ML13308C426

Relief from the Requirements of the American Society of Mechanical Engineers Code Regarding Extension of Pressure Retaining Boundary During System Leakage Test

ML13308C426
Person / Time
Site:	Saint Lucie
Issue date:	11/25/2013
From:	Lingam S P, Quichocho J F Plant Licensing Branch II
To:	Nazar M Florida Power & Light Co
Lingam S P
References
TAC MF0711
Download: ML13308C426 (8)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Mano Nazar Executive Vice President, and Chief Nuclear Officer Florida Power & Light Company P.O. Box 14000 Juno Beach, FL 33408-0420 November 25, 2013

SUBJECT:

ST. LUCIE PLANT, UNIT 2-RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE REGARDING EXTENSION OF PRESSURE RETAINING BOUNDARY DURING SYSTEM LEAKAGE TEST (TAC NO. MF0711)

Dear Mr. Nazar:

By letter dated February 13, 2013, as supplemented by letter dated July 3, 2013, and email dated August 14, 2013, Florida Power & Light Company (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for the use of an alternative to a certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI requirements at St. Lucie Plant (St. Lucie), Unit 2. The licensee proposed an alternative to a certain requirement of the ASME Code,Section XI. The licensee submitted Request for Alternative (RFA) 9 for the St. Lucie Plant (St. Lucie), Unit 2, and relates to the inservice inspection (lSI) requirement of IWB-5222(b) of the ASME Code 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), Part 50, Section 50.55a(a)(3)(ii), the licensee proposed an alternative 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 has reviewed the licensee's submittal and has determined that RFA 9 will provide reasonable assurance of structural integrity or leak tightness of the subject components and that 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 9 at St. Lucie, Unit 2, for the remainder of the third 1 0-year lSI interval that will end on August 7, 2014. All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. 9 REGARDING EXTENSION OF PRESSURE RETAINING BOUNDARY

1.0 INTRODUCTION

DURING SYSTEM LEAKAGE TEST FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT, UNIT 2 DOCKET NO. 50-389 By letter dated February 13, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13056A203), as supplemented by letter dated July 3, 2013, and email dated August 14, 2013 (ADAMS Accession Nos. ML 13192A326 and ML 13247A226, respectively), Florida Power and Light Company (the licensee) submitted for the U.S. Nuclear Regulatory Commission (NRC) approval Request for Alternative (RFA) 9. The licensee proposed an alternative to a certain requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. RFA 9 is submitted for the St. Lucie Plant (St. Lucie), Unit 2, and relates to the inservice inspection (lSI) requirements of IWB-5222(b) of the ASME Code 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), Part 50, Section 50.55a(a)(3)(ii), the licensee proposed an alternative pressure retaining boundary for the system leakage test on 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 Part 50 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 Enclosure 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 1 0 CFR 50.55a(a)(3) of 1 0 CFR Part 50 states, in part, that alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the 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 3.1 The licensee's Request for Alternative The ASME Code components affected by this request include:

• Class 1 piping-pressure retaining boundary in Table IWB-2500-1, Examination Category B-P, Item No. 815.50, of the reactor coolant system (RCS) and the safety injection and shutdown cooling as noted in Tables 1 and 2 of RFA 9.

• Class 1 valves-pressure retaining boundary in Table IWB-2500-1, Examination Category B-P, Item No. 815.70, of the RCS and the safety injection and shutdown cooling as noted in Tables 1 and 2 of RFA 9. The components for which an alternative is proposed are listed in Tables 1 and 2 of RFA 9. Table 1 of RFA 9 identifies small bore Class 1 RCS vent or drain piping segments.

Table 2 of RFA 9 identifies large bore Class 1 hot leg shutdown cooling suction and high pressure safety injection (HPSI) hot leg injection lines. The Code of record for the third 1 0-year lSI interval at St. Lucie, Unit 2, is the 1998 edition through 2000 addenda of the ASME Code. IWB-2500, Table IWB-2500-1, Examination Category B-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 test 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 test 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).

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 boundaries 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." ASME Code Case N-798 permits the boundaries of IWB-5222(a) to apply. 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 boundaries specified in ASME Code Case N-800 "Alternative Pressure Testing Requirements for Class 1 Piping between the First and Second Injection Valves." ASME Code Case N-800 permits the system leakage test to be conducted by pressurizing the Class 1 volume using the Class 2 safety system pressure.

ASME Code Cases N-798 and N-800 have not been accepted by the NRC in Regulatory Guide 1.147, Rev. 16. The basis for this request is provided in Sections 4 and 5 of RFA 9. During the proposed system leakage test, the piping segments identified in Table 1 of RFA 9 will remain in their normal operating configuration and will not be pressurized.

During the proposed system leakage test, the piping segments identified in Table 2 of RFA 9 will remain in their normal operating configuration and will not be pressurized to the RCS system pressure, but will be examined at full operating pressures commensurate with their respective safety functions.

The licensee submitted this request for the remainder of the third 1 0-year lSI interval that ended on August 7, 2013. By email dated August 14, 2013, the licensee clarified that the applicable requirements of IWA-2430 have been utilized to extend the third 1 0-year lSI interval to August 7, 2014. 3.2 NRC Staff Evaluation The NRC staff has evaluated RFA 9 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.

The NRC staff notes that the current RFA 9 is similar to the Request No. 29 that was authorized by the NRC on March 4, 2008, for the third 1 0-year lSI interval of St. Lucie, Unit 1 (ADAMS Accession No. ML080500075).

Within context of RFA 9, the NRC staff determined that the licensee provided adequate description and technical information to support the basis for a hardship or unusual difficulty.

The licensee's bases for hardship are as follows.

• To extend the pressure retaining boundary during system pressure test to all Class 1 pressure retaining components within the system boundary as required by IWB-5222(b), the licensee would have to make a number of unusual and temporary system alterations to pressurize the piping segments of Tables 1 and 2 of RFA 9. Examples of these temporary configurations include installation of temporary piping such as hard-pipe jumpers, altering the control logic, or connecting an external pressure source. Since the ASME Code requires that the Class 1 system pressure testing is conducted at a pressure not less than the pressure corresponding to 100 percent rated reactor power, the above mentioned alterations to these lines could conflict with Technical Specifications (TS).

• The isolation devices such as inboard isolation valves and outboard blind flanges, two valves, or two check valves in Class 1 piping segments of Tables 1 and 2 of RFA 9 provide double isolation barrier for the reactor coolant pressure boundary (RCPB). During normal plant operation, these valves are maintained in the closed position.

For vent and drain lines of Table 1 of RFA 9, the downstream pipe is not normally pressurized.

For safety injection and shutdown cooling piping segments of Table 2 of RFA 9, the downstream pipe is normally pressurized at pressure lower than normal operating pressure. Pressurizing these piping segments per IWB-5222(b) would necessitate manually opening the inboard isolation valves which defeats the double isolation barrier and reduces the margin of safety for personnel performing the test. Alternatively, connecting an external pressure source at each pipe segment location by way of installing temporary connections to pressurize the lines could conflict with TS.

• By establishing and restoring temporary configurations or defeating double isolation barriers in order to test these lines in accordance with IWB-5222(b), potential hazards to personnel safety could be introduced.

Examples include occupational hazards, risk for spills, contaminations, and additional radiation exposure.

• By letter dated July 3, 2013, the licensee submitted an estimate of personnel radiation exposure as a result of compliance with IWB-5222(b).

Personnel would receive approximately 3.872 person-rem for activities such as opening (normally closed) valves to pressurize the lines, performing additional tests, and restoring the system following the tests. Additional radiation exposure of approximately 0.513 person-rem would be received for connecting an external pump at four locations of the Class 1 safety injection piping to pressurize these lines to normal operating pressure.

The NRC staff finds that the above technical information constitutes a justifiable hardship or unusual difficulty if the ASME Code requirement were to be imposed upon the licensee.

By letter dated July 3, 2013, the licensee provided information on operating experience (OE) specifically related to potential degradation of welded connections in the piping segments of Tables 1 and 2 of RFA 9 due to fatigue or stress corrosion cracking (SCC). The licensee's OE review from St. Lucie, Units 1 and 2, and the fleet did not show any issues with sec or fatigue in the welded connections of the subject piping segments or similar piping configurations.

The external nuclear facilities contacted for OE indicated that they had not experienced any significant issues in the welded connections of similar piping configurations.

The NRC staff has determined that the small bore Class 1 RCS vent and drain lines identified in Table 1 of RFA 9 are not pressurized past the first isolation valve during normal operation.

These lines are equipped with the inboard isolation valve and an outboard blind flange or two valves in series that provide double isolation of the RCPB. Pressurizing these lines past the first isolation valve to perform a system leakage test in accordance with IWB-5222(b) defeats the double isolation requirements, results in hardship or unusual difficulties to the licensee, and presents occupational hazards and radiation exposure for those performing the test. In lieu of the requirement, the licensee will inspect the entire length of these lines by the required VT-2 visual examinations as part of the Class 1 system leakage test at the end of each refueling outage while these piping segments remain in their normal operating configurations.

The NRC staff finds that there would be no compensating increase in the level of quality and safety by pressurizing the lines past the first isolation valve during the system leakage test to comply with the ASME Code requirements because these piping segments are not pressurized during normal operation.

In addition, the licensee's OE review indicated that there has not been any evidence of potential degradations by sec or fatigue in the welded connections of the subject lines. Therefore, the NRC staff finds the licensee's proposed alternative for the RCS vent and drain lines identified in Table 1 of RFA 9 acceptable and provides reasonable assurance of structural integrity or leak tightness of these piping segments.

The NRC staff has determined that the large bore class 1 hot leg shutdown cooling suction piping segments identified in Table 2 of RFA 9 are equipped with two valves. These valves are interlocked to prevent opening until reactor coolant system pressure is below 350 pounds per square inch gauge and administratively controlled to be closed at pressures in excess of 275 pounds per square inch absolute to avoid over-pressurization of the shutdown cooling system. These lines are not pressurized past the first isolation valve but it is possible that the piping becomes pressurized due to minor leakage past the first isolation valve. To pressurize these lines past the first isolation valve to conduct a system leakage test in accordance with IWB-5222(b), a number of unusual temporary system configurations are needed that could conflict with TS, result in hardship or unusual difficulties to the licensee, and present occupational hazards and radiation exposure for those performing the test. In lieu of the requirement, the licensee will inspect the entire length of these lines by the required VT -2 visual examinations as part of the Class 1 system leakage test at the end of each refueling outage while these piping segments remain in their normal operating configurations.

In addition, these lines are visually examined when the shutdown cooling system is in service. These system leakage tests will provide assurance that the combined first and second isolation valves and piping between the valves are effective in maintaining the RCPB at normal operating temperature and pressure.

In addition, the licensee's OE review indicated that there has not been any evidence of potential degradations by sec or fatigue in the welded connections of the subject lines. Therefore, the NRC staff finds the licensee's proposed alternative for the hot leg shutdown cooling suction piping segments identified in Table 2 of RFA 9 is acceptable and provides reasonable assurance of structural integrity or leak tightness of these piping segments.

The NRC staff has determined that the large bore Class 1 HPSI/hot leg injection piping segments identified in Table 2 of RFA 9 are equipped with two check valves oriented to flow into the RCS that provide the design-required double isolation barrier for the RCPB. The licensee will not pressurize these lines past the first check valves. To pressurize these lines past the first check valve to conduct a system leakage test in accordance with IWB-5222(b), the licensee has to perform a number of unusual temporary system configurations.

These temporary configurations would challenge both the header check valves in the auxiliary building and the loop check closure at the RCS connection, and violate the design requirement for two primary coolant pressure boundary isolation devices. Furthermore, this results in hardship or unusual difficulties to the licensee and reduces the margin of safety for occupational hazards and radiation exposure for those performing the test. In lieu of the requirement, the licensee will inspect the entire length of these lines by the required VT -2 visual examinations as part of the Class 1 system leakage test at the end of each refueling outage while these piping segments remain in their normal operating configurations.

These lines are also visually examined during the HPSI system functional pressure test conducted at HPSI pump discharge pressure in accordance with the requirement to examine systems at their highest operating pressure.

In addition, the OE review indicated that there has not been any evidence of potential degradations by SCC or fatigue in the welded connections of the subject lines. Therefore, the NRC staff finds the licensee's proposed alternative for the HPSI/hot leg injection piping segments identified in Table 2 of RFA 9 acceptable and provides reasonable assurance of structural integrity or leak tightness of these piping segments.

As clarified by the licensee in the email dated August 14, 2013, the NRC staff notes that the third 1 0-year lSI interval has been extended by 1 year to August 7, 2014, in accordance with the applicable requirements of IWA-2430 of the ASME Code. On the basis of the above evaluation, the NRC staff finds that the proposed alternative in RFA 9 is acceptable for the remainder of the third 1 0-year lSI interval of St. Lucie, Unit 2, which will end on August 7, 2014.

4.0 CONCLUSION

As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity or leak tightness of the subject components 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 9 at St. Lucie, Unit 2, for the remainder of the third 1 0-year lSI interval which will end on August 7, 2014. 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:

Ali Rezai Date: November 25, 2013 M. Nazar If you have any questions, please contact the Project Manager, Mr. Siva P. Lingam by phone at 301-415-1564 or via e-mail at Siva.Linqam@

nrc.gov. Docket No. 50-389

Enclosure:

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

PUBLIC LPL2-2 r/f RidsNrrDorllpl2-2 RidsNrrDeEpnb RidsAcrsAcnw_MaiiCTR RidsNrrLABCiayton RidsNrrPMStLucie RidsRgn2MaiiCenter V. Campbell, EDO All A. Rezai, NRR Sincerely, IRA/ Jessie F. Quichocho, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ADAMS Accession No.: ML 13308C426

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LPL2-2/BC NAME Slingam BCiayton Tlupold JQuichocho DATE 11/05/13 11/05/13 9/17/13 11/25/13 OFFICIAL RECORD COPY