Safety Evaluation Accepting Rev 2 of Third Interval Inservice Testing Program for Pumps & Valves for James a FitzPatrick Nuclear Power PlantML20202J089 |
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02/03/1999 |
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NRC (Affiliation Not Assigned) |
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ML20202J061 |
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NUDOCS 9902090036 |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20216D9541999-07-28028 July 1999 Safety Evaluation Authorizing Proposed Alternatives for Second 10-year Interval Pursuant to 10CFR50.55a(a)(3)(ii) ML20202J0891999-02-0303 February 1999 Safety Evaluation Accepting Rev 2 of Third Interval Inservice Testing Program for Pumps & Valves for James a FitzPatrick Nuclear Power Plant ML20196J3501998-12-0404 December 1998 SER Accepting License Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20196F9251998-11-25025 November 1998 Safety Evaluation Re Thrid 10-year Interval Inservice Insp Program Relief Requests for Plant ML20195K4211998-11-17017 November 1998 Safety Evaluation Authorizing Proposed Alternative in Relief Request VRR-05 Per 10CFR50.55a(a)(3)(i) & PRR-01,PRR-02R1, PRR-03,PRR-04,VRR-02,VRR-03 & VRR-04 Per 10CFR50.55a(a)(3)(ii) ML20195E1051998-11-13013 November 1998 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves, Issued 950817 ML20155H5801998-11-0303 November 1998 Safety Evaluation Authorizing Postponement of Beginning of Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) Re ASME Code,Section XI ML20155H5321998-11-0303 November 1998 Safety Evaluation Authorizing Alternative to ASME Code Requirements for CRD Bolting ML20154L6591998-10-14014 October 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Snubber Visual Inservice Exam Intervals & Sampling Rates Requirements Contained in ASME Code,Section Xi,Subsection Iwf,Article IWF-5000 ML20153D2591998-09-21021 September 1998 SER Accepting Proposed Alternative Testing of Containment Following ECCS Suction Strainer Replacement ML20236V8181998-07-29029 July 1998 Safety Evaluation Accepting Request for Relief from Implementation of Requirements of 10CFR50.55a Related to Containment Repair & Replacement Activities for James a FitzPatrick Nuclear Power Plant ML20236X5881998-07-29029 July 1998 Safety Evaluation Supporting Amend 245 to License DPR-59 ML20154L9201998-07-10010 July 1998 SER Accepting Rev to Reactor Vessel Surveillance Capsule Withdrawal Schedule for James a Fitzpatrick Nuclear Power Plant ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively ML20198S3031998-01-14014 January 1998 Supplemental SE Accepting RG 1.97,rev 2 Recommendations for Containment Isolation Valve Position Indication Instrumentation at NPP ML20128G2441993-02-0909 February 1993 Correction to NRC SE Associated W/Ts Amend 184,dtd 921217. SE Restates Portion of Section 2.0 ML20126F4771992-12-23023 December 1992 Safety Evaluation Granting Licensee Relief from ASME Code Requirements for Repair of RWCU Equalizing Line Until Next Refueling Outage ML20059K4331990-09-13013 September 1990 Safety Evaluation Accepting Util 881110,890328 & 900129 Submittals Re IGSCC Insp & Repair for Facility Reload 8/ Cycle 9 Refuel Outage ML20056B4011990-08-20020 August 1990 Safety Evaluation Approving Licensee Relief Request R14 & Denying Requests R15 & R5A Re Hydrostatic Test Requirements ML20058P4331990-08-13013 August 1990 Safety Evaluation Accepting ATWS Recirculation Pump Trip Sys Design Mod ML20206F5701988-11-18018 November 1988 Safety Evaluation Re Compliance w/10CFR50.62 ATWS Rule Re Alternate Rod Injection & Recirculating Pump Trip Sys ML20206D5231988-11-10010 November 1988 Safety Evaluation Supporting 880309 Request for Relief from Hydrostatic Test Requirement for HPCI & Rcic,Provided That Alternative Testing Performed ML20148B1001988-03-14014 March 1988 Safety Evaluation Accepting Util Justification for Deviations from Reg Guide 1.97 for post-accident Monitoring Variables ML20236G5801987-10-27027 October 1987 Safety Evaluation Supporting Util 850930,860827 & 1208 Submittals of Second 10-yr Inservice Insp Program Plan & Associated Relief Requests from ASME Code Insp Requirements ML20238E1461987-09-0808 September 1987 Safety Evaluation of Util 870415 Proposed Design for Standby Liquid Control Sys.Design Meets Requirements of 10CFR50.62 Re ATWS ML20237L5241987-09-0101 September 1987 Safety Evaluation Supporting Util 831109,840629 & 850702 Responses to Generic Ltr 83-28,Items 2.1 & 4.5.2 Re Equipment Classification & Vendor Interface & Reactor Trip Sys Reliability, Respectively ML20236J8151987-07-30030 July 1987 Safety Evaluation Re Insps for & Repairs of IGSCC During Reload 7/Cycle 8 Refueling Outage.Facility Can Be Safely Operated for One 18-month Fuel Cycle in Present Configuration ML20212F9421986-12-31031 December 1986 Safety Evaluation Supporting Amend to License DPR-59, Changing Tech Specs Re Second Level of Undervoltage Protection ML20214R5871986-11-24024 November 1986 Safety Evaluation Accepting Util Actions to Ensure Structural Integrity of Vacuum Breakers in Mark I Containments ML20210T3101986-10-0202 October 1986 Safety Evaluation Accepting Util 860228 Submittal of Rev 2 to Offsite Dose Calculation Manual on Interim Basis ML20205E3601986-08-0606 August 1986 Safety Evaluation on Util 830806,1109 & 840330 Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1. Licensee Complied W/All Items ML20210K5571986-04-18018 April 1986 Safety Evaluation Supporting Util Request for Relief from First 10-yr Inservice Insp Requirements for Class 1,2 & 3 Components ML20137S9901985-09-26026 September 1985 Safety Evaluation Accepting MSIV Leakage Control Sys,Per GDC 54, Piping Sys Penetrating Containment ML20134D2071985-08-0909 August 1985 Safety Evaluation of Util 831109 & 840629 Responses to Generic Ltr 83-28,Item 1.1, Post-Trip Review (Program Description & Procedure). Program & Procedures Acceptable ML20133F0011985-07-30030 July 1985 Safety Evaluation Accepting Util 831109 & 840629 Responses to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing ML20129E1721985-07-0101 July 1985 Safety Evaluation Re Radiological Consequences of Hypothetical LOCA While Purging Containment at Power. Radiological Consequences Acceptable ML20129E1411985-07-0101 July 1985 Safety Evaluation Supporting Demonstration of Containment Purge & Vent Valve Operability.Info Submitted Demonstrates Ability of 20-inch & 24-inch Purge & Vent Valves to Close Against Buildup of Containment Pressure During Dba/Loca ML20127E7221985-06-17017 June 1985 SER Supporting Util 840629 Response to Generic Ltr 84-09, Recombiner Capability Requirements of 10CFR50.44(c)(3)(ii) ML20127B3251985-06-10010 June 1985 Interim Safety Evaluation Approving Util 830630 Procedures Generation Package (PGP) for Emergency Operating Procedures Upon Resolution of Exceptions Noted in Section 2.PGP Submitted Per Generic Ltr 82-33 Re Suppl 1 to NUREG-0737 ML20127C7961985-06-0606 June 1985 Safety Evaluation Re Insp & Repair of RCS Piping.Plant Can Be Safely Returned to Operation in Present Configuration for Duration of Cycle 7 ML20140G5731975-07-15015 July 1975 Safety Evaluation Supporting Tech Spec Changes to License DPR-59 to Revise Suppression Pool Water Temp Limits 1999-07-28
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARJAFP-99-0277, Monthly Operating Rept for Sept 1999 Fpr Ja FitzPatrick Nuclear Power Plant.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 Fpr Ja FitzPatrick Nuclear Power Plant.With ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data JAFP-99-0261, Monthly Operating Rept for Aug 1999 for Jafnpp.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Jafnpp.With JAFP-99-0236, Monthly Operating Rept for July 1999 for Ja FitzPatrick Npp. with1999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Ja FitzPatrick Npp. with ML20216D9541999-07-28028 July 1999 Safety Evaluation Authorizing Proposed Alternatives for Second 10-year Interval Pursuant to 10CFR50.55a(a)(3)(ii) ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept JAFP-99-0211, Monthly Operating Rept for June 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Ja FitzPatrick Nuclear Power Plant.With JAFP-99-0175, Annual Summary of Changes,Tests & Experiments for 1997/1998. with1999-06-0202 June 1999 Annual Summary of Changes,Tests & Experiments for 1997/1998. with JAFP-99-0181, Monthly Operating Rept for May 1999 for Jafnpp.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Jafnpp.With JAFP-99-0166, Monthly Operating Rept for Apr 1999 for Ja FitzPatrick Nuclear Plant.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Ja FitzPatrick Nuclear Plant.With JAFP-99-0142, Monthly Operating Rept for Mar 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Ja FitzPatrick Nuclear Power Plant.With JAFP-99-0092, Monthly Operating Rept for Feb 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Ja FitzPatrick Nuclear Power Plant.With ML20207E9311999-02-26026 February 1999 Part 21 Rept Re Sprague Model TE1302 Aluminum Electrolytic Capacitors with Date Code of 9322H.Caused by Aluminum Electrolytic Capacitors.Affected Capacitors Replaced ML20202J0891999-02-0303 February 1999 Safety Evaluation Accepting Rev 2 of Third Interval Inservice Testing Program for Pumps & Valves for James a FitzPatrick Nuclear Power Plant ML20199M0891999-01-22022 January 1999 Part 21 Rept Re Failure of Square Root Converters.Caused by Failed Aluminum Electrolytic Capacitory Spargue Electric Co (Model Number TE1302 with Mfg Date Code 9322H).Sent Square Root Converters Back to Mfg,Barker Microfarads,Inc JAFP-99-0011, Monthly Operating Rept for Dec 1998 for Ja FitzPatrick Nuclear Power Plant.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Ja FitzPatrick Nuclear Power Plant.With ML20198F9991998-12-0404 December 1998 Assessment of Licensing Basis for Use of Containment Overpressure Credit for Net Positive Suction Head Analyses Power Authority of State of New York,James a Fitzpatrick Nuclear Power Plant ML20196J3501998-12-0404 December 1998 SER Accepting License Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves JAFP-98-0396, Monthly Operating Rept for Nov 1998 for Ja FitzPatrick Npp. with1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Ja FitzPatrick Npp. with ML20196F9251998-11-25025 November 1998 Safety Evaluation Re Thrid 10-year Interval Inservice Insp Program Relief Requests for Plant ML20195J7521998-11-18018 November 1998 Rev 7 to Jaf Colr ML20195K4211998-11-17017 November 1998 Safety Evaluation Authorizing Proposed Alternative in Relief Request VRR-05 Per 10CFR50.55a(a)(3)(i) & PRR-01,PRR-02R1, PRR-03,PRR-04,VRR-02,VRR-03 & VRR-04 Per 10CFR50.55a(a)(3)(ii) ML20195E1051998-11-13013 November 1998 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves, Issued 950817 ML20197G6221998-11-0606 November 1998 Non-proprietary Rev 7 to HI-971661, Licensing Rept for Reracking of Ja FitzPatrick Sfp ML20155H5321998-11-0303 November 1998 Safety Evaluation Authorizing Alternative to ASME Code Requirements for CRD Bolting ML20155H5801998-11-0303 November 1998 Safety Evaluation Authorizing Postponement of Beginning of Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) Re ASME Code,Section XI JAFP-98-0360, Monthly Operating Rept for Oct 1998 for Ja FitzPatrick Nuclear Power Plant.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Ja FitzPatrick Nuclear Power Plant.With ML20155C2821998-10-30030 October 1998 Non-proprietary Rev 0 to GENE-187-30-1598 Np, CRD Bolting Flaw Evaluation for Ja FitzPatrick Nuclear Power Plant ML20154L6591998-10-14014 October 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Snubber Visual Inservice Exam Intervals & Sampling Rates Requirements Contained in ASME Code,Section Xi,Subsection Iwf,Article IWF-5000 JAFP-98-0322, Monthly Operating Rept for Sept 1998 for Ja Fitzpatrick Nuclear Power Plant.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Ja Fitzpatrick Nuclear Power Plant.With ML20153D2591998-09-21021 September 1998 SER Accepting Proposed Alternative Testing of Containment Following ECCS Suction Strainer Replacement ML20153B5611998-09-0101 September 1998 Rev 1 to JAF-SE-98-013, RHR & Core Spray Suppression Pool Suction Strainer Replacement ML20151X6891998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Ja FitzPatrick Nuclear Power Plant ML20237E8361998-08-25025 August 1998 Rev 6 to Colr ML20237E9471998-08-0808 August 1998 Rev 6 to Colr JAFP-98-0264, Monthly Operating Rept for July 1998 for Ja FitzPatrick Nuclear Plant1998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Ja FitzPatrick Nuclear Plant ML20236X5881998-07-29029 July 1998 Safety Evaluation Supporting Amend 245 to License DPR-59 ML20236V8181998-07-29029 July 1998 Safety Evaluation Accepting Request for Relief from Implementation of Requirements of 10CFR50.55a Related to Containment Repair & Replacement Activities for James a FitzPatrick Nuclear Power Plant ML20153B5781998-07-28028 July 1998 Rev 0 to JAF-SE-98-025, HPCI & RCIC Suppression Pool Suction Strainer Replacement ML20236X3831998-07-14014 July 1998 Rev 2 to JAF-RPT-MULTI-02671, Summary of Detailed Evaluation for NRC Generic Ltr 96-06 ML20154L9201998-07-10010 July 1998 SER Accepting Rev to Reactor Vessel Surveillance Capsule Withdrawal Schedule for James a Fitzpatrick Nuclear Power Plant JAFP-98-0222, Monthly Operating Rept for June 1998 for Ja FitzPatrick Nuclear Power Plant1998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Ja FitzPatrick Nuclear Power Plant JAFP-98-0193, Monthly Operating Rept for May 1998 for Ja FitzPatrick Nuclear Plant1998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Ja FitzPatrick Nuclear Plant ML20248F3531998-05-21021 May 1998 Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted JAFP-98-0168, Monthly Operating Rept for Apr 1998 for Ja FitzPatrick Nuclear Power Plant1998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Ja FitzPatrick Nuclear Power Plant JAFP-98-0128, Monthly Operating Rept for Mar 1998 for Ja FitzPatrick Nuclear Power Plant1998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Ja FitzPatrick Nuclear Power Plant ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively JAFP-98-0091, Monthly Operating Rept for Feb 1998 for JAFNPP1998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for JAFNPP ML20202G9081998-02-0606 February 1998 Safety Evaluation Re Amend to License DPR-59 to Revise TS Tables 3.2-2 & 4.2-2 JAFP-98-0058, Monthly Operating Rept for Jan 1998 for Ja FitzPatrick Nuclear Power Plant1998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Ja FitzPatrick Nuclear Power Plant 1999-09-30
[Table view] |
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.p y m& UNITED STATES g j NUCLEAR REGULATORY COMMISSION i WASHINGTON. D.C. 20556-0001
. . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION INSERVICE TESTING PROGRAM REllEF REQUESTS POWER AUTHORITY OF THE STATE OF NEW YORK !
i JAMES A. FITZPATRICK NUCLEAR POWER PLANT l DOCKET NUMBER 50-333
1.0 INTRODUCTION
i l
The Code of Federal Reaulations,10 CFR 50.55a, requires that inservice testing (IST) of certain American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 pumps and valves be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (ASME Code) and applicable addenda, except where alternatives have been authorized or relief has been requested by the licensee and granted by the Commission pursuant to Sections (a)(3)(i), (a)(3)(ii), or (f)(6)(i) of 10 CFR 50.55a. In proposing alternatives or requesting relief, the licensee must demonstrate that: (1) the proposed alternatives provide an acceptable level of quality and safety; (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for its facility. Section 50.55a authorizes the Commission to approve alternatives and to grant relief from ASME Code requirements upon making the necessary findings.
Guidance related to the development and implementation of inservice testing programs is given in Generic Letter (GL) 89-04," Guidance on Developing Acceptable Inservice Testing Programs," issued April 3,1989, and its Supplement 1 issued April 4,1995. Guidance is also provided in NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power Plants," and NUREG/CR-6396," Examples, Clarifications, and Guidance on Preparing Requests for Relief from Pump and Valve inservice Testing Requirements."
The 1989 Edition of the ASME Code is the latest edition incorporated by reference in Paragraph (b) of Section 50.55a. Subsection IWV of the 1989 Edition, which gives the requirements for IST of valves, references Part 10 of the American National Standards Institute /ASME Operations and Maintenance Standards (OM-10) as the rules for IST of valves. OM-10 replaces specific requirements in previous editions of Section XI, Subsection IWV, of the ASME Code. Subsection IWP of the 1989 Edition, which gives the requirements for IST of pumps, references Part 6 of the American National Standards Institute /ASME Operations and Maintenance Standards (OM-6) as the rules for IST of pumps. OM-6 replaces specific requirements in previous editions of Section XI, Subsection IWP, of the ASME Code.
On November 2,1998, the Power Authority of the State of New York (the licensee, also known as the New ' York Power Authority) submitted the Inservice Testing Program for Pumps and Valves, Third Interval Plan, Revision 2 for the James A. FitzPatrick Nuclear Power Plant.
Included in the submittaiis one new Relief Request, PRR-06, and one revised Relief Request, VRR-06R1. VRR-06 was originally submitted in the October 21,1997, third ten-year IST program update, but was retracted by the licensee's July 30,1998 letter. The staff's 9902090036 990203 PDR ADOCK 05000333 P PDR Enclosure
)
l November 17,1998, safety evaluation (SE) of the licensee's original IST program for the third interval (submitted on October 21,1997) did not evaluate this relief request. Additionally, one relief request that was evaluated in the staff's SE, VRR-05, has been deleted in Revision 2.
The other changes in Revision 2 are mainly editorial and do not affect the conclusions of the staff's November 17,1998 SE. l 2.0 EVALUATION j 2.1 Pumo Relief Reauest PRR-06 For the residual heat removal service water (RHRSW)/ emergency service water (ESW) system !
pumps (10P-1 A,10P-1B,10P-10,46P-2A, and 46P-2B), the licensee requasts relief from the requirements of OM-6 Section 4.6.1.1. This section specifies that instruments used for ;
pressure measurement be accurate to within 2% of the full-scale reading on the instrument. i 1
2.1.1 Licensee's Basis for Reauestina Relief The licensee states:
The RHRSW and ESW pumps are of a vertical submerged open line shaft design. There is no installed instrument for direct measurement of the inlet pressure. Instead, the i minimum pumping levelis monitored to ensure adequate NPSH [ net positive suction head) is available for pump operation. Since the forebay water level is not expected to change significantly during the testing of these pumps, only one measurement per test is required.
During each test, the difference in elevation between the forebay water level and the pump discharge pressure gauge will be determined by measurement. This value will be verified to be less than or equal to the value corresponding to the minimum water level required for pump operation and will also be used to calculate pump differential pressure. This calculation method is in accordance with OM-6, Section 4.6.2.2, and NUREG-1482, Section 5.5.3.
Due to limitations of human factors related to measuring the elevation between the forebay water level and the pump discharge pressure gauge, the accuracy of differential pressure calculation (s) cannot be verified to within i2% as required by the Code.
2.1.2 Attemate Testina The licensee proposes:
In accordance with the guidance provided in NUREG-1482, Section 5.5.3, Differential Pressure for the RHRSW and ESW pumps will be measured as follows:
For each pump, the pump correction value will be determined by measuring the difference in elevation between the forebay water level and the pump discharge pressure gauge, and then calculated in accordance with the procedure. The discharge pressure of the pump will be recorded and then added to the pump correction value to determine the Differential Pressure. This value will be recorded during the performance of each test and then
( evaluated in accordance with analysis and evaluation criteria specified in OM-6, Section 6, as applicable.
2.1.3 Evaluation i The RHRSW/ESW pumps provide cooling water for safety-related heat loads during a design basis loss-of-coolant accident. OM-6, Section 4.6.1.1 requires that the instruments used for )
pressure measurement be accurate to within t2% of the full-scale reading on the instrument.
Pressure gauges are not installed in the inlet of these vertical line shaft pumps. So, instead of directly measuring the inlet pressure for use in determining pump differential pressure, the licensee proposes to use a calculational method.
Guidance on using forebay level to calculate differential pressure is contained in NUREG-1482, Section 5.5.3 and NUREG/CR-6396, Section 3.2. The staff has determined that, if the licensee uses a bay level to calculate the inlet pressure, the calculation should be included in the implementing procedure. The proposed attemate testing method meets this criterion. i However, the guidance also states that the licensee should verify that the reading scale for measuring the level and the calculational method yield an accuracy within *2% The licensee's
- statement that "[d]ue to limitations of human factors related to measurlag the elevation between i j the forebay water level and the pump discharge pressure gauge, the accuracy of differential i j pressure calculation (s) cannot be verified to within *2% as required by the Code" does not provide sufficient justification for granting relief based on impracticality. In Appendix A to NUREG 1482, Response to Question Group 105, the staff stated that in determining what is ,
" practical within the limitation of design, geometry, and materials of construction of the l component," the staff considers modifications such as the instal!ation of instrumentation to be practical as used in 10 CFR 50.55a(f)(4). Therefore, the licensee should either comply with the Code or develop and justify another method of evaluating the hydraulic performance of these pumps.
2.1.4 Conclusion l
- The proposed alternative is authorized for an interim period of 1 year pursuant to 10 CFR 50.55a(a)(3)(ii) based on the determination that immediate compliance with the requirements would result in a hardship without a compensating increase in the level of quality and safety, given the assurance of operational readiness provided by the proposed alternative during this interim period. This interim period will provide the licensee sufficient time to adequately address differential pressure calculational accuracy. The licensee should either 1) resubmit this relief request with a more detailed description of how the human factors considerations preclude meeting the Code accuracy requirement and provide an indication of the differential pressure calculation accuracy that can reasonably be achieved with the installed system configuration, or 2) install instrumentation to directly measure the pump inlet pressure to the level of accuracy required by the Code.
2.2 Valve Relief Reauest VRR-06R1 The licensee has requested relief from the requirements of OM-10, Section 4.2.1.4 for the l service water / emergency service water valves,70TCV-120A,70TCV-1208,70TCV-121 A, L
1 _
4 70TCV-121B,67PCV-101. This section of the Code describes the stroke testing requirements for power operated valves.
2.2.1 Licensee's Basis for Reouestina Relief The licensee states:
These valves have no position indication or manual control switches. Valve operation is controlled by temperature switches or pressure controllers. Stoke timing these valves would be extremely difficult and require an abnormal system configuration to obtain consistent stroke time results. Performing a stroke time test of these valves is impractical without a compensating level of quality and safety. !
2.2.2 Alternate Testina The licensee proposes:
In accordance with the guidance provided in NUREG-1482 adequate assessment of the 1 operational readiness of these valves is achieved as follows: !
All valvca are fail safe tested on a quarterly frequency. Prior to the test the valves are verified to not be in the full open position. During conduct of the test the valve air or 4 electrical control is interrupted and the valve operation is observed locally to verify proper operation and movement to the fait safe full open position.
Valves 70TCV-121 A,B are also stroked once per operating cycle per Technical Specification 4.11.B.2 during the calibration of their associated. instrumentation control loop.
Valves 70TCV-120A,B are also stroked once per operating cycle during the calibration of their associated instrumentation control loop.
2.2.3 Evaluation The normal function of the temperature control valves 70TCV-120A,B and 70 TOV-121 A,B are to modu! ate the flow of chilled water in order to maintain appropriate air temperature and relative humidity in the Operations Office, Control Room, and Relay Room. The safety function of the valves is the same except that failure of the valve actuator mechanism results in valve movement to the maximum cooling water flow position.
The normal function of valve 67PCV-101 is to maintain a backpressure at the common service
, water return header for the cable tunnel and electric bay coolers. The safety function of this valve is to fail open on loss of air.
Control valves that perform a safety function are required to be tested in accordance with the Code requirements so that the valves can be monitored for degrading conditions. Paragraph 4.2.1.4 of OM-10 requires that power-operated valves be stroke tested. A limiting value of ful!-
stroke time of each valve will be specified, and the stroke time will be measured to the nearest
5-1 second and compared to the limiting value. Any abnormalities will be evaluated for possible corrective actions.
These valves have no position indication or manual control switches. It is possible to stroke time the valves. However, it would be very difficult and require an abnormal system configuration. This would result in a hardship for the licensee and doing so may compromise a level of quality and safety. Instead of conforming to Code-required testing, the licensee proposes an attemative that is consistent with the guidance in NUREG-1482, Section 4.2.9.
In NUREG-1482, the staff recommends that licensees investigate attematives that include stroke-timing with acoustic or other nonintrusive methods, stroke timing with local observation or obuvation of system conditions, enhanced maintenance with a periodic stroke which may not be timed, stroke timing and fail safe testing during cold shutdowns or refueling outages that involve bypassing control signals and a control system signal calibration to verify the stroke times of the valves. The licensee's proposed alternative to quarterly fail-safe test the valves while locally observing valve operation, along with stroking the valves once per operating cycle during a calibration of their associated instrument controlloop, should provide reasonable assurance of the valves' operational readiness.
2.2.4 Conclusion The proposed alternative to the requirements of OM-10 Paragraph 4.2.1.4 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii). Compliance with the specific requirements of this section would result in hardship without a compensating increase in the level of quality and safety.
3.0
SUMMARY
The proposed alternative described in Relief Request VRR-06R1 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii), on the basis that compliance with the specific Code requirements would -
result in unusual difficulty without a compensating increase in the level of quality and safety.
The proposed alternative described in Relief Request PRR-06 is authorized for an interim period of 1 year from the date of the letter forwarding this SE pursuant to 10 CFR 50.55a(a)(3)(ii), based on the determination that immediate compliance with the requirements would result in a hardship without a compensating increase in the level of quality and safety.
After that time, the licensee should either 1) resubmit this relief request with a more detailed description of how the human factors considerations preclude meeting the accuracy requirements of the ASME Code, and provide an indication of the differential pressure calculation accuracy that can reasonably be achieved with the installed system configuration, or
- 2) install instrumentation to directly measure the pump inlet pressure to the level of accuracy required by the ASME Code.
Principal Contributor: M. Kotzalas Date: February 3, 1999
. .- o