ML14122A197

Relief Requests for the Fourth 10-Year Inservice Testing Interval (TAC Nos. MF2905 Through MF2912 and MF2915)

ML14122A197
Person / Time
Site:	Susquehanna
Issue date:	05/22/2014
From:	Meena Khanna Plant Licensing Branch 1
To:	Rausch T Susquehanna
Whited J, NRR/DORL/LPLI-2
References
TAC MF2905, TAC MF2906, TAC MF2907, TAC MF2908, TAC MF2909, TAC MF2910, TAC MF2911, TAC MF2912, TAC MF2915
Download: ML14122A197 (29)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 r*,1ay 22, 2014

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2-RELIEF REQUESTS FOR THE FOURTH 10-YEAR INSERVICE TESTING INTERVAL (TAC NOS. MF2905 THROUGH MF2912 AND MF2915)

Dear Mr. Rausch:

By a letter dated December 12, 2013, which superseded the previous letter dated October 28, 2013, and as supplemented by letters dated February 26, 2014, and March 31, 2014, PPL Susquehanna, LLC, (the licensee) submitted alternative requests 1 RR-01, 2RR-01, 1 RR-02, 2RR-02, 1 RR-03, 2RR-03, 1 RR-04, 2RR-04, and 1 RR-05 to the Nuclear Regulatory Commission (NRC) for review and approval. In these requests, the licensee proposed alternatives to certain inservice test (1ST) requirements of the 2004 Edition through 2006 Addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), for the 1ST program at Susquehanna Steam Electric Station (SSES), Units 1 and 2, for the fourth 1 0-year 1ST program interval.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(i),

the licensee requested to use the proposed alternatives in 1 RR-01, 2RR-01, 1 RR-02, 2RR-02, 1 RR-03, 2RR-03, and 1 RR-05 on the basis that the alternatives provide an acceptable level of quality and safety. Pursuant to 10 CFR Part 50, Section 50.55a(a)(3)(ii), the licensee requested to use the proposed alternatives in 1 RR-04 and 2RR-04 on the basis that the alternatives provide reasonable assurance that the components are operationally ready and imposing the ASME OM Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The NRC staff determined that the proposed alternatives described in alternative requests 1 RR-01, 2RR-01, 1 RR-03, and 2RR-03 provide an acceptable level of quality and safety. The NRC staff determined that the proposed alternatives, described in alternative requests 1 RR-02, 2RR-02, and 1 RR-05 provide reasonable assurance that valves listed in Tables 2 and 4 of the enclosed Safety Evaluation are operationally ready and imposing the Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff also determined that alternative requests 1 RR-04 and 2RR-04 provide reasonable assurance of operational readiness of pumps and valves subject to the ASME OM Code 1ST, and imposing the Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Accordingly, the NRC staff concludes, as stated in the enclosed Safety Evaluation, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i) for alternative requests 1 RR-01, 2RR-01, 1 RR-03 and 2RR-03, and as set forth in 10 CFR 50.55a(a)(3)(ii) for requests 1 RR-02, 2RR-02, 1 RR-04, 2RR-04, and 1 RR-05 and is in compliance with the ASME OM Code requirements.

Therefore, the NRC staff authorizes the proposed alternatives in requests 1 RR-01, 2RR-01, 1 RR-02, 2RR-02, 1 RR-03, 2RR-03, 1 RR-04, 2RR-04, and 1 RR-05 for the fourth 1ST interval at SSES, Units 1 and 2, which is currently scheduled to begin on June 1, 2014, and end on May 31, 2024. All other ASME OM Code requirements for which relief was not specifically requested and approved in the subject request remain applicable.

If you have any questions, please contact the SSES Project Manager, Mr. Jeffrey A Whited, at jeffrey.whited@nrc.gov or 301-415-4090.

Docket Nos. 50-387 and 50-388

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ Sincerely, Meena K. Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REGARDING RELIEF REQUESTS 1 RR-01, 2RR-01. 1 RR-02, 2RR-02.

1 RR-03. 2RR-03, 1 RR-04. 2RR-04, AND 1 RR-05 ASSOCIATED WITH THE FOURTH 10-YEAR INSERVICE TEST INTERVAL PPLSUSQUEHANNA.LLC SUSQUEHANNA STEAM ELECTRIC STATION. UNITS 1 AND 2 DOCKET NOS. 50-387 AND 50-388

1.0 INTRODUCTION

By a letter dated December 12, 2013, 1 which superseded the previous letter dated October 28, 2013, 2 and as supplemented by letters dated February 26, 2014, 3 and March 31, 2014,4 PPL Susquehanna, LLC, (the licensee) submitted alternative requests 1 RR-01, 2RR-01, 1 RR-02, 2RR-02, 1 RR-03, 2RR-03, 1 RR-04, 2RR-04, and 1 RR-05 to the Nuclear Regulatory Commission (NRC) for review and approval. In these requests, the licensee proposed alternatives to certain inservice test (1ST) requirements of the 2004 Edition through 2006 Addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), for the 1ST program at Susquehanna Steam Electric Station (SSES or Susquehanna), Units 1 and 2, for the fourth 1 0-year 1ST program interval.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(i), the licensee requested to use the proposed alternatives in 1 RR-01, 2RR-01, 1 RR-02, 2RR-02, 1 RR-03, 2RR-03, and 1 RR-05 on the basis that the alternatives provide an acceptable level of quality and safety. Pursuant to 10 CFR Part 50, Section 50.55a(a)(3)(ii), the licensee requested to use the proposed alternatives in 1 RR-04 and 2RR-04 on the basis that the alternatives provide reasonable assurance that the components are operationally ready and imposing the ASME OM Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

2.0 REGULATORY EVALUATION

The fourth 1 0-year 1ST interval at SSES, Units 1 and 2, begins on June 1, 2014, and is currently scheduled to end on May 31, 2024. The applicable ASME OM Code edition and addenda for the fourth 1 0-year 1ST Interval at SSES, Units 1 and 2, is the 2004 Edition through the 2006 Addenda.

1 Agencywide Documents Access and Management System (ADAMS) Accession No. ML133478233.

2 ADAMS Accession No. ML13282A554.

3 ADAMS Accession No. ML14059A084.

4 ADAMS Accession No. ML14090A507.

Based on the above, the NRC staff finds regulatory authority exists for the licensee to request, and the NRC to authorize, the proposed alternatives to the ASME OM Code subject to the evaluation given below.

The regulations in 10 CFR 50.55a(f), "lnservice Testing Requirements," requires, in part, that 1ST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda, except where alternatives have been authorized pursuant to paragraphs (a)(3)(i) or (a)(3)(ii).

In proposing alternatives, a licensee must demonstrate that the proposed alternatives provide an acceptable level of quality and safety as outlined in 10 CFR 50.55a(a)(3)(i), or compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety as outlined in 10 CFR 50.55a(a)(3)(ii).

3.0 TECHNICAL EVALUATION

All of the requested alternatives outlined below are requested for the duration of the fourth 1 0-year 1ST Interval which begins on June 1, 2014, and is currently scheduled to end on May 31, 2024. The applicable ASME OM Code Edition and Addenda for SSES, Units 1 and 2, during the fourth 1 0-year 1ST Interval is the 2004 Edition through the 2006 Addenda.

3.1 Licensee's Alternative Requests 1 RR-01 and 2RR-01 3.1.1 Applicable Code Requirements ASME OM Code Subparagraph ISTC-3522(c), "Category C Check Valves," states, "[i]f exercising is not practicable during operation at power and cold shutdowns, it shall be performed during refueling outages."

ASME OM Code Paragraph ISTC-3700, "Position Verification Testing," states, in part, that

"[v]alves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated."

3.1.2 ASME Code Components Affected The licensee requested using the alternative for the following instrumentation line excess flow check valves (EFCVs):

Table 1 Valve 10 Valve ID System Cat Class (Unit 1)

(Unit 2)

XV141F009 XV241F009 Nuclear Boiler c

1 XV141F070A XV241F070A Nuclear Boiler c

1 XV141F070B XV241F070B Nuclear Boiler c

1 XV141F070C XV241F070C Nuclear Boiler c

1 XV141F070D XV241F070D Nuclear Boiler c

1 XV141F071A XV241F071A Nuclear Boiler c

1 XV141F071B XV241F071B Nuclear Boiler c

1 XV141F071C XV241F071C Nuclear Boiler c

1 Table 1 Valve ID Valve 10 System Cat Class (Unit 1)

(Unit 2)

XV141F071D XV241F071D Nuclear Boiler c

1 XV141F072A XV241F072A Nuclear Boiler c

1 XV141F072B XV241F072B Nuclear Boiler c

1 XV141F072C XV241F072C Nuclear Boiler c

1 XV141F072D XV241F072D Nuclear Boiler c

1 XV141F073A XV241F073A Nuclear Boiler c

1 XV141F073B XV241F073B Nuclear Boiler c

1 XV141F073C XV241F073C Nuclear Boiler c

1 XV141F073D XV241F073D Nuclear Boiler c

1 XV14201 XV24201 Nuclear Boiler c

1 XV14202 XV24202 Nuclear Boiler c

1 XV142F041 XV242F041 Nuclear Boiler c

1 XV142F043A XV242F043A Nuclear Boiler c

1 XV142F043B XV242F043B Nuclear Boiler c

1 XV142F045A XV242F045A Nuclear Boiler c

1 XV142F045B XV242F045B Nuclear Boiler c

1 XV142F047A XV242F047A Nuclear Boiler c

1 XV142F047B XV242F047B Nuclear Boiler c

1 XV142F051A XV242F051A Nuclear Boiler c

1 XV142F051B XV242F051B Nuclear Boiler c

1 XV142F051C XV242F051C Nuclear Boiler c

1 XV142F051D XV242F051D Nuclear Boiler c

1 XV142F053A XV242F053A Nuclear Boiler c

1 XV142F053B XV242F053B Nuclear Boiler c

1 XV142F053C XV242F053C Nuclear Boiler c

1 XV142F053D XV242F053D Nuclear Boiler c

1 XV142F055 XV242F055 Nuclear Boiler c

1 XV142F057 XV242F057 Nuclear Boiler c

1 XV142F059A XV242F059A Nuclear Boiler c

1 XV142F059B XV242F059B Nuclear Boiler c

1 XV142F059C XV242F059C Nuclear Boiler c

1 XV142F059D XV242F059D Nuclear Boiler c

1 XV142F059E XV242F059E Nuclear Boiler c

1 XV142F059F XV242F059F Nuclear Boiler c

1 XV142F059G XV242F059G Nuclear Boiler c

1 XV142F059H XV242F059H Nuclear Boiler c

1 XV142F059L XV242F059L Nuclear Boiler c

1 XV142F059M XV242F059M Nuclear Boiler c

1 XV142F059N XV242F059N Nuclear Boiler c

1 XV142F059P XV242F059P Nuclear Boiler c

1 XV142F059R XV242F059R Nuclear Boiler c

1 XV142F059S XV242F059S Nuclear Boiler c

1 XV142F059T XV242F059T Nuclear Boiler c

1 XV142F059U XV242F059U Nuclear Boiler c

1 XV142F061 XV242F061 Nuclear Boiler c

1 Table 1 Valve 10 Valve 10 System Cat Class (Unit 1)

(Unit 2)

XV143F003A XV243F003A Reactor Recirculation c

1 XV143F003B XV243F003B Reactor Recirculation c

1 XV143F004A XV243F004A Reactor Recirculation c

1 XV143F004B XV243F004B Reactor Recirculation c

1 XV143F009A XV243F009A Reactor Recirculation c

1 XV143F009B XV243F009B Reactor Recirculation c

1 XV143F009C XV243F009C Reactor Recirculation c

1 XV143F009D XV243F009D Reactor Recirculation c

1 XV143F010A XV243F010A Reactor Recirculation c

1 XV143F010B XV243F010B Reactor Recirculation c

1 XV143F010C XV243F010C Reactor Recirculation c

1 XV143F010D XV243F010D Reactor Recirculation c

1 XV143F011A XV243F011A Reactor Recirculation c

1 XV143F011B XV243F011B Reactor Recirculation c

1 XV143F011C XV243F011C Reactor Recirculation c

1 XV143F011D XV243F011 D Reactor Recirculation c

1 XV143F012A XV243F012A Reactor Recirculation c

1 XV143F012B XV243F012B Reactor Recirculation c

1 XV143F012C XV243F012C Reactor Recirculation c

1 XV143F012D XV243F012D Reactor Recirculation c

1 XV143F040A XV243F040A Reactor Recirculation c

1 XV143F040B XV243F040B Reactor Recirculation c

1 XV143F040C XV243F040C Reactor Recirculation c

1 XV143F040D XV243F040D Reactor Recirculation c

1 XV143F057A XV243F057A Reactor Recirculation c

1 XV143F057B XV243F0578 Reactor Recirculation c

1 XV14411A XV24411A Reactor Water Cleanup c

1 XV14411B XV24411B Reactor Water CleanuQ c

1 XV14411C XV24411C Reactor Water Cleanup c

1 XV14411D XV24411D Reactor Water Cleanup c

1 XV144F046 XV244F046 Reactor Water CleanuQ c

1 XV149F044A XV249F044A Reactor Core Isolation Cooling c

1 XV149F044B XV249F044B Reactor Core Isolation Cooling c

1 XV149F044C XV249F044C Reactor Core Isolation Cooling c

1 XV149F044D XV249F044D Reactor Core Isolation Cooling_

c 1

XV155F024A XV255F024A High Pressure Coolant Injection c

1 XV155F024B XV255F024B High Pressure Coolant lnlection c

1 XV155F024C XV255F024C High Pressure Coolant Injection c

1 XV155F024D XV255F024D High Pressure Coolant Injection c

1 XV15109A XV25109A Residual Heat Removal c

1 XV151098 XV251098 Residual Heat Removal c

1 XV15109C XV25109C Residual Heat Removal c

1 XV15109D XV25109D Residual Heat Removal c

1 XV152F018A XV252F018A Core Spray c

1 XV152F0188 XV252F0188 Core Spsay c

1 3.1.3 Licensee's Reason for Request In Attachments 1 and 6 of the alternative request submitted by letter dated December 12, 2013, the licensee stated, in part, that:

These valves are instrumentation line excess flow check valves (EFCVs) provided in each instrument line process line that penetrates primary containment in accordance with Regulatory Guide [RG] 1.11 [Instrument Lines Penetrating the Primary Reactor Containment5]. The EFCVs are designed to close upon rupture of the instrument line downstream of the EFCV and otherwise remain open. The lines are sized and/or orificed such that off-site dose will be substantially below 10 CFR 100 limits in the event of a rupture.

Basis for Relief Pursuant to 1 OCFR 50.55a, "Codes and Standards," paragraph (a)(3), relief is requested from the requirements of ASME OM Code ISTC-3522(c) and ISTC-3700. The basis of the relief request is that the proposed alternative would provide an acceptable level of quality and safety.

Testing the subject valves quarterly or during cold shutdown is not practicable, based on plant conditions. These valves have been successfully tested throughout the life of the Susquehanna Steam Electric Station Unit[s] 1 [and 2]

and they have shown no degradation or other signs of aging.

The technology for testing these valves is simple and has been demonstrated effectively during the operating history of Susquehanna Steam Electric Station Unit[s] 1 [and 2]. The basis for this alternative is that testing a sample of EFCVs each refueling outage provides a level of safety and quality equivalent to that of the [ASME OM] Code-required testing.

Excess flow check valves are required to be tested in accordance with ISTC-3522, which requires exercising check valves nominally every three months to the positions required to perform their safety functions. ISTC-3522(c) permits deferral of this requirement to every reactor refueling outage. Excess flow check valves are also required to be tested in accordance with ISTC-3700, which requires remote position verification at least once every 2 years.

The EFCVs are classified as ASME Code Category C and are containment isolation valves. However, these valves are excluded from 10 CFR 50 Appendix J Type C leak rate testing, due to the size of the instrument lines and upstream orificing. Therefore, they have no safety-related seat leakage criterion.

The excess flow check valve is a simple device. The major components are a poppet and spring. The spring holds the poppet open under static conditions.

The valve will close upon sufficient differential pressure across the poppet.

Functional testing of the valve is accomplished by venting the instrument side of 5 ADAMS Accession No. ML100250396.

the valve. The resultant increase in flow imposes a differential pressure across the poppet, which compresses the spring and decreases flow through the valve.

Functional testing is required by Technical Specification Surveillance Requirement 3.6.1.3.9. System design does not include test taps upstream of the EFCV. For this reason, the EFCVs cannot be isolated and tested using a pressure source other than reactor pressure.

The testing described above requires removal of the associated instrument or instruments from service. Since these instruments are in use during plant operation, removal of any of these instruments from service may cause a spurious signal, which could result in a plant trip or an unnecessary challenge to safety systems. Additionally, process liquid will be contaminated to some degree, requiring special measures to collect flow from the vented instrument side and will contribute to an increase in personnel radiation exposure.

Industry experience as documented in [Boiling Water Reactor Owners Group Licensing Topical Report) NED0-32977-A, [Excess Flow Check Valve Testing Relaxation]6 indicates the ECFVs have a very low failure rate. At Susquehanna, the failure rate has been approximately 1% [percent]. Only half of these failures have resulted in replacement of the EFCV. The Susquehanna test history shows no evidence of common mode failure. This Susquehanna test experience is consistent with the findings of NEDO. The NEDO indicates similarly that many reported test failures at other plants were related to test methodologies and not actual EFCV failures. Thus, the ECFVs at Susquehanna, consistent with the industry, have exhibited a high degree of reliability, availability, and provide an acceptable level of quality and safety.

Testing on a cold shutdown frequency is impractical considering the large number of valves to be tested and the condition that reactor pressure greater than 500 psig is needed for testing. In this instance, considering the number of valves to be tested and the conditions required for testing, it is also a hardship to test all these valves during refueling outages. Recent improvements in refueling outage schedules minimized the time that is planned for refueling and testing activities during the outages.

The appropriate time for performing excess flow check valve test is during refueling outages in conjunction with vessel hydrostatic testing. As a result of shortened outages, decay heat levels during hydrostatic tests are higher than in the past. If the hydrostatic test were extended to test all EFCVs, the vessel could require depressurization several times to avoid exceeding the maximum bulk coolant temperature limit. This is an evolution that challenges the reactor operators and thermally cycles the reactor vessel. This evolution should be avoided if possible. Also, based on past experience, excess flow check valve testing during hydrostatic testing becomes the outage critical path and could possibly extend the outage by two days if all EFCVs were to be tested during this time frame.

6 ADAMS Accession No. ML003729011.

Proposed Alternative As an alternative to testing all EFCVs during the refueling outage, a sampling plan will be implemented. This plan will test certain excess flow check valves immediately preceding the refueling outage while the reactor is at power, while also instituting the appropriate conditions for testing (reactor press > 500 psig).

This alternative provides an acceptable level of quality and safety. Performance of this excess flow check valve testing prior to the outage will be scheduled such that, in the event of a failure, the resulting action statement and limiting condition of operation will encompass the planned shutdown for the refueling outage.

Using this strategy, unplanned, unnecessary plant shutdowns as a result of excess flow cheek valve testing will be avoided.

Functional testing with verification that flow is checked will be performed per Technical Specification 3.6.1.3.9, either immediately preceding a planned refueling outage or during the refueling outage for certain EFCVs. For those valves tested prior to the refueling outage, appropriate administrative and scheduling controls will be established.

Surveillance Requirement 3.6.1.3.9 allows a "representative sample" of EFCVs to be tested every 24 months, such that each EFCV will be tested at least once every ten years (nominal).

The EFCVs have position indication in the control room. Check valve remote position indication is excluded from Regulatory Guide 1.97 as a required parameter for evaluating containment isolation. The remote position indication will be verified in the closed direction at the same frequency as the exercise test, which will be performed at the frequency prescribed in Technical Specification Surveillance Requirement 3.6.1.3.9. After the close position test, the valve will be reset, and the remote open position indication will be verified. Although inadvertent actuation of an EFCV during operation is highly unlikely due to the spring poppet design, Susquehanna verifies the EFCVs indicate open in the control room at a frequency greater than once every two years.

In summary, considering the extremely low failure rate along with personnel and plant safety concerns to perform testing, the alternative sampling plan proposed provides an acceptable level of quality and safety.

3.1.4

NRC Staff Evaluation

The licensee proposed an alternative test in lieu of the requirements found in the 2004 Edition through 2006 Addenda of the ASME OM Code Section ISTC-3522(c) and ISTC-3700 for SSES, Units 1 and 2, instrument process line excess flow check valves listed in Table 1. Specifically, the licensee proposed to functionally test and verify the EFCVs per Technical Specification (TS)

Surveillance Requirement (SR) 3.6.1.3.9. SR 3.6.1.3.9 allows a representative sample of EFCVs to be tested every 24 months, such that each EFCV will be tested at least once every 10 years.

EFCVs in reactor instrumentation lines are used to limit the release of fluid from the reactor coolant system in the event of an instrument line break. EFCVs are not required to close in response to a containment isolation signal and are not postulated to operate under post loss-of-coolant accident (LOCA) conditions. The EFCVs were installed following the guidance of RG 1. 11, which states, in part, that the instrumentation lines penetrating the primary containment that are part of the reactor coolant boundary should be sized or orifice installed in such a manner as to ensure that, in the event of any breach, the leakage is reduced to the maximum extent practical and that the rate and extent of coolant loss are within the capability of the normal reactor coolant makeup system. Should an excess flow check valve fail to close when required, the main flow path through the valve has a resistance to flow at least the equivalent of a sharp-edged orifice of 0.375 inch diameter. Valve position indication and excess flow alarm are provided in the control room.

Changes toTS SR 3.6.1.3.9 were initially proposed during the licensee's second 1 0-year 1ST program interval and approved by letter dated April11, 2001.7 In addition to the TS SR proposal, the licensee submitted relief request RR-23 for adapting the test interval of TS SR 3.6.1.3.9, in lieu of the requirements of the 1987 Edition with 1988/1989 Addenda of ASME OM Code Part 10, Section 4.3.2.1, which states that Category C check valves shall be exercised nominally every 3 months. The 1987 Edition with 1988/1989 Addenda of ASME OM Code Part 10, Section 4.3.2.1 is equivalent to the 2004 Edition through 2006 Addenda ASME OM Code Section ISTC-3522.

In the safety evaluation for relief request RR-23, dated April 11, 2001, 8 the NRC staff concluded that the impact of the increase in EFCV surveillance test intervals to 10 years would result in an increase in the release frequency of about 2.02E-04/year from the current release frequency estimate (for a 24-month surveillance test interval) of about 5.07E-05/year. The NRC staff considered this estimate to be sufficiently low. The NRC staff also noted that the consequence of such an accident is unlikely to lead to core damage. The NRC staff concluded that the consequences of the steam release from the depicted events is bounded by an existing Updated Final Safety Analysis Report analysis and that the increase in risk associated with the licensee's request for relaxation of EFCV surveillance testing is low. A review of today's measures and standards yields no changes to the previous conclusions.

The licensee also requested to use TS SR 3.6.1.3.9 test interval in lieu of the requirements of 2004 Edition through 2006 Addenda of the ASME OM Code Section ISTC-3700, which states that valves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated. As noted in the discussion above, EFCVs are not required to close in response to a containment isolation signal and are not postulated to operate under post LOCA conditions. Also, check valve remote position indication is excluded from RG 1.97, "Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants."9 The check valve position indications are monitored on a daily basis. Abnormal position indications are addressed via the corrective action system. The overall performance of the EFCVs has been consistent with industry data yielding very low failure rates, no evidence of common mode failure, and has exhibited a high degree of reliability and availability. Testing and remote position verification of a representative sample of EFCVs every 24 months, such that each EFCV will be tested at least once every 10 years, per TS SR 3.6.1.3.9, provides reasonable assurance that the EFCVs will perform their design function when called upon.

Based on the above, the NRC staff has determined that proposed alternatives 1 RR-01 and 2RR-01 provide an acceptable level of quality and safety.

7 ADAMS Accession No. ML010960024.

8 ADAMS Accession No. ML010960041.

9 ADAMS Accession No. ML061580448.

3.2 Licensee's Alternative Request 1 RR-02 and 2RR-02 3.2.1 Applicable Code Requirements ASME Code Subparagraph ISTC-3630, "Leakage Rate for Other Than Containment Isolation Valve," states, "Category A valves with a leakage requirement not based on an Owner's 10 CFR 50, Appendix J program, shall be tested to verify their seat leakages are within acceptable limits.

Valve closure before seat leakage testing shall be by using the valve operator with no additional closing force applied."

ASME Code Subparagraph ISTC-3630(a), "Frequency," states, "[t]ests shall be conducted at least once every 2 years."

3.2.2 ASME Code Components Affected The licensee requested the alternative for the following Residual Heat Removal (RHR), Low Pressure (LP) Coolant Injection (LPCI), Core Spray, and Reactor Head Spray valves:

Table 2 Valve ID Valve ID System Cat Class App (Unit 1)

(Unit 2)

J HV151F008 HV251F008 RHR Shutdown Cooling Suction A

1 Yes Outboard Isolation Valve HV151F009 HV251 F009 RHR Shutdown Cooling Suction A

1 Yes Inboard Isolation Valve HV151F015A HV251F015A RHR Loop A Injection Outboard A

1 Yes Isolation Valve HV151F015B HV251F015B RHR Loop B Injection Outboard A

1 Yes Isolation Valve HV151F022 HV251F022 RHR Head Spray Inboard Shutoff A

1 Yes HV151F023 HV251F023 RHR Reactor Head Spray Flow A

2 Yes Control Valve HV151F050A HV251F050A RHR LP A Testable Check Valve NC 1

No HV151F050B HV251F050B RHR LP B Testable Check Valve NC 1

No HV151F122A HV251F122A RHR/LPCI Injection Testable A

1 No Check Bypass Valve HV151F122B HV251F122B RHR/LPCIInjection Testable A

1 No Check Bypass Valve HV152F005A HV252F005A Core Spray Loop A Inboard A

1 Yes Injection Shutoff Valve HV152F005B HV252F005B Core Spray Loop B Inboard A

1 Yes Injection Shutoff Valve HV152F006A HV252F006A Core Spray Loop A Testable NC 1

Yes Check Valve HV152F006B HV252F006B Core Spray Loop B Testable NC 1

Yes Check Valve HV152F037A HV252F037A Core Spray Loop A Testable A

1 Yes Check Valve Bypass Air-Operated Valve Table 2 Valve 10 Valve 10 System Cat Class App (Unit 1)

(Unit 2)

J HV152F037B HV252F037B Core Spray Loop B Testable A

1 Yes Check Valve Bypass Air-0_2_erated Valve 3.2.3 Licensee's Reason for Request In Attachments 2 and 7 of the alternative request submitted by letter dated December 12, 2013, the licensee stated, in part, that:

These valves are the Category A and A/C Pressure isolation Valves (PIVs) for Residual Heat Removal System (RHR), Low Pressure Coolant Injection System (LPCI), Core Spray and Reactor head Spray for Susquehanna Steam Electric Station (SSES) Unit[s] 1 [and 2]. They provide isolation and prevent over pressurization of the low pressure piping between the Emergency Core Cooling System (ECCS) and Reactor Coolant System (RCS) boundaries.

Basis for Relief Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (a)(3)(i), relief is requested from the requirement of ASME OM Code ISTC-3630(a). The basis of the relief request is that the proposed alternative would provide an acceptable level of quality and safety.

ISTC-3630 requires that leakage rate testing for PIVs be performed at least once every 2 years. PIVs are not specifically included in the scope for performance-based testing as provided for in 10 CFR Part 50, Appendix J, Option B, while the motor operated PIVs and check valves HV152F006A/B [HV252F006A/B] affected by this request are CIVs [containment isolation valves] and tested in accordance with the 10 CFR 50 Appendix J Program. Check valve PIVs HV151 F050A/B

[HV251 F050A/B] and HV151 F122A/B [HV251 F122A/B] are not within the Appendix J scope.

The concept behind the Option B Alternative for containment isolation valves is that licensees should be allowed to adopt cost effective methods for complying with regulatory requirements. Additionally, [Nuclear Energy Institute] NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," describes the risk-informed basis for the extended test intervals under Option B. That justification shows that for valves which have demonstrated good performance by passing their associated leak rates tests for two consecutive cycles, further failures appear to be governed by the random failure rate of the component. NEI 94-01 also presents the results of a comprehensive risk analysis, including the statement that "the risk impact associated with increasing [leak rate] test intervals is negligible (less than 0.1 percent of total risk)." The valves identified in this relief request are all water applications. The PIV testing is performed with water pressurized to normal plant operating pressures in accordance with ISTC-3630. This relief request is intended to provide for a performance-based scheduling of PIV tests at SSES.

The reason for requesting this relief is dose reduction I [as low as reasonably achievable] ALARA. Recent historical data was used to identify that PIV testing alone each refuel outage incurs total dose of approximately 500 milliRem.

Assuming all of the PIVs remain classified as good performers the extended test intervals would provide for a savings of approximately 1.0 Rem over the 4-year period.

NUREG 0933, "Resolution of Generic Safety Issues," Issue 105 (Interfacing Systems LOCA at LWRs [light-water reactors]) discussed the need for PIV leak rate testing based primarily on three pre-1980 historical failures of applicable valves industry-wide. These failures all involved human errors in either operations or maintenance. None of these failures involved in service equipment degradation. The performance of PIV leak rate testing provides assurance of acceptable seat leakage with the valve in a closed condition. Typical PIV testing does not identify functional problems, which may inhibit the valves ability to re-position from opened to closed. For check valves, such functional testing is accomplished per ASME OM Code ISTC-3522 and ISTC-3520. Power-operated valves are routinely full stroke tested per ASME OM Code to ensure their functional capabilities. At SSES, these functional tests for motor operated PIVs are performed on a cold shutdown frequency. The functional testing of the PIV check valve is performed in accordance with ISTC-5221 "Valve Obturator Movement." Performance of separate 2 year PIV leak rate testing does not contribute any additional assurance of functional capability; it only determines the seat tightness of the closed valves.

PIV testing is performed with water pressurized to normal plant operating pressures in accordance with ISTC-3630. The intent of this relief request is to allow for a performance-based approach to the scheduling of PIV leakage testing. It has been shown that Interfacing System Loss of Coolant Accident (ISLOCA) represents a small risk impact to BWRs such as SSES.

NUREG/CR-5928, "Final Report of the NRC-sponsored ISLOCA Research Program" (ADAMS Accession No. ML072430731 ), evaluated the likelihood and potential severity of ISLOCA events in Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR). The BWR design used as a reference for this analysis was a BWR-4 with Mark I containment. SSES is listed as a similar plant. The BWR systems were individually analyzed and in each case the report concluded that the system was "judged to not be an important consideration with respect to ISLOCA risk." Section 4.3 of the report concluded the BWR portion of the analysis by saying "ISLOCA is not a risk concern for the BWR plant examined here."

The functional tests for PIVs are performed only at a cold shutdown frequency.

Such testing is not performed online in order to prevent any possibility of an inadvertent ISLOCA condition. The functional testing of the PIVs is adequate to identify any abnormal condition that might affect closure capability.

Proposed Alternative SSES proposes to perform PIV testing at intervals ranging from every refuel to every third refuel. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the Containment Isolation valve (CIV) process under 1 OCFR50 Appendix J, Option B, program guidance. The test frequency will be established such that if any of the valves, subject to a CIV and a PIV test, fail either test, the test interval for both tests will be reduced to once every 24 months until they can be re-classified as good performers per the performance evaluation requirements of Appendix J, Option B. The test intervals for the valves with a PIV only function will be determined in a similar manner as is done for CIV testing under Option B. The test interval may be extended upon completion of two consecutive periodic PIV tests with results within prescribed acceptance criteria. Any PIV test failure will require a return to the initial interval until good performance can again be established.

3.2.4

NRC Staff Evaluation

The licensee proposed an alternative test in lieu of the requirements found in the 2004 Edition through 2006 Addenda of the ASME OM Code Section ISTC-3630(a) for 16 pressure isolation valves (PIV) per unit. Twelve of the 16 valves, at each unit, also function as containment isolation valves (CIV). Specifically, the licensee proposes to functionally test and verify the leakage rate of 16 PIVs, per unit, using 10 CFR 50 Appendix J, Option B, performance-based schedule. Valves would initially be tested at the required interval schedule, which is currently every refueling outage (RFO) RFO or 2 years, as specified by ASME OM Code Section ISTC-3630(a). Valves that have demonstrated good performance for two consecutive cycles may have its test interval extended to every 3 RFO, not to exceed 6 years. Any PIV leakage test failure would require the component to return to the initial interval of every RFO or 2 years until good performance can again be established.

Pressure isolation valves (PIVs) are defined as two valves in series within the reactor coolant pressure boundary, which separate the high-pressure reactor coolant system from an attached lower pressure system. Failure of a PIV could result in an over-pressurization event, which could lead to a system rupture and possible release of fission products to the environment. This type of failure event was analyzed under NUREG/CR-5928, "Inter System Loss of Coolant Accident (ISLOCA) Research Program." The purpose of NUREG/CR-5928 was to quantify the risk associated with an ISLOCA event. NUREG/CR-5928 analyzed BWR and PWR designs.

Specifically, NUREG/CR-5928 reviewed the BWR-4 design, which included SSES. The conclusion of the analysis resulted in ISLOCA not being a risk concern for the BWR-4 design.

Appendix J, Option B of 10 CFR is a performance-based leakage test program. Guidance for implementation of acceptable leakage rate test methods, procedures, and analyses is provided in RG 1.163, "Performance Based Containment Leak Test Program."10 RG 1.163 endorses Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 0, "Industry Guideline for Implementing Performance Based Option of 10 CFR 50, Appendix J," dated July 26, 1995, 11 with the limitation that Type C components test interval cannot extend greater than 60 months. The 10 ADAMS Accession No. ML003740058.

11 ADAMS Accession No. ML11327A025.

current version of NEI 94-01 is Revision 3-A, 12 which allows Type C containment isolation valves test intervals to be extended to 75 months with a permissible extension for non-routine emergent conditions of 9 months (84 months total). The NRC staff finds that the guidance in NEI 94-01, Revision 3-A, is acceptable with the following conditions:

1) Extended interval for Type C local leak rate tests (LLRTs) may be increased to 75 months with the requirement that a licensee's post outage report include the margin between Type 8 and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level when the licensee's regulatory limit is not met. Extensions of up to 9 months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. This provision (9 month extension) does not apply to valves that are restricted and/or limited to 30 month intervals in Section 10.2 (such as 8WR main steam isolation valves) or to valves held to the base interval (30 months) due to unsatisfactory LLRT performance.

2) When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type 8 and C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

The 16 PIVs per unit are currently being leak tested every RFO or 2 years. Performance of the leakage test of the 16 PIVs per unit places a burden on test personnel being exposed to radiation.

Overall completion of leak test requirements averages a dose of 500mRem. The valves have maintained a history of good performance. Extending the leakage test interval based on good performance and the low risk factor as noted in NUREG/CR-5928 is a logical progression to a performance based program.

The licensee requested relief on the basis of 10 CFR 50.55a(a)(3)(i}, the proposed alternative would provide an acceptable level of quality and safety. However, the licensee's reason for the alternative proposal was dose reduction/ALARA, which is considered a hardship. To maintain the current RFO or 2 year leakage test interval represents an undue hardship without an increase in the level of quality and safety, which is applicable to 10 CFR 50.55a(a)(3)(ii). Testing low risk valves on a performance-based schedule provides reasonable assurance that the component is operationally ready.

The licensee is authorized to implement a performance-based program, as outlined above, for the 16 PIVs per unit at SSES. The performance-based program interval shall not exceed 3 RFOs or 75 months. Non-routine emergent conditions may extend the program interval 9 months.

12 ADAMS Accession No. ML12221A202.

3.3 Licensee's Alternative Requests 1 RR03 and 2RR03 3.3.1 Applicable Code Requirements ASME OM Code Paragraph ISTC-3200, "lnservice Testing," states, "[i]nservice testing in accordance with this Subsection shall commence when the valves are required to be operable to fulfill their required function(s)."

ASME OM Code Subparagraph ISTC-5240, "Safety and Relief Valves," states, "[s]afety and relief valves shall meet the inservice test requirements of Mandatory Appendix 1."

Mandatory Appendix I, Subparagraph 1-1320 (a), states, "Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested within each interval; however, a minimum of 20 percent of the valves from each valve group shall be tested within any 24-month interval. This 20 percent shall consist of valves that have not been tested during the current 5-year interval, if they exist. The test interval for any individual valve shall not exceed 5 years."

ASME OM Code Case OMN-17, "Alternate Rules for Testing ASME Class 1 Pressure Relief/Safety Valves," from the 2009 Edition of ASME OM Code, allows an extended test interval of 6 years, for testing these relief valves, plus an additional 6-month grace period, provided the licensee disassembles and inspects each valve after as-found set-pressure testing to verify that valve parts are free of defects resulting from time-related degradation or service-induced wear.

3.3.2 ASME Code Components Affected The licensee requested using alternative testing for the following Class 1, Category C, Main Steam Safety Relief Valves (MSRVs):

Table 3 Valve 10 Valve 10 System Cat Class (Unit 1)

(Unit 2)

PSV141F013A PSV241F013A Nuclear Boiler c

1 PSV141F013B PSV241 F013B Nuclear Boiler c

1 PSV141F013C PSV241 F013C Nuclear Boiler c

1 PSV141F013D PSV241 F013D Nuclear Boiler c

1 PSV141F013E PSV241 F013E Nuclear Boiler c

1 PSV141F013F PSV241F013F Nuclear Boiler c

1 PSV141 F013G PSV241F013G Nuclear Boiler c

1 PSV141F013H PSV241 F013H Nuclear Boiler c

1 PSV141F013J PSV241 F013J Nuclear Boiler c

1 PSV141 F013K PSV241F013K Nuclear Boiler c

1 PSV141F013L PSV241 F013L Nuclear Boiler c

1 PSV141 F013M PSV241 F013M Nuclear Boiler c

1 PSV141 F013N PSV241 FO 13N Nuclear Boiler c

1 PSV141F013P PSV241F013P Nuclear Boiler c

1 PSV141 F013R PSV241 F013R Nuclear Boiler c

1 PSV141 F013S PSV41F013S Nuclear Boiler c

1 3.3.3 Licensee's Reason for Request In Attachments 3 and 8 of the alternative request submitted by letter dated December 12, 2013, the licensee stated, in part, that:

These valves are Main Steam Safety/Relief Valves. They provide overpressure protection for the reactor coolant pressure boundary to prevent unacceptable radioactive release and exposure to plant personnel.

Basis for Relief In accordance with 10 CFR 50.55a(a)(3)(i), the licensee's relief request seeks approval of an alternative to the 5-Year Test Interval requirements of ASME OM Code, Appendix I, Section l-1320(a), for the Susquehanna Main Steam Safety/Relief Valves (MSRVs) for Unit[s] 1 [and 2]. Susquehanna requests that the test interval be increased from 5 years to 72 months in accordance with ASME OM Code Case, OMN-17, "Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves," so that the test interval for any individual valve that is in service shall not exceed 72 months except that a 6-month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods. The basis of the relief request is that the proposed alternative would provide an acceptable level of quality and safety.

Without [ASME OM] Code relief for 24-month fuel cycles, strict [ASME OM] Code compliance would restrict Susquehanna's operating philosophy to not operate with weeping MSRVs as [ASME OM] Code testing would be required to be completed within 5 years. This testing strategy does not account for any leaking valves that may need to be refurbished. Since Susquehanna's philosophy is to share spare valves between both units, (the valves that are removed from one unit are installed in the other unit's next refueling outage), this testing strategy is less than adequate. This strategy could only be accomplished if a large population of MSRVs are tested each outage or additional spare valves are purchased. More than 8 valves would need to be sent to the offsite testing facility during a refueling outage. The testing and return of these valves would have to be completed expeditiously in order to not impact the refuel outage schedule duration. For this reason, additional expenditures would be incurred to purchase and test a greater number of valves each outage. Without [ASME OM]

Code relief, the additional outage work would be contrary to the principles of ALARA and could compromise radiation safety. Because of the location of certain MSRVs in the containment, interferences exist that would require the removal of more valves and piping for those valves that must be removed for the sample testing. This results in more radiation exposure to the maintenance personnel than is desirable.

With [ASME OM] Code relief, the 16 MSRVs per unit can be tested within 6 years to complete the [ASME OM] Code required testing for the total population and accommodate any weeping MSRVs. The increased testing over only 2 refuel cycles would result in no additional safety benefit to the plant. Susquehanna has had excellent performance with MSRVs over the last 10 years. Since 1987, Susquehanna has imposed a more conservative as-left leakage criterion on the testing facility than was specified in the General Electric [GE] Specification and incorporated in the PPL Specification for testing Crosby style relief valves. The criterion imposed on the test lab is 0 ml/5 minutes (via the purchase order) compared to a GE Specification "as-left" leakage criterion of 38 ml/5 minutes.

Additionally, a review of the set point testing results (for both units) from initial operation to the present shows that the average of the set point drifts percentages is approximately -0.91% [percent]. This indicates that, in general, the MSRVs Set Pressure tends to drift slightly downward, not upward. The calculated standard deviation from the average for the data was determined to be approximately 1.68%.

Also, the testing history shows that since commercial operation, Susquehanna has had only two "as-found" set pressure test acceptance criteria failures (above

+3%) of the tested valves, which required additional MSRVs to be tested.

Proposed Alternate Testing For the fourth ten-year interval, Susquehanna proposes to remove at least 20%

of the 16 Main Steam Safety/Relief Valves (MSRVs) plus weeping valves detected during the previous operating cycle and any valves required to be removed to access scheduled or weeping valves up to a maximum of 8 valves during each refueling outage.

Additional valves above the Code required minimum 20% will be tested if the as-found setpoint exceeds +3%, -5% (as approved per Technical Specification Amendment No. 257) of the nameplate. The additional valves tested will be from the initial population removed that are in excess of the 20% [ASME OM] Code required minimum. If one of these valves fails, then all the MSRVs would be removed and tested.

The proposed alternative will provide for disassembly and inspection of the MSRVs to verify parts are free from defects resulting from time-related degradation or maintenance induced wear. This maintenance will also help to reduce the potential for set point drift, and increase the reliability of these Safety Relief Valves to perform their design requirement functions. Consistent with the special maintenance requirement in [ASME OM] Code Case OMN -17, critical components will be inspected for wear and defects.

Completion of [ASME OM] Code testing will be accomplished over a period of 3 refuel cycles or 6 years. This approach results in maintenance and operational flexibility with the following benefits:

Provides the ability to both test the [ASME OM] Code required valves out of the population not yet tested and replace any weeping MSRVs.

Maintains relatively leak-free MSRVs, thus minimizing the necessary run time of ECCS systems that provide suppression pool cooling.

Consistent application of ALARA principles.

Enhances equipment reliability.

Results in minimal impact on outage durations.

The MSRVs will be tested such that a minimum of 20% of the valves (previously untested, if they exist) are tested every 24 months, such that all the valves will be tested within 3 refuel cycles. This proposal utilizes the same maintenance and testing approach that was applied in 18-month refuel cycles. This alternative frequency will continue to provide assurance of the valve operational readiness and provides an acceptable level of quality and safety.

Additionally, any failures, either seat leakage or pressure set point, occurring at the test facility, as well as weeping MSRVs that develop during the operating cycle will be documented by the corrective action program, evaluated and dispositioned accordingly.

3.3.4

NRC Staff Evaluation

MSRVs are ASME Code Class 1 and Category C pressure relief valves that provide overpressure protection for the reactor coolant pressure boundary to prevent unacceptable radioactive release and exposure to plant personnel. ASME OM Code, Mandatory Appendix I requires that Class 1 pressure relief valves be tested at least once every 5 years. However, Mandatory Appendix I does not require that pressure relief valves be disassembled and inspected prior to the start of the 5-year test interval. In lieu of the 5-year test interval, the licensee proposed to implement ASME OM Code Case OMN-17, which allows a test interval of 6 years plus a 6-month grace period. The ASME Committee on OM developed Code Case OMN-17 and published it in the 2009 Edition of the ASME OM Code. ASME OM Code Case OMN-17 imposes a special maintenance requirement to disassemble and inspect each pressure relief/safety valve to verify that parts are free from defects resulting from time-related degradation or service-induced wear prior to the start of the extended test interval. The purpose of this maintenance requirement is to reduce the potential for pressure relief valve set-point drift.

ASME OM Code Case OMN-17 has not been added to RG 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code,"13 or included in 10 CFR 50.55a by reference.

However, the NRC has allowed licensees to use ASME OM Code Case OMN-17, provided all requirements in the Code Case are met. Consistent with the special maintenance requirement in ASME OM Code Case OMN-17, each MSRV at SSES will be disassembled and inspected to verify that internal surfaces and parts are free from defects or service induced wear prior to the start of the next test interval. This maintenance will also help to reduce the potential for set point drift, and increase the reliability of these MSRVs to perform their design requirement functions.

Consistent with the special maintenance requirement in ASME OM Code Case OMN-17, critical components will be inspected for wear and defects. This process is consistent with ASME OM Code Case OMN-17 paragraphs (d) and (e).

Furthermore, ASME OM Code Case OMN-17 is performance-based, in that it requires that MSRV be tested more frequently if test failures occur. For example, ASME OM Code Case OMN-17 requires that two additional valves be tested when a valve in the initial test group exceeds the set pressure acceptance criteria. At SSES, Units 1 and 2, two additional valves will be tested if the as-found setpoint exceeds +3% or -5%, of the name plate. This was approved per SSES TS Amendments 257and 237, which were issued by letter dated November 17, 2011. 14 All remaining valves in the group are required to be tested if one of the additional valves tested exceeds its set 13 ADAMS Accession No. ML030730430.

14 ADAMS Accession No. ML11292A137.

pressure acceptance criteria. Therefore, the MSRV test frequency would be equivalent to the current test frequency, if test failures occur.

Additionally, a review of the set point testing results (for both units) from initial operation to the present shows that the average of the set point drifts percentages is approximately -0.91%. This indicates that, in general, the MSRVs set pressure tends to drift slightly downward, not upward.

The testing history shows that since commercial operation, SSES has had only two "as-found" set pressure test acceptance criteria failures (above +3%) of the tested valves, which required additional MSRVs to be tested, and that the subject valves have historically exhibited very limited susceptibility to time-related degradation or set-point drift.

Based on the historical performance of the set-point testing of MSRVs and disassembly and inspection of the MSRVs prior to use, the NRC staff finds that implementation of the ASME OM Code Case, OMN-17, for the testing of the designated MSRVs, in lieu of the requirements of the 2004 Edition through the 2006 Addenda and the Mandatory Appendix I, Section 1320 of the ASME OM Code, provides an acceptable level of quality and safety.

3.4 Licensee's Alternative Requests 1 RR04 and 2RR04 3.4.1 Applicable Code Requirements This request applies to the frequency specifications of the ASME OM Code. The frequencies for tests given in the ASME OM Code include the following, but do not include a tolerance band:

ASME OM Code Subparagraph ISTA-3120, "lnservice Test Interval," (a) states,

"[e]xamination and test frequency shall be in accordance with the requirements of Section I ST."

ASME OM Code Paragraph ISTB-3400, "Frequency of lnservice Tests," states, "[a]n inservice test shall be run on each pump as specified in Table ISTB-3400-1."

Table ISTB-3400-1, "lnservice Test Frequency," notes that Group A and Group 8 pump tests are to be conducted quarterly and comprehensive pump tests are to be conducted biennially.

ASME OM Code Subparagraph ISTC-351 0, "Exercising Test Frequency," states, "[a]ctive Category A, Category 8, and Category C check valves shall be exercised nominally every 3 months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221, and ISTC-5222. Power-operated valves shall be exercise tested once per fuel cycle."

ASME OM Code Subparagraph ISTC-3540, "Manual Valves," states, "[m]anual valves shall be full-stroke exercised at least once every 2 years, except where adverse conditions may require the valve to be tested more frequently to ensure operational readiness. Any increased testing frequency shall be specified by the Owner. The valve shall exhibit the required change of obturator position."

ASME OM Code Subparagraph ISTC-3630, "Leakage Rate for Other Than Containment Isolation Valves," (a) "Frequency," states, "[t]ests shall be conducted at least once every 2 years."

ASME OM Code Paragraph ISTC-3700, "Position Verification Testing," states, in part, that

"[v]alves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated."

ASME OM Code Subparagraph ISTC-5221, "Valve Obturator Movement," (c)(3), states, "[a]t least one valve from each group shall be disassembled and examined at each refueling outage; all valves in each group.shall be disassembled and examined at least once every 8 years."

Mandatory Appendix I, "lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants," 1-1320, "Test Frequencies, Class 1 Pressure Relief Valves," (a), "5-Year Test Interval," states, in part, that "Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation."

Mandatory Appendix I, 1-1330, "Test Frequency, Class 1 Non-reclosing Pressure Relief Devices," states, "Class 1 non-reclosing pressure relief devices shall be replaced every 5 years unless historical data indicates a requirement for more frequent replacement."

Mandatory Appendix I, 1-1340, "Test Frequency, Class 1 Pressure Relief Valves That Are Used for Thermal Relief Application," states, "[t]ests shall be performed in accordance with 1-1320, Test Frequencies, Class 1 Pressure Relief Valves."

Mandatory Appendix I, 1-1350, "Test Frequency, Classes 2 and 3 Pressure Relief Valves," (a),

"10-YearTest Interval," states, in part, that "Class 2 and 3 pressure relief valves, with the exception of PWR main steam safety valves, shall be tested every ten years, starting with initial electric power generation."

Mandatory Appendix I, 1-1360, "Test Frequency, Classes 2 and 3 Non-reclosing Pressure Relief Devices," states, "Classes 2 and 3 non-reclosing pressure relief devices shall be replaced every 5 years, unless historical data indicates a requirement for more frequent replacement."

Mandatory Appendix I, 1-1370, "Test Frequency, Classes 2 and 3 Primary Containment Vacuum Relief Valves," states, "(a) Tests shall be performed on all Classes 2 and 3 containment vacuum relief valves at each refueling outage or every 2 years, whichever is sooner, unless historical data requires more frequent testing. (b) Leak tests shall be performed on all Classes 2 and 3 containment vacuum relief valves at a frequency designated by the Owner in accordance with Table ISTC-3500-1."

Mandatory Appendix I, 1-1380, "Test Frequency, Classes 2 and 3 Vacuum Relief Valves, Except for Primary Containment Vacuum Relief Valves," states, "[a]ll Classes 2 and 3 vacuum relief valves shall be tested every 2 years, unless performance data suggest the need for a more appropriate test interval."

Mandatory Appendix I, 1-1390, "Test Frequency, Classes 2 and 3 Pressure Relief Devices That Are Used for Thermal Relief Application," states, "[t]ests shall be performed on all Classes 2 and 3 relief devices used in thermal relief application every 10 years, unless performance data indicate more frequent testing is necessary. In lieu of tests the Owner may replace the relief devices at a frequency of every 10 years, unless performance data indicate more frequent replacements are necessary."

Mandatory Appendix II, "Check Valve Condition Monitoring Program," 11-4000, "Condition-Monitoring Activities," (a), "Performance Improvement Activities," (1 ), states, in part, that "[i]f sufficient information is not currently available to complete the analysis required in 11-3000, or if this analysis is inconclusive, then the following activities shall be performed at sufficient intervals over an interim period of the next 5 years or two refueling outages, whichever is less, to determine the cause of failure or the maintenance patterns."

Mandatory Appendix II, 11-4000, (b), "Optimization of Condition-Monitoring Activities," (1)(e),

states, "Identify the interval of each activity. Interval extensions shall be limited to one fuel cycle per extension. Intervals shall not exceed the maximum intervals shown in Table 11-4000-1. All valves in a group sampling plan must be tested or examined again, before the interval can be extended again, or until the maximum interval would be exceeded. The requirements of ISTA-3120, lnservice Test Interval, do not apply."

ASME OM Code Case OMN-20, "lnservice Test Frequency," which addresses alternatives that may be applied to the test frequencies for pumps and valves specified in ASME OM Division 1, Section 1ST, 2009 Edition through OMa-2011 Addenda and all earlier edition and addenda.

3.4.2 Licensee's Reason for Request In Attachments 4 and 9 of the alternative request submitted by letter dated December 12, 2013, the licensee stated, in part, that:

Basis for Relief Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (a)(3)(ii), relief is requested from the frequency specifications of the ASME OM Code. The basis of the relief request is that the [ASME OM] Code requirement represents an undue hardship without a compensating increase in the level of quality and safety.

ASME OM Code Section 1ST establishes the inservice test frequency for all components within the scope of the [ASME OM] Code. The frequencies (e.g.,

quarterly) have always been interpreted as "nominal" frequencies (generally as defined in the Table 3.2 of NUREG-1482, [Guidelines for lnservice Testing at Nuclear Power Plants] Revision 1 [2]15) and Owners routinely applied the surveillance extension time period (i.e., grace period) contained in the plant Technical Specification (TS) Surveillance Requirements (SRs). The TS typically allow for a less than or equal to 25% extension of the surveillance test interval to accommodate plant conditions that may not be suitable for conducting a TS surveillance (SR 3.0.2). However, regulatory issues have been raised concerning the applicability of the TS "Grace Period" to ASME OM Code required 1ST frequencies irrespective of allowances provided under TS Administrative Controls (i.e., TS 5.5.6, "lnservice Testing Program," invokes SR for various OM Code frequencies).

The lack of a tolerance band on the ASME OM Code [1ST] frequency restricts operational flexibility. There may be a conflict where [1ST] could be required (i.e.,

its frequency could expire), but where it is not possible or not desired that it be performed until sometime after a plant condition or associated Limiting Condition for Operation is within its applicability. Therefore, to avoid this conflict, the [1ST]

should be performed.when it can be and should be performed.

The NRC recognized this potential issue in the TS by allowing a frequency tolerance as described in TS SR 3.0.2. The lack of a similar tolerance applied to the [ASME OM] Code testing places an unusual hardship on the plant to adequately schedule work tasks without operational flexibility.

Thus, just as with TS required surveillance testing, some tolerance is needed to allow adjusting [ASME] OM Code testing intervals to suit the plant conditions and other maintenance and testing activities. This assures operational flexibility 15 ADAMS Accession No. ML13295A020.

when scheduling [1ST] that minimizes the conflicts between the need to complete the testing and plant conditions.

Proposed Alternative Testing Susquehanna proposes to use the ASME OM Code Case OMN-20, from the 2012 Edition of the ASME OM Code, as an alternative for grace period associated with lnservice Testing Requirements.

ASME OM Code establishes component test frequencies that are based either on elapsed time periods (e.g., quarterly, 2 years, etc.) or on the occurrence of plant conditions or events (e.g., cold shutdown, refueling outage, upon detection of a sample failure, following maintenance, etc.).

a. Components whose test frequencies are based on elapsed time periods shall be tested at the frequencies specified in ASME Code Section 1ST with a specified time period between tests as shown in the following table.

Frequency Specified Time Period Between Tests (all values are 'not to exceed'; no minimum _periods are s~ecified)

Quarterly 92 days (or every 3 months)

Semiannually 184 days (or every 6 months)

Annually 366 days (or every year) x Years x calendar years where 'x' is a whole number of years <:: 2

b. The specified time period between tests may be reduced or extended as follows:

i.

For periods specified as less than 2 years, the period may be extended by up to 25% for any given test. This is consistent with SSES TS Section 5.5.6, "lnservice Testing Program."

ii. Period extensions may be also be applied to accelerated test frequencies (e.g., pumps in Alert Range).

iii. For periods specified as greater than or equal to 2 years, the period may be extended by up to 6 months for any given test.

c.

Components whose test frequencies are based on the occurrence of plant conditions or events (e.g., cold shutdown, refueling outage, upon detection of a sample failure, following maintenance, etc.) may not have their period between tests extended except as allowed by ASME OM Code.

3.4.3

NRC Staff Evaluation

Historically, licensees have applied and the NRC staff has accepted the standard TS definitions for 1ST intervals (including allowable interval extensions) to ASME OM Code required testing (Reference NUREG-1482 Revision 1, Section 3.1.3). Recently, the NRC staff reconsidered the allowance of the TS testing intervals and interval extensions, for 1ST not associated with TS SRs.

As noted in Regulatory Issue Summary (RIS) 2012-10, "NRC Staff Position on Applying Surveillance Requirements 3.0.2 and 3.0.3 to Administrative Controls Program Tests,"16 the NRC determined that programmatic test frequencies cannot be extended in accordance with the TS SR 3.0.2. This includes all 1ST described in the ASME OM Code not specifically required by the TS SRs.

Following this development, the NRC staff sponsored and co-authored an ASME OM Code inquiry and Code Case to modify the ASME OM Code to include TS-Iike test interval definitions and interval extension criteria. The ASME OM Code Case OMN-20 was approved by the ASME Operation and Maintenance Standards Committee and published in the 2012 Edition of the ASME OM Code. The licensee proposes to use the ASME OM Code Case OMN-20 from the 2012 Edition of the ASME OM Code for grace period associated with 1ST requirements.

However, the NRC staff has noticed that the wording in the above proposed alternative is incomplete and certain restrictions/limitations are omitted. Below are the restrictions/limitations for use of ASME OM Code Case OMN-20:

Period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g., performance of the test would cause an unacceptable increase in the plant risk profile due to transient conditions or other ongoing surveillance, test, or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified.

Period extensions may also be applied to accelerated test frequencies (e.g., pumps in alert range) and other fewer than 2-yr test frequencies not specified in Table 1.

Period extensions may not be applied to the test frequency requirements specified in Subsection ISTD, "Preservice and lnservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants," as Subsection ISTD contains its own rules for period extensions.

Implementation of ASME OM Code Case OMN-20 requires implementation of this code case in its entirety, including the above stated restrictions/limitations, without exceptions.

Requiring the licensee to meet the ASME OM Code requirements, without an allowance for defined frequency and frequency extensions for 1ST of pumps and valves, results in a hardship without a compensating increase in the level of quality and safety. Based on the prior acceptance by the NRC staff of the similar TS test interval definitions and interval extension criteria, the staff finds that implementation of the test interval definitions and interval extension criteria contained in ASME OM Code Case OMN-20 is acceptable. Allowing usage of ASME Code Case OMN-20 provides reasonable assurance of operational readiness of pumps and valves subject to the ASME OM Code 1ST.

16 ADAMS Accession No. ML12079A393.

3.5 Licensee's Alternative Request 1 RR-05 3.5.1 Applicable ASME OM Code Requirements ASME OM Code Subparagraph ISTC-351 0, "Exercising Test Frequency," states, in part, that "Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3560, ISTC-5221, and ISTC-5222."

ASME OM Code Subparagraph ISTC-3522(c), "Category C Check Valves," states, "[i]f exercising is not practicable during operation at power and cold shutdowns, it shall be performed during refueling outages."

3.5.2 ASME OM Code Components Affected The licensee requested alternative testing for the following Control Structure Chilled Water (CSCW) and Emergency Service Water (ESW) valves:

Table 4 Valve ID System Cat Class 086018 cscw c

3 086118 cscw c

3 086241 ESW c

3 086341 ESW c

3 3.5.3 Licensee's Reason for Request In Attachment 5 of the alternative request submitted by letter dated December 12, 2013, the licensee stated, in part, that:

Valves 086018 and 086118 are six (6) inch Emergency Condenser Pump OP171A/B discharge check valves. They have an open safety function to provide a flow path from the Emergency Condenser Pump to the Control Structure Chiller Condenser. These check valves have no closed safety function. Valves 086241 and 086341 are two (2) inch Emergency Service Water (ESW) keepfill check valves. These valves are considered part of the Control Structure Chilled Water (CSCW) system. They are keepfill check valves that allow Service water to maintain the Emergency Condenser Water Circulating (ECWC) subsystem full.

The ECWC subsystem is fed from the ESW system. These check valves have a closed safety function to prevent diversion of ESW from the ECWC subsystem when operating under emergency conditions. The check valves have no open safety function.

Basis for Relief Pursuant to 1 OCFR50.55a, "Codes and Standards," paragraph (a)(3), relief is requested from the requirements of ASME OM Code ISTC-3522(c). The basis of the relief request is that the proposed alternative would provide an acceptable level of quality and safety.

The components listed above are check valves with no external means for exercising and no external position indication. The only means to verify closure is by leak testing. This involves setup of test equipment and system configuration changes that are a hardship without a compensating increase in quality or safety on a quarterly or cold shutdown basis. The leak testing can be performed at intervals other than refueling outages such as during system outage windows...

Leak testing check valves and other periodic work activities in the CSCW (and ESW) system(s) will cause CSCW (and ESW) to become INOPERABLE in accordance with Technical Specifications (TS) and Technical Requirements Manual (TRM). In accordance with TS 3. 7.3, operation with one Control Room Emergency Outside Air Supply (CREOAS) subsystem INOPERABLE is permitted for up to 7 days. In accordance with TS 3.7.4, operation with one control room floor cooling system INOPERABLE is permitted for up to 30 days. In accordance with TRM 3.7.9, operation with a single division of the Control Structure Chilled Water system INOPERABLE is permitted for up to 30 days. In accordance with TRM 3.8.6 (Unit 1 only), operation with a one required Emergency Switchgear Room Cooling subsystem INOPERABLE is permitted for up to 30 days. Leak testing of CSCW check valves takes between 4 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, which would typically be accomplished within a 24-hour system work window...

Proposed Alternative Pursuant to 1 OCFR50.55a(a)(3)(i), SSES 1 [... ] proposes an alternative testing frequency for performing inservice testing of the valves identified above. The valves will be closure tested by leak testing on a frequency of at least once per operating cycle in lieu of once each refueling outage as currently allowed by ASME OM Code, 2004 Edition through 2006 Addenda ISTC-3522(c), "Category C Check Valves." The open safety function of check valves 086018 and 086118 will be demonstrated quarterly in conjunction with the Control Structure Chilled Water flow verification (inservice pump test). The open function of check valves 086241 and 086341 is demonstrated continuously through the keepfill function.

In Attachment 1 of the response to request for additional information submitted by letter dated March 31, 2014, the licensee stated, in part, that:

Check valves 086241 (Emergency Service Water (ESW) Control Structure Chiller (CSC) Loop A Keepfill) and 086341 (ESW CSC Loop B Keepfill) are tested on a 24 month frequency. Each of these check valves have a separate surveillance test which accomplishes their required inservice testing. The requested alternative requires performance of the closure test at least once per operating cycle, and the scheduled test interval is required to be once every 24 months.

This does not allow scheduling this test to be at the beginning of one operating cycle and at the end of the next operating cycle. Therefore, the maximum scheduled interval between tests will remain 24 months.

Check valves 086018 (Emergency Condenser Water Circulating Pump A Discharge Check Valve) and 086118 (Emergency Condenser Water Circulating Pump B Discharge Check Valve) have an open safety function to provide a flow path from the emergency condenser water pump to the chiller condenser. These check valves are exercised tested open on a quarterly frequency.

The valves 086018 and 086118 will close to prevent backflow through an idle pump. The function is not required for safe shutdown or accident mitigation. The requested alternative requires performance of the closure test at least once per operating cycle, and the scheduled test interval is required to be once every 24 months. This does not allow scheduling this test to be at the beginning of one operating cycle and at the end of the next operating cycle. Therefore, the maximum scheduled interval between tests will remain 24 months.

3.3.2

NRC Staff Evaluation

The valves in Table 4 are Safety Class 3 Category C check valves. ASME OM Code ISTC-3510 states, in part, that "Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221, and ISTC-5222." ASME OM Code ISTC-3522(a) states, in part, that "During operation at power, each check valve shall be exercised or examined in a manner that verifies obturator travel by using the methods in ISTC-5221. Each check valve exercise test shall include open and close tests."

Check valves 086018 and 086118 have an open safety function only. ASME OM Code ISTC-5221 (a)(2) states that "Check valves that have a safety function in only the open direction shall be exercised by initiating flow and observing that the obturator has traveled either the full open position or to the position required to perform its intended function(s) (see ISTA-1100), and verify closure." The licensee states that these valves will maintain their quarterly exercise to verify the open safety function.

Check valves 086241 and 086341 have a close safety function only. ASME OM Code ISTC-5221 (a)(3) states, in part, that "Check valves that have a safety function in only the close direction shall be exercised by initiating flow and observing that the obturator has traveled at least the partially open position, and verify that on cessation or reversal of flow, the obturator has traveled to the seat." The licensee states that these valves are keepfill valves that allow Service Water to maintain the Emergency Condenser Water Circulating (ECWC) subsystem full. The keepfill function verifies the check valve open position.

The licensee states that verification of closure for all check valves would entail a leak test.

Performance of a leakage test would require the plant to enter an LCO for a minimum of 4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Based on maintenance history, this represents a hardship in meeting the quarterly exercise requirements of ASME OM Code ISTC-3510, "Exercising Test Frequency," without a compensating increase in the level of quality and safety. This request does not represent a case for impractical condition. Examples of impractical conditions can be found in NUREG-1482 Revision 2, Section 2.4.5, "Deferring Valve Testing to Cold Shutdown or Refueling Outages."

The licensee requested relief on the basis of 10 CFR 50.55a(a)(3)(i), the proposed alternative would provide an acceptable level of quality and safety. However, the licensee's reason for the alternative proposal was the need to enter an LCO condition per each valve for 4 - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, which is considered a hardship. To maintain the quarterly exercise test frequency requirement represents an undue hardship without an increase in the level of quality and safety which is applicable to 10 CFR 50.55a(a)(3)(ii).

The licensee has requested that the close verification leak test be performed on a frequency of at least once per operating cycle in lieu of once each refueling outage. Each check valve has its own surveillance test that will be performed on a 24 month frequency interval. Execution of the close verification leakage test will be scheduled and completed using their in-house program governing maintenance. This program follows the requirements of 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants." Prior to a system being taken out of service, its effect on risk is evaluated. The NRC staff considers this to be an acceptable alternative that provides reasonable assurance that the components are operationally ready.

4.0 CONCLUSION

As set forth above, the NRC staff determined that the proposed alternatives described in requests 1 RR-01, 2RR-01, 1 RR-03, and 2RR-03 provide an acceptable level of quality and safety for valves listed in Tables 1 and 3. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i) for alternative requests 1 RR-01, 2RR-01, 1 RR-03 and 2RR-03, and is in compliance with the ASME OM Code requirements.

As set forth above, the NRC staff determined that the proposed alternatives, described in requests 1 RR-02, 2RR-02, and 1 RR-05, provide reasonable assurance that valves listed in Tables 2 and 4 are operationally ready. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii) for alternative requests 1 RR-02, 2RR-02, and 1 RR-05, and is in compliance with the ASME OM Code requirements.

As set forth above, the NRC staff determined that the proposed alternatives described in requests 1 RR-04 and 2RR-04, using ASME OM Code Case OMN-20 provides reasonable assurance of operational readiness of pumps and valves subject to the ASME OM Code 1ST, and imposing the Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii) for requests 1 RR-04 and 2RR-04, and is in compliance with the ASME OM Code requirements.

All other ASME OM Code requirements for which relief was not specifically requested and approved in the subject request remain applicable.

Principle Contributors: M. Farnan, NRR J. Huang, NRR Date: May 22, 2014

Therefore, the NRC staff authorizes the proposed alternatives in requests 1 RR-01, 2RR-01, 1 RR-02, 2RR-02, 1 RR-03, 2RR-03, 1 RR-04, 2RR-04, and 1 RR-05 for the fourth 1ST interval at SSES, Units 1 and 2, which is currently scheduled to begin on June 1, 2014, and end on May 31, 2024. All other ASME OM Code requirements for which relief was not specifically requested and approved in the subject request remain applicable.

If you have any questions, please contact the SSES Project Manager, Mr. Jeffrey A. Whited, at jeffrey.whited@nrc.gov or 301-415-4090.

Sincerely, IRA/

Meena K. Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ DISTRIBUTION:

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