ML15030A501

Issuance of Amendment No. 230, Revise Technical Specifications for One-Time Extension of Completion Time Required Actions Related to Residual Heat Removal System B Inoperability (Emergency Circumstances)

ML15030A501
Person / Time
Site:	Columbia
Issue date:	02/01/2015
From:	Andrea George Plant Licensing Branch IV
To:	Reddemann M Energy Northwest
George A
References
TAC MF5635
Download: ML15030A501 (22)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 1, 2015 Mr. Mark E. Reddemann Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)

Richland, WA 99352-0968

SUBJECT:

COLUMBIA GENERATING STATION -ISSUANCE OF AMENDMENT RE:

REVISION TO TECHNICAL SPECIFICATIONS FOR ONE-TIME EXTENSION OF COMPLETION TIMES RELATED TO RESIDUAL HEAT REMOVAL SYSTEM B INOPERABILITY (EMERGENCY CIRCUMSTANCES) (TAC NO. MF5635)

Dear Mr. Reddemann:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 230 to Renewed Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the Technical Specifications (TS) in response to your application dated January 30, 2015.

The amendment makes a one-time revision toTS 3.5.1, "ECCS [Emergency Core Cooling System]- Operating," TS 3.6.1.5, "Residual Heat Removal (RHR) Drywell Spray," and TS 3.6.2.3, "Residual Heat Removal (RHR) Suppression Pool Cooling," to extend the Completion Time (CT) of Required Actions specifically associated with RHR System B inoperability from 7 days to 14 days. This extension will allow completion of a system modification, required testing, and system restoration. This amendment was necessitated by emergent issues that have delayed completion of activities to modify the 24-inch Division 2 (Loop B) RHR suction piping.

The license amendment is issued under emergency circumstances as provided in the provisions of paragraph 50.91 (a)(5) of Title 10 of the Code of Federal Regulations due to the time critical nature of the amendment. In this instance, an emergency situation exists in that the proposed amendment is needed to allow the licensee to preclude a plant shutdown.

M. Reddemann A copy of the related Safety Evaluation is also enclosed. The safety evaluation describes the emergency circumstances under which the amendment was issued and the final no significant hazards determination. A Notice of Issuance addressing the final no significant hazards determination and opportunity for a hearing associated with the emergency circumstances, will be included in the Commission's next biweekly Federal Register notice.

Sincerely, Andrea E. George, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397

Enclosures:

1. Amendment No. 230 to NPF-21

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 230 License No. NPF-21

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Energy Northwest (licensee), dated January 30, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-21 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 230 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The license amendment is effective as of its date of issuance and shall be implemented immediately upon issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION Eric R. Oesterle, Acting Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. NPF-21 and Technical Specifications Date of Issuance: February 1, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 230 RENEWED FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of the Renewed Facility Operating License No. NPF-21 and Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Facility Operating License REMOVE INSERT Technical Specification REMOVE INSERT 3.5.1-1 3.5.1-1 3.6.1.5-1 3.6.1.5-1 3.6.1.5-2 3.6.2.3-1 3.6.2.3-1 3.6.2.3-2 3.6.2.3-2

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 230 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

a. For Surveillance Requirements (SRs) not previously performed by existing SRs or other plant tests, the requirement will be considered met on the implementation date and the next required test will be at the interval specified in the Technical Specifications as revised in Amendment No. 149.

(3) Deleted.

(4) Deleted.

(5) Deleted.

(6) Deleted.

(7) Deleted.

(8) Deleted.

(9) Deleted.

(10) Deleted.

(11) Shield Wall Deferral (Section 12.3.2, SSER #4, License Amendment #7)

The licensee shall complete construction of the deferred shield walls and window as identified in Attachment 3, as amended by this license amendment.

(12) Deleted.

(13) Deleted.

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed License No. NPF-21 Amendment No. 230

ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure s 150 psig.

ACTIONS

---

N0 T E-----------------------------------------------------------

LCO 3.0.4.b is not applicable to HPCS.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days' 1) injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.

B High Pressure Core B.1 Verify by administrative Immediately Spray (HPCS) System means RCIC System is inoperable. OPERABLE when RCIC System is required to be OPERABLE.

AND B.2 Restore HPCS System to 14 days OPERABLE status.

1

' ) The Completion Time that one train of RHR (RHR-8) can be inoperable as specified by Required Action A.1 may be extended beyond the 7 day completion time up to 7 days to support restoration of RHR-8 from the modification activity. Upon successful restoration of RHR-8, this footnote is no longer applicable and will expire at 05:00 PST on February 9, 2015.

Columbia Generating Station 3.5.1-1 Amendment No. 37 ~ 230

RHR Drywell Spray 3.6.1.5 3.6 CONTAINMENT SYSTEMS 3.6.1.5 Residual Heat Removal (RHR) Drywell Spray LCO 3.6.1.5 Two RHR drywell spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR drywell spray A.1 Restore RHR drywell spray 7 days<1 )

subsystem inoperable. subsystem to OPERABLE status.

B. Two RHR drywell spray B.1 Restore one RHR drywell 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> subsystems inoperable. spray subsystem to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 1

( ) The Completion Time that one train of RHR (RHR-B) can be inoperable as specified by Required Action A.1 may be extended beyond the 7 day completion time up to 7 days to support restoration of RHR-B from the modification activity. Upon successful restoration of RHR-B, this footnote is no longer applicable and will expire at 05:00 PST on February 9, 2015.

Columbia Generating Station 3.6.1.5-1 Amendment No. 4-W ~ 230

RHR Drywell Spray 3.6.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify each RHR drywell spray subsystem manual, 31 days power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

SR 3.6.1.5.2 Verify each spray nozzle is unobstructed. 10 years Columbia Generating Station 3.6.1.5-2 Amendment No. 230

RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling LCO 3.6.2.3 Two RHR suppression pool cooling subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.1 Restore RHR suppression 7 days( 1 l pool cooling subsystem pool cooling subsystem to inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Two RHR suppression pool cooling subsystems inoperable.

1

( l The Completion Time that one train of RHR (RHR-B) can be inoperable as specified by Required Action A.1 may be extended beyond the 7 day completion time up to 7 days to support restoration of RHR-B from the modification activity. Upon successful restoration of RHR-B, this footnote is no longer applicable and will expire at 05:00 PST on February 9, 2015.

Columbia Generating Station 3.6.2.3-1 Amendment No. 4-W ~ 230

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling 31 days subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.

SR 3.6.2.3.2 Verify each RHR pump develops a flow rate In accordance

~ 7100 gpm through the associated heat exchanger with the lnservice while operating in the suppression pool cooling Testing Program mode.

Columbia Generating Station 3.6.2.3-2 Amendment No.~~ 230

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 230 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397

1.0 INTRODUCTION

By application dated January 30, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15031A002), Energy Northwest (the licensee) requested changes to the Technical Specifications (TSs), Appendix A to Renewed Facility Operating License No. NPF-21, for the Columbia Generating Station (CGS).

The amendment would modify TS 3.5.1, "ECCS [Emergency Core Cooling System]-

Operating," TS 3.6.1.5, "Residual Heat Removal (RHR) Drywell Spray," and TS 3.6.2.3, "Residual Heat Removal (RHR) Suppression Pool Cooling," to make a one-time extension of the Completion Time (CT) of Required Actions specifically associated with RHR System B inoperability. This amendment was necessitated by emergent issues that have delayed completion of activities to modify the 24-inch Division 2 (Loop B) RHR suction piping. The proposed amendment would revise the TS CT for the Required Actions associated with RHR-B system inoperability from 7 days to 14 days to allow for the completion of installation of Phase 2 of the Fuel Pool Cooling Assist Modification.

The licensee requested that the Nuclear Regulatory Commission (NRC) staff process this submittal as an emergency amendment.

2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act (Act) requires applicants for nuclear power plant operating licenses to include TSs as part of the license. These TSs are derived from the plant safety analyses.

In Section 50.36, "Technical specifications," of Title 10 of the Code of Federal Regulations (1 0 CFR), the NRC established its regulatory requirements related to the content of TSs.

Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and Enclosure 2

limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant's TSs.

The regulations in 10 CFR 50.36(c)(2) state that LCOs are the lowest functional capability or performance level of equipment required for safe operation of the facility and when LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the LCO can be met.

Licensees may propose revisions to the TSs. The NRC staff reviews proposed changes and will generally issue changes provided that the plant-specific review supports a finding of continued adequate protection of public health and safety because: (1) the change is editorial, administrative, or provides clarification (i.e. no requirements are materially altered), (2) the change is more restrictive than the licensee's current requirement, or (3) the change is less restrictive than the licensee's current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards. The detailed application of this general framework, and additional specialized guidance, is discussed in Section 3.0 of this safety evaluation in the context of the proposed TS changes contained in the licensee's LAR.

The regulations in 10 CFR 50.46(b) establish acceptance criteria for ECCS evaluations for light-water nuclear power reactors, as summarized below:

• Peak cladding temperature- the calculated maximum fuel element cladding temperature shall not exceed 2,200 degrees Fahrenheit (°F).

• Maximum cladding oxidation - the calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

• Maximum hydrogen generation -the calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

• Coolable geometry - calculated changes in core geometry shall be such that the core remains amenable to cooling.

• Long-term cooling - after any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

The 10 CFR 50, Appendix A, General Design Criteria (GDC) applicable to this license amendment request are discussed below:

• GDC 35, "Emergency core cooling," applies insofar as it requires:

A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

• GDC 38, "Containment heat removal," insofar as it requires, in part, that:

A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of coolant accident and maintain them at acceptably low levels.

The regulatory guidance that the NRC staff used in its review of the risk information submitted in support of the LAR consisted of the following:

• Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment

[PRA] in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

Revision 2, May 2011 (ADAMS Accession No. ML100910006), describes an acceptable method for licensees and the NRC to use for assessing the nature and impact of proposed changes to the licensing basis by considering engineering issues and applying risk insights. This regulatory guide also provides risk-acceptance guidelines for evaluating the results of such evaluations.

• RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 1, May 2011 (ADAMS Accession No. ML100910008),

describes methods acceptable to the NRC for assessing the nature and impact of proposed permanent TS changes, including allowed outage times, by considering engineering issues and applying risk insights. This regulatory guide also provides risk-acceptance guidelines for evaluating the results of such assessments.

3.0 TECHNICAL EVALUATION

3.1 Description The purpose of the emergency core cooling system (ECCS) at CGS is to mitigate design basis loss-of-coolant accidents (LOCAs) and satisfy the requirements of 10 CFR 50.46 for ECCS performance criteria. The ECCS system at CGS is comprised of the high-pressure core spray (HPCS) system, the low-pressure core spray (LPCS) system, the automatic depressurization system (ADS), and the low pressure coolant injection (LPCI) mode of the RHR system. The

RHR system also supports LOCA containment heat removal functions, in that, following a LOCA, the RHR system provides containment spray and/or suppression pool cooling functions.

In Section 4.1 of its LAR, the licensee stated that the proposed TS change does not alter the redundant design capability of the CGS RHR system to accomplish its specified ECCS functions as required by the 10 CFR 50, Appendix A, GDC.

Additionally, when the reactor is shut down, the shutdown cooling mode of RHR is used to remove residual heat and decay heat from the vessel. During normal at-power operations (when reactor pressure is above RHR system design pressure), the low-pressure portions of the RHR system are isolated from full reactor pressure.

On January 26, 2015, at 05:00a.m. Pacific Standard Time (PST), the licensee commenced work on a modification to its RHR System 8 piping. In Section 2.2 of its license amendment request (LAR), the licensee stated, in part, that:

This modification would allow a means to provide alternate fuel pool and core cooling during refueling operations and facilitate future work on a degraded valve in the RHR shutdown cooling suction line by establishing a cross connection between the RHR and Fuel Pool Cooling (FPC) System piping.

The implementation plan for this modification was designed in 2 phases, as depicted in Figure 1, to minimize the time during which RHR Loop 8 would be required to be inoperable. Phase 1 of the modification was previously completed and the existing RHR-8 outage is associated with Phase 2. Phase 2 specifically involves cutting the 24" RHR Loop 8 suction side piping and welding a new pipe section connecting to the 10" FPC piping installed in Phase 1. The time required for completion of the work required in Phase 2 was estimated to be approximately 4 days.

3.2 Proposed TS Changes The current CT for Condition A, Required Action A.1 of the following TS LCOs 3.5.1, 3.6.1.5, and 3.6.2.3 states "7 days." The licensee proposed a one-time amendment to add a footnote to the current CT, which would state:

(1) The Completion Time that one train of RHR (RHR-8) can be inoperable as specified by Required Action A.1 may be extended beyond the 7 day completion time up to 7 days to support restoration of RHR-8 from the modification activity. Upon successful restoration of RHR-8, this footnote is no longer applicable and will expire at 05:00 PST on February 9, 2015.

3.3 NRC Staff Evaluation 3.3.1 Shutdown and Transient Risk with One RHR System Inoperable The NRC staff reviewed the proposed change to the TSs, and determined that the proposed footnote is less restrictive than the licensee's current TS requirements, because the Required Action A.1 CTs forTS 3.5.1, 3.6.1.5, and 3.6.2.3 would be extended from 7 days to 14 days.

In Section 2.2 of its LAR, the licensee stated, in part, that:

Since this work impacts the shutdown cooling system, a system relied upon for decay heat removal when the plant is shutdown, the decision was made to perform the work while on-line, thus minimizing station risk.

When the station is online, more ECCS systems and/or subsystems are available to mitigate a design-basis accident (e.g., HPCS, LPCS, LPCI mode of RHR, and ADS), than when the station is in a shutdown condition (only RHR available). In the current situation, with RHR system B inoperable, if the station reached its current CT of 7 days and was required by TS to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (shutdown), the situation would exist where only one RHR system would be operable for mitigation of a LOCA until the modification is complete.

Based on the currently degraded condition of the RHR B system at CGS, the NRC staff concludes that the station reduces its risk exposure by staying at power while completing the modification and restoring RHR system B to operable status. Additionally, the NRC staff notes that, should the station have been required to shut down, the inherent risk in transitioning between operational modes (operating to shutdown) is avoided by staying in the current operational mode (Mode 1, at power) to complete the modification activities.

3.3.2 Probabilistic Risk Assessment In Section 3 of its LAR, the licensee stated that the basis for the proposed TS CT changes is "a deterministic/qualitative analysis and will rely on defense-in-depth measures and configuration management measures." Therefore, it is not a risk-informed LAR. However, in Section 3.1 of its LAR, the licensee provided a detailed risk evaluation. As part of this risk evaluation, the licensee calculated the change in core damage frequency (~CDF), the change in large early release frequency (~LERF), incremental conditional core damage probability (ICCDP), and incremental conditional large early release probability (ICLERP) for the plant-specific configuration that includes the proposed extended CT for Low Pressure ECCS injection systems associated with RHR System B inoperability. Based on the licensee's risk evaluation, the

~CDF, ~LERF, ICCDP, and ICLERP are 1.04E-7, 4.32E-9, 1.91 E-7, and 7.94E-9, respectively.

All of these risk values are below the risk acceptance guidelines in RGs 1.174 and 1.177.

Based on the information provided by the licensee and the evaluation above, the NRC staff concludes that the licensee's risk evaluation for the plant configuration which includes the proposed CT extension for RHR system B is reasonable, and that the results of the selected risk metrics are consistent with the acceptance ~uidelines in RG 1.174 and RG 1.177.

3.3.3 Compensatory Measures In Section 3.2 of its LAR, the licensee proposed a number of compensatory measures that will be in place during the duration of the extended TS LCO CT. The licensee stated that it will implement the compensatory measures in Section 3.2 of its LAR in order to provide additional margin of safety to mitigate the unavailability of RHR System B. The licensee also stated that the compensatory actions consist of protecting systems that support the plant's critical safety functions and protecting the non-safety related systems with the potential to cause a plant transient.

The NRC staff has reviewed the licensee's proposed compensatory measures and concludes, based on information provided by the licensee, that they are appropriate to reduce the risk of unnecessary plant transients, protect systems needed for accident mitigation, and raise operator awareness of necessary recovery actions with one RHR system inoperable.

In Attachment 4 of its LAR, the licensee provided a regulatory commitment regarding the compensatory measures listed in Section 3.2 of its LAR. For more information regarding the NRC staff's review of this regulatory commitment, please see Section 6.0 of this safety evaluation.

3.3.4 Conclusion Based on the information provided by the licensee and the analysis in Section 3.0 of this safety evaluation, the NRC staff concludes that while the licensee's proposed TS changes are less restrictive than the licensee's current TS requirements, the proposed changes still provide adequate assurance of safety when judged against current regulatory standards. The licensee's proposed CTs extension do not have any impact on the licensee's compliance with the 10 CFR, Appendix A General Design Criteria (as described in the licensee's Final Safety Analysis Report), 10 CFR 50.46 ECCS performance criteria, or the 10 CFR 50.36 TS requirements. The NRC staff also concludes, based on information provided by the licensee, that the changes in risk associated with extending the associated CTs are less than that of the guidance thresholds in RG 1.177 and RG 1.174, and, therefore, support the extension of the CTs associated with the inoperability of RHR system B from 7 to 14 days. Therefore, the NRC staff concludes that the licensee's proposed one-time extension of the CTs associated with the inoperability of RHR system B from 7 to 14 days is acceptable.

4.0 EMERGENCY CIRCUMSTANCES The NRC's regulations in 10 CFR 50.91 (a)(5) state that where the NRC finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the NRC will publish a notice of issuance under 10 CFR 2.1 06, providing for opportunity for a hearing and for public comment after issuance.

In its LAR dated January 30, 2015, the licensee requested that the amendment be treated as an emergency amendment. In Section 2.2 of its LAR, the licensee provided, in part, the following justification of emergency circumstances:

Residual heat removal pump RHR-P-28 was declared inoperable to support completion of the final phase (Phase 2) of a modification to the suction side of the pump at 0500 on January 26, 2015. Unanticipated delays occurred with the installation of the RHR modification. This was due primarily to issues in fit-up and installation of a sweep-o-let connection to the 24" suction line of the RHR B pump. On the morning of January 29, 2015 the results of a radiographic test (RT) on welds to install the sweep-o-let revealed a number of minor flaws that

needed to be repaired. Work is ongoing per a revised completion schedule which added time to perform the weld repairs. However, there is little margin to the existing 7 day completion. The proposed one time extension would allow sufficient time to complete repairs should additional problems be presented.

The implementation plan for this modification was designed in 2 phases, as depicted in Figure 1, to minimize the time during which RHR Loop 8 would be required to be inoperable. Phase 1 of the modification was previously completed and the existing RHR-8 outage is associated with Phase 2. Phase 2 specifically involves cutting the 24" RHR Loop 8 suction side piping and welding a new pipe section connecting to the 10" FPC piping installed in Phase 1. The time required for completion of the work required in Phase 2 was estimated to be approximately 4 days.

The pre-planned implementation schedule was designed to maximize margin towards the 7 day allowed CT and to afford time to address unforeseen complications. The project implementation plan included many provisions to ensure timely execution of the work including the use of experienced personnel, pre-assembled components, pre-staging of equipment and training mock-ups.

Therefore, efforts were made to minimize the likelihood for delays due to job planning or preparation. Contingencies were developed and carried out for the existing problems. Any further challenge could result in the station reaching the 7 day completion time, thus requiring a shutdown with only one division of shutdown cooling available.

The NRC staff reviewed the licensee's basis for processing the proposed amendment as an emergency and agrees that an emergency situation exists. The licensee could not have foreseen multiple delays due to the complexity of the modification, given the considerable pre-planning, training, weld practice, and pre-staging of equipment and personnel. The NRC staff concludes that the licensee's actions were reasonable and that the identification of the problem, the actions of the licensee to inform the NRC staff, the implementation of a revised modification schedule, and the testing of the modification are being addressed in a reasonable amount of time and that the emergency situation could not have been avoided. The NRC staff also concludes that failure to issue this license amendment would result in the shutdown of CGS.

The NRC staff agrees that an emergency situation exists consistent with the provisions in 10 CFR 50.91 (a)(5). The NRC staff determined that: (1) the licensee used its best efforts to make a timely application; (2) the licensee could not reasonably have avoided the situation; and (3) the licensee has not abused the provisions of 10 CFR 50.91 (a)(5). Based on these findings, and the determination that the amendment involves no significant hazards consideration as discussed below, the NRC staff has determined that a valid need exists for issuance of the license amendment using the emergency provisions of 10 CFR 50.91 (a)(5).

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

Pursuant to 10 CFR 50.91 (a)(5), if the Commission finds that an emergency situation exists, in that failure to act in a timely way would result in shutdown of a nuclear power plant, it may issue a license amendment involving no significant hazards consideration without prior notice and

opportunity for a hearing or for public comment. As noted in Section 4.0 of this safety evaluation, the NRC staff has concluded that an emergency situation does exist, in that failure to issue the amendment would result in the shutdown of CGS. Therefore, a final finding of no significant hazards consideration follows.

The Commission has made a final determination that the amendment request involves no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As required by 10 CFR 50.91 (a), in its letter dated January 30, 2015, the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment does not increase the probability of an accident because the RHR system cannot initiate an accident. The RHR system provides coolant injection to the reactor core, cooling of the suppression pool water inventory, and drywell sprays following a design basis accident.

The proposed one time 14 day CT change does not alter the conditions, operating configurations, or minimum amount of operating equipment assumed in the safety analysis for accident mitigation. No changes are proposed in the manner in which the ECCS provides plant protection or which create new modes of plant operation. In addition, a PSA

[probabilistic safety assessment] evaluation concluded that the risk contribution of the increased CT is a very small increase in risk. The proposed change in CT will not affect the probability of any event initiators. There will be no degradation in the performance of, or an increase in the number of challenges imposed on, safety related equipment assumed to function during an accident situation. There will be no change to normal plant operating parameters or accident mitigation performance. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment will not create the possibility of a new or different kind of accident because inoperability of one RHR subsystem is not an accident precursor. There are no hardware changes nor are there any changes in the method by which any plant system performs a safety function. This request does not affect the normal method of plant operation. The proposed amendment does not introduce new equipment, or new way of operation of the system which could create a new or different kind of accident. No new external threats, release pathways, or equipment failure modes are created. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this request. Therefore, the implementation of the proposed amendment will not create a possibility for an accident of a new or different type than those previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Columbia's ECCS is designed with sufficient redundancy such that a low pressure ECCS subsystem may be removed from service for maintenance or testing and the remaining subsystems are capable of providing water and removing heat loads to satisfy the FSAR requirements for accident mitigation or plant shutdown. A PSA evaluation concluded that the risk contribution of the CT extension is within allowable limits. There will be no change to the manner in which safety limits or limiting safety system settings are determined nor will there be any change to those plant systems necessary to assure the accomplishment of protection functions. For these reasons, the proposed amendment does not involve a significant reduction in a margin of safety.

Accordingly, the NRC staff has determined that this amendment involves no significant hazards determination.

6.0 REGULATORY COMMITMENTS In its letter dated January 30, 2015, the licensee made the following regulatory commitment:

Compensatory measures outlined in section 3.2 of this letter will be implemented during the period of the addition 7 day completion time.

The NRC staff concludes that reasonable controls for the implementation of and/or evaluation of proposed changes to the above regulatory commitment are best provided by the licensee's

administrative processes, including its commitment management program. Therefore, the NRC staff concludes that the above regulatory commitment does not warrant the creation of a regulatory requirement (an item requiring prior NRC approval of any subsequent changes).

7.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Washington State official was notified of the proposed issuance of the amendment. The State official had no comments as of the date of issuance of this amendment.

8.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation and use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

9.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) the amendment does not (a) involve significant increase in the probability or consequences of an accident previously evaluated or, (b) create the possibility of a new or different kind of accident from any previously evaluated or, (c) involve a significant reduction in a margin of safety and, therefore, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (3) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (4) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: A George, NRR/DORL H. Hamzehee, NRR/DRA/APLA M. Hamm, NRR/DSS/STSB Date: February 1, 2015

ML15030A501 *via email OFFICE NRRIDORULPL4-1 /PM NRR/DORULPL4-1 /LA NRRIDSS/STSB/BC NRR/DSS/SRXB/BC NAME A George JBurkhardt* REIIiott* CJackson*

DATE 1/31/15 1/31/15 2/1/15 2/1/15 OFFICE NRRIDRAIAPLA/BC OGC- NLO NRRIDORL/LPL4-1 /BC(A) NRR/DORULPL4-1 /PM NAME HHamzehee* MSpencer* EOesterle A George DATE 2/1/15 2/1/15 2/1/15 2/1/15