L-09-711, Response to Request for Additional Information Related to Inservice Testing Program Requests in Support of the Third Ten-Year Interval
| ML091560034 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 05/28/2009 |
| From: | Bezilla M FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-09-7117, TAC ME0191, TAC ME0195, TAC ME0196, TAC ME0197, TAC ME0198 | |
| Download: ML091560034 (16) | |
Text
FENOC,. Perry Nuclear Power Station 10 Center Road FirstEnergyNuclear OperatingCompany Perry, Ohio 44081 Mark B. Bezilla 440-280-5382 Vice President Fax: 440-280-8029 May 28, 2009 L-09-7117 10 CFR 50.55a
,ATTN:'Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Perry.Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Response to Request for Additional Information Related to InseNvice Testing Program Requests in Support of the Third Ten-Year Interval (TAC Nos. ME0191, ME 0195, ME0196, ME0197, and ME0198)
By correspondence dated November 18, 2008, FirstEnergy Nuclear Operating Company (FENOC) submitted eight separate proposed alternatives to requirements associated with the Perry Nuclear Power Plant Inservice Testing Program. By letter dated March 20, 2009, the Nuclear Regulatory Commission (NRC) staff requested additional information to complete its review. The attachment provides responses to the NRC staffs questions, as modified during a teleconference between FENOC and NRC staff on May 7, 2009. Enclosed is 10 CFR 50.55a Request VR-4, modified as discussed during teleconferences with NRC staff on January 29 and February 6, 2009.
Additionally, FENOC has also elected to withdraw 10 CFR 50.55a Request VR-5.
There are no regulatory commitments contained in this submittal. If there are any questions or additional information is required, please contact Mr. Thomas A. Lentz, Manager -..Fleet Licensing, at (330) 761-6071.
Sincerely,
<Mar B.-*Bez~ti& fp17kjT 14 ZF..9ZIAje. PLvT" S
Attachment:
Response to Request for Additional Information Related to Inservice Testing Program Requests in Support of the Third Ten-Year Interval
Enclosure:
10 CFR 50.55a Request VR-4, Revision 0 cc: NRC Region III Administrator , -7 NRC Resident Inspector NRC Project Manager KA4
- I-I..
Attachment 1 L-09-117 Response to Request for Additional Information Related to Inservice Testing Program Requests in Support of the Third Ten-Year Interval Page 1 of 8 The Nuclear Regulatory Commission (NRC) staff has requested additional information regarding several of the proposed alternatives to requirements associated with the Perry Nuclear Power Plant (PNPP) Inservice Testing Program. The FirstEnergy Nuclear Operating Company (FENOC) responses for PNPP are provided below. The NRC staff's questions are presented in bold, followed by FENOC's responses.
Request PR-I PR-I-1 What are the rated flows and discharge pressures for the waterleg pumps?
Response
All four waterleg pumps listed in Request PR-I are identical. Per the waterleg pump manufacturer's specifications, rated flow at 75 feet of head is 40 gallons per minute (gpm).
Note that in the PNPP keepfill (recirculation) application, pump flow rates are approximately 10 gpm with a discharge pressure of approximately 45 to 48 pounds per square inch (psig), when pump suction is aligned to the suppression pool. When pump suction is aligned to the condensate storage tank (High Pressure Core Spray and Reactor Core Isolation Cooling Waterleg Pumps only), pump flow rates are approximately 10 gpm with a discharge pressure of approximately 66 to 69 psig due to additional net positive suction head available (NPSHA). As described in Request PR-i, full flow can only be achieved at PNPP by realigning the discharge path, which is only done during cold shutdown.
PR-1-2 Are there annunciator alarms in the control room that alarm if the pump discharge header pressure drops below a preset value? If so, at what percentage of the pumps' differential pressure do they alarm?
Response
Yes. For the systems serviced by the waterleg pumps, an annunciator alarms in the control room to alert reactor operators of a low pressure condition indicative of a waterleg pump malfunction or any other condition that allows pressure to degrade (e.g.,
excessive leakage beyond waterleg pump make-up capabilities). Operator actions following receipt of low pressure alarms include alarm validation, verifying whether the waterleg pumps are running, and initiating a system fill and vent, as necessary.
As discussed during a May 7, 2009 teleconference, the NRC staff no longer needs information relative to the pumps' differential pressure.
L-09-117 Page 2 of 8 PR-1-3 Is there a Technical Specification Surveillance Requirement to periodically verify that the headers are filled with water by venting the piping at high point vents?
Response
As discussed during a May 7, 2009 teleconference, the piping for each required Emergency Core Cooling System (ECCS) injection/spray subsystem is periodically verified to be adequately filled with water, and that having an adequately filled piping system is necessary to meet the Technical Specifications (TS) definition of OPERABILITY. Currently, a Surveillance Requirement (SR) states to verify the piping is filled with water from the pump discharge valve to the injection valve. Discussions are ongoing with the NRC through resolution of Generic Letter (GL) 2008-01 to determine necessary TS controls in the future.
PR-1-4 Are the waterleg pumps currently being monitored in a vibration monitoring program? What are the reference values of vibration velocity for these pumps?
Response
Yes. Vibration monitoring of the waterleg pumps is performed quarterly while in the keepfill (recirculation) application and under full flow conditions when the plant is in cold shutdown.
The waterleg pumps' reference values of vibration velocity are as follows:
1E12-C003, Residual Heat Removal (RHR) B & C Waterleg Pump Quarterly Cold Shutdown Horizontal velocity: 0.070 in/sec 0.061 in/sec Vertical Velocity: 0.080 in/sec 0.046 in/sec Axial Velocity: 0.070 in/sec 0.041 in/sec 1E21 -C002, Low Pressure Core Spray (LPCS) & RHR A Waterleg Pump Quarterly Cold Shutdown Horizontal velocity: 0.040 in/sec 0.040 in/sec Vertical Velocity: 0.030 in/sec 0.040 in/sec Axial Velocity: 0.033 in/sec 0.020 in/sec 1E22-C003, High Pressure Core Spray (HPCS) Waterleg Pump Quarterly Cold Shutdown Horizontal velocity: 0.040 in/sec 0.070 in/sec Vertical Velocity: 0.052 in/sec 0.040 in/sec Axial Velocity: 0.034 in/sec 0.040 in/sec L-09-117 Page 3 of 8 1E51-C003, Reactor Core Isolation Cooling (RCIC) Waterleg Pump Quarterly Cold Shutdown Horizontal velocity: 0.060 in/sec 0.040 in/sec Vertical Velocity: 0.052 in/sec 0.026 in/sec Axial Velocity: 0.086 in/sec 0.022 in/sec Request VR-3 VR-3-1 Licensee reason for request states "Code Case OMN-1 has been determined by the NRC to provide an acceptable level of quality and safety when implemented in conjunction with the conditions imposed in RG 1.192. Since the NRC staff recommends licensees implement ASME Code Case OMN-1, PNPP proposed to implement Code Case OMN-1, Revision 1 in lieu of the stroke-time provisions specified in ISTC-5120 for MOVs as well as the position verification testing in ISTC-3700."
The staff review of ASME OM Code cases noted that OMN-1 was first published with the 1995 edition of the code. Minor changes were made when it was re-published with the 2002 addenda. OMN-1 was later revised in its entirety in the 2006 addenda publication. However, the title of OMN-1 in any of the code case publication years has no reference to current revision number status. Does PNPP's request to apply Code Case OMN-1 Revision 1 refer to the version of OMN-1 published with the 2001 edition through 2003 addenda, or the major revision of OMN-1 published with the 2006 addenda?
Response
The code case requested for use is OMN-1, as adopted by the OMb-2006 Addenda of the Code.
Request VR-4 VR-4-1 Perry relief request VR-4 proposes to use a performance-based testing program for various-valves. The performance-based testing program is patterned on an approved methodology that was developed for Containment Leak Rate testing. 10 CFR Part 50 Appendix J details the testing requirements for Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. There are two options in Appendix J for meeting the containment leakage test requirements.
Option A is a prescriptive requirement and Option B is a performance-based requirement. Nuclear Regulatory Guide 1.163 "Performance-Based Containment Leak-Test Program" approved the use of the Option B method. Are the valves listed in VR-4 considered to be applicable to 10 CFR Part 50 Appendix J? If not, please explain why the Appendix J Option B methodology and the supporting
. . . .; ; * -:V.,. :. * ..... . 1.*i**
L-209-117 Page 4 of 8 analysis for Containment System Leakage Testing Requirements would also apply to valves that are not considered part of the'Appendix J scope.
Response
The majority of the valves listed within Request VR-4 are not designated as containment isolation valves, but the concepts of a performance-based program for valve leakage testing is still applicable to each of the valves listed therein.
The concepts that led to creation of Appendix J Option B are described in NUREG-1493, "Performance-Based Containment Leak-Test Program." Section 4, "Leakage Rate Test Experience" explains that the examination of leak-tested component performance indicates that excessive leakages are more frequent early in plant life and decrease with time. The reason for the observed behavior is generally understood - when repeated failures of certain components are observed, the problems are remedied by changing design, materials, or replacing the troublesome component with a different design, or improved repair procedures. After such deficiencies are corrected, subsequent failures are governed by random failure rates until the component reaches the wear-out portion of its life. The NUREG notes that performance-based testing alternatives that are predicated on components passing two successive tests before extending the testing interval (such as is the case with the proposed PNPP ASME Performance-Based Testing Program) will minimize testing of good performers and will thus focus on those components that suffer some kind of deficiency or reach wear-out.
For the valves that are leak tested per both Appendix J and ASME Code Category A or AC requirements, the concepts of the Appendix J methodology apply. Valves that exhibit consistently acceptable leakage rates when tested per both programs are allowed to have their test frequency extended per Appendix J, but are not currently allowed extensions per the Code. Approval of Request VR-4 would eliminate this dichotomy.
For the remaining valves that only receive leak tests per Code requirements, the concept that performance-based testing is acceptable, for the reasons described above, equally applies. For these remaining valves it was determined that if performance-based testing was acceptable for 10 CFR 50 Appendix J leak tested valves, which maintain containment integrity and offsite dose release rates within those specified in the regulations, then a similar performance-based methodology would be acceptable for valves that are leak tested for purposes other than containment isolation (with the exception of Technical Specification required pressure isolation valves (PIVs), which are not included in the scope of Request VR-4).
L-09-117 Page 5 of 8 Request VR-5 VR-5-1 The relief request proposes timing the slowest valve in a group to demonstrate functionality of all valves in the group. If the slowest valve exceeds the maximum stroke time of two seconds, then the valves will be re-timed to ensure the failure is properly attributed to the correct valve. ISTC-5153(b) requires that "Valves with measured stroke times that do not meet the acceptance criteria of ISTC-5152 shall be immediately retested or declared inoperable... If the second set of data meets the acceptance criteria, the cause of the initial deviation shall be analyzed and the results documented in the record of tests." Please describe plans for positively identifying which individual valve or valves exceeded the limiting stroke time during the original test, so the cause of the initial deviation can be determined as required.
Response
FENOC has chosen to maintain the existing valve stroke time testing methodology. As a result, FENOC has elected to withdraw PNPP 10 CFR 50.55a Request VR-5 and a response to this question is no longer necessary.
Request VR-6 VR-6-1 The ASME Code has developed a Code Case addressing test frequencies of ASME Class 1 pressure relief/safety valves. The Code Case provides a 72 month test interval with a 6 month grace period to accommodate extended shutdown periods provided certain requirements are implemented. One of the requirements is that each valve is disassembled and inspected after as-found set pressure testing to verify that parts are free of defects resulting from time related degradation or service induced wear. Based on this inspection, the owner shall determine the need for additional inspections or testing to address any generic concerns. Please discuss the feasibility of implementing the criteria established by the ASME OM Code for extending the SRV test interval beyond 5 years.
Response
As described in Updated Safety Analysis Report (USAR) Section 5.2.2.10, following as-found set pressure determination, the safety/relief valves are externally inspected. The valves are then disassembled to inspect internal surfaces and parts for wear, damage, or erosion. During this process, anomalies or damage are identified and dispositioned for resolution. Damaged or worn parts, and gaskets and seals are replaced as necessary. The valves are lubricated and the valve seats relapped, if necessary.
Following reassembly, the valve's set pressure is recertified.
L-09-117 Page 6 of 8 This existing process aligns with the described requirements of the 72-month test interval. Additionally, extending the test interval to 72 months better aligns with the current PNPP 24-month operating cycles, reduces the number of valves removed and tested each outage, and reduces cumulative occupational radiation exposure.
VR-6-2 The relief request states that four as-found setpoint tests exceeded the Code tolerance of plus or minus 3%. Please provide the as-found setpoint data for the four failed tests, identify the cause of the failure if known, and identify any corrective actions taken to improve valve performance following the failed tests.
Response
In PNPP's first refueling outage (1989), valve 1B21-F047D (serial number 160874) lifted at 1139 PSIG, approximately 3.4% below its set pressure setpoint of 1180 PSIG. An investigation did not reveal any other equipment anomalies or damage. The cause was attributed to setpoint drift. The valve was subsequently rebuilt and recertified. No additional corrective actions were taken.
In PNPP's third refueling outage (1992), valve 1B21-F051D (serial number 160881) had experienced excessive seat leakage during the preceding operating cycle and its set pressure setpoint was unable to be determined due to the seat leakage. An investigation did not reveal any other equipment anomalies or damage. The valve had been cycled many times during the operating cycle, potentially exacerbating the seat leakage. The valve was subsequently rebuilt and recertified. No additional corrective actions were taken.
In PNPP's fourth refueling outage (1994), valve 1B21-F047H (serial number 160875) had experienced excessive seat leakage during the preceding operating cycle and its set pressure setpoint was unable to be determined due to the seat leakage. An investigation did not reveal any other equipment anomalies or damage. The valve had been cycled several times during start-up, potentially exacerbating the seat leakage.
The valve was subsequently rebuilt and recertified. No additional corrective actions were taken.
In PNPP's fourth refueling outage (1994), valve 1B21-FO51G (serial number 160863) had experienced excessive seat leakage during the preceding operating cy(le and its set pressure setpoint was unable to be determined due to the seat leakage. An investigation did not reveal any other equipment anomalies or damage. The valve had been cycled several times during start-up, potentially exacerbating the seat leakage.
The valve was subsequently rebuilt and recertified. No additional corrective actions were taken.
Since 1994 (RFO4), PNPP has used a different vendor to test and refurbish the valves; and since 2005 (RFO10), uncoupled actuator testing that eliminates stroking the valves during post-refueling outage plant start-ups has been implemented. Both of these actions have resulted in improved seat leakage performance of the valves.
L-09-117 Page 7 of 8 VR-6-3 Please provide a summary of the SRV testing conducted for the last 5 refueling outages (valve, date tested, as-found setpoint, interval since previous test).
Response
The last five refueling outages, with outage/testing time periods include:
RFO7 3/27/99 through 5/3/99 RFO8 2/17/01 through 3/23/01 RFO9 4/5/03 through 5/31/03 RFO10 2/22/05 through 5/6/05 RFO11 4/2/07 through 5/13/07 The following table provides the requested information. Serial numbers of individual valves have been provided in the table because PNPP replaces a portion of the safety/relief valves each refueling outage with pretested valves. Each installed valve is currently removed at least once per 5-year interval, tested, inspected, and recertified during the same outage period. Valve as-found set pressure results for refueling outage RFO12 are not included; however, none of the nine valves removed and tested exceeded the +/- 3 percent setpoint tolerance.
L-09-117 Page 8 of 8
SUMMARY
OF TEST RESULTS - MAIN STEAM SAFETY / RELIEF VALVES DESIGN SET AS- AS- AS- AS- AS-LOCATION PRESSURE RFO7 FOUND RFO8 FOUND RFO9 FOUND RFO10 FOUND RFO11 FOUND 1B21-F041A 1165 160883 N/A 160883 1172 160846 N/A 160846 1162 160900 N/A 1B21-F041B 1165 160848 1188 160871 N/A 160871 1179 160848 N/A 160848 1148 1B21-F041C 1165 160866 N/A 160866 1163 160885 N/A 160885 1175 160885 N/A 1B21-F041D 1165 160886 1169 160849 N/A 160849 1168 160886 N/A 160886 1171 1B21-F041E 1165 160887 N/A 160887 1150 160850 N/A 160850 1157 160850 N/A 1B21-F041F 1165 160869 1152 160851 N/A 160851 1170 160869 N/A 160869 1174 1B21-F041G 1165 160865 N/A 160865 1163 160852 N/A 160852 1152 160852 N/A 1B21-F041K 1165 160889 1163 160888 N/A 160888 1190 160889 NT 160888 '-1169 1B21-F047B 1180 160872 1183 160853 N/A 160853 1190 160891 N/A 160891 1182 1B21-F047C 1180 160895 N/A 160895 1173 160854 N/A 160854 1187 160854 N/A 1B21-F047D 1180 160890 1174 160894 N/A 160894 1198 160890 N/A 160890 1175 1B21-F047F 1180 160873 1170 160856 N/A 160856 1196 160873 N/A 160873 1182 1B21-F047G 1180 160876 N/A 160876 1173 .160893 N/A 160893 1176 160893 N/A 1B21-F047H 1180 160877 1171 160875 N/A 160875 1168 160870 N/A 160870 1191 1B21-F051A 1190 160878 N/A 160878 1200 160859 N/A 160859 1211 160878 N/A 1B21-F051B 1190 160860 1186 160882 N/A 169882 1192 160860 N/A 160860 1186 1B21-F051C 1190 160880 N/A 160880 1181 160861 N/A 160861 1182 160861 IN/A 1B21-F051D 1190 160896 1196 160862 1194 160896 1192 160857 N/A 160857 1167 1B21-F051G 1190 160899 N/A 160899 1200 160863 N/A 160863 1190 160899 N/A NOTES:
Listed set pressure and as-found pressure values are in pounds per square inch (PSIG)
N/A listed when valve was not required to be removed for testing NT listed when valve was not required to be tested, but was removed and replaced for other reasons (e.g., in RFO10, valve 160889 was not scheduled for testing, but was removed due to known seat leakage and replaced with valve 160888)
i'~~2 ~
Perry Nuclear Power Plant Unit 1 10 CFR 50.55a Request VR-4, Revision 0 Page 1 of 7 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)
--Alternative' Provides Acceptable Level of Quality and Safety--
- 1. ASME Code Component(s) Affected 1B21-F024A, B, C, & D - Inboard Main Steam Isolation Valve (MSIV) Accumulator Supply Check Valves (Class 3) 1B21-F029A, B, C, & D - Outboard MSIV Accumulator Supply Check Valves (Class 3) 1P57-F555A & F556A - Safety-Related Air "A" Accumulator Supply Check Valves (Class 3) 1 P57-F555B & F556B - Safety-Related Air "B" Accumulator Supply Check Valves (Class 3) 1P57-F572B & F574B - Outboard MSIV Accumulator Normal Supply Check Valves (Class 3) 1B21-RO11 A-F, RO11A-G, RO11B-F, RO11B-G, RO11C-F, RO11C-G, RO11D-F, &
R01 1D-G - Reactor Vessel Reference Level Backfill Supply Check Valves (Class 2) 1 D23-FO1 A & B, 1D23-FO20A & B, 1 D23-FO30A & B, 1 D23-FO40A & B, 1D23-F050
- Containment Atmosphere Monitoring Instrument Line Isolation Valves (Class 2) 1M17-F055 & 1M17-F065 - Containment Vacuum Relief A & B Instrument Line Isolation Valves (Class 2) 1G43-FO50A & B, 1G43-F060 - Suppression Pool Make-up A, B & C Wet Leg Instrument Line Isolation Valves (Class 2) 1C41-F033A & B - Standby Liquid Control Pump Discharge Check Valves (Class 2) 1E12-F019 - Residual Heat Removal (RHR) Head Spray Inboard Isolation Check Valve (Class 1) 1E12-F023 - RHR Head Spray Outboard Isolation Motor Operated Valve (MOV)
(Class 1) 1E12-FO50A & B - RHR Shutdown Cooling Isolation Check Valves (Class 2) 1E12-F053A & B - RHR Shutdown Cooling Isolation MOVs (Class 2) 1N27-F739A & B, 1 N27-F742A & B - Feedwater Leakage Control System (FWLCS)
Supply Inboard Isolation Check Valves (Class 2) 1N27-F737 & F740 - FWLCS Supply Outboard Isolation MOVs (Class 2)
Page 2 of 7 For Category A or AC valves, seat leakage is limited to a specific maximum amount in the closed position for fulfilling the required function of shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident. These valves are grouped by the safety function requiring a seat leakage limit, including:
- 2) instrumentation leakage isolation (failure of backfill, sensing transmitters, cabinets, etc.)
- 3) high-to-low system interface, not including designated reactor coolant system (RCS) pressure isolation valves (PIVs)
- 4) parallel pump bypass flow
2. Applicable Code Edition and Addenda
ASME OM Code-2001, with Addenda through OMb-2003
3. Applicable Code Requirement
ISTC-3630, "Leakage Rate for Other Than Containment Isolation Valves."
ISTC-3630(a), "Frequency," requires leakage tests to be conducted at least once every 2 years.
4. Reason for Request
A performance-based testing program has been developed for selected Perry Nuclear Power Plant (PNPP) valves that would eliminate prescriptive test frequency requirements and allow test intervals based on system and component performance.
Through its own regulatory improvement program, the Nuclear Regulatory Commission (NRC) staff has institutionalized an ongoing effort to eliminate requirements marginal to safety and reduce regulatory burdens on utilities. A performance-based testing program, utilizing an extended testing interval based on the successful completion of two or more consecutive leakage rate tests, would be consistent with the findings of NUREG-1493, "Performance-Based Containment Leak-Test Program." The conclusions drawn by the NUREG are that if a component does not fail within two operating cycles, further failures are governed by the random failure rate of the component. The NUREG also states that any testing scheme considered should require a failed component to pass at least two consecutive tests before allowing an extended test interval.
The PNPP performance-based testing program for ASME valves requiring non-Appendix J leakage tests was developed in much the same manner as the Option B Program for Appendix J testing, which was permitted by amendment of the Code of Federal Regulations (CFR) on October 26, 1995.
Page 3 of 7 In studies performed in support of the CFR change, it was concluded that performance-based testing is feasible without significant risk (NUREG 1493). Also, EPRI Research Project Report TR-1 04285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," reaffirmed this position by stating that changes in leakage testing frequencies are feasible without significant risk.
The development of this performance-based testing program started with the generation of leakage test histories for each valve considered for inclusion in the program. Then, a review of each valve's test history was conducted to establish if a minimum of two consecutive tests had passed and whether any erratic behavior could be detected. The valves were then placed into a type category (check, globe, gate, etc.) to establish which types may be more prone to failure. By performing this, a direct comparison could be made of like valves in like systems to determine if some of those valves should maintain their original test frequencies, even if they have acceptable test histories. Valves that pass a minimum of two consecutive tests will be put on an extended interval of four years or two refueling cycles, whichever is longer. Any valve not meeting the minimum threshold requirement will be left on a two-year or every refueling cycle test interval until at least two consecutive tests are acceptable. In addition, if a failure occurs on any extended interval valve, the initial test frequency of two years or every refueling cycle must be re-established until two consecutive tests pass.
- 5. Proposed Alternative and Basis for Use Leakage rate testing of Category A and AC valves will be performed in accordance with the PNPP ASME Performance-Based Testing Program. Valves that have met the threshold of passing two consecutive tests without erratic behavior and are not considered to be suspect valves, as described above, will be permitted to be tested every four years or two refueling cycles, whichever is longer. Valves that fail their acceptance criterion will be tested every two years or every refueling cycle until they acceptably pass a minimum of two consecutive tests.
All valves placed on an extended testing interval for seat leakage will still have all other associated code required testing (specifically, exercising and position verification), performed at the required frequency by the Inservice Testing Program.
Using the provisions of this request as an alternative to the leakage test frequency specified by ISTC-3630(a) provides an acceptable level of quality and safety.
Included below is a listing of valve groups to be included in the PNPP ASME Performance-Based Testing Program. Valves affected by this request are categorized by the safety function requiring a seat leakage limit.
Each valve or combination of valves is assigned an operational frequency rating that is indicative of the expected frequency at which the valve would perform its active function of opening or closing. The operational frequency assigned would be the inverse to the expected rate of valve degradation (that is, valves seldom exercised would not be expected to lose their valve seat integrity as rapidly as those valves
Page 4 of 7 exercised more frequently). The operational frequency ratings will be assigned as follows: seldom, infrequent, occasional and frequent.
SELDOM - Maintenance or convenience type valves in which operation is seldom desired or required.
INFREQUENT - Valves in which operation would be expected at a cold shutdown or greater frequency for testing or other evolutions.
OCCASIONAL - Valves in which operation would be expected at a quarterly frequency for testing or other evolutions.
FREQUENT - Valves in which operation is expected during normal plant operation, for reasons other than testing. Valves assigned this rating would be considered for exclusion from the performance-based testing program.
ACCUMULATOR PRESSURE BOUNDARY LEAKAGE Safety-related components might rely upon an accumulator as a pressure source for actuation or backup actuation. The following valves are used in the isolation integrity of the accumulator's pressure boundary.
Inboard Main Steam Isolation Valve (MSIV) Accumulator - INFREQUENT:
Inboard accumulator supply check valves (1B21-F024A, B, C, and D) are an integral portion of the inboard MSIV accumulator pressure boundary. These accumulators are supplied by the nonsafety instrument air system and the supply check valves must close and limit seat leakage to a specific maximum amount upon loss of the air supply. Maintaining the seat leakage below, the specified limit ensures that the accumulator will maintain sufficient pressure for proper cycling of the inboard MSIV.
Seat leakage is currently measured by the feed rate required to maintain test pressure in the test volume.
Outboard Main Steam Isolation Valve (MSIV) Accumulator - INFREQUENT:
Outboard accumulator supply check valves (1B21-F029A, B, C, and D) are an integral portion of the outboard MSIV accumulator pressure boundary. These accumulators are supplied by the nonsafety instrument air system and the supply check valves must close and limit seat leakage to a specific maximum amount upon loss of the air supply. Post-accident, these accumulators are also supplied by the safety-related air system after it is manually initiated, at which point these valves are no longer required to maintain seat leakage to a specific maximum amount.
Maintaining the seat leakage below the specified limit ensures that the accumulator will maintain sufficient pressure for proper cycling of the outboard MSIV until the safety-related instrument air supply can be aligned. Seat leakage is currently measured by the feed rate required to maintain test pressure in the test volume.
Page 5 of 7 Safety-Related Air "A" Accumulator - SELDOM:
Safety-related air "A" accumulator supply check valves (1 P57-F555A and 1P57-F556A) are an integral portion of the safety-related air system pressure boundary.
The accumulator is supplied by the nonsafety-related portion of the safety-related air system and the supply check valves must close and limit seat leakage to a specific maximum amount upon loss of the air supply. Maintaining the seat leakage below the specified limit ensures that the accumulator will maintain sufficient pressure for proper cycling of automatic depressurization system (ADS) safety/relief valves. Seat leakage measurements are currently satisfied by measuring leakage through a downstream telltale connection while maintaining test pressure on one side.
Safety-Related Air "B" Accumulator - SELDOM Safety-related air "B" accumulator supply check valves (1 P57-F555B and 1P57-F556B) are an integral portion of the safety-related air system pressure boundary.
The accumulator is supplied by the nonsafety-related portion of the safety-related air system and the supply check valves must close and limit seat leakage to a specific maximum amount upon loss of the air supply. Maintaining the seat leakage below the specified limit ensures that the accumulator will maintain sufficient pressure for proper cycling of ADS safety/relief valves. Seat leakage measurements are currently satisfied by measuring leakage through a downstream telltale connection while maintaining test pressure on one side.
Additionally, outboard MSIV accumulator normal supply check valves (1 P57-F572B and 1 P57-F574B) are an integral portion of the safety-related air system pressure boundary after manual initiation of safety-related air "B." The outboard MSIV accumulators are supplied by the instrument air system and the supply check valves must close and limit seat leakage to a specific maximum amount upon loss of the air supply. Maintaining the seat leakage below the specified limit ensures that the accumulators will maintain sufficient pressure for proper cycling of ADS safety relief valves, as well as closure force for the outboard MSIVs. Seat leakage measurements are currently satisfied by measuring leakage through a downstream telltale connection while maintaining test pressure on one side.
INSTRUMENTATION LEAKAGE ISOLATION RPV Water Level Instruments Continuous Backfill - INFREQUENT:
Reactor vessel reference level backfill supply check valves (1 B21-RO1 1A-F, R011A-G, R011B-F, R011B-G, R011C-F, R011C-G, R011D-F and R011D-G) provide a makeup flow path from the control rod drive (CRD) system to each of the reactor pressure vessel level sensing line reference legs. This backfill is used to prevent errors in reactor vessel level indication during normal and transient operating conditions by preventing the buildup of non-condensable gasses. The portion of the CRD system which supplies the makeup flow is nonsafety-related and the supply check valves must close and limit seat leakage to a specific maximum
,.j Page 6 of 7 amount upon loss of the CRD system. Maintaining the seat leakage below the specified limit ensures proper reactor vessel level indication and minimizes a potential reactor coolant leakage path. Seat leakage measurements are currently satisfied by measuring leakage through a downstream telltale connection while
- maintaining test pressure on one side.
Containment Atmosphere and Water Level Instrumentation - OCCASIONAL:
Containment atmosphere and water level instrument isolation valves (1D23-FOl0A &
B, F020A & B, F030A & B, F040A & B, 1D23-F050, 1M17-F055, 1M17-F065, 1G43-FO50A & B, and 1G43-F060) isolate instrument lines that are considered closed-loops outside containment. These solenoid isolation valves are normally-open and remotely closed in the unlikely event of a failure of the closed-loop portion (i.e., instrumentation) of the system outside containment in order to limit seat leakage to a specific maximum amount. Maintaining the seat leakage below the specified limit ensures the proper isolation of the primary containment pressure boundary if an instrument line failure were to occur. Seat leakage is currently measured by the feed rate required to maintain test pressure in the test volume.
HIGH-TO-LOW SYSTEM INTERFACE PRESSURE ISOLATION High-to-low system interface PIVs are defined as two normally-closed valves in series that isolate a high pressure liquid system from an attached low pressure liquid system. Valve testing prevents the unlikely condition of excessive seat leakage from causing a system overpressure condition. These valves remain closed during normal plant operation to limit seat leakage to a specific maximum amount.
Maintaining the seat leakage below a specified limit ensures that proper intersystem isolation exists between systems. Reactor coolant system PIVs are not included in this group. Seat leakage is currently measured by the feed rate required to maintain test pressure in the test volume.
Residual Heat Removal (RHR) Head Spray Line - INFREQUENT:
RHR head spray inboard isolation check valve (1E12-F019) and outboard isolation motor operated valve (1 E12-F023) are the head spray line intersystem pressure isolation valves used to prevent over-pressurization of a low pressure safety-related system (RHR).
Residual Heat Removal (RHR) Shutdown Coolinq Return Lines - INFREQUENT:
RHR shutdown cooling isolation check valves (1E12-FO50A & B) and isolation motor operated valves (1E12-F053A & B) are the shutdown cooling return lines intersystem pressure isolation valves used to prevent over-pressurization of a low pressure safety-related system (RHR).
Page 7 of 7 Feedwater Leakage Control System (FWLCS) Supply Lines - INFREQUENT:
FWLCS supply inboard isolation check valves (1 N27-F739A & B, and 1 N27-F742A
& B) and outboard isolation motor operated valves (1 N27-F737 and 1 N27-F740) are supply line intersystem pressure isolation valves used to prevent over-pressurization of low pressure safety-related systems (Low Pressure Core Spray and RHR).
PARALLEL PUMP BYPASS FLOW Systems that require an active isolation of the parallel pump loop to perform the desired safety function may require the isolation feature to maintain minimum system leakage.
Standby Liquid Control (SLC) System Parallel Pump Bypass Flow - OCCASIONAL:
SLC pump discharge check valves (1C41-F033A & B) allow flow of borated coolant to the reactor vessel upon system activation and prevent pump bypass flow upon closure. Maintaining seat leakage below the specified limit ensures minimum pump bypass flow. Seat leakage is currently measured by the feed rate required to maintain test pressure in the test volume.
- 6. Duration of Proposed Alternative The proposed alternative identified in this request shall be utilized during the third ten-year Inservice Testing Program (ISTP) interval, which began May 18, 2009.
- 7. Precedent Perry Nuclear Power Plant, Docket No. 50-440, Safety Evaluation Report (SER) dated August 9, 1999, Safety Evaluation of the Inservice Testing Program Second Ten-Year Interval for Pumps and Valves - Perry Nuclear Power Plant (TAC No.
MA3328). Previously approved as VR-1 0 in the aforementioned SER; refer to Attachment 3. This SER did not authorize relief for Category A and AC Reactor-Coolant System (RCS) Pressure Isolation Valves (PIVs); therefore, these types of valves are not included in this request.
- 8. References
- 1. NUREG-1493, "Performance-Based Containment Leak-Test Program," January 1995.
- 2. EPRI Research Project Report TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals."
- 3. Nuclear Energy Institute (NEI 94-01), "Industry Guidelines for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."