Text

Issuance of Amendment No. 275 Revision to Technical Specification 3.3.1.1 (Emergency Circumstances)
	ML24096A120
Person / Time
Site:	Cooper
Issue date:	04/05/2024
From:	Byrd T; NRC/NRR/DORL/LPL4
To:	Dent J; Nebraska Public Power District (NPPD)
References
	EPID L-2024-LLA-0037
	Download: ML24096A120 (1)
	v • d • e

April 5, 2024 John Dent, Jr.

Executive Vice President and Chief Nuclear Officer Nebraska Public Power District Cooper Nuclear Station 72676 648A Avenue P.O. Box 98 Brownville, NE 68321

SUBJECT:

COOPER NUCLEAR STATION - ISSUANCE OF AMENDMENT NO. 275 RE:

REVISION TO TECHNICAL SPECIFICATION 3.3.1.1 (EMERGENCY CIRCUMSTANCES) (EPID L-2024-LLA-0037)

Dear John Dent:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 275 to Renewed Facility Operating License No. DPR-46 for the Cooper Nuclear Station (Cooper). The amendment consists of changes to the Technical Specifications (TS) in response to your application dated April 1, 2024, as supplemented by letter dated April 3, 2024.

The amendment changes TS 3.3.1.1, Reactor Protection System (RPS) Instrumentation, specifically by adding a temporary footnote to TS Table 3.3.1.1-1, Reactor Protection System Instrumentation, that is applicable to Function 8, Turbine Stop Valve - Closure. The current design configuration of the turbine stop valve position switches that input to the RPS does not meet the channel independence criteria. This temporary footnote allows the licensee to not enter the associated actions in TS 3.3.1.1, Condition A or B for the channel independence condition.

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

J. Dent, Jr.

The NRC staffs related safety evaluation is also enclosed. The safety evaluation describes the emergency circumstances under which the amendment is issued, and the NRCs final no significant hazards determination and an opportunity for a hearing associated with the issuance of this amendment under emergency circumstances will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Thomas J. Byrd, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

1. Amendment No. 275 to DPR-46

2. Safety Evaluation cc: Listserv

NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 275 Renewed License No. DPR-46

1.

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

A.

The application for amendment by Nebraska Public Power District (the licensee) dated April 1, 2024, as supplemented by letter dated April 3, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

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. DPR-46 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 275, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-46 and the Technical Specifications Date of Issuance: April 5, 2024 Jennivine K.

Rankin Digitally signed by Jennivine K. Rankin Date: 2024.04.05 16:28:17 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 275 RENEWED FACILITY OPERATING LICENSE NO. DPR-46 COOPER NUCLEAR STATION DOCKET NO. 50-298 Replace the following pages of Renewed Facility Operating License No. DPR-46 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 3.3-8 3.3-8

Amendment No. 275 (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2419 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 275, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: Cooper Nuclear Station Safeguards Plan, submitted by letter dated May 17, 2006.

NPPD shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The NPPD CSP was approved by License Amendment No. 238 as supplemented by changes approved by License Amendments 244 and 249.

(4) Fire Protection NPPD shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated April 24, 2012 (and supplements dated July 12, 2012, January 14, 2013, February 12, 2013, March 13, 2013, June 13, 2013, December 12, 2013, January 17, 2014, February 18, 2014, and April 11, 2014), and as approved in the safety evaluation dated April 29, 2014.

Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if

FUNCTION

7. Scram Discharge Volume Water Level - High a.

Level Transmitter b.

Level Switch 8.

Turbine Stop Valve - Closure 9.

Turbine Control Valve Fast Closure, DEH Trip Oil Pressure -

Low

10. Reactor Mode Switch -

Shutdown Position

11. Manual Scram RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2 1,2

29.5% RTP

29.5% RTP 1,2 1,2 REQUIRED CHANNELS PER TRIP SYSTEM 2

2 2

CONDITIONS REFERENCED FROM REQUIRED SURVEILLANCE ALLOWABLE ACTION D.1 REQUIREMENTS VALUE G

SR 3.3.1.1.4 s; 90 inches SR 3.3.1.1.9 H

G H

E E

G H

SR 3.3.1.1.12(a,b}

SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.4 SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.4 SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.4 SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.13 SR 3.3.1.1.15 s; 90 inches s; 90 inches s; 90 inches SR 3.3.1.1.4 s; 10% closed SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.4

[ 1018 psig SR 3.3.1.1.9 SR 3.3.1.1.12(a,b}

SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.9 SR 3.3.1.1.13 NA NA NA NA (a) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(b) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (L TSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the L TSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-found and the as-left tolerances are specified in the Technical Requirements Manual.

(c) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) The Turbine Stop Valve - Closure function does not meet channel independence criteria. Conditions A and B are not required to be entered for this condition. NPPD will implement compensatory measures described in NPPD letter NLS2024024, dated April 1, 2024 (ML24092A376), during the applicable specified condition for the position switches until startup from RE33. If the position switches are returned to OPERABLE status prior to startup from RE33, then these compensatory measures are no longer required.

Cooper 3.3-8 Amendment No. 275

SAFETY EVALUATION BY THE OFFICE NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

By letter dated April 1, 2024 (Reference 1), as supplemented by letter dated April 3, 2024 (Reference 2), Nebraska Public Power District (NPPD, the licensee) submitted an emergency license amendment request (LAR) to revise the technical specifications (TS) for Cooper Nuclear Station (Cooper).

The amendment would change TS 3.3.1.1, Reactor Protection System (RPS) Instrumentation.

Specifically, a temporary footnote would be applied to Function 8, Turbine Stop Valve -

Closure, of TS Table 3.3.1.1-1, Reactor Protection System Instrumentation. The current design configuration of the Turbine Stop Valve (TSV) position switches that input to the RPS does not meet the channel independence criteria. This temporary footnote would allow NPPD to not enter the TS 3.3.1.1, Condition A or B for the channel independence condition for a period ending no later than startup from Refuel Outage 33 (RE33).

The LAR is a follow up to the notice of enforcement discretion (NOED) request (Reference 3) from TS 3.3.1.1, Conditions A and B, in accordance with U.S. Nuclear Regulatory Commission (NRC) Enforcement Manual, Appendix F, Notices of Enforcement Discretion (Reference 4).

Cooper entered the actions for TS 3.3.1.1, Conditions A and B for Function 8 due to a concern bringing into question the channel independence associated with both sets of TSV RPS position switches on March 29, 2024. Subsequently, the NRC granted verbal approval to the NPPDs enforcement discretion request on March 29, 2024, and issued a letter detailing the approval on April 2, 2024 (Reference 5).

1.1

System Description

The Cooper unit is a General Electric design Type 4 Boiling Water Reactor. The RPS for Cooper is described in Chapter VII, section 2.0 of the Updated Safety Analysis Report (USAR)

(Reference 6). The RPS system provides timely protection against the onset and consequences of conditions that threaten fuel barrier integrity and challenge the reactor coolant pressure boundary (RCPB). The RPS system provides protection from uncontrolled release of radioactive

material from the fuel and RCPB by initiating an automatic scram to terminate excessive neutron flux and pressure increases. The input parameters to the RPS scram logic are from instrumentation that monitors reactor vessel water level, reactor vessel pressure, neutron flux, main steam line isolation valve position, turbine control valve fast closure, TSV position, drywell pressure, and scram discharge volume water level.

The RPS is arranged as two separately powered trip systems (trip system A and trip system B).

Each trip system has redundancy within it. Actuation of at least one channel from within each trip system (designated as RPS Channels A1, A2, B1, and B2) is combined to produce an automatic trip signal. The outputs of the automatic logic channels in a trip system are combined in a one-out-of-two logic so that either channel can trip the associated trip system, which creates a half-scram condition. The tripping of both trip systems A and B will produce a reactor scram. The operability of RPS is dependent on the operability of individual instrumentation channel functions specified in TS Table 3.3.1.1-1.

1.1.1 Turbine Stop Valve Closure The purpose of the TSV closure scram signal is to protect the reactor whenever its link to the heat sink is in the process of being removed. The valve stem position of each TSV is monitored by limit switches. At Cooper, two limit switches on each TSV provide input to each RPS trip system. The RPS A1 and A2 limit switches monitor the position of TSV1 and the RPS B1 and B2 limit switches monitor the position of TSV2. Closure of the main TSV with the reactor at high power can result in a significant addition of positive reactivity to the core as the resulting rise in reactor vessel pressure collapses steam voids. The TSV closure scram, which is used to sense loss of heat sink condition, will initiate a scram earlier than either the neutron high flux monitors or reactor vessel high pressure instruments. This provides a satisfactory margin below core thermal hydraulic limits for this category of abnormal operational transients. The scram counteracts the addition of positive reactivity due to pressure by inserting negative reactivity with the control rods. Although the reactor vessel high pressure scram, in conjunction with the pressure relief system, is adequate to preclude overpressurizing the reactor vessel, the TSV closure scram provides additional margin to the reactor vessel pressure limit.

Each RPS trip system for the Turbine Stop Valve-Closure Function receives two Turbine Stop Valve-Closure channel inputs from a TSV, each consisting of one position (i.e., limit) switch assembly with two contacts, each inputting to a relay. The relays provide a parallel logic input to an RPS trip logic channel. The logic for the Turbine Stop Valve-Closure Function requires that both TSVs must be less than 90 percent open to produce a reactor scram initiation. A single TSV closure will only produce a half scram.

1.2 Proposed Changes The existing Cooper TS Table 3.3.1.1-1, Function 8, Turbine Stop Valve - Closure, requires two channels per trip system. If one or more required channels are inoperable, Condition A of Limiting Condition for Operation (LCO) 3.3.1.1 is entered, which requires placing either the channel or associated trip system in trip status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If one or more functions with one or more required channels are inoperable in both trip systems, Condition B is entered which requires placing either a channel in one trip system or one trip system in trip status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If both Condition A and Condition B completion times are not met, then for Function 8, Condition E is entered requiring the thermal power to be reduced to less than 29.5 percent of rated thermal power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The configuration of the mechanical trip input to the TSV position switches in its current design configuration does not meet channel independence criteria. The two position switches on each TSV are mechanically connected to actuate together on TSV closure. The specification of the LCO within TS Table 3.3.1.1-1, Function 8, requires the operability of two channels per trip system. This configuration of not meeting the channel independence criteria for RPS inputs is unique to the TSVs.

The licensee proposed to add a temporary Note (d) to TS Table 3.3.1.1-1 that applies only to Function 8, which states:

(d) The Turbine Stop Valve - Closure function does not meet channel independence criteria. Conditions A and B are not required to be entered for this condition. NPPD will implement compensatory measures described in NPPD letter NLS2024024, dated April 1, 2024 (ML24092A376), during the applicable specified condition for the position switches until startup from RE33. If the position switches are returned to OPERABLE status prior to startup from RE33, then these compensatory measures are no longer required.

1.3 Emergency LAR Basis As a result of the unforeseen concern with channel independence, at 1437 on March 29, 2024, Cooper entered a 12-hour action statement under TS 3.3.1.1, Condition A, and a 6-hour action statement under TS 3.3.1.1 Condition B, related to TS Table 3.3.1.1-1 Function 8. The licensee stated in the LAR, as supplemented, that resolution of the design concern cannot be accomplished within the completion times without requiring a reduction in rated thermal power below 29.5 percent, with no commensurate benefit in nuclear safety.

The licensee-proposed changes to TS Table 3.3.1.1-1 would allow the licensee to not enter Conditions A and B for the channel independence issue until it is resolved. The licensee stated in the LAR that a resolution would not be able to be implemented until RE33 in the fall of 2024.

This change is needed sooner than can be issued under exigent circumstances because the Notice of Enforcement Discretion approved on March 29, 2024, will expire at 0237 on April 6, 2024.

2.0 REGULATORY EVALUATION

2.1 Applicable Regulatory Requirements Section 50.36, Technical specifications, of the Title 10 of the Code of Federal Regulations (10 CFR) establishes the regulatory requirements related to the content of TSs. The regulations in 50.36(a)(1) requires an application for an operating license to include proposed TSs. The regulations further state that, A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs. The regulation in 10 CFR 50.36(c)(2) states, in part, that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met. The regulation in 10 CFR 50.36(c)(3) states that surveillance requirements are requirements relating to test, calibration, or inspection to assure

that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The regulations in 10 CFR 50.91(a)(5) states, in part, that Where the Commission 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.

Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (hereinafter referred to as GDC), establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components (SSCs) important to safety.

The NRC staff found the following GDC applicable to the subject LAR:

GDC 10, Reactor design, states that, The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 21, Protection system reliability and testability, states that, The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed.

Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated.

The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

GDC 22, Protection system independence states that, The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis.

Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.

Cooper was not licensed to the 10 CFR Part 50, Appendix A, GDC. Cooper was designed and constructed to meet the principal design criteria (PDC) described in the Atomic Energy Commissions (AEC) proposed rule, General Design Criteria for Nuclear Power Plant Construction Permits, published in the Federal Register on July 11,1967 (32 FR 10213). The conformance to the 1967 proposed GDC is described in Appendix F, Conformance to AEC Proposed General Design Criteria, to the Cooper USAR.

The Cooper PDC requirements as compared to the applicable GDCs for the proposed change are listed below:

Criterion 14, Core Protection Systems, states, Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

Criterion 20, Protection Systems Redundancy and Independence states that, Redundancy and independence designed into protection systems shall be sufficient to assure that no single failure or removal from service of any component or channel of a system will result in loss of the protection function.

The redundancy provided shall include, as a minimum, two channels of protection for each protection function to be served. Different principles shall be used where necessary to achieve true independence of redundant instrumentation components.

Pursuant to 10 CFR 50.55a(a)(2)(ii), the Cooper USAR references IEEE Std 279, Criteria for Protection Systems for Nuclear Power Generating Stations. Chapter VII, section 1.7, of the Cooper USAR describes a compliance comparison of the RPS and emergency core cooling system design with each design requirement of the proposed IEEE Std 279, by making reference to the Atomic Power Equipment Division (APED) Topical Report NEDO-10139, Compliance of Protection Systems to Industry Criteria: General Electric BWR Nuclear Steam Supply System, dated June 1970. The USAR states that this NEDO report is applicable to the Cooper Plant.

Chapter VII, section 2.3.6.4 of the Cooper USAR states that the switches on each valve are mechanically and electrically separated and satisfy IEEE Std 279. Channel Functional Testing is performed by manipulation of each stop valve during periods of low (or no) power or manual actuation of the limit switches at any power level.

Clause 4.2, Single failure, of IEEE Std 279-1971, states:

Any single failure within the protection system shall not prevent proper protective action at the system level when required.

Clause 4.6, Channel Independence, of IEEE Std 279-1971, states:

Channels that provide signals for the same protective function shall be independent and physically separated to accomplish decoupling of the effects of unsafe environmental factors, electric transients, and physical accident consequences documented in the design basis, and to reduce the likelihood of

interactions between channels during maintenance operations or in the event of channel malfunction.

2.2 Applicable Regulatory Guidance NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water] Edition, Chapter 15, Revision 3, Transient and Accident Analysis (Reference 7), was used for specific review criteria for the proposed request.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the emergency LAR, as supplemented, to determine whether the proposed changes are consistent with the regulatory requirements, the plant-specific design and the licensing basis. The NRC staff evaluations are detailed below.

3.1 RPS Channel Independence and Functionality Evaluation In the current configuration, the two position switches on each TSV are mechanically connected to actuate together on TSV closure. The upper TSV position switch (A1 on TSV1 and B1 on TSV2) is physically actuated by the TSV, while the lower position switch (A2 on TSV1 and B2 on TSV2) is only actuated by the movement of the upper switch by way of the mechanical linkage. Both switches on each valve currently provide a switching action to drop out normally energized relays in their respective RPS trip channels when the lever arm of each switch is rotated. In their current configuration, these protection system sensors do not meet the intent of the design criteria in PDC 20 and IEEE Std 279-1971, Clause 4.6, on channel independence.

The LAR describes that a failure of the mechanical linkage between the position switches would only prevent the lower position switch from actuating on TSV closure. Therefore, with a failure of the mechanical linkage, the one-out-of-two logic per TSV would still be met to produce a half-scram condition in the RPS logic. However, the LAR further describes that if a failure occurred that prevents rotation of the upper position switch about the fulcrum (splined shaft), this could prevent actuation of both switches. The licensee states in the LAR that this failure mode has never been observed to occur throughout the Cooper service life. The NRC staff notes that although this configuration of the TSV limit switch sensors with mechanical linkage does not meet the design criteria for independence in PDC 20 and IEEE Std 279-1971, the current configuration appears to be performing reliably. That is, of the failure modes that have occurred over the years, failure of the mechanical linkage is not one of them. In the LAR, the licensee states that Over the course of the plant's operating history, the station has not experienced this failure; the freedom of movement and proper rotation are verified by periodic surveillance of the switches.

At present, both limit switches on each TSV are functional and are responsive to the closure of the TSV, although one of the two switches on each TSV is indirectly monitoring the position of the TSV. By letter dated April 3, 2024, the licensee responded to the NRC staffs request for additional information (RAI) (Reference 8), that both TSV1 position switches (RPS Channels A1 and A2) were replaced and calibrated in January 2024. The licensee also stated that the TSV2 position switches (RPS Channels B1 and B2) were last calibrated during refueling outage RE32 (November 9, 2022) per Cooper Surveillance Procedure 6.RPS.303. This calibration is performed by stroking the valves individually and verifying that the applicable RPS relays drop out when the valve is at <10 percent closed.

In addition, the licensee stated in response to an RAI that a functional test of the RPS A1, A2, B1, and B2 position switches was last performed in March 2024. This functional test is performed on a 13-week frequency (during quarterly downpowers) using Cooper Surveillance Procedure 6.RPS.302, Main Turbine Stop Valve Closure and Steam Valve Functional Test. By letter dated April 3, 2024, the licensee responded to an RAI, in part:

Surveillance Procedure 6.RPS.302 is performed on a quarterly frequency to complete the channel functional test of the TSV position switches, RPS Channels A1, A2, B1, and B2. This is done by either manual actuation of the position switch linkage and a visual inspection of the linkage or by stroking of the individual TSV one at a time. Both methods verify the applicable RPS relays drop out. To perform the method of stroking of the TSV, power must be less than or equal to 1693 MWt. Manual actuation of the TSV position switches can be performed at any power but is typically performed during a downpower to reduce dose. The required frequency for this surveillance as described in Appendix B of the TRM 3.3.1.1.9 Function 8, is 92 days.

The licensee noted that they are currently scheduled to perform these functional tests during the May 2024 and August 2024 quarterly downpowers, which represents two additional periodic surveillances that will be conducted for each switch before the fall 2024 refueling outage.

The NRC staff notes that although the current design configuration of the TSV position switches that input to the RPS does not meet the channel independence criteria of PDC 20 and IEEE Std 279-1971, Clause 4.6, all the TSV position switches are currently functional and will continue to be functionally tested and visually inspected by the licensee every 13 weeks until the fall refueling outage. Therefore, the staff has reasonable assurance that the TSV position switches and the RPS input channels for the Turbine Stop Valve - Closure Function can be demonstrated to continue to function reliably until the outage, and that any mis-operation or degradation in the switch lever mechanisms will be readily detected and corrected, if necessary.

3.2 Pressure Boundary Evaluation The LAR states that the scrams based on turbine governor valve fast closure and TSV fast closure are anticipatory in nature, and they provide additional margin to overpressurization of reactor coolant system boundary. It states that the reactor vessel high pressure scram, in conjunction with the pressure relief system, is adequate to preclude overpressurizing the reactor vessel. The LAR states that the RCPB analysis is based on main steam isolation valve (MSIV) closure event with a flux scram and the pressure values for this event are below the applicable limits when evaluating the shortest closure time of 3 seconds.

In response to the NRC staff RAI, the licensee provided a response stating that the overpressure analysis in section 3.3 of the LAR is based on the closure of all MSIVs. This event is described in Chapter IV, section 4.9.3 of the Cooper USAR. The licensee stated that event assumes the scram off the position of the MSIVs (direct scram) does not occur and the neutron flux signal scrams the reactor. Chapter XIV, section 4.4.2 of the Cooper USAR confirms that this hypothesized overpressure protection event is more severe than any of the anticipated operational occurrences (AOOs) described in Chapter XIV. Hence, the AOOs will not result in nuclear system stress more than that allowed for transients by applicable industry codes.

Chapter XIV, section 5.1 of the Cooper USAR lists the analyses of the AOOs resulting in nuclear system pressure increase, including the TSV closure event. The analyses performed

show that a full closure of all MSIV with direct scram is the limiting pressure increase AOO for the RCPB.

The LAR states that the TSVs close via depressurization of the stop valve emergency trip header, which will also depressurize the control valve trip header leading to a separate scram signal. This scram signal is the basis for turbine control valve fast closure (generator load rejection event). The turbine control valve fast closure is explicitly analyzed in Cooper USAR, Chapter XIV. Thus, in case of failure of a scram signal from the TSV position, the reactor will scram on turbine control valve fast closure event.

Based on the analysis in Chapter IV and Chapter XIV of the Cooper USAR, as well as the information presented by the licensee, the NRC staff agrees with the licensee statement that the reactor vessel high pressure scram, in conjunction with the pressure relief system, is adequate to preclude overpressurizing of the reactor vessel. Hence, the NRC staff finds that the failure of a scram signal from the TSV position does not pose risk to the reactor coolant system boundary.

3.3 Fuel Damage Evaluation The LAR states that fuel damage is avoided by setting the most limiting operating minimum critical power ratio (MCPR) analyzed for the cycle, which is the Inadvertent high pressure coolant injection (HPCI) event from the date of the LAR to the end of Cycle 33. Chapter XIV, section 5.2 of the Cooper USAR has a cycle-specific reload analysis that demonstrated the limiting inadvertent HPCI event MCPR would not exceed the safety limit MCPR.

In response to the NRC staff RAI, the licensee stated that a sensitivity case for an inadvertent HPCI event, without crediting the reactor scram on TSV position, was performed for Cycle 33.

The results of the sensitivity case performed round to the same Operating Limit MCPR values as provided in the Cooper Cycle 33 Supplemental Reload Licensing Report.

Based on the fact that (a) turbine trip without bypass (initiated by TSV closure) event is not limiting for the remainder of Cycle 33, (b) the limiting inadvertent HPCI event, without crediting the reactor scram on TSV position, does not exceed the safety limit MCPR, and (c) the reactor will scram on turbine control valve fast closure in the event of TSV closure with a failure of a scram signal from the TSV position, the NRC staff finds the licensee request does not pose a fuel damage risk.

3.4 Compensatory Measures The licensee specifies two sets of compensatory measures with respect to the performance of the Cooper turbine stop valves. First, the licensee will use its protected equipment program to ensure heightened sensitivity and risk management actions associated with equipment with elevated potential for initiating a main turbine trip. The licensees process will ensure management review and approval for activities that might impact the applicable equipment.

Second, the licensee will apply its Critical Evolution Meeting process to Cooper Procedure 6.RPS.302 to ensure additional management oversight and approval are applied to activities that exercise the main turbine stop valve position switches. In its supplemental letter dated April 3, 2024, the licensee confirmed that the compensatory measures will continue until startup from Refueling Outage 33 with the clarification that the compensatory measures are required when the TSV position switches are required to be operable (i.e., greater than 29.5 percent rated thermal power). The licensee noted that during low power and shutdown

conditions, the compensatory measures are not necessary because the position switches are not required to be operable.

In the supplemental letter dated April 3, 2024, the licensee provided a revision to proposed Note (d) to TS Table 3.3.1.1-1 to include the reference to the compensatory measures, which designates them as requirements. The licensee also stated that the proposed TS change will not impact the inservice testing program activities at Cooper. Based on the revision to proposed Note (d) to TS Table 3.3.1.1-1, the NRC staff finds that the compensatory measures will provide reasonable assurance of proper oversight of activities related to the turbine stop valves until the next refueling outage.

3.5 Technical Conclusion The NRC staff has reviewed the licensees proposed change to Cooper TS Table 3.3.1.1-1 to temporarily allow the licensee to not enter the TS 3.3.1.1, Condition A or B, for the TSV closure channel independence condition for a period ending no later than startup from RE33. Based on the NRC staffs evaluation in sections 3.1 through 3.4 of this safety evaluation, the staff concludes that proposed Note (d) to Table 3.3.1.1-1 is acceptable, and the TSV closure position switches historical operational and test data, analyses presented by the licensee, and proposed compensatory measures provide reasonable assurance that Cooper has the capability for safe shutdown during the period of operation from March 29, 2024, through RE33 in fall 2024.

Therefore, the NRC staff concludes that the proposed change will not impact the licensee's continuous compliance with the regulatory requirements in 10 CFR 50.36(c)(2) and 10 CFR 50.36(c)(3).

4.0 EMERGENCY SITUATION The NRCs regulations in 10 CFR 91(a)(5) state that where the NRC finds that an emergency situation exists, in that failure to act in a timely manner would result in the derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to a plants licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for public comment or for public hearing. In such a situation, the NRC will publish a notice of issuance under 10 CFR 2.106, providing for opportunity for a hearing and for public comment after issuance.

In the application letter dated April 1, 2024 (Reference 1), and in response to the NRCs RAI dated April 3, 2024 (Reference 2), the licensee requested that the NRC process the proposed amendment on an emergency basis. The licensee stated that because of a concern with channel independence, at 1437 central standard time (CST) on March 29, 2024, Cooper entered a 12-hour action statement under TS 3.3.1.1, Condition B, related to TS Table 3.3.1.1-1, Function 8. Resolution of the design concern could not be accomplished within the completion time of the TS LCO required action without requiring a reduction in thermal power below 29.5 percent without a commensurate benefit in nuclear safety. The requested action was also sooner than could be accomplished under exigent circumstances because the prior issued Notice of Enforcement Discretion (NOED) expires at 0237 April 6, 2024 (Reference 5), thus making this license amendment timely due to the unplanned nature of this condition.

The NRC staff reviewed the licensees basis for processing the proposed amendment as an emergency amendment (as discussed above) and determines that an emergency situation exists consistent with the provisions of 10 CFR 50.91(a)(5). Furthermore, the NRC staff determined that:

(1) the licensee used its best efforts to provide a timely application, (2) the licensee could not have reasonably avoided this 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

The NRCs regulations in 10 CFR 50.92(c) state that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would 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), by letter dated April 1, 2024, the licensee provided its analysis of the issue of no significant hazards consideration, which is presented below:

1)

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

Response: No The proposed amendment is to allow operation of Cooper Nuclear Station for the remainder of Cycle 33 with the Reactor Protection System (RPS) function for the Turbine Stop Valve (TSV) - Closure not meeting channel independence criteria. The TSV position switches that initiate the RPS function for the TSV closure are not accident initiators. The RPS functions to prevent fuel damage, limit system pressure and thus restrict the release of radioactive material.

Scrams based on turbine governor valve fast closure and TSV fast closure signals are anticipatory in nature in order to provide additional margin to over pressurization of the reactor coolant system boundary.

The high reactor vessel pressure scram in conjunction with the pressure relief system is sufficient to maintain reactor coolant boundary pressure below applicable limits. The reactor coolant boundary analysis is based on the Main Steam Isolation Valve closure event with flux scram. This event does not assume TSV or control valve fast closure scrams.

Fuel damage is avoided by setting the most limiting operating Minimum Critical Power Ratio based on the most limiting event analyzed for the cycle. The limiting pressurization events from the cycle reload analysis are inadvertent injection of High Pressure Coolant Injection, generator load rejection without bypass, and feedwater controller failure (maximum demand). Since the Turbine Stop Valve - Closure function is not limiting for the cycle, the consequences of an accident are not increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2)

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

Response: No The proposed amendment is to allow operation of Cooper Nuclear Station for the remainder of Cycle 33 with the RPS function for the Turbine Stop Valve - Closure not meeting channel independence criteria. This amendment does not change the design function or operation of the RPS instrumentation. No changes in processes are being proposed that would affect the operation of the RPS instrumentation or affect its ability to perform its design function.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3)

Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed amendment is to allow operation of Cooper Nuclear Station for the remainder of Cycle 33 with the RPS function for the Turbine Stop Valve - Closure not meeting channel independence criteria. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the Reactor Coolant System, and the containment system. Notwithstanding that Criterion 20,Protection Systems Redundancy and Independence, of the 1967 proposed General Design Criteria is not fully met, the proposed amendment will not challenge the acceptability of any analytical limits under normal, transient, and accident conditions. All applicable design and safety limits will continue to remain satisfied such that the fission product barriers will continue to perform their design functions.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

6.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Nebraska State official was notified of the proposed issuance on April 2, 2024. The State official had no comments.

7.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or 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. As discussed in section 5.0 of this SE, the Commission has determined that the amendment involves no significant hazards consideration. 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.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

9.0 REFERENCES

1.

Dia, K., NPPD, letter to the NRC, Emergency License Amendment Request to Revise Technical Specifications 3.3.1.1, Reactor Protection (RPS)

Instrumentation Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46, dated April 1, 2024 (Agencywide Documents Access and Management System (Agencywide Documents Access and Management System (ADAMS) Accession No ML24092A376).

2.

Dia, K., NPPD, letter to the NRC, Response to Nuclear Regulatory Commission Request for Additional Information Regarding Application for the Emergency License Amendment Request to Revise Technical Specifications 3.3.1.1, Reactor Protection (RPS) Instrumentation Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46, dated April 3, 2024 (ML24094A331).

3.

Dia, K., NPPD, letter to NRC, Request for Notice of Enforcement Discretion for Technical Specifications 3.3.1.1, Reactor Protection (RPS) Instrumentation Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46, dated March 31, 2024 (ML24091A003).

4.

NRC, Appendix F, Notices of Enforcement Discretion, dated December 26, 2023 (ML23362A014).

5.

Miller, G., NRC, letter to K. Dia, NPPD, Notice of Enforcement Discretion for Cooper Nuclear Station (EPID: L-2024-LLD-0003), EA-24-036, dated April 2, 2024 (ML24093A228).

6.

NPPD, Cooper Nuclear Station Submittal of Revision 31 to Updated Safety Analysis Report, dated April 20, 2023 (Package ML23129A261).

7.

NRC, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, NUREG-0800, Chapter 15, Revision 3 Transient and Accident Analysis, dated March 2007 (ML070710376).

8.

Byrd, T., NRC, email to L. Dewhirst, NPPD, Re: Cooper - RAI Emergency LAR Turbine Stop Valve Limit Switch TS 3.3.1.1 (EPID L-2024-0LLA-0037), dated April 3, 2024 (ML24094A246).

Principal Contributors: D. Rahn S. Darbali S. Bhatt K. West T. Scarbrough T. Byrd Date: April 5, 2024

ML24096A120

concurrence via email NRR-058 OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA*

NRR/DSS/STSB/BC*

NRR/DSS/SNSB/BC*

NAME TByrd PBlechman SMehta PSahd DATE 04/04/2024 04/04/2024 04/04/2024 04/04/2024 OFFICE NRR/DEX/EICB/BC*

NRR/DEX/EMIB/BC*

OGC - NLO*

NRR/DORL/LPL4/BC*

NAME FSacko SBailey MCarpentier JRankin DATE 04/04/2024 04/04/2024 04/05/2024 04/05/2024 OFFICE NRR/DORL/LPL4/PM*

NAME TByrd DATE 04/05/2024

v t e Letter Sequence Approval
EPID:L-2024-LLA-0037, Request for Notice of Enforcement Discretion for Technical Specifications 3.3.1.1, Reactor Protection System (RPS) Instrumentation (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions 3 April 2024 Answers 3 April 2024 Results Approval Other:
MONTHYEAR2024-04-01 [Table View]