ML14133A112

Relief Request 11-SPT-1 for Alternative System Leakage Test

ML14133A112
Person / Time
Site:	Sequoyah
Issue date:	05/20/2014
From:	Lisa Regner Plant Licensing Branch II
To:	James Shea Tennessee Valley Authority
Hon A DORL/LPL2-2 301-415-8480
References
TAC MF2327, TAC MF2328
Download: ML14133A112 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801 May 20, 2014

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNIT 1 AND 2-RELIEF REQUEST 11-SPT-1 FOR ALTERNATIVE SYSTEM LEAKAGE TEST (TAC NOS. MF2327 AND MF2328)

Dear Mr. Shea:

By letter dated June 26, 2013 (Agencywide Documents Access and Management Systems (ADAMS) Accession No. ML13178A280), as supplemented by letter dated September 25, 2013 (ADAMS Accession No. ML13273A269), Tennessee Valley Authority (the licensee) requested relief from a certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), for the third 10-year lnservice Inspection (lSI) Program at Sequoyah Nuclear Plant (SQN), Units 1 and 2. The licensee submitted Relief Request RFA 11-SPT -1 as an alternative for system leakage testing conducted at or near the end of the inspection interval.

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

Section 50.55a(a)(3)(ii), the licensee requested relief from article IWB-5222(b) of Section XI of the ASME Code regarding pressure retaining boundary for the system leakage test, on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The NRC staff reviewed the subject request and concludes, as set forth in the enclosed safety evaluation that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.552(a)(3)(ii). Therefore, the NRC staff authorizes the use of RFA 11-SPT-1 at SQN, Units 1 and 2, for the third 1 0-year lSI interval that commenced on June 1, 2006, and will end on April30, 2016. Use of ASME Code Cases N-798 and N-800 is authorized and their use is limited to the end of the third 1 0-year lSI interval or until such time as ASME Code Cases N-798 and N-800 are published in a future version of Regulatory Guide (RG) 1.147 and incorporated by reference in 10 CFR 50.55a(b), whichever occurs earlier. At that time, if the licensee intends to continue implementing these ASME Code cases, it must follow all provisions of ASME Code Cases N-798 and N-800 with conditions as specified in RG 1.147 and 10 CFR 50.55a(b)(4), (b)(5), and (b)(6), if any.

J.Shea All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.

If you have any questions, please contact the Project Manager, Andrew Hon at 301-415-8480.

Docket Nos. 50-327 and 50-328

Enclosure:

Safety Evaluation cc w/enclosure: Distribution via ListServ Lisa M. Regner, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE 11-SPT-1: PRESSURE RETAINING BOUNDARY DURING SYSTEM LEAKAGE TEST SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NUMBERS 50-327 AND 50-328

1.0 INTRODUCTION

By letter dated June 26, 2013 (Agencywide Documents Access and Management Systems (ADAMS) Accession No. ML13178A280), as supplemented by letter dated September 25, 2013 (ADAMS Accession No. ML13273A269), Tennessee Valley Authority (the licensee) requested relief from a certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), for the third 10-year lnservice Inspection (lSI) Program at Sequoyah Nuclear Plant (SQN), Units 1 and 2. The licensee submitted Relief Request RFA 11-SPT-1 as an alternative for system leakage testing conducted at or near the end of the inspection interval.

Specifically, pursuant to Title 1 0 of the Code of Federal Regulations ( 1 0 CFR),

Section 50.55a(a)(3)(ii), the licensee requested relief from article IWB-5222(b) of Section XI of the ASME Code regarding pressure retaining boundary for the system leakage test, on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

2.0 REGULATORY EVALUATION

Section 50.55a(g)(4) of 10 CFR specifies that ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 1 0-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, incorporated by reference in 10 CFR 50.55a(b ), 12 months prior to the start of the 120-month interval, subject to the conditions listed therein.

Section 50.55a(a)(3) of 10 CFR states that alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC), if the licensee demonstrates (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

RELIEF REQUEST 11-SPT-1 The components affected by this request include ASME Code Class 1, Table IWB-2500-1, Examination Category 8-P, Item No. 815.10, pressure retaining components of the safety injection system (SIS), chemical volume and control system (CVCS), the residual heat removal (RHR) system, and the reactor coolant system (RCS).

The components for which an alternative is proposed are identified in Tables 1, 2, 3, and 4 of RFA 11-SPT-1. Table 1 identifies piping segments in the SIS that consist of the cold leg accumulator (CLA) piping and the emergency core cooling system (ECCS) piping. Table 2 identifies piping segments in the eves that consist of the reactor coolant pump (RCP) seal injection piping, the normal and alternate charging piping, and the pressurizer spray piping.

Table 3 identifies piping segments in the RHR system. Table 4 identifies piping segments in the RCS that consist of the RCS loop drain piping and the reactor vessel vent piping.

The code of record for the third 1 0-year lSI interval at SQN, Units 1 and 2, is the 2001 Edition through 2003 Addenda of the ASME Code.

ASME Code,Section XI, IWB-2500, Table IWB-2500-1, Examination Category 8-P, establishes requirements to conduct the system leakage test and the VT-2 visual examination in accordance with IWB-5220 and IWA-5240, respectively, prior to plant startup following each refueling outage. In accordance with IWB-5221 (a), the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100 percent rated reactor power. In accordance with IWB-5222(a), the pressure retaining boundary during system leakage testing shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup. The visual examination shall, however, extend to and include the second closed valve at the boundary extremity. In accordance with IWB-5222(b ), the pressure retaining boundary during system leakage testing conducted at or near the end of each inspection interval shall extend to all Class 1 pressure retaining components within the system boundary.

The licensee proposed an alternative to IWB-5222(b ). The proposed alternative is to use the boundary prescribed by IWB-5222(a). For portions of Class 1 piping between the first and the second vent, drain, and test isolation devices that normally remain closed during plant operation, the licensee proposed to use the requirements specified in ASME Code Case N-798 "Alternative Pressure Testing Requirements for Class 1 Piping between the First and Second Vent, Drain, and Test Isolation Devices." For portions of the Class 1 boundary between the first and the second isolation valves in the injection and return path of standby safety systems, the licensee proposed to use the requirements specified in ASME Code Case N-800 "Alternative Pressure Testing Requirements for Class 1 Piping Between the First and Second Injection Valves." The NRC staff has not approved ASME Code Cases N-798 and N-800 in Regulatory Guide (RG) 1.147, Rev. 16, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1." Pursuant to 10 CFR 50.55a(a)(3), licensees are permitted to submit alternatives such as Code Cases N-798 and N-800 to the ASME Code requirements as presented in the subject relief request. Use of the above code cases as an alternative to the ASME Code,Section XI, requires the NRC pre-authorization.

The licensee submitted RFA 11-SPT-1 on the basis of hardship or unusual difficulty. The basis for hardship for each piping segment is discussed below.

TheCLA piping segments identified in Table 1 of RFA 11-SPT-1 will be pressurized to 650 pounds per square inch gauge (psig) as a result of the CLA pressure required by the Technical Specification (TS) Limiting Condition for Operation (LCO) 3.5.1.1 when the RCS is at 100 percent rated reactor power. These piping segments cannot be raised to full RCS pressure and temperature because this would require isolation of the CLA during plant operation.

Isolation of the CLA would be required to prevent overpressurization of the accumulator tank above the relief valve setpoint of 700 psig, and would require entry into the TS LCO 3.5.1.1 action statement to restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Therefore, failure of the lone isolation valve or leakage past the valve during the test, and compliance with the TS requirements could ultimately result in an unplanned reactor shutdown. The licensee estimated that extending the boundary as required by the ASME Code,Section XI, for testing these piping segments would require two people working 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to erect scaffolding, perform the leak inspection, and remove the scaffolding. In addition, it would require one person working 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to isolate the accumulator tanks and realign the valves following completion of the testing. In total, performing work on these piping segments would expose the worker to an accrual of 1.2 roentgen equivalent man (rem) of additional radiological dose.

The remaining piping segments identified in Table 1 of RFA 11-SPT-1 are associated with ECCS injection and are at static head pressure and temperature when the RCS is at normal operating pressure and temperature, and are separated from the RCS conditions by self-operating check valves. The only method of achieving full RCS pressure conditions in the Class 1 piping between the primary and secondary isolation check valves would be by:

(1) installing temporary hoses around the primary check valve that would defeat the double isolation requirement for the reactor coolant pressure boundary or (2) making permanent modifications to the plant and installing qualified piping solely for the purpose of conducting the infrequently performed ASME Code,Section XI, pressure test. The licensee estimated that extending the boundary for the ASME Code testing of these piping segments would require four people working 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to erect scaffold, install temporary test hoses, perform the leak inspection, disassemble the scaffolding, and return the configuration to normal. In total, work on the piping segments would result in accrual of 4.4 rem of additional radiological dose.

The RCP seal injection, the normal and alternate charging, and the pressurizer spray piping segments identified in Table 2 of RFA 11-SPT-1 will be pressurized to 2400 psig (the charging pump discharge pressure) and 110 degrees Fahrenheit (°F) when the RCS is at normal operating pressure and temperature. These segments are at a pressure higher than the RCS pressure but below the RCS temperature. The temperature is maintained below the RCS temperature to protect the RCP seals from premature failure, preclude failure of the lower radial bearing, and prevent warping of the RCP shaft. At full power, the temperature of the RCP seal injection piping remains at 110 °F. If the temperature exceeds 179 °F, the main control room receives an alarm for high temperature. Above 230 °F, operating procedures require the pump to be tripped. In addition, the design temperature for the volume control tank (seal flow suction source) is 250 °F. Increasing the temperature to nominal RCS temperature would violate the design criteria for the tank. Therefore, applying the ASME Code,Section XI, required temperature to the piping segments would adversely affect plant safety by requiring actions and procedural provisions intended to safely operate plant equipment and prevent equipment damage to be bypassed during the conduct of the testing.

The RHR piping segment identified in Table 3 of RFA 11-SPT-1 removes the heat from the RCS. When shutting down the plant, plant procedures place the RHR system in service when the RCS temperature and pressure is less than 235 °F and 350 psig, respectively. When starting-up the plant, RHR remains in service until the RCS temperature is greater than 200 °F and RCS pressure is between 325 psig and 350 psig. The SQN Updated Final Safety Analysis Report specified safety system pressure interlocks prevent the RHR supply valves (SQN-1/2-FCV-74-1 and -2) from being opened until RCS pressure is less than 380 psig. In addition, the circuit breakers to the valves are opened and administratively locked prior to raising RCS temperature above 350 °F. These interlock actions have been put in place for the following reasons: to prevent overpressurization of the downstream RHR piping that has the design pressure and temperature ratings of 600 psig and 400 °F, respectively, and to prevent loss of the RCS inventory to the refueling water storage tank in the event that the valves do not isolate. Pressurizing the piping between these valves to nominal RCS operating pressure and temperature by opening the SQN-1/2-FCV-74-1 or installing test jumpers would require defeating a safety feature and the redundancy afforded by double isolation valves, thus reducing plant safety. The licensee estimated that extending the boundary as required by the ASME Code,Section XI, for this piping segment would require two people working 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to install temporary test hoses (or open the inboard valve), perform the leak inspection, and return the configuration to normal. In total, this piping segment would result in accrual of 0.56 rem of additional radiological dose.

The Class 1 RCS vents and drains piping segments identified in Table 4 of RFA 11-SPT-1 are equipped with inboard isolation valves and outboard isolation valves (or blind flanges). The valves are maintained in the closed position during normal plant operation, and the piping downstream of the inboard valve is not normally pressurized. In order to pressurize those piping segments as required by IWB-5222(b ), it would be necessary to manually open the inboard valves to pressurize the piping and connections. Pressurization by this method defeats the double isolation and reduces the margin of personnel safety for those performing the test, and could result in overpressurization of the Reactor Coolant Drain Tank if the lone isolation valve were to leak during the test. In addition, there is currently no available method to depressurize the downstream piping following the test completion; thus, placing the plant in an abnormal condition. The licensee estimated that extending the boundary as required by the ASME Code,Section XI, for these piping segments would require five people working 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to open the manual valves, perform the leak inspection, and restore the valves to their normal configuration. In total, these piping segments would result in accrual of 0.52 rem of additional radiological dose.

In a letter dated September 25, 2013, the licensee stated that several nuclear facilities, including St. Lucie, Turkey Point, Point Beach, SQN, and Watts Bar power plants, were surveyed for information regarding known issues with stress corrosion cracking (SCC) or fatigue in socket or butt welds in similar piping. They had not experienced any significant issues with these degradation mechanisms in similar piping configurations. The welds in the piping segments described in Tables 1 through 4 of RFA 11-SPT-1 are also included in the SQN risk informed (RI )-lSI program during the second and third 1 0-year lSI intervals. No failure in the welded connections has been identified by the RI-ISI program. Based on review of the results under the SQN Corrective Action Program and the Boric Acid Corrosion Control Program, no pressure boundary leaks were identified to have occurred during the third 1 0-year lSI interval that involved the piping segments described in Table 1 through 4 of RFA 11-SPT-1.

The licensee submitted this request for SQN, Units 1 and 2, for the third 1 0-year lSI interval that commenced on June 1, 2006, and will end on April 30, 2016.

NRC Staff Evaluation

The NRC staff has evaluated RFA 11-SPT-1 pursuant to 10 CFR 50.55a(a)(3)(ii). The NRC staff focuses on whether compliance with the specified requirements of 10 CFR 50.55a(g), or portions thereof, would result in hardship or unusual difficulty, and if there is a compensating increase in the level of quality and safety despite the hardship.

Safety injection system (Table 1 of RFA 11-SPT-1)

The NRC staff determined that requiring the licensee to comply with IWB-5222(b) and conduct system leakage test of the CLA and ECCS piping segments at a pressure corresponding to 100 percent rated reactor power would result in hardship. The CLA and ECCS piping segments with isolation valves and check valves were designed to serve as double isolation barrier to the reactor coolant pressure boundary. Opening and bypassing the valve to pressurize the CLA and ECCS piping to perform the required pressure test defeats the double isolation criteria, and reduces safety of the plant operation. For testing of the CLA piping, opening the valve also creates a conflict with the plant TSs and puts the plant in abnormal conditions that may ultimately lead to a reactor shutdown. Both opening the valve and bypassing it by making design modification to the valve's existing configuration to accommodate system leakage testing of the CLA and ECCS piping exposes the licensee's personnel to a radiation dose, which would be an as low as reasonably achievable (ALARA) concern.

For the CLA piping segments, the NRC staff finds the proposed alternative to the system leakage testing in accordance with IWB-5222(a) acceptable because the proposed testing is performed when the piping segments are subjected to a pressure and a temperature that they are designed to operate. Specifically, the CLA piping will be tested at the accumulator tank operating pressure and temperature since the tank is designed to inject at a maximum pressure of 650 psig and a maximum temperature of 130 °F. Within the above normal operating parameters, the required VT-2 visual examinations that accompany the system leakage testing would detect any leakage if it originated from an existing flaw in the CLA piping segments and their associated connections. For the ECCS piping segments, the NRC staff finds the proposed system leakage testing acceptable because the piping will be tested at a pressure and a temperature that are within the normal operating condition of the piping. Specifically, the ECCS piping segments are pressurized to their normal operating pressure of 1500 psig when the safety injection pump is in operation. The ECCS pipe operates at an ambient temperature since the suction to the pump is from the refueling water storage tank maintained at ambient temperature. Within the above normal operating parameters, the required VT-2 visual examinations that accompany the system leakage testing would detect any leakage, if it originated from an existing flaw in the ECCS piping segments and their associated connections.

Furthermore, the NRC staff has not identified any documented operational experience (OE) regarding the SCC and fatigue in the socket and butt welds of the subject CLA and ECCS or similar piping configurations. The NRC's OE review has not identified any pressure boundary leaks in the CLA and ECCS piping segments either. Therefore, the NRC staff finds that the proposed test provides a reasonable assurance of structural integrity and leak tightness of the CLA and ECCS piping segments.

Chemical volume and control system (Table 2 of RFA 11-SPT-1)

The NRC staff determined that conducting system leakage testing in accordance with IWB-5222(b) of the RCP seal injection, the normal and alternate charging, and the pressurizer spray piping segments at 100 percent rated reactor power would create hardship to the licensee. To comply with the ASME Code requirement, the licensee has to subject the RCP seals to the RCS temperature that is higher than the seal's normal operating temperature. This places the plant in abnormal and unsafe operating conditions that would cause the pump to be tripped to prevent equipment damage.

For the RCP seal injection, the normal and alternate charging, and the pressurizer spray piping segments, the NRC staff finds the proposed system leakage testing acceptable because the piping will be subjected to a pressure and a temperature that are designed for normal operation.

Specifically, the system leakage testing will be performed at the pipe normal operating pressure of 2400 psig and temperature of 110 °F. Within the above normal operating parameters, the required VT-2 visual examinations that accompany the proposed system leakage testing would detect any leakage if it originated from an existing flaw in the subject piping segments and their associated connections. Furthermore, the NRC staff has not identified any documented issues regarding the SCC and fatigue in the socket and butt welds of the subject piping segments or similar piping configurations. The NRC staff has not identified any pressure boundary leaks in the subject piping segments either. Therefore, the NRC staff finds that the proposed test provides a reasonable assurance of structural integrity and leak tightness of the RCP seal injection, the normal and alternate charging, and the pressurizer spray piping segments.

Residual heat removal system (Table 3 of RFA 11-SPT-1)

The NRC staff determined that it would be a hardship for the licensee to conduct IWB-5222(b) required system leakage testing of the RHR piping segments. In order to comply with IWB-5222(b ), the licensee has to either open the inboard isolation valve or bypass the valve by making design modifications to the existing piping configuration. Both opening and bypassing the valve defeat the double isolation criteria. Furthermore, additional radiation dose incurred by the licensee's personnel performing activities such as opening or bypassing the valve and conducting the IWB-5222(b) system leakage test would be an ALARA concern.

For the RHR piping segments, the NRC staff finds the proposed system leakage testing acceptable because the piping will be subjected to a pressure and a temperature that are designed for normal operation. Specifically, the system leakage testing is performed when the RHR system is in service for heat removal (i.e., when starting up and shutting down the plant).

Within the normal operating parameters of RHR system during heat removal, the required VT-2 visual examinations that accompany the proposed system leakage testing would detect any leakage if it originated from an existing flaw in the subject piping segments and their associated connections. Furthermore, the NRC staff's OE reviews have not identified any documented issues with the sec and fatigue in the socket and butt welds of the subject RHR or similar piping configurations. The NRC staff has not identified any pressure boundary leaks in the RHR piping segments either. Therefore, the NRC staff finds that the proposed test provides a reasonable assurance of structural integrity and leak tightness of the RHR piping segments.

Reactor coolant system (Table 4 of R FA 11-SPT-1)

The NRC staff determined that requiring the licensee to conduct the IWB-5222{b) required system leakage test of the RCS vents and drains piping segments at 1 00 percent rated reactor power would result in hardship. By imposing the IWB-5222(b) requirement, the licensee has to manually open the inboard valve to pressurize the RCS vent and drain piping, which places the plant in abnormal conditions. The inability to depressurize the piping after completion of the test is an additional difficulty and a concern to personnel safety. Opening of the inboard valve defeats the double isolation criteria. Furthermore, both opening the valve and performing the ASME Code,Section XI, compliant test would expose the licensee's personnel to additional radiation dose, which would be an ALARA concern.

For the RCS vents and drains piping segments, the NRC staff finds the proposed system leakage testing acceptable because the piping between the first and second isolation valve normally remains closed during plant operation and the downstream of the inboard valve is not pressurized. The NRC staff finds that performing the proposed system leakage testing with all valves in the position required during plant startup, and extending the required VT-2 visual examinations to include the second (outboard) closed valve, provides reasonable assurance of structural integrity and leak tightness of the subject RCS vents and drains piping segments.

Furthermore, the NRC staff's OE review has not identified any documented issues with SCC and fatigue that resulted in pressure boundary leaks in the socket and butt welds of the subject RCS vents and drains or similar piping configurations.

Piping segments in Table 1 through 4 of RFA 11-SPT-1 In addition, the SON RI-ISI program includes in its risk informed process, the piping segments described in Tables 1 through 4 of RFA 11-SPT -1. This means that the welds in the subject piping have also been evaluated and periodically examined in accordance with the requirements of the SON RI-ISI program. The risk informed evaluation provides an additional assurance of the structural integrity and leak tightness of the piping identified in Tables 1 through 4 of this request.

Use of ASME Code Cases N-798 and N-800 The ASME Code committees approved the ASME Code Cases N-798 and N-800. However, the NRC has not yet accepted ASME Code Cases N-798 and N-800 in RG 1.147 by rulemaking (1 0 CFR 50.55a). The NRC staff authorizes use of ASME Code Cases N-798 and N-800 and their use is limited to the end of the third 1 0-year lSI interval or until such time as these code cases are published in a future version of RG 1.147 and incorporated by reference in 10 CFR 50.55a(b), whichever occurs earlier. At that time, if the licensee intends to continue implementing these code cases, it must follow all provisions of ASME Code Cases N-798 and N-800 with conditions as specified in RG 1.147 and 10 CFR 50.55a(b)(4), (b)(5), and (b)(6), if any.

On the basis of the above evaluation, the NRC staff finds that complying with the requirement specified in IWB-5222(b) would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff also finds that the licensee's proposed alternative is acceptable and provides a reasonable assurance of the structural integrity and leak tightness of the subject piping segments and their associated welded connections.

CONCLUSION As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity and leak tightness of the subject piping segments and the associated welded connections, and complying with the specified requirement 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).

Therefore, the NRC staff authorizes the use of RFA 11-SPT-1 at SQN, Units 1 and 2, for the third 1 0-year lSI interval that commenced on June 1, 2006, and will end on April 30, 2016. Use of ASME Code Cases N-798 and N-800 is authorized and their use is limited to the end of the third 10-year lSI interval or until such time as ASME Code Cases N-798 and N-800 are published in a future version of RG 1.147 and incorporated by reference in 10 CFR 50.55a(b),

whichever occurs earlier. At that time, if the licensee intends to continue implementing these ASME Code cases, it must follow all provisions of ASME Code Cases N-798 and N-800 with conditions as specified in RG 1.147 and 10 CFR 50.55a(b)(4), (b)(5), and (b)(6), if any.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.

Principal Contributor: A. Rezai Date: May 20, 2014

J.Shea All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.

If you have any questions, please contact the Project Manager, Andrew Hon at 301-415-8480.

Docket Nos. 50-327 and 50-328

Enclosure:

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

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