BSEP 16-0110, Request for Alternatives in Support of the Fifth 10-Year Inservice Testing (IST) Program

Request for Alternatives in Support of the Fifth 10-Year Inservice Testing (IST) Program

ML16350A064
Person / Time
Site:	Brunswick
Issue date:	12/14/2016
From:	Pope A Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 16-0110
Download: ML16350A064 (14)

Text

Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 December 14, 2016 Serial: BSEP 16-0110 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Request for Alternatives in Support of the Fifth 10-Year Inservice Testing (IST)

Program

References:

1. Letter from Annette H. Pope (Duke Energy) to the U.S. Nuclear Regulatory Commission Document Control Desk, Fifith 10-Year Inservice Testing (IST)

Program, dated October 27, 2016

2. Letter from Thomas H. Boyce (NRC) to Benjamin Waldrep (Carolina Power &

Light Company), Brunswick Steam Electric Plant, Units 1 and 2 - Relief Requests for the Brunswick Steam Electric Plant, Units 1 and 2 Fourth 10-Year Pump and Valve Inservice Testing Program (TAC Nos. MD7425, MD7426, MD7428, MD7429, MD7430, MD7431, MD7432, MD7433, MD7434, MD7435, MD7436, MD7437, MD7438, MD7440, and MD7441),

dated May 8, 2008, ADAMS Accession Number ML081130002 Ladies and Gentlemen:

On October 27, 2016 (i.e., Reference 1), Duke Energy Progress, LLC (Duke Energy), notified the NRC that the start date for the next Inservice Testing (IST) Program interval for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2, has been revised to begin November 1, 2017.

The purpose of this letter is to provide, for NRC review and approval, the 10 CFR 50.55a alternative requests needed to support implementation of the next 10-year IST Program. The proposed 10 CFR 50.55a alternatives are provided in the enclosure to this submittal.

The fifth 10-year IST Program will follow the American Society of Mechanical Engineers (ASME)

Code for Operation and Maintenance of Nuclear Power Plants (i.e., referred to herein as the OM Code), 2004 Edition with Addenda through OM-2006.

Duke Energy requests approval of the enclosed 10 CFR 50.55a alternatives by November 1, 2017, to coincide with implementation of the fifth 10-year IST Program.

U.S. Nuclear Regulatory Commission Page 2of2 Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

Sincerely, A+/-:.--t;:pe)t:(JJ~

Director - Organizational Effectiveness Brunswick Steam Electric Plant WRM/wrm

Enclosure:

10 CFR 50.55a Alternatives Supporting the lnservice Testing (IST) Program for the Fifth 10-Year Interval cc: U.S. Nuclear Regulatory Commission, Region II ATTN: Ms. Catherine Haney, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Ms. Michelle P. Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 Andrew.Hon@nrc.gov Chair - North Carolina Utilities Commission (Electronic Copy Only) 4325 Mail Service Center Raleigh, NC 27699-4300 swatson@ncuc.net Mr. Cliff Dautrich, Bureau Chief North Carolina Department of Labor Boiler Safety Bureau 1101 Mail Service Center Raleigh, NC 27699-1101

Enclosure BSEP 16-0110 10 CFR 50.55a Alternatives Supporting the Inservice Testing (IST) Program for the Fifth 10-Year Interval

Enclosure BSEP 16-0110 Page 1 of 11 DUKE ENERGY IST PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(z)(1)

VRR-01 PLANT/UNIT: Brunswick Steam Electric Plant, Unit Nos. 1 and 2.

INTERVAL: Fifth 10-year interval beginning November 1, 2017, and ending October 31, 2027.

COMPONENTS 1-B21-F013A thru 1-B21-F013L AFFECTED: 2-B21-F013A thru 2-B21-F013L CODE EDITION ASME OM Code 2004 Edition with Addenda through OM-2006.

AND ADDENDA:

REQUIREMENTS: ISTC-5113(c) states that the stroke time of all valves shall be measured to at least the nearest second.

REASON FOR Pursuant to 10 CFR 50.55a, Codes and Standards, paragraph (z)(1),

REQUEST: Duke Energy Progress, LLC, is requesting a proposed alternative from stroke time requirements of the ASME OM Code. The proposed alternative provides an acceptable level of quality and safety.

This alternative is a re-submittal of NRC approved fourth interval alternative VRR-01 that was based on the ASME OM Code-2001 Edition. This fifth 10-year interval request for relief, VRR-01, is based on the ASME OM Code-2004 Edition with Addenda through OM-2006.

There have been no substantive changes to this alternative, to the OM Code requirements, or to the basis for use, which would alter the previous NRC Safety Evaluation conclusions.

The functions of the primary steam line safety/relief valves are to:

(1) open upon receipt of an Automatic Depressurization System (ADS) signal to blow down the reactor vessel (i.e., for the ADS valves only).

(2) act as primary system safety valves actuating on high system pressure or by manual actuation from the Control Room, and (3) to close to maintain the primary system pressure boundary and prevent uncontrolled de-pressurization of the reactor (i.e., stuck open relief valve). The function of the solenoid valves is to energize upon receipt of a manual or ADS actuation signal and, in so doing, open the associated pilot valve to allow venting of the area behind the main piston resulting in the associated main valve disc opening.

The valves are sent to a vendor (i.e., NTS Technologies) and as-found tested which includes visual inspection, leakage testing, stroke time testing, and set pressure testing. The stroke time of the main disc is measured by using accelerometers. The acceptance criteria is set at

<100 milliseconds. This verifies the valves will perform their desired function. The valves are full stroke exercised and remote position verified, in accordance with ASME OM Code and Technical Specification 3.4.3, Safety/Relief Valves (SRVs). Temperature sensors and acoustic monitors downstream of the valves discharge nozzles

Enclosure BSEP 16-0110 Page 2 of 11 are used to provide a positive valve position indication.

The proposed alternative testing above, together with the extensive preventative maintenance requirements for these valves, gives adequate assurance that these valves will perform satisfactorily and reliably. This position and alternate testing conforms to the recommendations presented in NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants, paragraph 4.3.2.1.

PROPOSED Each of these valves will be exercised open and closed, and proper ALTERNATIVE operation will be ascertained by observing the response and changes AND BASIS: in main steam parameters within a specified time period and observation of the outputs of the downstream temperature and acoustic sensors. Specific as-found stroke times, visual inspections, set pressure and leakage testing will be measured by the vendor.

DURATION: The proposed alternative will be used for the entire fifth 10-year interval which begins November 1, 2017, and ends October 31, 2027.

PRECEDENTS: This request was approved as VRR-01 for the fourth 10-year IST interval (i.e., ADAMS Accession Number ML081130002).

Enclosure BSEP 16-0110 Page 3 of 11 DUKE ENERGY IST PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(z)(1)

VRR-02 PLANT/UNIT: Brunswick Steam Electric Plant, Unit Nos. 1 and 2.

INTERVAL: Fifth 10-year interval beginning November 1, 2017, and ending October 31, 2027.

COMPONENTS 1-B21-F010A and 1-B21-F010B AFFECTED: 2-B21-F010A and 2-B21-F010B CODE EDITION ASME OM Code 2004 Edition with Addenda through OM-2006.

AND ADDENDA:

REQUIREMENTS: ISTC-3510 requires all active Category A, Category B, and Category C check valves be exercised nominally every 3 months.

ISTC-5224 requires a check valve to be declared inoperable if it fails to exhibit the required change of obturator position. In addition, a retest showing acceptable performance shall be run following any required corrective action before the valve is returned to service.

REASON FOR Pursuant to 10 CFR 50.55a, Codes and Standards, paragraph (z)(1),

REQUEST: Duke Energy Progress, LLC, is requesting a proposed alternative from the check valve exercise and obturator requirements of the ASME OM Code. The proposed alternative provides an acceptable level of quality and safety.

This alternative is a re-submittal of NRC approved fourth interval alternative VRR-02 that was based on the ASME OM Code-2001 Edition. This fifth Interval proposed alternative VRR-02 is based on the ASME OM Code-2004 Edition with Addenda through OM-2006. There have been no substantive changes to this alternative, to the OM Code requirements, or to the basis for use, which would alter the previous NRC Safety Evaluation conclusions.

These feedwater line-to-reactor pressure vessel check valves open to provide flow paths for normal feedwater flow as well as High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) flow into the reactor vessel. These are simple check valves, with no external means of exercising or external determination of disk position.

Thus, the only practical method of exercising these valves to their open position and confirming full open operation per the guidance of NUREG-1482, Revision 2, is with flow from the reactor feedwater system, or from the HPCI or RCIC systems themselves. The HPCI Technical Specification flow requirement is 4250 gpm, and RCIC Technical Specification flow requirement is 400 gpm. Injecting water directly from either the HPCI or RCIC systems to the reactor is impractical during plant operation due to the possibility of creating an unacceptable reactor vessel water level transient, thermal shock to reactor vessel nozzles, a reactivity excursion, or upsetting reactor

Enclosure BSEP 16-0110 Page 4 of 11 water chemistry. Under normal shutdown conditions, steam is unavailable to operate the HPCI and RCIC turbines and there is a potential for over-pressurizing the reactor vessel. Thus, the only practical way of exercising these valves is with reactor feedwater flow during power operation.

During normal plant operation, the feedwater flow is approximately 15,000 gpm per loop, which is greater than the maximum Technical Specification flow of either HPCI or RCIC through these check valves.

The Reactor Feedwater system arrangement is such that flow indication can be obtained for each of the individual feedwater loops.

Thus, flow measurement through each check valve can be made to verify proper opening of the subject check valve. This method complies with NUREG-1482, Revision 2, paragraph 4.1.3, which allows verification of the check valve full stroke from closed to open position by measuring feedwater flow to the reactor vessel during normal plant operation. ISTC-3550 of ASME OM Code, 2004 Edition with 2006 Addenda, titled Valves in Regular Use, also provides allowances that satisfy the exercising provided that observations otherwise required for testing are made and analyzed during operation and recorded in the plant record at no greater than specified within the OM Code. Monitoring of feedwater flow monthly and as required during post-refueling start-up test program, in accordance with plant procedures, complies with the above guidance.

PROPOSED Exercising of these valves open will only be performed to the extent ALTERNATIVE that adequate reactor feedwater flow is available. Full accident flow AND BASIS: through each feedwater injection leg will be confirmed by monitoring A-loop and B-loop flow through feedwater flow venturis 1/2-C32-FE-N001A/B during power operation.

Feedwater flow is a critical input to the reactor heat balance and is monitored by Operations continuously via the plant process computer.

Where maintenance or corrective action has been performed on a valve during a shutdown period, the subject valve will not be flow tested (i.e., opened) prior to being placed in service.

DURATION: The proposed alternative will be used for the entire fifth 10-year interval which begins November 1, 2017, and ends October 31, 2027.

PRECEDENTS: This proposed alternative was approved as VRR-02 for the fourth 10-year IST interval (i.e., ADAMS Accession Number ML081130002).

Enclosure BSEP 16-0110 Page 5 of 11 DUKE ENERGY IST PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(z)(1)

VRR-03 PLANT/UNIT: Brunswick Steam Electric Plant, Unit Nos. 1 and 2.

INTERVAL: Fifth 10-year interval beginning November 1, 2017, and ending October 31, 2027.

COMPONENTS 1/2-B21-F008 1/2-B21-F014A 1/2-B21-F014B 1/2-B21-F014C AFFECTED: 1/2-B21-F014D 1/2-B21-F014E 1/2-B21-F014F 1/2-B21-F014G 1/2-B21-F014H 1/2-B21-F014J 1/2-B21-F014K 1/2-B21-F014L 1/2-B21-F014M 1/2-B21-F014N 1/2-B21-F014P 1/2-B21-F014R 1/2-B21-F014S 1/2-B21-F040 1/2-B21-F042A 1/2-B21-F042B 1/2-B21-F044A 1/2-B21-F044B 1/2-B21-F046A 1/2-B21-F046B 1/2-B21-F047C 1/2-B21-F047D 1/2-B21-F048A 1/2-B21-F048B 1/2-B21-F049C 1/2-B21-F049D 1/2-B21-F050A 1/2-B21-F050B 1/2-B21-F050C 1/2-B21-F050D 1/2-B21-F052A 1/2-B21-F052B 1/2-B21-F052C 1/2-B21-F052D 1/2-B21-F054 1/2-B21-F056 1/2-B21-F058A 1/2-B21-F058B 1/2-B21-F058C 1/2-B21-F058D 1/2-B21-F058E 1/2-B21-F058F 1/2-B21-F058G 1/2-B21-F058H 1/2-B21-F058L 1/2-B21-F058M 1/2-B21-F058N 1/2-B21-F058P 1/2-B21-F058R 1/2-B21-F058S 1/2-B21-F058T 1/2-B21-F058U 1/2-B21-F060 1/2-B21-IV-2149 1/2-B21-IV-2196 1/2-B21-IV-2455 1/2-B21-IV-2456 1/2-B32-F005A 1/2-B32-F005B 1/2-B32-F006A 1/2-B32-F006B 1/2-B32-F039A 1/2-B32-F039B 1/2-B32-F039C 1/2-B32-F039D 1/2-B32-F041A 1/2-B32-F041B 1/2-B32-F041C 1/2-B32-F041D 1/2-B32-F042A 1/2-B32-F042B 1/2-B32-F042C 1/2-B32-F042D 1/2-B32-F058A 1/2-B32-F058B 1/2-E21-F017A 1/2-E21-F017B 1/2-E41-F023A 1/2-E41-F023B 1/2-E41-F023C 1/2-E41-F023D 1/2-E51-F043A 1/2-E51-F043B 1/2-E51-F043C 1/2-E51-F043D CODE EDITION ASME OM Code 2004 Edition with Addenda through OM-2006.

AND ADDENDA:

REQUIREMENTS: ISTC-3510 requires all active Category A, Category B, and Category C check valves be exercised nominally every 3 months.

ISTC-3700 requires valves with remote position indicators to be observed locally at least once every 2 years to verify that valve operation is accurately indicated.

REASON FOR Pursuant to 10 CFR 50.55a, Codes and Standards, paragraph (z)(1),

REQUEST: Duke Energy Progress, LLC, is requesting a proposed alternative from the check valve exercise and remote position observation requirements of the ASME OM Code. The proposed alternative provides an acceptable level of quality and safety.

This alternative is a re-submittal of NRC approved fourth Interval relief request VRR-04 that was based on the ASME OM Code-2001 Edition.

This fifth Interval request for relief, VRR-03, is based on the ASME OM

Enclosure BSEP 16-0110 Page 6 of 11 Code-2004 Edition with Addenda through OM-2006. There have been no substantive changes to this alternative, to the OM Code requirements, or to the basis for use, which would alter the previous NRC Safety Evaluation conclusions.

Because of the design of excess flow check valves, verifying their closure indication requires a simulated instrument line break. Based on the burden and costs associated with testing these excess flow check valves, Duke Energy is proposing to perform the exercise tests and valve position verification tests on a sampling basis (i.e.,

approximately an equal number of excess flow check valves every 24 months such that each excess flow check valve is tested at least once every 10 years).

Duke Energy has determined that alternative excess flow check valve testing will provide an acceptable level of quality and safety for the following reasons:

1. Excess flow check valves are a simple and reliable device. The major components are a poppet and spring. The spring holds the poppet open only under static conditions, such that the valve will close upon sufficient differential pressure across the poppet.

Functional testing of the valve is accomplished by venting the instrument side of the tube. The resultant increase in flow imposes a differential pressure across the poppet, which compresses the spring and decreases flow through the valve.

2. The Boiling Water Reactor Owners Group (BWROG) has developed a basis, documented in Topical Report B21-00658-01, Excess Flow Check Valve Testing Frequency Relaxation, dated November 1998, for reducing the excess flow check valve testing frequency. This report was initially submitted to the NRC as part of a Duane Arnold Energy Center proposed license amendment on April 12, 1999. The BWROG report was supplemented by BWROG letter dated January 6, 2000, Generic Response to NRC Request For Additional Information on Lead Plant Technical Specification Change Request Regarding Excess Flow Check Valve Surveillance Requirements. The report was approved for use by an NRC Safety Evaluation dated March 14, 2000. Additionally, issues raised by the NRC in the March 14, 2000, Safety Evaluation were addressed in the issuance of General Electric Topical Report NEDO-32977-A (i.e., BWROG Topical Report B21-00658-01), Excess Flow Check Valve Testing Relaxation, dated June 2000.

The BWROG topical report concluded that the change in excess flow check valve test frequency has an insignificant impact on excess flow check valve reliability. The topical report evaluated the reliability of

Enclosure BSEP 16-0110 Page 7 of 11 excess flow check valves at various boiling water reactor plants, including BSEP, based on information covering a 10-year period.

Industry experience with excess flow check valves indicate that they have very low failure rates. A large portion of the reported test failures at other plants was related to test methodologies and not actual valve failures.

On October 4, 2001, the NRC issued License Amendments 215 and 242 for BSEP Units 1 and 2, respectively, revising the BSEP Technical Specifications to incorporate excess flow check valve testing requirements consistent with TSTF-334.

Excess flow check valves have been extremely reliable throughout the industry.

An orifice is installed on each of the affected instrument lines. The orifice limits leakage to a quantity where the integrity and functional performance of secondary containment and the associated safety systems are maintained. The process fluid loss for a postulated rupture of an instrument line is within the capability of the reactor coolant makeup systems.

The reduced testing associated with the alternative will result in an increase in the availability of the associated instrumentation during plant refueling outages. The reduced testing associated with the alternative will also reduce occupational radiological exposure.

PROPOSED Duke Energy proposes to test a representative sample of excess flow ALTERNATIVE check valves consisting of an approximately equal number of excess AND BASIS: flow check valves every 24 months, such that each excess flow check valve will be tested at least once every 10 years. In addition, Duke Energy proposes to verify the open position indication at a frequency more often than what the ASME Code requires, but verify the close position indication in conjunction with excess flow check valve exercise tests.

DURATION: The proposed alternative will be used for the entire fifth 10-year interval which begins November 1, 2017, and ends October 31, 2027.

PRECEDENTS: 1. This proposed alternative was approved as VRR-04 for the fourth 10-year IST interval (i.e., ADAMS Accession Number ML081130002).

2. NRC Regulatory Guide 1.11, Instrument Lines Penetrating the Primary Reactor Containment.

Enclosure BSEP 16-0110 Page 8 of 11 DUKE ENERGY IST PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(z)(1)

VRR-04 PLANT/UNIT: Brunswick Steam Electric Plant, Unit Nos. 1 and 2.

INTERVAL: Fifth 10-year interval beginning November 1, 2017, and ending October 31, 2027.

COMPONENTS Pumps and Valves contained within the Inservice Testing Program AFFECTED: scope.

CODE EDITION ASME OM Code 2004 Edition with Addenda through OM-2006.

AND ADDENDA:

REQUIREMENTS: This request applies to the following examination frequency requirements of the ASME OM Code.

ISTA-3120(a) "The frequency for the inservice testing shall be in accordance with the requirements of Section IST."

ISTB-3400 Frequency of lnservice Tests ISTC-3510 Exercising Test Frequency ISTC-3540 Manual Valves ISTC-3630(a) Frequency ISTC-3700 Position Verification Testing ISTC-5221 (c)(3) "At least one valve from each group shall be disassembled and examined at each refueling outage; all valves in a group shall be disassembled and examined at least once every 8 years."

Appendix I, I -1320 Test Frequencies, Class I Pressure Relief Valves Appendix I, 1-1330 Test Frequencies, Class I Nonreclosing Pressure Relief Devices Appendix I, 1- I 340 Test Frequencies- Class I Pressure Relief Devices That Are Used for Thermal Relief Application Appendix I, 1-1350 Test Frequencies- Class 2 and 3 Pressure Relief Valves Appendix I, 1-1360 Test Frequencies- Class 2 and 3 Nonreclosing Pressure Relief Devices Appendix I, 1-1370 Test Frequencies- Class 2 and 3 Primary Containment Vacuum Relief Valves Appendix I, 1- I 380 Test Frequencies- Class 2 and 3 Vacuum Relief Valves Except for Primary Containment Vacuum Relief Valves Appendix I, I -1390 Test Frequencies- Class I Pressure

Enclosure BSEP 16-0110 Page 9 of 11 Relief Devices That Are Used for Thermal Relief Application Appendix II. 11-4000(a)( I) Performance Improvement Activities Interval Appendix II, 11-4000(b)( I )(e) Optimization of Condition Monitoring Activities Interval REASON FOR Pursuant to 10 CFR 50.55a, Codes and Standards, paragraph (z)(1),

REQUEST: Duke Energy Progress, LLC, is requesting a proposed alternative from the examination frequency requirements of the ASME OM Code. The proposed alternative provides an acceptable level of quality and safety.

ASME OM Code, Section IST, establishes the inservice test frequency for all components within the scope of the Code. The frequencies (e.g., quarterly) have always been interpreted as "nominal" frequencies (i.e., generally as defined in the Table 3.2 of NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants) and Owners routinely applied the surveillance extension time period (i.e., grace period) contained in the plant Technical Specifications (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 the surveillance (i.e., SR 3.0.2). However, regulatory issues have been raised concerning the applicability of the TS "Grace Period" to ASME OM Code required inservice test frequencies irrespective of allowances provided under TS Administrative Controls (i.e., TS 5.5.6, lnservice Testing Program, invokes SR 3.0.2 for various OM Code frequencies).

The lack of a tolerance band on the ASME OM Code inservice test frequency restricts operational flexibility. There may be a conflict where a surveillance test could be required but where it is not possible or not desired that it be performed until sometime after a certain restricted plant condition is cleared. Therefore, to avoid this conflict, the surveillance test should be performed as soon as it is practicable.

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 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 extending OM Code testing intervals. Interval 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). Such extensions are not intended to be used repeatedly merely as an operational

Enclosure BSEP 16-0110 Page 10 of 11 convenience to extend test intervals beyond those specified.

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

a. Components whose test frequencies are based on elapsed time periods shall undergo lnservice Testing at frequencies as specified in the Brunswick Steam Electric Plant Technical Specifications (i.e., TS 5.5.6) and shown in the following table:

Frequency Specified Time Period Between Tests Weekly At least once per 7 days Monthly At least once per 31 days Quarterly At least once per 92 days Semiannually At least once per 184 days Yearly/Annually At least once per 366 days

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

i. For periods specified as less than 2 years, the period may be extended by up to 25 percent for any given test.

ii. 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 the ASME OM Code.

d. 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.

e. Period extensions of 25 percent may also be applied to accelerated test frequencies (i.e., pumps in Alert Range) and

Enclosure BSEP 16-0110 Page 11 of 11 other less than two year test frequencies not specified in the table above.

This proposed alternative is requested citing the guidance found in ASME approved Code Case OMN-20. The proposed alternative provides an acceptable level of quality and safety, and thus should be granted pursuant to 10 CFR 50.55a(z)(1).

DURATION: The proposed alternative will be used for the entire fifth 10-year interval which begins November 1, 2017, and ends October 31, 2027.

PRECEDENTS: 1. Quad Cities Relief Request RV-01, Safety Evaluation dated February 14, 2013, ADAMS Accession Number ML13042A348.

2. Callaway Relief Request PR-04, Safety Evaluation dated July 15, 2014, ADAMS Accession Number ML14178A769.

3. Calvert Cliffs Relief Request IST-RR-01, Safety Evaluation dated September 24, 2014, ADAMS Accession Number ML14247A555.

4. Three Mile Island Station, Unit 1 Relief Request VR-02, Safety Evaluation dated August 15, 2013, ADAMS Accession Number ML13227A024.

5. Dresden Relief Request RV-01, Safety Evaluation dated October 31, 2013, ADAMS Accession Number ML13297A515.

6. Plant Hatch Relief Request RR-V-08, Safety Evaluation dated December 30, 2015, ADAMS Accession Number ML15310A406.

REFERENCES:

1. NRC Regulatory Issue Summary 2012-10, NRC Staff Position on Applying Surveillance Requirements 3.0.2 and 3.0.3 to Administrative Controls Program Tests

2. ASME OM Code Case OMN-20, Inservice Test Frequency