - Business & Co-Owners Affairs (N18)

Onsite Safety Review Engineer - Hope Creek (Hil)

Manager - Salem Licensing (X09)

Manager - Hope Creek Licensing (XO9)

Supervisor - Operations Assessment (X09)

Senior VP and General Counsel, E. Selover. (Newark, SA)  !

Perry Robinson, Esq.

Records Management (N21)

Microfilm Copy File Nos. 1.2.1 3.2 (NRC Bulletin 96-03) i i

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REF: LR-N970092 l.

L STATE OF NEW JERSEY )

) SS.  !

COUNTY OF SALEM )

i. E. C. Simpson, being duly sworn according to law deposes and says: l I am Senior Vice President - Nuclear Engineering of Public' '

Service Electric and Gas Company, and as such, I find the matters  ;

set forth in the above referenced letter, concerning the Hope j Creek Generating Station, are true to the best of my knowledge, l information and belief. i

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Subscribed and Sworn o before me this day of //)[Uf , 1996 I I 2 lbAlu Y sv c) Notary Publih ofh Jersey l KIMBERLY JO BROWN l NOT ARY PUBLIC 0F NEW JERSEY My Comm'ission expires on My Commission bpires AD'il 21,1998 1;

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I ATTACHMENT I l

l IIOPE CREEK GENERATING STATION ECCS STRAINER DESIGN

Background

i To ensure long term operability of the emergency core cooling l systems (ECCS), PSE&G plans to install large capacity passive l strainers on the Core Spray and Residual Heat Removal (RHR) system pump suction lines during the fall 1997 outage. l Modifications are not required for the current High Pressure I Coolant Injection (HPCI) system strainers. The methods described in the Boiling Water Reactor Owners Group (BWROG) Utility Resolution Guidance (URG) serve as the basis for determining the size of the new strainers.

ECCS Pump NPSH Hope Creek is a BWR 4 with a Mark I containment. The present ECCS suction strainers are a stacked disk design with a single strainer for each suction line. The RHR strainers are 32 inches in diameter and 22 inches long. The Core Spray strainers are 23 inches in diameter and 16 inches long. The strainers are sized to screen out particles greater than 1/8 inch in diameter.

The current design basis for ECCS pump net positive suction head (NPSH) assumes that the pump suction strainers are 50 percent plugged. Suppression pool temperature assumed for NPSH is 212 F, the maximum expected average temperature following a LOCA with a loss of offsite ptwer. No credit is taken for containment pressurization in accordance with Regulatory Guide 1.1.

The' current calculated accident flow rates, NPSH required and NPSH available are as follows: I Flow Rate Required NPSH Available NPSH (50% plugging)

RHR 10,500 gpm 4.5 feet 6.0 feet Core Spray 4,015 gpm 10.0 feet 11.2 feet The new strainers will be designed for the same flow rates and maximum suppression pool temperature as the current strainers. No credit will be taken for containment pressurization.

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i ATTACHMENT LR-N970092 Plant Specific Method for Siring New Strainers '

Pipe Break Locations Drywell insulation debris sources and quantities available for )

transport to the suppression pool are based on postulated pipe break locations described in the current Hope Creek licensing l basis. As noted in the Hope Creek Updated Final Safety Analysis

! Report (UFSAR) section 3.6.2, postulated pipe rupture locations inside the containment are based upon Branch Technical Position l MEB 3-1, " Postulated Rupture Locations in Fluid System Piping l Inside and outside Containment," with clarifications as discussed l in UFSAR section 3.6.2.7. Among the approximately 120 pipe break l locations postulated inside the drywell, the evaluation shows that the recirculation pipe break at the RPV nozzle creates the largest volume of insulation debris.

l PSE&G used Method 3, " Break Spec 3fic Analysis Using Break Dependent Zones of Influence," described in URG Section l 3.2.1.2.3.3 to determine the zones of influence (ZOI). The evaluation conservatively treated all break jets as double jets ]

(i.e., the break is fed from both sides and a spherical ZOI is l established). In addition, all breaks were treated as )

l unrestrained and fully offset. l I

1 Insulation Debris Generation and Transport The drywell contains approximately 4500 ft3 of fiberglass wool pad insulation (NUKON or equal) in stainless steel jacketing. In

, evaluating postulated break locations as potential sources of l insulation debris, PSE&G used the transport factors recommended in URG section 3.2.3.2.5. Specifically, 28 percent was used for

! debris generated above the lowest level of drywell grating; 78 l

percent was used for debris generated below that elevation. The I worst-case break was determined to produce about 1402 ft3 of damaged insulation with approximately 393 ft3 transported to the suppression pool.

i The reactor pressure vessel insulation behind the biological shield wall is a stainless steel reflective metallic insulation (RMI). Debris from behind the biological shield wall would be limited because the bottom of the reactor pedestal is closed.

The paths available for debris transport (up and over the biological shield wall or through the biological shield door openings) would limit the quantity of debris released to the drywell.

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o ATTACHMENT LR-N970092 Other Drywell Debris The evaluation accounted for other particulate materials as potential debris sources in accordance with URG recommendations.

The following values were used in the evaluation:

Debris Type Quantity URG Section Dirt / Dust 150 lbm 3.2.2.2.1 Rust 50 lbm 3.2.2.2.2 Paint Chips 85 lbm 3.2.2.2.1.1

The evaluation assumed that all the non-insulation debris is l

transported to the suppression pool.

Suppression Pool Debris Corrosion products (sludge) in the suppression pool combine with

( LOCA-generated debris from the drywell as the principal contributors to ECCS suction strainer blockage. The evaluation

! currently assumes 300 lbm. of sludge present in the suppression pool based on the URG sludge generation rate of 150 lbm. per year

for two years. PSE&G believes the assumed rate of sludge generation is conservative based upon Hope Creek operating experience. Accessible areas of the torus were inspected and sections with appreciable accumulation were cleaned during Hope Creek's fifth refueling outage. The inspection showed that the amount of foreign material transported to the torus since initial plant operation was minimal. The assumed quantity of sludge in the suppression pool may be re-evaluated after the final strainer design is selected.

l PSE&G has procedures in place for torus inspection and cleaning.

l These procedures will be revised if required to ensure the assumed level of suppression pool sludge is not exceeded.

1 No other fibrous debris in the suppression pool is assumed. Any other fibrous debris is expected to have an insignificant impact due to the large amount of fibrous insulation already considered.

This is consistent with the results of previous samples which indicated a very low number of small fibers in the suppression pool.

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ATTACHMENT LR-N970092 ECCS IDCA Analysis Over the spectrum of postulated break sizes and locations, the most limiting large break occurs in the recirculation pump suction. For both large and small breaks, the limiting single failure is failure of the channel A DC source which leaves two l Core Spray pumps and three RHR pumps available.

l l For long term (10 minutes after initiation signal) cooling, in 7

addition.to the reactor coolant pressure boundary piping failure l that initiated the loss of coolant event, the ECCS can sustain l one failure, either active or passive, and still have at least one low pressure ECCS injection loop (LPCI or Core Spray)

operating for vessel makeup, and 100 percent SACS water flow to l the Residual Heat Removal (RHR) heat exchanger operating for heat  !

l removal.

I Strainer Design The new strainers will have sufficient capacity to ensure adequate NPSH with three ECCS pumps running (two Core Spray and j

one RHR). This combination produces the bounding debris loading that would result from either a single active or passive failure during the long term primary containment cooling mode of recovery from a LOCA.

The new strainers will be capable of withstanding the loads 1 l imposed by missiles, debris accumulation, seismic events and LOCA-induced hydrodynamic loads. Analysis of hydrodynamic loads will conform to the current licensing basis.

The strainer vendor will determine the head loss correlation to )'

be used and the basis for its applicability to the Hope Creek .

design. If necessary, this information will be supplied to the

NRC after PSE&G has selected a strainer vendor.

No credit for suppression pool settling will be taken. The debris loading on each strainer will be determined by assuming the debris in the suppression pool is accumulated on the ,

operating strainers in proportion to the flow through each I strainer.

l PSE&G does not plan to install new strainers on the high pressure coolant injection (HPCI) system suction line. The HPCI system l functions to maintain reactor vessel inventory after small breaks that do not depressurize the reactor vessel. These small breaks would have a very small ZOI and would not generate sufficient Page 4 of 5

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, o ATTACHMENT LR-N970092 debris to threaten HPCI pump NPSH. The primary suction supply for the HPCI pump is the condensate storage tank (CST). In addition, the HPCI system does not function to provide long term cooling after a LOCA.

Surveillance Requirements Since the proposed resolution approach uses only passive .

structures and components, PSE&G does not plan to propose Technical Specification surveillance requirements for the i strainers. No surveillances are currently required for the ECCS pump suction strainers in the Hope Creek Technical Specifications. In addition, the improved Standard Technical Specifications for General Electric BWR 4 plants (NUREG-1433, Rev. 1) do not include any surveillance requirements for ECCS l pump suction strainers.

As described in PSE&G's response to NRC Bulletin 95-02, suppression pool cleaning and inspection are controlled by procedure in a manner similar to other preventive maintenance activities.

l PSE&G is evaluating the need for a change to Technical -1 Specification 3.5.3 (Suppression Chamber) to account for the volume of water displaced by the replacement strainers. This l evaluation will be completed after the final strainer design is

! determined. If a change is required, PSE&G will make application l for a license amendment in accordance with the requirements of 10 CFR 50.90.

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