LR-N09-0143, License Amendment Request to Relocate the Reactor Recirculation System Motor-Generator (MG) Set Scoop Tube Stop Setting Surveillance

License Amendment Request to Relocate the Reactor Recirculation System Motor-Generator (MG) Set Scoop Tube Stop Setting Surveillance

ML092230345
Person / Time
Site:	Hope Creek
Issue date:	07/30/2009
From:	Barnes G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR H09-02, LR-N09-0143
Download: ML092230345 (13)

Text

PSEG Nuclear LLC P.O. Box 236,, Hancocks Bridge, NJ 08038-0236 0 PSEG Nuclear L.L. C.

10 CFR 50.90 SUUL 3 0 2009 LR-N09-0143 LAR H09-02 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

LICENSE AMENDMENT REQUEST TO RELOCATE THE REACTOR RECIRCULATION SYSTEM MOTOR-GENERATOR (MG) SET SCOOP TUBE STOP SETTING SURVEILLANCE In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear, LLC (PSEG) requests an amendment to the facility operating license listed above. The proposed amendment relocates the Technical Specification surveillance requirements for the Reactor Recirculation System MG Set Scoop Tube Stop setpoints to the Technical Requirements Manual (TRM). The proposed change is consistent with changes previously approved by the NRC for other reactor licensees and with Boiling Water Reactor (BWR) Improved Technical Specifications NUREG-1433, Rev. 3. provides a detailed description of the proposed change, background and technical analysis, No Significant Hazards Consideration Determination, and the Environmental Review Consideration. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 which provides the marked up TS Bases is included for information only.

No regulatory commitments are contained in this submittal.

PSEG requests approval of this proposed amendment by September 1, 2010 with a 60 day implementation period. The proposed changes have been reviewed by the Plant Operations Review Committee. In accordance with the requirements of 10 CFR 50.91(b)(1), a copy of this request for amendment has been sent to the State of New Jersey.

OJgj2

Document Control Desk Page 2 LR-N09-01 43 JUL 3 0 2009 If you have any questions or require additional information, please contact Mr. Jeffrie Keenan at (856) 339-5429.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on (Date)

Sin erely, George P. Barnes Site Vice President Hope Creek Generating Station Attachments (3)

S. Collins, Regional Administrator - NRC Region I R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector - Salem P. Mulligan, Manager IV, NJBNE Commitment Tracking Coordinator - Hope Creek PSEG Commitment Tracking Coordinator - Corporate

ATTACHMENT 1 LR-N09-0143 LAR H09-02 ATTACHMENT 1 EVALUATION OF THE PROPOSED CHANGE:

LICENSE AMENDMENT TO RELOCATE THE REACTOR RECIRCULATION SYSTEM MOTOR-GENERATOR (MG) SET SCOOP TUBE STOP SETTING SURVEILLANCE TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION

5.0 REGULATORY ANALYSIS

5.1 Precedent 5.2 Significant Hazards Considerations 5.3 Applicable Regulatory Requirements/Criteria

7.0 REFERENCES

1 of 6

ATTACHMENT 1 LAR H09-02 LR-N09-0143

1.0 DESCRIPTION

This letter requests an amendment to Operating License NPF-57 (Docket 50-354) for Hope Creek Generating Station (HCGS). The proposed change involves relocating the Technical Specification (TS) Surveillance Requirement (SR) 4.4.1.1.3 and associated bases to the Technical Requirements Manual (TRM). This amendment request is consistent with changes previously approved by the NRC for other reactor licensees.

SR 4.4.1.1.3 requires that each motor generator (MG) Set scoop tube mechanical and electrical stop shall be demonstrated OPERABLE with overspeed set points less than or equal to 109% and 107% respectively, of rated core flow, at least once per 18 months.

2.0 PROPOSED CHANGE

The proposed License change will relocate the following TS and associated Bases to the TRM:

TS SR 4.4.1.1.3 - Each MG Set scoop tube mechanical and electrical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 109% and 107%, respectively, of rated core flow, at least once per 18 months.

The proposed change is consistent with changes previously approved by the NRC for other reactor licensees and with Boiling Water Reactor (BWR) Improved Technical Specification NUREG-1433, Rev 3.

3.0 BACKGROUND

The Recirculation MG Scoop Tube stop setpoints are part of TS at HCGS. The MG Set scoop tube stop setpoints are required to limit the maximum core flow achievable due to a postulated reactor recirculation dual pump slow flow run out transient, which is not terminated by a reactor scram. The requirement to demonstrate the OPERABILITY of the MG set scoop tube stops is not included in NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4." During the development of NUREG-1433, the scoop tube stop SR was identified as a detail of system design not required to be included in Technical Specifications. Specifically, the MG set stop settings are inputs to the cycle-specific reload analysis. PSEG is requesting to relocate the MG Set stop setting requirement to the TRM.

The adjustable scoop tube mechanism converts an electrical input signal from the speed controller into a mechanical scoop tube position. The positioner has both mechanical (stop block) and electrical (cam operated switch) stops that limit recirculation flow by limiting the MG Set speed. The electrical stop actuates first. The mechanical stop is designed to prevent the scoop tube motion if the electrical stop fails or to mitigate overshoot of the electrical stop. The electrical stops are not an input for any accident or transient event reload evaluation. However, the mechanical stop settings are an input used in the determination of the flow dependent Minimum Critical Power Ratio (MCPR) and flow dependent Linear Heat Generation Rate (LHGR) or Average Planar Linear Heat Generation Rate (APLHGR) Operating Limits.

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ATTACHMENT 1 LAR H09-02 LR-N09-0143 Additional description of the Recirculation Flow Control System is in Updated Final Safety Analysis Report (UFSAR) section 7.7, Control Systems Not Required for Safety, subsection 7.7.1.2 Recirculation Flow Control System (RFCS).

4.0 TECHNICAL ANALYSIS

The purpose of the mechanical stops is to limit the maximum core flow achievable due to a postulated reactor recirculation dual pump slow flow run out transient which is not terminated by a reactor scram. SR 4.4.1.1.3 ensures that the maximum possible core flow is less than or equal to the maximum core flow assumed for the establishment of the flow dependent Minimum Critical Power Ratio (MCPR) and the flow dependent Linear Heat Generation Rate (LHGR) or Average Planar Linear Heat Generation Rate (APLHGR) Operating Limits (either LHGR or APLHGR may be used to provide protection for the fuel thermal mechanical limits). This event stabilizes at a new core power level corresponding to the maximum possible core flow which is dictated by the MG set scoop tube mechanical stop. The flow dependent MCPR and LHGR or APLHGR Operating Limits are established to protect the fuel cladding by considering the reactor power increase which would result from a postulated increase in recirculation flow to the maximum core flow allowable by the mechanical stops due to a dual pump run out. The flow dependent MCPR and LHGR or APLHGR Operating Limits are established with consideration of this event such that the MCPR Safety Limit or fuel thermal mechanical limits are not violated. The NRC staff review of the analyses of the increase in core flow events for HCGS, including the slow recirculation increase event, is documented in Reference 7.1. Thus, although the verification of scoop tube stop setting is located in the Reactor Coolant System section of the TS, it is more correctly associated with the requirements for the MCPR and the LHGR or APLHGR Operating Limits.

The MCPR and LHGR or APLHGR Operating Limits are specified on a cycle specific basis in the Core Operating Limits Report (COLR). Implicit in the establishment of the MCPR and LHGR or APLHGR Operating Limits is that the plant is operated and configured in accordance with the plant design and licensing basis contained in the TS, TRM and UFSAR. Upon approval of the proposed change details of the configuration or routine activities needed to give reasonable assurance that the limits are satisfied will be within the TRM, which is controlled via the 10 CFR 50.59 process.

This provides the flexibility for licensee control of these details under an appropriate framework of regulatory control. The TS requirement to operate consistent with the MCPR and LHGR or APLHGR Operating Limits provides adequate assurance that the plant will be operated with adequate protection of the public health and safety.

5.0 REGULATORY ANALYSIS

In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear, LLC (PSEG) requests an amendment to Facility Operating Licensing No. NPF-57 for Hope Creek Generating Station (HCGS). The proposed amendment relocates the Technical Specification surveillance requirements for the Reactor Recirculation System MG Set Scoop Tube Stop setpoints to the Technical Requirements Manual (TRM). This proposed change is consistent with changes previously approved by the NRC for other 3 of 6

ATTACHMENT 1 LAR H09-02 LR-N09-0143 reactor licensees and with Boiling Water Reactor (BWR)

Improved Technical Specifications NUREG-1433, Rev. 3.

5.1 Precedent The proposed change is similar to changes previously approved in Amendment 130 to Facility Operating License No. NPF-43 for Fermi 2, dated February 8, 1999.

5.2 No Significant Hazards Consideration PSEG Nuclear, LLC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment for HCGS by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment" as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

HCGS SR 4.4.1.1.3 contains implementation detail that is more adequately and more appropriately controlled in accordance with 10 CFR 50.59 process. The detail in SR 4.4.1.1.3 does not fall within the requirements of 10 CFR 50.36(c)(3). The change does not eliminate the necessary surveillance testing of the MG Set mechanical and electrical stops.

As previously stated the electrical stops are not an input for any accident or transient event reload evaluation. The recirculation pump MG Set mechanical stop settings will continue to be maintained at or below the values assumed for the establishment of the MCPR and LHGR or APLHGR Operating Limits. The MCPR and LHGR or APLHGR Operating Limits are established and specified in the COLR in accordance with TS 6.9.1.9 and operation within these limits is required by TS 3.2.1, 3.2.3, and 3.2.4. The changes described will therefore have no impact on the probability or consequences of an accident previously evaluated.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed TS change does not result in any changes to the design (equipment / configuration) or operation of the plant and will thus not create a new failure mode or common mode failure. The MG Set mechanical and electrical stops will continue to operate as intended and as designed. These changes will therefore not create the possibility of a new or different kind of accident, from any accident previously evaluated.

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ATTACHMENT 1 LAR H09-02 LR-N09-0143

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Relocating the TS SR 4.4.1.1.3 requirement to set the MG Set stops to the TRM will not involve a significant reduction in the margin of safety because setting the stops is still required per the TRM to ensure the stops are set consistent with the values assumed for the establishment of the MCPR and LHGR or APLHGR Operating Limits. Additionally, the MCPR and LHGR or APLHGR Operating Limits are specified in the COLR, and operation within these limits is still required by TS 3.2.1, 3.2.3, and 3.2.4. Therefore, the margin of safety as defined in the bases of any TS is not reduced by relocating the surveillance requirement from the TS to the TRM. In addition to the above, relocation of the TS is consistent with the Boiling Water Reactor (BWR) Improved Technical Specifications NUREG-1433, Rev. 3.

Based upon the above evaluation, PSEG has concluded that the proposed change does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.3 Applicable Regulatory Requirements/Criteria The applicable regulation 10 CFR 50.36(c)(3), states that "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operationwill be met." PSEG concludes that SR 4.4.1.1.3 is not required to be in TS per 10 CFR 50.36(c)(3). The proposed license amendment request is to relocate SR 4.4.1.1.3 from TS to the TRM. The setting of the stops by itself does not assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, or that the limiting conditions for operation are met. Rather, the set point of the mechanical stop is used as an input to determine the flow dependent MCPR and LHGR or APLHGR operating limits. The MCPR and LHGR or APLHGR Operating Limits are specified in the COLR and operation within these limits is required by TS 3.2.1, 3.2.3, and 3.2.4.

Operation within these limits with MG Set mechanical stops as an input will assure that the necessary quality of system and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation are met.

As described in the technical analysis section, relocating the MG Set stop surveillance to the TRM will still require the setting of the mechanical and electrical stops. The recirculation pump MG Set speed stop settings will continue to be maintained at or below the values assumed for the establishment of the MCPR and LHGR or APLHGR Operating Limits. Hence, the nuclear quality and operation within safety limits is maintained by compliance with TS 3.2.1, 3.2.3 and 3.2.4. LCO 3.4.1.1 requires two reactor coolant system recirculation loops to be in operation. SR 4.4.1.1.3 is not required 5 of 6

ATTACHMENT 1 LAR H09-02 LR-N09-0143 to assure LCO 3.4.1.1 will be met. Therefore, SR 4.4.1.1.3 is not required to be in TS per 10 CFR 50.36(c)(3).

The proposed change is similar to changes previously approved in Amendment 130 Fermi 2, dated February 8, 1999.

Based on the considerations discussed above, (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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 NRC letter to William Levis, PSEG Nuclear LLC, dated May 14, 2008, "Hope Creek Generating Station - Issuance of Amendment Re: Extended Power Uprate (TAC No.

MD3002)"

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ATTACHMENT 2 LR-N09-0143 LAR H09-02 ATTACHMENT 2 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES:

LICENSE AMENDMENT TO RELOCATE THE REACTOR RECIRCULATION SYSTEM MOTOR-GENERATOR (MG) SET SCOOP TUBE STOP SETTING SURVEILLANCE The following Technical Specifications for HCGS (Facility Operating License NPF-57) are affected by this change request:

TS No.

Title Page 4.4.1.1.3 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 3/4 4-2a 1 of 2

ATTACHMENT 2 LR-N09-0143 LAR H09-02 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 With one reactor coolant system recirculation loop not in operation at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:

a.

Reactor THERMAL POWER is

• 60.86% of RATED THERMAL POWER, and

b.

The recirculation flow control system is in the Local Manual

mode, and C.

The speed of the operating recirculation pump is less than or equal to 90% of rated pump speed.

4.4.1.1.2 With one reactor coolant system recirculation loop not in operation, within no more than 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is

• 38% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is S 50% of rated loop flow:

a.

145*F between reactor vessel steam space coolant and bottom head drain line coolant, and

b.

50*F between the reactor coolant within the loop notin operation and the coolant in the reactor pressure vessel, and

c.

5 50'F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements or Specifications 4.4.1.1.2b and 4.4.1.1.2c do not apply when the loop not in operation is isolated from the reactor pressure vessel.

0 HOPE CREEK 3/4 4-2a Amendment No.

174 2 of 2

ATTACHMENT 3 LR-N09-0143 LAR H09-02 ATTACHMENT 3 (Information Only)

TECHNICAL SPECIFICATION BASES PAGES WITH PROPOSED CHANGES:

LICENSE AMENDMENT TO RELOCATE THE REACTOR RECIRCULATION SYSTEM MOTOR-GENERATOR (MG) SET SCOOP TUBE STOP SETTING SURVEILLANCE The following Technical Specification Bases for HCGS (Facility Operating License DPR-

70) are affected by this change request:

Bases No. I Title Page 3/4.4.1 RECIRCULATION SYSTEM B 3/4 4-1 and 2 1 of 3

ATTACHMENT 3 LAR H09-02 LR-N09-0143 3/4.4 REACTOR COOLAT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single loop operation is permitted if the MCPR fuel cladding Safety Limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2 respectively.

APLHGA limits are decreased by the factor given in the CORE OPERATING LIMITS RE-PORT (COLR),

LHGR limits are decreased by the factor given in the COLR, and MCPR operating limits are adjusted as specified in the COLR.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive core internals vibration.

The surveillance on differential temperatures below 38%

THERMAL POWER or 501 rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.

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ATTACHMENT 3 LR-N09-0143 3/4A.-4 RECTOR COOLANT SYSTEM LAR H09-02 BAS ES 3/4.4.1 RECIRCOLAT-IONN SYSTEM (continued) cuvcl2angot ncdo to be performad mere-than nr.nzi r.tza-*-ye-

,oeof the phenomena, to "06idtr -rt: :) Th tor flow Nosistanec may deerease diaring the c~rer-ting eyele, reuIring lewer reairoulnticrn pump

.th b gi-tizn,

2) Significaat ohanges 4.rn fue! #.sign oam affect the r.- l a t i qn s h i p o f r e ei v¢ -_ak~ it n d r !

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ilow, and e) tgltn c*t*. oyetm itruop od-*aibr4tin4 ' 0an lgs cts.;,

f racrciation loopn inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core, thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance or. a prescribed schedule for significant degradation.

Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA:analysis design criteria for two recirculatioi loop operation.

The limits-will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.' In the case where the mismatch limits cannot be maihtained'during two loop operation, continued operation is permitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to startup 'of an idle loop.

The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pumpo and recirculation nozzles.

Sudden equalization'of a temperature difference > 145*F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

3/4.4.i2 SAETY/RZLIEF VALVES The safety valve function of the safety/telief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1375 psig in accordance with the ASME Code.

A total of 13 OPERAbLE safety/relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case transient.

Demonstration of the safety relief valve lift settings occurs only during, shutdown.

The safety relief valve pilot stage assemblies are set pressure tested in accordance with the recomnendations of General Electric HOPZ CREEK B 3/4 4-2 50.59 # HC-09-056 (PSEG Issued) 3 of 3