ML120750481

Safety Evaluation in Support of 10 CFR 50.55a Request RR-A35, Proposed Alternative to System Leakage Test Requirements

ML120750481
Person / Time
Site:	Davis Besse
Issue date:	04/09/2012
From:	Jacob Zimmerman Plant Licensing Branch III
To:	Allen B FirstEnergy Nuclear Generation Corp
mahoney m
References
TAC ME6056
Download: ML120750481 (10)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 9, 2012 Mr. Barry S. Allen Site Vice President FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Mail Stop A-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 - SAFETY EVALUATION IN SUPPORT OF 10 CFR 50.55a REQUEST RR-A35, PROPOSED ALTERNATIVE TO SYSTEM LEAKAGE TEST REQUIREMENTS (TAC NO. ME6056)

Dear Mr. Allen:

By letter to the Nuclear Regulatory Commission (NRC) dated April 15, 2011, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML11109A119) as supplemented by letter dated December 2, 2011, (ADAMS Accession No. ML113390047),

FirstEnergy Nuclear Operating Company, the licensee, submitted a proposed alternative to certain requirements associated with the inservice testing proposed for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS).

The NRC staff has reviewed the licensee's submittals and concludes that the licensee's proposed alternative provides reasonable assurance of structural integrity and leak tightness, and that complying with the specified ASME Code,Section XI 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 Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(ii). Therefore, the NRC staff authorizes the proposed alternative for DBNPS for the duration of the third 1O-year inservice inspection interval, currently scheduled to end on September 20, 2012.

The NRC staff's safety evaluation is enclosed.

B.Allen -2 Please contact the DBNPS Project Manager, Michael Mahoney at (301) 415-3867 if you have any questions on this action.

Sincerely,

( \ . OJK ) .../.

\ P LA! AJA/) S-&V-acob I. Zimmerman, Chief lant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST RR-A35, PROPOSED ALTERNATIVE TO SYSTEM LEAKAGE TEST REQUIREMENTS FIRSTENERGY NUCLEAR OPERATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 DOCKET NO. 50-346

1. 0 INTRODUCTION By letter to the Nuclear Regulatory Commission (NRC) dated April 15, 2011, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML11109A119), as supplemented by letter dated December 2, 2011, (ADAMS Accession No. ML113390047),

FirstEnergy Nuclear Operating Company (the licensee) submitted relief request RR-A35, "Proposed Alternative to System Leakage Test Requirements" for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS).

The licensee requests relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, IWB-5222(b), required extent of the reactor coolant pressure boundary (RCPB) system leakage test to be conducted at or near the end of the inservice inspection (lSI) interval on certain Class 1 piping segments. The licensee states that performing the system leakage test to the extent required by IWB-5222(b) would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

2.0 REGULATORY EVALUATION

Pursuant to Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50). paragraph 55a(g)(4), "Inservice Inspection Requirements", ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice 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 10-year lSI interval and subsequent 1O-year lSI 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 inspection interval, subject to the limitations and modifications listed therein.

Enclosure

-2 Paragraph 55a(a}(3} of 10 CFR 50 states that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an accepta_ble 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. The NRC staff reviewed and evaluated the licensee's request pursuant to 10 CFR 50.55a(a)(3)(ii}.

The code of record for the DBNPS third-10-year lSI interval, scheduled to end on September 20, 2012, is the 1995 Edition through the 1996 Addenda of Section XI of the ASME Code.

3.0 TECHNICAL EVALUATION

3.1 Test Zone/Component(s) For Which Relief Is Requested ASME Class 1 components (piping and valves) in the following DBNPS pressure test zones:

Test Zone Components Auxiliary pressurizer spray line components at the following location:

DH16 Between RC51, Auxiliary Spray from Decay Heat Pump 2 Check Valve, and DH2735, Decay Heat Auxiliary Spray Stop Valve.

Reactor coolant system (RCS) components at the following three locations:

Between HP48, High Pressure Injection (HPI) Train 1-1 Stop Check Valve, and HP50, HPI Train 1-1 Check Valve RC01 Between HP49, HPI Train 1-2 Stop Check Valve, and HP51, HPI Train 1-2 Check Valve Between HP57, HPI Train 2-1 Stop Check Valve, and HP59, High Pressure Injection Train 2-1 Check Valve.

3.2 ASME Code Requirements ASME Code,Section XI, Table IWB-2500-1, Examination Category B-P, Item B15.10 requires that all Class 1 pressure retaining components be visually examined (VT -2) each refueling outage, and a system leakage test be conducted in accordance with IWB-5220.

IWB-5221 requires that the system leakage test be conducted at a pressure not less than nominal pressure associated with normal system operation.

IWB-5222(a) requires that the pressure retaining boundary during the system leakage test correspond to the reactor coolant pressure boundary, with all valves in the position required for normal reactor operation startup, with the VT-2 examination extending to and including the second closed valve at the boundary extremity.

-3 IWB-5222(b) requires that the pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval extend to all Class 1 pressure retaining components within the system boundary.

3.3 Licensee's Basis for Requesting Relief (as stated)

Normal RCS [operating] pressure is approximately 2155 [pounds per square inch, gauge] psig. For Test Zone DH16 and RC01, the components cited above are separated from RCS pressure by check valves and are not normally subjected to 2155 psig.

For Test Zone DH 16, an alternate test pressure rig is required to pressurize the piping between the first and second isolation valves. The first valve off the reactor coolant system is a check valve. A pressure differential between the RCS and the test boundary is required to ensure the check valve remains closed. For this check valve to remain closed, the maximum attainable test pressure would be less than the 2155 psig required by IWB-5221 (a). Maintaining a differential pressure to ensure no fluid intrusion of non-borated water into the RCS (reactivity control concern), while meeting the IWB-5221 (a) requirements, is considered unusually difficult.

External pressurization also requires the use of a non-Code test pressure rig.

This device and its temporary system connections are not qualified to meet ASME Code Class 1 requirements should seat leakage occur at the first isolation valve during testing. Should a test rig failure occur, there is a potential for personnel injury, and manual isolation of the leak in an adverse environment would be required. Use of non-qualified materials to maintain reactor coolant pressure boundary (RCPB) integrity between the first and second isolation valves, at normal RCS pressure, also conflicts with the double isolation principle noted in 10 CFR 50.55a(c)(2)(iii).

Hydrostatic testing of this segment of the auxiliary pressurizer spray line, with the RCS out of service, was also considered and involves the following in-plant actions:

1. Remove check valve [RC51] insulation;

2. Grind off cover to casing seal weld;

3. Disassemble the valve and install a hydrostatic plug;

4. Temporarily reassemble the valve;

5. Perform the system leakage test;

6. Disassemble the valve and remove the hydrostatic plug;

7. Reassemble the valve;

8. Seal weld cover to casing; and

9. Reinstall valve insulation.

The radiological dose rate at check valve RC51 is approximately 80 milli[Roentgen equivalent man]Rem/hour. It is estimated that the identified

actions require approximately 11 man-hours to complete, resulting in a cumulative radiological exposure of approximately 0.9 man-Rem.

For Test Zone RC01, the HPI check valves are welded back-to-back; there are no test connections that would permit pressurization between the two check valves.

Pressurization upstream of the check valves would require abnormal system lineups and would likely cause a thermal transient cycle on the respective train's HPI nozzles, which are required to count against the rapid RCS depressurization transient design cycle limits. By analysis, there are a limited number of transient cycles available throughout the life of the plant.

Hydrostatic testing of the HPI check valves, with the RCS out of service, was also considered and involves the following in-plant actions:

1. Erect scaffold to access the inner check valve [HP50, HP51, or HP59};

2. Remove valve insulation;

3. Disassemble the valve and install a hydrostatic plug;

4. Temporarily reassemble the valve;

5. Perform the system leakage test;

6. Disassemble the valve and remove the hydrostatic plug;

7. Reassemble the valve;

8. Reinstall valve insulation; and

9. Remove scaffold.

Radiological dose rates at check valves HP50, HP51, and HP59 are approximately 50,50, and 130 milliRem/hour for trains 1-1,1-2, and 2-1, respectively. It is estimated that the identified in-plant actions require approximately 23 man-hours per check valve to complete resulting, in a cumulative radiological exposure of approximately 5.3 man-Rem.

Based on the reasons noted above, the licensee believes that compliance with the specified ASME Code requirements, for the Test Zone DH16 and RC01 Class 1 components cited above, would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The licensee also cites the following precedents where other licensees received authorization for use of similar alternatives.

1. Letter dated June 14, 2010, "McGuire Nuclear Station, Units 1 and 2 - Relief 09-Mn-005 for Alternative Leakage Testing for Various American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), Class 1 Piping and Components During the Third 10-Year Inservice Inspection (lSI) Interval (TAC Nos. ME1732 and ME1733)," ADAMS Accession Number ML101580422

2. Letter dated November 2, 2010, "North Anna Power Station (NAPS), Unit No.2, Fourth 10-Year Inservice Inspection (lSI) Interval, Relief Requests SPT-001 Through SPT-006 (TAC Nos. ME3311 through ME3316)," ADAMS Accession Number ML102510218

- 5 3.4 Licensee's Proposed Alternative The licensee proposes to use VT-2 examinations and reduced pressure testing as an alternative for the Code-required pressure testing described in Section 3.2 of this safety evaluation (SE).

The Test Zone DH16 and RC01 components cited above will be VT-2 examined for leakage each refueling outage as part of the Class 1 system leakage test with all valves in the positions required for normal reactor operation startup. The VT-2 examination will extend to and include the second closed valve at the test boundary extremity.

Within Test Zone DH16, the area between the check valve and stop valve is not normally pressurized. This piping segment will be VT-2 examined during the Class 2 decay heat removal (DHR) system leakage test conducted at or near the end of the 10-year lSI interval. This segment is pressurized to approximately 77 pounds per square inch, absolute (psia) during this test.

Within Test Zone RC01, the area between the HPI check valves is not normally pressurized.

The HPI check valves will be VT -2 examined during the Class 2 HPI system leakage test conducted at or near the end of the 10-year lSI interval. This segment is pressurized to approximately 58 psia during this test.

3.5 NRC Staff Evaluation ASME Code,Section XI, Table IWB-2500-1, Examination Category B-P, requires that pressure retaining components be tested in accordance with IWB-5220. IWB-5222(b) requires that the pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval extend to all Class 1 pressure retaining components within the system boundary. The licensee has proposed an alternative to the system leakage test requirements of the ASME Code for the components detailed in Section 3.1 of this SE.

Test Zone DH16 The licensee has identified two possible methods of pressure testing Test Zone DH16. The first method is to pressurize the segment to the required pressure while the RCS pressure is 2155 psig. The licensee states that the first valve off the RCS is a check valve and a pressure differential between the RCS and the test boundary is required to ensure the check valve remains closed. For this check valve to remain closed, the maximum attainable test pressure would be less than the 2155 psig required by IWB-5221 (a). Maintaining a differential pressure across the check valve to prevent non-borated water fluid intrusion, while meeting the IWB-5221 (a) requirements, is considered unusually difficult. Non-borated water fluid intrusion into the RCS is a reactivity control concern which would produce an off-normal plant transient.

In addition, pressurization to 2155 psig would require external, non-Code pressurization equipment. Use of such equipment could introduce non-qualified materials into the RCS pressure boundary. The NRC staff finds that pressurizing the subject segment to the required 2155 psig pressure during plant operation would constitute a hardship as the result of the possibility of an off-normal plant transient.

-6 The second method of pressurizing the segment to the required pressure would require significant modification of check valve RC51, performance of the test then again modifying the valve to its original condition. The effort to perform these modifications is accompanied by a cumulative radiological exposure of approximately 0.9 man-Rem. The NRC staff finds that the radiological dose which accompanies this test method also would constitute a hardship.

In lieu of performing the required pressure test at full system pressure, the licensee proposes to perform a VT-2 examination for leakage each refueling outage with all valves in the positions required for normal reactor operation startup as part of the Class 1 system leakage test. In addition, the licensee proposes to perform a VT-2 examination of this piping segment when the segment is at a pressure of 62 psig during the Class 2 DHR system leakage test conducted at or near the end of the 10-year lSI interval. The VT-2 examination will extend to and include the second closed valve at the test boundary extremity. The licensee states that the subject piping is 1.5 inch nominal pipe size stainless steel that has a design pressure of 2500 psig and does not contain any primary water stress corrosion cracking (PWSCC) susceptible Alloy 600/82/182 materials.

The NRC staff is generally concerned that the subject segment could be susceptible to fatigue at socket welds and stress corrosion cracking (SSC). In response to the NRC staffs request for additional information, the licensee states that the subject segment is 127 feet long, contains 42 socket welds and is expected to operate at an average pressure of 82 psig when the RCS pumps are stopped and the DHR pump is operating. The licensee further states that review of the corrective action program did not identify any plant-specific history of degradation of this segment due to SCC or fatigue, and review of lSI results noted no stress corrosion or fatigue-related leaks. The NRC staff notes that the subject segment between the check valve and stop valve is not normally pressurized thus would not experience flow-induced vibration which could lead to fatigue of socket welds.

The NRC staff finds that the large margin between design pressure (2500 psig) and operating pressure (82 psig), lack of plant-specific evidence of SCC or fatigue, and performance of the VT-2 examination of the subject segment at or near the end of the 10-year lSI interval at 62 psig (a pressure that is nearly equal to the expected operating pressure) will provide reasonable assurance of structural integrity of the auxiliary pressurizer spray system segments, while maintaining personnel radiation exposure as low as reasonably achievable (ALARA).

Test Zone RC01 The licensee states that the Test Zone RC01 HPI check valves are welded back-to-back with no piping or test connections between them that would permit pressurization between the valves, and that pressurization upstream of the check valves would require abnormal system lineups that would likely result in a thermal transient cycle on the respective train's HPI nozzles. The NRC staff finds that the abnormal system lineup and the possibility of producing a thermal transient cycle on the respective train's HPI nozzles, which would count against the rapid RCS depressurization design cycle limits, would constitute a hardship.

The performance of the pressure test at check valves HP50, HP51, and HP59 could also be performed with the RCS out of service. There would be a significant effort to modify the valves for the pressure test then to modify them again to perform their design function. Accompanying

-7 this modification effort would be a significant cumulative radiological exposure, approximately 5.3 man-Rem. The NRC staff finds that the radiological dose that accompanies modification, testing, and restoring of the subject valves to service conditions would constitute a hardship.

Within Test Zone RC01, the area between the HPI check valves is not normally pressurized.

The licensee proposes to perform a VT-2 examination of the HPI check valves during the Class 2 HPI system leakage test conducted at or near the end of the 1O-year lSI interval. This segment is pressurized to approximately 43 psig during this test. While the test pressure is significantly less than the RCPB pressure, a mitigating factor in accepting the Class 2 system leakage test pressure in lieu of the Code-required test pressure is based on the fact that there is no known degradation mechanism, such as intergranular stress corrosion cracking (lGSCC),

PWSCC, or thermal fatigue, that is likely to affect the subject welds. The NRC staff finds that a pressure test of the subject welds conducted at a reduced pressure will effectively detect leakage, albeit at a lower leak rate than that at RCPB pressure. The NRC staff finds that the plant-specific lack of evidence of SCC, and performance of the VT-2 examination of the subject welds at a Class 2 HPI system leakage test pressure (43 psig) will provide reasonable assurance of structural integrity of the weld in the back-to-back welded check valves while maintaining personnel radiation exposure with consideration to ALARA principles.

Based on the materials of construction, low usage service conditions, and the Code-compliant VT-2 examination of the segments performed each outage and during the Class 2 HPI system leakage test conducted at or near the end of the 1O-year lSI interval, the NRC staff finds that there is reasonable assurance of structural integrity of the subject Test Zones. The NRC staff concludes that imposition of the ASME Code requirement to extend pressure retaining boundary to the subject Class 1 components for the system leakage test at the end of the lSI interval would result in hardship without a compensating increase in the level of quality and safety.

4.0 CONCLUSION

As set forth above in Section 3.5, the NRC staff has determined that the proposed alternative provides reasonable assurance of structural integrity and leak tightness, and that complying with the specified ASME Code,Section XI 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). Therefore, the NRC staff authorizes the proposed alternative for DBI\IPS for the duration of the third 1O-year lSI interval, currently scheduled to end on September 20, 2012.

All other requirements of the ASIVIE Code for which relief has not been specifically requested and authorized remain applicable, including a third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: J. Wallace, NRR Date of issuance: April 9, 2012

B.Allen 2 Please contact the DBNPS Project Manager, Michael Mahoney at (301) 415-3867 if you have any questions on this action.

Sincerely, IRAJ Joel S. Wiebe for Jacob I. Zimmerman, Chief Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosure:

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

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