JAFP-09-0132, Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves

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Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves
ML093310431
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 11/23/2009
From: Peter Dietrich
Entergy Nuclear Northeast, Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-09-0132, TAC ME1818
Download: ML093310431 (22)
v • d • e


Text

-En tergy Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. Fitzpatrick NPP P.O. Box 110 Lycoming, NY 13093 Pete Dietrich Site Vice President JAFP-09-0132 November 23, 2009 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

SUBJECT:

Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59

REFERENCE:

NRC Letter, Nancy Salgado to Vice President, Operations Entergy Nuclear Operations, Inc. James A. FitzPatrick Nuclear Power Plant, James A. FitzPatrick Nuclear power Plant - Relief Request VRR-06, Revision 1 From The Requirements of the OM Code RE: Inservice Testing of Safety Relief Valves (TAC NO, ME-1818), dated October 1, 2009

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby requests an amendment to the Technical Specifications (TS) for the James A. FitzPatrick Nuclear Power Plant (JAF).

This license amendment submittal requests modifying the TS requirements for testing of the Safety/Relief Valves (SRVs) by deleting the current requirement to manually actuate each SRV during plant startup. This request is consistent with the relief from the OM Code approved in reference 1. provides the Application for Amendment to Modify the Technical Specifications Requirements On Testing of Safety/Relief Valves. provides the proposed TS changes as marked up pages. provides the proposed TS changes in final typed format with change bars. provides the proposed TS Bases changes as marked up pages.

The TS Bases changes are provided for NRC information only. The final TS Bases pages will be submitted with a future update in accordance with TS 5.5.11, "Technical Specifications (TS)

Bases Control Program."

Entergy requests NRC approval of the proposed TS amendment by July 31, 2010, with the amendment being implemented within 60 days from approval.

ADQLj 7

JAFP-09-0132 Page 2 of 3 In accordance with 10 CFR 50.91, a copy of this application, with the associated attachments, is being provided to the designated New York State official.

There are no new commitments made in this letter.

Questions concerning this report may be addressed to Mr. Joseph Pechacek, Licensing Manager, at (315) 349-6766.

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

Executed on the 23 N ber 2009.

Site Vice President PD/JP/ed Attachments: 1. Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves

2. Proposed Technical Specification Changes (Marked up)
3. Proposed Technical Specification Changes (Final Typed)
4. Proposed Technical Specification Bases Changes (Marked up) (Information Only) cc: next page

JAFP-09-0132 Page 3 of 3 cc:

Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Resident Inspector's Office U.S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, NY 13093 Mr. Bhalchandra Vaidya, Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C2A Washington, DC 20555-0001 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 101h Floor Albany, NY 12223 Mr. Francis J. Murray Jr., President NYSERDA 17 Columbia Circle Albany, NY 12203-6399

JAFP-09-0132 Attachment 1 Application for Amendment to Modify the Technical Specifications Requirements for Testing of Safety/Relief Valves (6 Pages)

JAFP-09-0132 Attachment 1

1.0 DESCRIPTION

The proposed amendment would modify the Technical Specifications (TS) requirements for testing of the Safety/Relief Valves (SRVs) by deleting the current requirement to manually actuate each SRV during plant startup. Elimination of the manual actuation requirement is desirable to decrease the potential for SRV leakage and spurious SRV openings. This elimination is consistent with the James A. FitzPatrick Nuclear Power Plant's (JAF) approved OM Code Relief Request VRR-06.

2.0 PROPOSED CHANGE

S Current TS Surveillance Requirement (TSSR) 3.4.3.2 states, "Verify each required SRV opens when manually actuated." TSSR 3.5.1.13 likewise states, "Verify each required ADS valve opens when manually actuated." The proposed amendment would delete these requirements in their entirety.

TS Bases associated with these Surveillance Requirements will also be deleted in their entirety. Revised Bases pages are attached for information, but do not require NRC approval.

3.0 BACKGROUND

SRVs installed at JAF are Target Rock model 7567F two-stage safety/relief valves.

Eleven SRVs are installed on the main steam lines between the reactor vessel and the inboard main steam isolation valves. Each SRV discharges via a separate tailpipe to a point below the water level in the suppression pool. SRVs open:

" In the safety mode on high reactor pressure, to provide primary overpressure protection to the reactor coolant pressure boundary.

" In the relief mode when actuated by the SRV Electric Lift logic on high reactor pressure, as a backup to the safety mode actuation.

  • In the relief mode when manually actuated by individual control switches in the Control Room, or by individual control switches in the Remote Shutdown system.

Experience in the industry and at JAF has shown that manual actuation of SRVs during plant operation leads to valve seat leakage. In particular, manual actuation testing has been a significant contributor to main stage seat leakage at JAF. SRV leakage is routed to the suppression pool; the increased heat and fluid additions to the suppression pool require more frequent suppression pool cooling and pump-down operations. Main stage seat leakage also tends to mask the indications of pilot stage seat leakage; pilot stage Page 1 of 6

JAFP-09-0132 Attachment 1 leakage can cause maloperation of the SRV, including spurious actuation and/or failure to reclose after actuation. Excessive leakage of either stage requires plant shutdown to replace the leaking SRV.

The Boiling Water Reactor Owners' Group (BWROG) Evaluation of NUREG-0737, "Clarification of TMI Action Plan Requirements," Item I1.K.3.16, "Reduction of Challenges and Failures of Relief Valves," recommends that the number of SRV openings be reduced as much as possible and that unnecessary challenges should be avoided. NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants,"

NUREG-0123, "Standard Technical Specifications for General Electric Boiling Water Reactors," and NUREG-0626, "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications" also recommend reducing the number of challenges to the SRVs.

The current Surveillance Testing Requirements are based on demonstrating Compliance with the ASME OM Code 2001 Edition to 2003 Addenda. JAF requested NRC Permission to use a portion of a later code Edition for testing of SRVs. That relief was approved on October 1, 2009.

4.0 TECHNICAL ANALYSIS

The manual actuation test currently prescribed in TSSRs 3.4.3.2 and 3.5.1.13 provides demonstration of the mechanical operation of the SRVs. This demonstration is used as part of the surveillance bases for establishing the OPERABILITY of the SRVs and demonstrates JAF compliance with the ASME Code for the-Fourth Inservice Testing (IST) Interval. The Code of record for the Fourth IST Interval is ASME OM Code 2001 Edition through 2003 Addenda. ASME OM Code 2001 Edition through 2003 Addenda, Appendix I, paragraph 1-3410(d), requires the plant "to verify open and close capability of the valve before resumption of electric power generation. This applies to valves that have been either maintained in place, or removed for maintenance and testing and reinstalled." However, ASME OM CODE 2004 Edition no Addenda, Appendix I, Paragraph 1-3410 (d) states "Each valve with an auxiliary actuating device that has been removed for maintenance or testing and reinstalled after meeting the requirements of I-3310, shall have the electrical and pneumatic connections verified either through mechanical / electrical inspections or test prior to the resumption of electrical power generation. Main disc movement and set pressure verification are not required."

JAF requested permission to use a portion of the later code edition for testing of SRVs.

The NRC approved that request in their letter from Nancy Salgado to the Vice President, Operations, Entergy Nuclear Operations, Inc., James A, FitzPatrick Nuclear Power Plant, James A. FitzPatrick Nuclear Power Plant - Relief Request VRR-06, Revision 1 From the Requirements of the OM Code RE: Inservice Testing of Safety Relief Valves (TAC NO. ME 1818), dated October 1, 2009 and in the associated Safety Evaluation Report.

Based on the approved relief the manual actuation is no longer required to comply with the Inservice Testing Program. Manual actuation testing is also not required to establish surveillance bases for OPERABILITY. TSSR 3.0.1 Bases states in part, "Surveillances Page 2 of 6

JAFP-09-0132 Attachment 1 may be performed by means of any series of sequential, overlapping, or total steps provided the entire Surveillance is performed within the specified frequency." The following testing is performed on the SRVs:

  • The simulated automatic actuation test specified in TSSR 3.5.1.11, and additional surveillances associated with LCOs 3.3.5.1 and 3.3.3.2 and TRM 3.3.1, demonstrate the ability of various logics and controls to actuate the SRVs up to the point of energizing the solenoids. These tests are performed once per operating cycle (two years).

" A solenoid valve functional test will be performed in situ for each SRV solenoid valve once per operating cycle. This test demonstrates that when the solenoid is energized, it applies pneumatic pressure to the SRV actuator.

  • An SRV actuator functional test will be performed at an offsite test facility as part of certification testing for each SRV pilot. Certification test intervals are determined in accordance with the Inservice Testing Program, which limits the maximum interval of service to six years under conditions described in Relief Request VRR-006; typically, two-stage pilots are in service for a maximum of one operating cycle (two years). The actuator test demonstrates that the actuator moves the pilot stem when pneumatic pressure is applied.
  • Setpoint testing is performed at the offsite test facility as part of certification testing for each SRV pilot, at intervals determined in accordance with the Inservice Testing Program. This test is the existing test required by TSSR 3.4.3.1. In addition to demonstrating that the SRV pilot stage will actuate on high steam pressure in the safety mode, this test overlaps with the actuator functional test to demonstrate that the pilot stage will actuate in the relief mode.
  • Main stage certification testing will be performed at the offsite test facility at intervals determined in accordance with the Inservice Testing Program, at least every six years per approved Relief Request VRR-006. Main stage certification testing demonstrates that the main stage will open and port steam when actuated by the installed pilot stage.

These tests are required by the Technical Specifications and the Fourth Interval Inservice Testing Program and provide an acceptable surveillance and testing basis for the demonstrating the OPERABILITY of the SRVs.

The requested amendment to the Technical Specifications is required to implement the approved code relief.

Page 3 of 6

JAFP-09-0132 Attachment 1 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change deletes a testing requirement that is no longer required to meet the American Society of Mechanical Engineers (ASME) Operating and Maintenance (OM) Code testing requirements associated with the Safety Relief Valves (SRVs)

Certain SRV malfunctions are included in the Final Safety Analysis Report (FSAR) safety analyses. Specifically, the plant safety analyses include the inadvertent opening of an SRV and a stuck open SRV. By not actuating the SRVs during plant operation, for testing, the incidence of pilot stage leakage of SRVs has been shown to decrease. That decrease results in improved material condition of the SRVs. Industry experience shows that improvements in material condition increases equipment reliability.

Existing analyses address events involving an SRV inadvertently opening or failing to reclose. Analyses also address the likelihood and consequences of failure of one or more SRVs to open. The SRV failures documented at JAF in Licensee Event Reports over the past 20 years have been related to setpoint drift. In all cases the main bodies have opened once the pilot stage actuated.

Based on these considerations, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes delete testing requirements that are no longer required by the OM Code. The current requirements in the code provide for testing when the SRVs are not installed in the plant. Once the SRVs have been tested and certified at an offsite facility the valves are handled and shipped by individuals who are trained in material handling, they are receipt inspected, and then are installed by qualified maintenance personnel who are trained to use site specific installation procedures. As required by the code after installation the electrical and pneumatic connections for each SRV that is worked on is verified either through mechanical / electrical inspections or test prior to the resumption of electrical power generation. These activities have been evaluated by ASME to provide a high level of confidence that the SRVs will perform their function.

Existing analyses address events involving an SRV inadvertently opening or failing to reclose. Analyses also address the likelihood and consequences of failure of one or Page 4 of 6

JAFP-09-0132 Attachment 1 more SRVs to open. The proposed changes do not introduce any new failure mode, and therefore, do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No.

Overpressure protection of the reactor coolant pressure boundary is based on the SRVs setpoint and total relief capacity. Setpoint is verified at an offsite testing facility; this requirement is not altered by the proposed change. Relief capacity of each SRV is determined by valve geometry, which is also not altered by the test methods. The margin of safety in the Loss of Coolant Accident analysis due to functioning of the Automatic Depressurization System (ADS) is also based on total relief capacity of the associated SRVs. The proposed deletion of tests that are no longer required by the OM Code does not alter the critical parameters that affect the margin of safety in analyses involving the SRV functions. Therefore, the proposed change does not involve a significant reduction in any margin of safety.

5.2 Applicable Regulatory Requirements / Criteria 10 CFR 50.36 requires in part that the operating license of a nuclear production facility include technical specifications. Paragraph (c)(2)(ii) of that part requires that a limiting condition for operation (LCO) of a nuclear reactor must be established for each item meeting one or more of four criteria. The SRV functions identified in LCOs 3.4.3 and 3.5.1 both meet Criterion 3, "A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." Paragraph (c)(3) further requires the establishment of surveillance requirements, "relating to test, calibration, or inspection to assure.. .that the limiting conditions for operation will be met." As discussed above, the proposed deletion of the surveillance requirements to open the SRVs is consistent with the 2004 Edition of the ASME OM Code. The testing and certification required by the OM Code will continue to be applicable to JAF.

In conclusion, 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 approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Page 5 of 6

JAFP-09-0132 Attachment 1 6.0 ENVIRONMENTAL ASSESSMENT A review has determined that the proposed changes 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 changes do not involve: (i) a significant hazards consideration; (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet 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 changes.

7.0 PRECEDENT ASME OM CODE 2004 Edition no Addenda, Appendix I, Paragraph 1-3410 (d) "Each valve with an auxiliary actuating device that has been removed for maintenance or testing and reinstalled after meeting the requirements of 1-3310, shall have the electrical and pneumatic connections verified either through mechanical / electrical inspections or test prior to the resumption of electrical power generation. Main disc movement and set pressure verification are not required. Use of this portion of the ASME OM Code 2004 Edition was approved at JAF by the NRC in their letter from Nancy Salgado to the Vice President, Operations, Entergy Nuclear Operations, Inc., James A, FitzPatrick Nuclear Power Plant, James A. FitzPatrick Nuclear Power Plant - Relief Request VRR-06, Revision 1 From the Requirements of the OM Code RE: Inservice Testing of Safety Relief Valves (TAC NO. ME 1818), dated October 1, 2009 and in the associated Safety Evaluation Report.

Page 6 of 6

JAFP-09-0132 Attachment 2 Proposed Technical Specification Changes (Marked up)

Pages 3.4.3-2 3.5.1-6 3.5.1-7

S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoint of the In accordance required S/RVs is 1145 +/- 34.3 psig. Following with the Inservice testing, lift settings shall be within +/- 1%. Testing Program R3.2OTE TE Not required to be pe~rmcO~nd until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressUrc and flow arc adequatet pe~fOrM the test.

Verify each rcqUircd S/Rk opens when manually 24 months on a aetuated. STAGGERED TEST BASIS for eaeh vave seleod JAFNPP 3.4.3-2 Amendment 2-7-4

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUNCY SR 3.5.1.9 ------------------- NOTE-----------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure < 165 psig, the 24 months HPCI pump can develop a flow rate

> 3400 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.10 -------------- NOTES----------

1. For the HPCI System, not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
2. Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.11 -------------- NOTE-----------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 24 months simulated automatic initiation signal.

SR 3.5.1.12 Verify each LPCI motor operated valve 24 months independent power supply inverter capacity is adequate to supply and maintain in OPERABLE status the required emergency loads for the design duty cycle.

l (onu~ed) I JAFNPP 3.5.1-6 Amendment 2-74

EGGS - pe~tng 3.5.1 S*URVI*L.J l--I*.LANC RIEQUI*iREMENTSI(--It I *P SURVEILLANCE FREQUENCY SR-54 NOTE Not required to be pe~rmfF~ed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to performn the test.

24 months on a Verify eah rneqir*ed ADS valve opens when STAGGERED manually actuated. TEST BASIS-for eaeh ale sv*eno1d nr 27

  • L R 4AN -I-' .i 7 irtn A pm c 274

JAFP-09-0132 Attachment 3 Proposed Technical Specification Changes (Final Typed)

Pages 3.4.3-2 3.5.1-6

S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoint of the In accordance required S/RVs is 1145 +/- 34.3 psig. Following with~the Inservice testing, lift settings shall be within +/- 1%. Testing Program I

JAFNPP 3.4.3-2 Amendment

ECCS-- Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 -------------- NOTE-----------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure < 165 psig, the 24 months HPCI pump can develop a flow rate

> 3400 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.10 -------------- NOTES----------

1. For the HPCI System, not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
2. Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.11 --------------- NOTE----------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 24 months simulated automatic initiation signal.

SR 3.5.1.12 Verify each LPCI motor operated valve 24 months independent power supply inverter capacity is adequate to supply and maintain in OPERABLE status the required emergency loads for the design duty cycle.

I JAFNPP 3.5.1-6 Amendment

JAFP-09-0132 Attachment 4 Proposed Technical Specification Bases Changes (Marked up)

(Information Only)

Pages B 3.4.3-4 B 3.5.1-15 B 3.5.1-16 B 3.5.1-17

S/RVs B 3.4.3 BASES SU RV.EI LLANCE REQUIR~EMENTS

-(Gentined) A manual atuation of each required S/R- is pe.formed while bypassing mnain steam flow to the condcnscr and obscR'ing Ž!10%

closure of the turbine bypass valves to verif; that, mcchanncally, the valve is functioning properly and no blockage exists in the valve discharge line. This can also be demonstrated by the response of the turbine control valves, by a changc in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to pcrforM this test to avoi damaging the valve. Also, adequate steam flow must be passing, through the Main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs dive~t steam flow upon opening. Sufficient time is therefore allowed after the FeqUired pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is 970 psig9 (the pressure consistent with Vendor recommendain) Adqute steam floW is represented by two Or mor~e turbine bypas vavsoen, or total steam flow Žt 106 lb,'hr. These conditions Will require the plant to be in MODE 1, which has been shown to be an acceeptable condition to performn this test. This test causes a small neutron flux transient which may cause a scram in MODE 2 while operating cloe to the Average Power Range Monitors Neutron Flux- High (Sta~tup)

Allowable Value. Plant sta~tup is allowed prior to per-forming this test because valve OPERABILIW and the setpoints for overpressure protection are verifined, per ASME Code requirements, prior to valve installation. T-herefOre, this SR is modified by a Note that states the Surveillance- is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to performn the test.

The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required steam pressure and flow arc reached is sufficeient to achieve stbc conditions for testing and provides a reasonable time to complete the SR. if a valve fails to actuate due only to the failure of the solenoid S,'RV isconsidered OPER6ABLtiE.

The 24 month on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each 5/RV is alternately tested. The 24 month Frequency was developed bassed on the S/RV tests req Uired by the ASME Boiler and PresSUrP e','sselCode, Sectlion X! (Ref. 7). Operating expriecehas shown that these components usually pass the Surveillancee when perform~ed at the 24 month Frequency. Therefore, the Frequency is acceptable fromA a reliability standpoint.

(continued)

JAFNPP B 3.4.3-4 Revision 0- 1

ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.10 (continued)

REQUIREMENTS and flow are reached is sufficient to achieve stable conditions for testing and provides reasonable time to complete the SR. Note-2 excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.

SR 3.5.1.11 The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. Asystem functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e., solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.13 and the The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function. The 24 month Frequency is based on the need to perform the Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation since the valves are individually tested in accordance with SR 3.5.1.13 during main stage certification testing at the vendor test facility. This prevents the possibility of an RPV pressure blowdown.

SR 3.5.1.12 A LPCI motor operated valve independent power supply subsystem inverter test is a test of the inverter's capability, as found, to satisfy the design requirements (inverter duty cycle). The discharge rate and test length correspond to the design duty cycle requirements.

(continued)

JAFNPP B 3.5.1-15 Revision 0

ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.12 (continued)

REQUIREMENTS The Frequency of 24 months is acceptable, given plant conditions required to perform the test and the other requirements existing to ensure adequate LPCI inverter performance during the 24 month interval. In addition, the Frequency is intended to be consistent with expected fuel cycle lengths.

SR-34 A manual actuation of eaeh required ADS valve is performed while bypassing main steam flow to the condenser and observing Ž!10%

cloSUre of the tubinc bypass valves to veint; that the valve and solenoid are functioning properly and that no blockage exist-s inth S,'RV.dischag lie.Tis can also be demonstrated by the response of thcie tu-rbhine conrol1 or bypass valve Or by a change in the measured flow Or by any other mnethod suitable to vent,' steam flow.

Adequate reactor steam dome pressure must be available to perform this test to avoid damagging the valve. Also, adequate steam flow must be passin throuh the mnain turbine Or turbine bypass valves to continu to- conro reactor pressure when the ADS valves diver steam flow uoopnng. Sufficient time is therefore allowed after the required prsuead flow are ac-hieved to performn this SR.

Adequate prsr at which this SR is to be perform~ed isŽt 970 psig (the) prsuecnsistent with vendor recommendations). Adequate steam flow is represented by at least two Or morM turbine bypass valves open Or total steam flow Ž! -O lb/hr. These conditions will require the plant to be in MODE 1, which has been shown to be an acceptable cnionto perform this test. This test causes a small neutron flux transient which may cause a scram in MODE 2 while operating close to the Average Power Range Monitors Neutron Flux High (Startup) Allowable Value. Reactor startup is allowed prior tO performing this SR because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance Is not required to be performed until 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow arc adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure and flow are reached is sufficient to achieve stal SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing 0t the assumed safety function.

(continued)

JAFNPP B 3.5.1-16 Revision 0

ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3...3(continued)

ThecFequency of 24 moenths on a STAGGERED TEST BASIS ensure.s that both solenoids for each ADS valve arc alterniately tested. The Frequency is based on the need to perform the Surveillan1e under the condition thtaply during a sta~tup fromA a plant outage.

Operating experience has shown that these components usually pas the SR when performed at the 24 moenth IFrequency, which is based on the refueling eycle. Therefore, the IFrequency was conoluded to b acceptable froM a reliability standpoint.

REFERENCES 1. UFSAR, Section 6.4.3.

2. UFSAR, Section 6.4.4.
3. UFSAR, Section 6.4.1.
4. UFSAR, Section 6.4.2.
5. NEDC-31317P, Revision 2, James A. FitzPatrick Nuclear Power Plant SAFER/GESTR-LOCA, Loss-of-Coolant Accident Analysis, April 1993.
6. UFSAR, Section 14.6.1.5. V
7. UFSAR, Section 14.6.1.3.
8. 10 CFR 50, Appendix K.
9. UFSAR, Section 6.5.
10. 10 CFR 50.46.
11. 10 CFR 50.36(c)(2)(ii).
12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),

Recommended Interim Revisions to LCOs for ECCS Components, December 1, 1975.

13. UFSAR, Section 4.4.5.

JAFNPP B 3.5.1-17 Revision 0

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