ML072841247

License Amendments Revision to Licensing Basis - Net Positive Suction Head Methodology for Emergency Core Cooling System Pumps

ML072841247
Person / Time
Site:	Salem
Issue date:	11/15/2007
From:	Richard Ennis NRC/NRR/ADRO/DORL/LPLI-2
To:	Levis W Public Service Enterprise Group
Ennis R, NRR/DORL, 415-1420
References
TAC MD6353, TAC MD6354
Download: ML072841247 (15)

Text

November 15, 2007 Mr. William Levis President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: REVISION TO LICENSING BASIS - NET POSITIVE SUCTION HEAD METHODOLOGY FOR EMERGENCY CORE COOLING SYSTEM PUMPS (TAC NOS. MD6353 AND MD6354)

Dear Mr. Levis:

The Commission has issued the enclosed Amendment Nos. 285 and 268 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station (Salem),

Unit Nos. 1 and 2, in response to your application dated August 15, 2007, as supplemented by letter dated September 6, 2007.

The amendments revise the licensing basis, as described in Appendix 3A of the Salem Updated Final Safety Analysis Report, regarding the method of calculating the net positive suction head available for the emergency core cooling system and containment heat removal system pumps.

These changes to the Salem licensing basis relate to issues associated with Generic Letter 2004-02, APotential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.@

A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/ra/

Richard B. Ennis, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosures:

1. Amendment No. 285 to License No. DPR-70

2. Amendment No. 268 to License No. DPR-75

3. Safety Evaluation

cc w/encls: See next page

ML072841247 OFFICE LPL1-2/PM LPL1-2/LA SCVB/BC SSIB/BC OGC LPL1-2/BC NAME REnnis ABaxter RDennig MScott SUttal HChernoff DATE 10/12/07 10/17/07 10/23/07 10/25/07 11/13/07 11/15/07 Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc:

Mr. Thomas Joyce Township Clerk Senior Vice President - Operations Lower Alloways Creek Township PSEG Nuclear Municipal Building, P.O. Box 157 P.O. Box 236 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Mr. Paul Bauldauf, P.E., Asst. Director Mr. Dennis Winchester Radiation Protection Programs Vice President - Nuclear Assessment NJ Department of Environmental PSEG Nuclear Protection and Energy P.O. Box 236 CN 415 Hancocks Bridge, NJ 08038 Trenton, NJ 08625-0415 Mr. Robert Braun Mr. Brian Beam Site Vice President - Salem Board of Public Utilities PSEG Nuclear 2 Gateway Center, Tenth Floor P.O. Box 236 Newark, NJ 07102 Hancocks Bridge, NJ 08038 Regional Administrator, Region I Mr. Carl Fricker U.S. Nuclear Regulatory Commission Vice President - Operations Support 475 Allendale Road PSEG Nuclear King of Prussia, PA 19406 P.O. Box 236 Hancocks Bridge, NJ 08038 Senior Resident Inspector Salem Nuclear Generating Station Mr. George Gellrich U.S. Nuclear Regulatory Commission Plant Manager - Salem Drawer 0509 PSEG Nuclear Hancocks Bridge, NJ 08038 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. James Mallon Manager - Licensing PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Steven Mannon Manager - Salem Regulatory Assurance PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038

PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 285 License No. DPR-70

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated August 15, 2007, as supplemented by letter dated September 6, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR), Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, this license amendment authorizes changes to Appendix 3A of the Updated Final Safety Analysis Report as shown in Attachment 2 to the licensee's letter dated September 6, 2007.

3. This license amendment is effective as of its date of issuance and shall be implemented by December 31, 2007.

FOR THE NUCLEAR REGULATORY COMMISSION

/ra/

Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: November 15, 2007

PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 268 License No. DPR-75

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated August 15, 2007, as supplemented by letter dated September 6, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR), Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, this license amendment authorizes changes to Appendix 3A of the Updated Final Safety Analysis Report as shown in Attachment 2 to the licensee's letter dated September 6, 2007.

3. This license amendment is effective as of its date of issuance and shall be implemented by December 31, 2007.

FOR THE NUCLEAR REGULATORY COMMISSION

/ra/

Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: November 15, 2007

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 285 AND 268 TO FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311

1.0 INTRODUCTION

By letter dated August 15, 2007, as supplemented by letter dated September 6, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession Nos.

ML072350209 and ML072610603, respectively), PSEG Nuclear LLC (PSEG or the licensee) submitted a license amendment request for Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. The proposed amendment would revise the licensing basis, as described in Appendix 3A of the Salem Updated Final Safety Analysis Report (UFSAR), regarding the method of calculating the net positive suction head (NPSH) available for the emergency core cooling system (ECCS) and containment heat removal system pumps.

The available NPSH (NPSHa) is a measure of the specific energy, or head, of the liquid available at the centerline of the pump suction flange. It indicates the capability of the flow into the pump to overcome losses within the pump and avoid pump cavitation.

The required NPSH (which is determined by test for a given pump) is the head at the pump suction flange that produces a given (acceptable) amount of cavitation (measured as reduction in pump discharge head). For satisfactory operation of a centrifugal pump, NPSHa should exceed or equal the required NPSH. The difference between the available and required NPSH of a pump is referred to as the NPSH margin.

NPSHa can be determined from the following equation:

NPSHa = hatm + hstatic - hloss - hvapor where:

hatm = head on the liquid surface. In a containment, hatm would be the head due to the pressure of the containment atmosphere.

hstatic = head due to the height of the liquid surface above the pump suction centerline.

hloss = head loss due to fluid friction and fittings (valves, bends in piping, etc.) in the flow path to the pump.

hvapor = head due to the vapor pressure of the liquid at the temperature of the liquid.

Appendix 3A of the Salem UFSAR describes the licensee positions regarding applicable Nuclear Regulatory Commission (NRC or Commission) Regulatory Guides (RGs). The proposed amendment would revise the portion of UFSAR Appendix 3A pertaining to RG 1.1 (Safety Guide 1), ANet Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps,@ dated December 1, 1970 (ADAMS Accession No. ML003739925).

Currently, the method described in UFSAR Appendix 3A for calculating the NPSHa assumes that the containment pressure is equal to the vapor pressure of the sump fluid, thus excluding the initial containment air pressure. The proposed amendment would change the methodology described in the UFSAR to credit the pre-accident partial air pressure of the containment atmosphere in determining the NPSHa and would include the static head of water above the containment floor in the NPSHa determination.

PSEGs application dated August 15, 2007, indicated that the change in methodology is needed as a result of ECCS containment sump strainer modifications and associated analysis performed by the licensee to address the issues in NRC Generic Letter (GL) 2004-02, APotential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors,@ dated September 13, 2004 (ADAMS Accession No. ML042360586).

During the injection phase of a loss-of-coolant accident (LOCA) the ECCS and containment heat removal system pumps take suction from the refueling water storage tank (RWST) and credit for containment pressure is not an issue. During the recirculation phase, only the residual heat removal (RHR) pumps take suction from the containment sump. Because of the possibility of increased head loss at the sump screens due to debris and chemical effects (increasing hloss) and the high temperature of the sump water (increasing hvapor), it is important to ensure adequate NPSHa is maintained.

The supplement dated September 6, 2007, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff=s original proposed no significant hazards consideration determination as published in the Federal Register on September 11, 2007 (72 FR 51866).

2.0 REGULATORY EVALUATION

General Design Criterion (GDC) 35, Emergency core cooling, in Appendix A of Title 10 of the Code of Federal Regulations (10 CFR) requires that the ECCS provide abundant core cooling so that fuel and clad damage that could interfere with continued effective core cooling is prevented, and so that clad metal-water action is limited to negligible amounts. Adequate NPSH margin is necessary to ensure abundant core cooling.

GDC 38, Containment heat removal, requires that a system to remove heat from the reactor containment be provided whose safety function is to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following a

LOCA. Adequate NPSH margin is necessary to ensure that the containment spray pumps can cool the post-LOCA containment atmosphere.

RG 1.1 provides the following guidance on determining NPSHa:

Emergency core cooling and containment heat removal systems should be designed so that adequate net positive suction head (NPSH) is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated loss of coolant accidents.

NUREG-0800, Standard Review Plan (SRP) Section 6.2.2, Containment Heat Removal Systems, Revision 5, dated March 2007 (ADAMS Accession No. ML070160661), states that, with respect to GDC 38, the licensees analysis should demonstrate that NPSHa is greater than or equal to the required NPSH. The SRP also cites RG 1.82, Revision 3 as describing methods acceptable to the NRC staff for evaluating the NPSH margin.

Regulatory Position 1.3.1.1 of RG 1.82, Water Sources for Long-Term Recirculation Cooling Following a Loss-Of-Coolant Accident, Revision 3, dated November 2003 (ADAMS Accession No. ML033140347), states, in part, that NPSHa for ECCS and containment heat removal system pumps should be determined assuming the maximum expected temperature of pumped fluid and no increase in containment pressure from that present prior to the postulated LOCA.

These considerations are consistent with RG 1.1.

Regulatory Position 1.3.1.1 of RG 1.82 also provides guidance with respect to one method to ensure that credit is not taken for containment pressurization during the transient (i.e., LOCA).

This method is to assume that the containment pressure equals the vapor pressure of the sump water (i.e., hatm equals hvapor in the above equation). This method is consistent with the current Salem licensing basis as described in UFSAR Appendix 3A.

Regulatory Position 1.3.1.2 of RG 1.82 states that for certain operating pressurized water reactors (PWRs) for which the design cannot be practicably altered, conformance with Regulatory Position 1.3.1.1 may not be possible. For these cases, credit for containment accident pressure may be allowed. Since the licensee is requesting credit for only the containment pressure present prior to the LOCA, the licensee is not proposing to credit containment accident pressure.

Nuclear Energy Institute (NEI) report NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology dated May 28, 2004 (ADAMS Accession No. ML041550661), including a revision dated July 13, 2004 (ADAMS Accession No. ML042030528), contain a discussion regarding the role of containment pressure in determining NPSHa. Specifically, Section 6.4.7.1, NPSH Available, states, in part, that:

Typically for PWRs that do not credit containment overpressure in the design basis analyses, the basic assumption is to conservatively assume that containment pressure equals the vapor pressure of the liquid in the sump. In reality, this assumes that there is no air partial pressure in containment prior to the event, or that the air pressure is non-mechanistically lost during the event. A more realistic assumption is that at the time of safety injection recirculation the containment partial steam pressure is equal to sump fluid vapor pressure plus an air partial pressure equal to the containment air pressure

prior to the event. The air pressure prior to the event is to be calculated assuming 100%

relative humidity at a containment temperature corresponding to the maximum normal temperature experienced at the plant. Alternatively, the pre-event minimum containment pressure minus the partial steam pressure at the dew point temperature for the cooling water temperature can be assumed for the air pressure. The recognition of the pre-event air pressure acknowledges the thermal-hydraulic condition of containment prior to the event without crediting containment overpressure based on the accident scenario.

The NRC staff safety evaluation (SE) on NEI 04-07 dated December 6, 2004 (ADAMS Accession No. ML043280007), found the discussion of NPSH with respect to containment pressure to be acceptable.

For both Salem units, the RHR pumps are the only ECCS pumps taking suction from the containment sump during recirculation. The Salem UFSAR (Section 6.2.2.1.2) states that recirculation of water from the containment sump is provided by the diversion of a portion of the recirculation flow from the discharge of the RHR heat exchangers to the containment spray header after injection from the RWST has been terminated.

The licensee application dated August 15, 2007, indicated that the proposed license amendment was being requested because, consistent with the requirements of 10 CFR 50.59(c)(2)(viii), a license amendment is necessary to implement a proposed change that would result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

3.0 TECHNICAL EVALUATION

3.1 Credit for Pre-accident Containment Pressure The current Salem licensing basis, with respect to the containment pressure assumed in determining the NPSHa for the ECCS and containment heat removal system pumps, is described in UFSAR Appendix 3A (page 3A-2) as follows:

It is assumed that the vapor pressure of the liquid in the sump (corresponding to the temperature of the liquid at the onset of recirculation) is equal to the containment pressure. This assumption assures that the actual available NPSH is always greater than the calculated value determined from the reduced equation:

NPSHavailable = (h)static head - (h)loss The licensee proposes to replace the above text with the following:

The containment pressure value will be equal to the initial air pressure in containment prior to the LOCA (i.e., the pre-accident partial air pressure in containment). However, when the containment sump vapor pressure exceeds the containment initial pressure, then the following will be assumed (h)containment pressure = (h)vapor pressure

The containment air pressure value used in the NPSHa calculation is based on the containment conditions prior to the accident only and does not include any credit for accident pressure conditions, is conservatively determined based on minimum containment initial pressure, and maximum temperature and relative humidity conditions.

The calculation also accounts for further reduction of the initial air pressure based on possible maximum cooldown of the containment environment post-LOCA.

The vapor pressure term used in the NPSHa for the sump water being pumped, is based on the highest temperature of the sump fluid for the condition being evaluated.

The NRC staff finds that the above proposed changes to the Salem NPSHa methodology are based on assumptions that: (1) will conservatively estimate the containment pressure prior to a postulated LOCA and (2) are consistent with the guidance in RG 1.1 and RG 1.82. Based on these findings, the NRC staff concludes that there is reasonable assurance that the ECCS and containment heat removal pumps will have adequate NPSH margin following a postulated LOCA. As such, the NRC staff also concludes that GDC 35 and GDC 38 will continue to be met since the ECCS and containment spray pumps will be capable of performing their design functions.

This NRC staff evaluation did not make a finding as to the acceptability of the calculation methods used to determine pre-LOCA containment pressure or post-LOCA sump temperature.

Only the criterion for determining NPSHa was evaluated.

3.2 Revision to Static Head Assumed in NPSHa Calculation In addition to the proposed change in crediting pre-accident containment pressure discussed in SE Section 3.1, the licensee also proposes to delete the following discussion from UFSAR Appendix 3A regarding how the static head term in the NPSHa calculation is determined:

Added conservatism is introduced into the NPSH calculation by calculating the static head from the elevation of the top of the sump instead of the available water level above the sump.

In place of the deleted text, the licensee proposes to add the following to UFSAR Appendix 3A:

The static head term in the NPSHa is calculated using the minimum available water inventory in the containment for recirculation operation. This minimum water inventory ensures that the containment sump strainers are fully submerged prior to initiation of the recirculation phase.

PSEGs application dated August 15, 2007, explained that since 1997 (and contrary to the current description in the UFSAR), the water level for the static head term has been determined as a minimum calculated value rather than the level at the top of the sump. Specifically, the licensees application stated that:

The current Salem RHR pump recirculation operation NPSHa calculation, originally issued in 1997, uses the minimum sump pool water level required for recirculation operation for determination of the static head available for NPSHa concerns. This water

level was determined to be at plant elevation 80'-10", which is approximately 2'-1" above the top of the ECCS containment sump curb. The ECCS containment sump water level is monitored with float type level instrumentation and the level is displayed in the control room. Salem Emergency Operating Procedures (EOP's) require that operations personnel verify that the ECCS containment sump level is adequate, i.e., it has reached this minimum level required (accounting for instrument uncertainty) to provide adequate NPSHa for operation of the RHR pumps in recirculation mode, prior to initiating the switchover of the ECCS system from injection to recirculation operation. Therefore, the added conservatism is not required and the current calculation deviation from the UFSAR does not impact the plant safety.

The new ECCS containment sump strainers, which were installed to meet GSI [Generic Safety Issue]-191 concerns, were designed and installed to ensure that with this minimum water level in the containment the new strainers would have a minimum of 3" of water cover. As a result of the previously existing configuration, and the design of the new ECCS containment sump strainers, the removal of this statement from the Salem UFSAR is acceptable.

Since the method of determining the static head term in the NPSHa calculation is based on a minimum available water inventory this method is conservative (i.e., smaller hstatic term results in smaller NPSHa). Therefore, the NRC staff finds that this change is acceptable.

The NRC staff did not review the licensee=s calculation of water level in containment as part of this amendment review.

Note, as discussed above, the licensees methodology for determining the static head term changed in 1997 (i.e., to use a minimum calculated value rather than the level at the top of the sump). However, it does not appear that a license amendment request for the change was submitted until the current application dated August 15, 2007. As such, the change was not reflected in the UFSAR consistent with the requirements in 10 CFR 50.71(e). It appears that the change in methodology was made as a result of GL 97-04, Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps, dated October 7, 1997 (ADAMS Accession No. ML031110062). This issue may be assessed further by NRC Region I staff.

3.3 Technical Evaluation Conclusion Based on the discussion in SE Sections 3.1 and 3.2, the NRC staff concludes that the proposed amendment is acceptable.

Note, this SE does not address the adequacy of the licensees corrective actions for GL 2004-02. The NRC staff will review those corrective actions in its review of the licensees response to the GL.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (72 FR 51866). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: R. Lobel R. Ennis Date: November 15, 2007