ET 10-0014, Application to Revise Technical Specification 3.3.2, Engineered Safety Feature Actuation System Instrumentation, Table 3.3.2-1

Application to Revise Technical Specification 3.3.2, Engineered Safety Feature Actuation System Instrumentation, Table 3.3.2-1

ML101100391
Person / Time
Site:	Wolf Creek
Issue date:	04/13/2010
From:	Garrett T Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ET 10-0014
Download: ML101100391 (25)

Text

WLF CREEK NUCLEAR OPERATING CORPORATION Terry J. Garrett Vice President Engineering April 13, 2010 ET 10-0014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Application To Revise Technical Specification 3.3.2, "Engineered Safety Feature Actuation System Instrumentation," Table 3.3.2-1 Gentlemen:

Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). The proposed amendment revises Table 3.3.2-1, Function 8.a.,

(ESFAS Interlocks, Reactor Trip, P-4) of Technical Specification (TS) 3.3.2, "Engineered Safety Feature Actuation System Instrumentation."

WCNOC is proposing to add footnote (m) to Function 8.a. to identify the enabled functions and the applicable MODES for the Reactor Trip, P-4 interlock Function.

Attachment I through IV provide the Evaluation, Markup of TSs, Retyped TS pages, and proposed TS Bases changes, respectively, in support of this amendment request. Attachment IV, proposed changes to the TS Bases, is provided for information only.

Final TS Bases changes will be implemented pursuant to TS 5.5.14, "Technical Specification (TS) Bases Control Program," at the time the amendment is implemented. Attachment V provides a List of Regulatory Commitments made by WCNOC in this submittal.

It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92.

Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

This amendment application was reviewed by the Plant Safety Review Committee.

In accordance with 10 CFR 50.91, a copy of this amendment application, with attachments, is being provided to the designated Kansas State official.

Ao/

P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET Pd

ET 10-0014 Page 2 of 3 WCNOC requests approval of the proposed amendment by March 1, 2011 to support activities during Refueling Outage 18 and subsequent Refueling Outages.

It is anticipated that the license amendment, as approved, will be effective upon issuance and will be implemented within 90 days from the date of issuance. Please contact me at (620) 364-4084 or Mr. Richard Flannigan at (620) 364-4117 for any questions you may have regarding this application.

Sincerely, SJ Garrett TJG/rlt Attachments:

IV IIIV V

Evaluation Proposed Technical Specification Changes (Mark-up)

Revised Technical Specification Pages Proposed TS Bases Changes (for information only)

List of Regulatory Commitments cc:

E. E. Collins (NRC), w/a T. A. Conley (KDHE), w/a G. B. Miller (NRC), w/a B. K. Singal (NRC), w/a Senior Resident Inspector (NRC), w/a

ET 10-0014 Page 3 of 3 STATE OF KANSAS COUNTY OF COFFEY

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Terry J. Garrett, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

Terry /darrett' Vice President Engineering SUBSCRIBED and sworn to before me this day ofdayof

,2010.

GAYLE SHEPHEARD1 Notary Public - $tate pf Kansas My Appt. Expires

'7/2

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201/ J Notary Foblic Expiration Date 71L12oI I

Attachment I to ET 10-0014 Page 1 of 12 EVALUATION 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusion

5.0 ENVIRONMENTAL CONSIDERATION

Attachment I to ET 10-0014 Page 2 of 12 EVALUATION 1.0

SUMMARY

DESCRIPTION The amendment application proposes changes to Wolf Creek Generating Station (WCGS)

Technical Specification (TS) Table 3.3.2-1, Function 8.a., (ESFAS Interlocks, Reactor Trip, P-4) of TS 3.3.2, "Engineered Safety Feature Actuation System Instrumentation."

Wolf Creek Nuclear Operating Corporation (WCNOC) is proposing to add footnote (m) to Function 8.a. to identify the enabled functions and the applicable MODES for the Reactor Trip, P-4 interlock function.

2.0 DETAILED DESCRIPTION Proposed changes to the TSs are as follows:

New Footnote (m) is added to Function 8.a. to identify the enabled functions and the applicable MODES for that function. The new footnote states:

(m)

The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolates MFW with coincident low Tavg - MODES 1 and 2 Allows manual block of the automatic reactuation of SI after a manual reset of SI - MODES 1, 2, and 3 Prevent opening of the MFIVs if closed on SI or SG Water Level - High High

- MODES 1, 2, and 3 WCNOC is proposing changes to the TSs to prevent unnecessary cycling of the main feedwater isolation valves (MFIVs) and the Auxiliary Feedwater (AFW) System and unnecessary trips of the main turbine during turbine warmup that adversely impacts startups and shutdown evolutions.

In MODE 3, surveillance testing activities such as rod drop testing and cross calibration of the wide and narrow range resistance temperature detectors utilize the closing and opening of the reactor trip breakers. Additionally, the proposed change will preclude confusion for compliance with TS Table 3.3.2-1, Function 8.a (ESFAS Interlock; Reactor Trip, P-4) as discussed below.

On August 19, 2009, WCGS was operating at approximately 100% power. At approximately 1549 hours0.0179 days <br />0.43 hours <br />0.00256 weeks <br />5.893945e-4 months <br />, WCGS experienced an unplanned automatic shutdown. The turbine and reactor trip occurred due to the momentary loss of offsite power (LOOP) to the onsite buses. On August 22, 2009 at 0540 hours0.00625 days <br />0.15 hours <br />8.928571e-4 weeks <br />2.0547e-4 months <br /> the Control Room defeated the feedwater isolation on low Tavg coincident with P-4 function using procedure SYS SB-122, "Enabling/Disabling P-4/Lo Tavg FWIS." This procedure was performed to restore main feedwater flow through the MFIVs to supply water to the steam generators during troubleshooting of the rod drive motor generator sets.

Attachment I to ET 10-0014 Page 3 of 12 TS 3.3.2, Table 3.3.2-1 specifies the applicable MODE for Function 8.a. (ESFAS Interlocks, Reactor Trip, P-4) as MODES 1, 2, 3. The TS Bases identifies the functions of the P-4 interlock as:

Trips the main turbine; Isolates MFW with coincident low Tay; Allows manual block of the automatic reactuation of SI after a manual reset of SI; and

" Allows arming of the steam dump valves and transfers the steam dump from the load rejection Tvg controller to the plant trip controller; and Prevents opening of the MFW isolation valves if they were closed on SI or SG Water Level - High High.

Subsequently, on August 25, 2009, the Nuclear Regulatory Commission (NRC) Senior Resident Inspector questioned the defeating of the feedwater isolation on low Tavg coincident with P-4 function while in MODE 3. On November 10, 2009, Nuclear Regulatory Commission (NRC)

Integrated Inspection Report 2009004 identified a Green noncited violation of Limiting Condition for Operation (LCO) 3.0.3 in which both trains of TS 3.3.2, "Engineered Safety Feature Actuations System (ESFAS) Instrumentation," Table 3.3.2-1, Function 8.a (ESFAS Interlocks, Reactor Trip, P-4) were defeated in accordance with a plant procedure.

In 1994, in an effort to expedite startup physics testing it was desired to perform rod drop testing by dropping an entire control bank vice individual rods. It was decided the dropping of an entire control bank could best be accomplished by opening the reactor trip beakers. However opening the reactor trip breakers leads to the generation of a feedwater isolation signal when below the Tavg setpoint associated with the P-4 interlock. When the reactor trip breakers are opened for this test in MODE 3, a feedwater isolation is not required to control RCS temperature, and results in unnecessary cycling of the MFIVs and the AFW System which adversely impacts startup and shutdown evolutions.

In order to suppress the feedwater isolation signal upon opening the reactor trip breakers, Reactor Engineering personnel proposed defeating the low Tag plus P-4 input to the feedwater isolation circuitry. In support of this request, an evaluation of the Updated Safety Analysis Report (USAR) Chapter 15 accident analyses was performed to ensure safe operation of the plant would be maintained. The evaluation assumed that only the feedwater isolation function would be defeated and that all other reactor trip P-4 signals would remain active and function normally. The evaluation concluded that there was no impact to the current USAR analyses or reduction in the analyzed margin of safety. This resulted in WCNOC considering the defeating of the feedwater isolation on low Tavg coincident with P-4 in MODE 3 to be acceptable.

Additionally, defeating turbine trip on reactor trip function of P-4 was not considered to be required for P-4 interlock OPERABILITY per TS Table 3.3.2-1. On January 26, 2010, a review of a revision to procedure SYS AC-120, "Main Turbine Generator Startup," identified that Step 5.6 allows leads to be lifted that defeated Reactor Trip, P-4 interlock for the turbine trip function when the plant is in MODE 3.

In November 2002, procedure SYS AC-120 was revised and included changes that allowed leads to be lifted at the turbine control panel that defeated the Reactor Trip, P-4 interlock for the turbine trip function when the plant is in MODE 3.

The change was made based on an evaluation of an outage critique item to allow warming of the main turbine during various plant activities that involve opening of the reactor trip breakers. The procedure change also required that the trip function be reinstated prior to entry into MODE 2.

Attachment I to ET 10-0014 Page 4 of 12 This resulted in WCNOC considering the defeating of the turbine trip on reactor trip function in MODE 3 to be acceptable.

3.0 TECHNICAL EVALUATION

3.1

System Description

The Solid State Protection System (SSPS) initiates the proper unit shutdown or engineered safety feature (ESF) actuation in accordance with the defined logic and based on the bistable outputs from the signal process control and protection system.

The SSPS equipment is used for the decision logic processing of outputs from the signal processing equipment bistables.

To meet the redundancy requirements, two trains of SSPS, each performing the same functions, are provided.

If one train is taken out of service for maintenance or test purposes, the second train will provide ESF actuation for the unit. If both trains are taken out of service or placed in test, a reactor trip will result.

The SSPS performs the decision logic for most ESF equipment actuation; generates the electrical output signals that initiate the required actuation; and provides the status, permissive, and annunciator output signals to the control room. The bistable outputs from the signal processing equipment are sensed by the SSPS equipment and combined into logic matrices that represent combinations indicative of various transients.

If a required logic matrix combination is completed, the system will send actuation signals via master and slave relays to those components whose aggregate function best serves to alleviate the condition and restore the unit to a safe condition.

The Reactor Trip, P-4 interlock functions are developed on a per train basis by reactor trip breaker cell switches and auxiliary contacts, which sense reactor trip and bypass breaker position, and the corresponding logic circuits in each SSPS train.

3.2 Justification for Change Provided below is an evaluation of the functions of Reactor Trip, P-4 interlock that are impacted by the proposed change to the TSs.

Trips the main turbine As discussed in USAR Section 7.2.1.1.1, the reactor trip system initiates a turbine trip signal whenever reactor trip is initiated to prevent the reactivity insertion that would otherwise result from excessive reactor system cooldown.

This eliminates unnecessary actuation of the engineered safety feature actuation system.

The main turbine is typically placed into normal operation at power levels around 15%. The turbine trip on reactor trip function provided by the P-4 interlock serves to limit the potential for an excessive cooldown of the reactor coolant system.

Following the reactor trip signal, the turbine is tripped by promptly stopping steam flow to the turbine. Should the operating turbine fail to trip after a reactor trip, continuous steam flow from the steam generators removes additional energy from the RCS. This results in a reduction of primary coolant temperature and pressure.

In the presence of a negative moderator temperature coefficient, the continuous

Attachment I to ET 10-0014 Page 5 of 12 cooldown results in an insertion of positive reactivity. If the most reactive rod control cluster assembly is assumed stuck in its fully withdrawn position after a reactor trip, there is possibility that the core will become critical and return to power. However, the core would be ultimately shut down by the boric acid solution delivered by the emergency core cooling system when a safety injection signal associated with ESFAS is actuated upon receipt of a low pressurizer pressure signal.

None of the USAR Chapter 15 accident analyses credit turbine trip from reactor trip (P-4 interlock function) for accident mitigation.

In MODE 3, the reactor is producing decay heat. A reactor trip from MODE 3 does not change the decay heat production. Opening the reactor trip breakers in MODE 3 initiates a turbine trip on reactor trip (a function of the P-4 interlock), which results in closure of the steam admission valves necessary for warming of the main turbine. Additionally, in MODE 3 the main turbine is not in operation.

The TS 3.3.2 Bases state, in part, that the ESFAS Instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). 10 CFR 50.36(c)(2)(ii) requires that (c) technical specifications will include items in the following categories:

(2) limiting conditions for operation: (ii) a technical specification limiting condition for operation must be established for each item meeting one or more of the following criterion: Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Defeating the turbine trip on reactor trip function of the P-4 interlock in MODE 3 will have no impact on any accidents previously evaluated in the USAR since the signal to be blocked is not credited in the USAR Chapter 15 accident analyses.

Isolates MFW with coincident low T.w The function of the Condensate and Feedwater System (Main Feedwater System) is to supply a sufficient quantity of feedwater to the steam generator secondary side inlet during normal operating conditions and to guarantee that feedwater will not be delivered to the steam generators when feedwater isolation is required. The condensate pumps take suction from the condenser hotwell and the main feedwater (MFW) pumps deliver water to the steam generators at elevated temperatures and pressures. Additional information can be found in Updated Safety Analysis Report (USAR) Section 10.4.7, "Condensate and Feedwater System."

The Auxiliary Feedwater (AFW) System automatically supplies feedwater to the steam generators to remove decay heat from the Reactor Coolant System upon the loss of normal feedwater supply. The motor driven AFW pumps start automatically on steam generator water level - low-low in any steam generator, on trip of both MFW pumps, upon actuation of ATWS Mitigation System Actuation Circuitry (AMSAC), and on actuation by the LOCA or shutdown sequencer. The turbine driven AFW pump is automatically started by steam generator water level -

low-low in any two steam generators, 4.16 kv safety related bus NB01 or NB02 undervoltage, and upon actuation of AMSAC.

All three AFW trains can also be manually actuated. Additional information can be found in USAR Section 10.4.9, "Auxiliary Feedwater System." Auxiliary feedwater does not provide preheating to the feedwater.

Using colder auxiliary feedwater increases steam generator shrink and swell.

Main feedwater can be provided to the steam generators using the Main Feedwater System or the AFW System. Feedwater preheating requires the Main Feedwater System in operation.

Feedwater preheating is used to minimize thermal stresses on the feedwater piping and steam

Attachment I to ET 10-0014 Page 6 of 12 generator feedwater nozzles. At low pressures the condensate pumps can provide sufficient pressure to provide flow to the generators. Above condensate pressure a feedwater pump must also be used. A motor driven startup feedwater pump is normally used to provide feedwater at low power (-1.5%) until there is adequate steam flow to operate the MFW pumps. The capacity of the startup feedwater pump is 210 klbm/h at 300-400 °F, which is a small fraction of the heat removal capacity of AFW flow (600 klbm/h) that could be as low as 35 OF.

The resultant cooldown for an overfeed event is greatly reduced by using main feedwater as opposed to AFW.

With the P-4 interlock enabled (reactor trip and bypass breakers open), feedwater isolation will occur if Tavg is < 564 OF.

The feedwater isolation on low Tavg coincident with P-4 provides a control function (not in the primary success path) against excessive cooldown events while the main feedwater system is operating.

For reactor trips in MODES 1 and 2, the feedwater isolation on low Tavg coincident with P-4 offsets the sudden decrease in reactor heat production by isolating main feedwater, which provides a large reduction in heat sink to prevent an excessive RCS cooldown. When reactor trip breakers are opened in MODE 3, there is no change in heat production from the reactor, and thus no need to isolate main feedwater. Additionally, in MODE 3 the MFW pumps are not in service so the driving force for feedwater is limited to the motor driven startup feedwater pump or motor driven AFW pumps.

In order to maintain SG water levels after feedwater isolation, the auxiliary feedwater (AFW) pumps may be needed until the feedwater isolation signal can be reset and the MFIVs can be reopened. If feedwater flow is not isolated while the RCS is cooling down, Tavg will undershoot the target value of no-load temperature. This undershoot could result subsequently in safety injection actuation on low RCS pressure. Consequently, the nuclear steam supply system was designed with an anticipatory feedwater isolation on low RCS Tavg coincident with reactor trip.

Feedwater isolation via this function (isolation on low Tavg coincident with PA4) is not modeled in any USAR Chapter 15 analyses.

In order to satisfy the licensing basis accident analyses, feedwater isolation capability must be provided whenever the Main Feedwater System is in service and automatic valve closure must be provided after initiation signals from safety injection and steam generator water level high-high. These events are analyzed with the plant at hot zero power, full power or part power conditions. Feedwater isolation would be actuated by a safety injection signal for the large and small break LOCA and steamline break accidents.

For the analysis of the Excessive Feedwater Flow in USAR Section 15.1.2, continuous addition of excessive feedwater is prevented by the steam generator high-high level trip, which initiates feedwater isolation and trips the turbine and MFW pumps. Defeating the feedwater isolation signal on low Tavg coincident with P-4 will have no impact on any accidents previously evaluated in the USAR since the signal to be blocked has not been credited.

The TS 3.3.2 Bases state, in part, that the ESFAS Instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). 10 CFR 50.36(c)(2)(ii) requires that (c) technical specifications will include items in the following categories:

(2) limiting conditions for operation: (ii) a technical specification limiting condition for operation must be established for each item meeting~one or more of the following criterion: Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission

Attachment I to ET 10-0014 Page 7 of 12 product barrier. Defeating the feedwater isolation signal low Taw coincident with P-4 in MODE 3 will have no impact on any accidents previously evaluated in the USAR since the signal to be blocked has not been credited.

Allows arming of the steam dump valves and transfers the steam dump from the load rejection T

controller to the plant trip controller USAR Section 7.7 identifies the Steam Dump System as a plant control system not required for safety. The general design objective of the plant control systems are:

1)

To establish and maintain power equilibrium between the primary and secondary system during steady state operation.

2)

To constrain operational transients so as to preclude unit trip and reestablish steady state unit operation.

3)

To provide the reactor operator with monitoring instrumentation that indicates all required input and output control parameters of the systems and provides the operator with the capability of assuming manual control of the system.

The Steam Dump System, together with control rod movement, is designed to accept a 50 %

loss of net load without tripping the reactor. The automatic Steam Dump System is able to accommodate this abnormal load rejection and to reduce the effects of the transient imposed upon the RCS. By bypassing main steam directly to the condenser, an artificial load is thereby maintained on the primary system.

Following a reactor trip, the load rejection steam dump controller is defeated by the Reactor Trip, P-4 interlock, and the plant trip steam dump controller becomes active. The steam dump valves are modulated by the plant trip controller to regulate the rate of removal of decay heat and thus gradually establish the equilibrium hot shutdown conditions.

The TS 3.3.2 Bases state, in part, that the ESFAS Instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). 10 CFR 50.36(c)(2)(ii) requires that (c) technical specifications will include items in the following categories:

(2) limiting conditions for operation: (ii) a technical specification limiting condition for operation must be established for each item meeting one or more of the following criterion: Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The instrumentation utilized to initiate transfer to the plant trip steam dump controller does not serve a primary protective function so as to warrant inclusion in the TS. The instrumentation does not serve to ensure that the plant is operated within the bounds of initial conditions assumed in design basis accident and transient analyses. Likewise, the transfer to the plant trip steam dump controller instrumentation does not serve as part of the primary success path of a safety sequence analysis used to demonstrate that the consequence of these events are within the appropriate acceptance criteria.

The specific instrumentation requirements related to transfer to the plant trip steam dump controller, are not required to be in the TS per the criteria in 10 CFR 50.36 and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. As such, this function of the P-4 interlock is excluded from the proposed footnote (m) to TS Table 3.3.2-1, Function 8.a.

Attachment I to ET 10-0014 Page 8 of 12 Accident Analysis Review In MODE 3, the reactor is subcritical and temperature greater than 350 OF. The only USAR Chapter 15 analyses that have initial subcritical conditions (MODE 3) are Section 15.4.1, Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low Power Startup Condition (Rod Withdrawal from Subcritical) and 15.4.6, Chemical and Volume Control System Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant (Boron Dilution). The Rod Withdrawal from Subcritical event relies totally on core reactivity parameters for termination of the event; thus modeling the secondary system is unnecessary.

The Boron Dilution event in MODE 3 relies on operators to detect and recover from an inadvertent boron dilution.

To be conservative, the impact of defeating these P-4 interlock functions on the safety analysis was assessed based on the USAR Chapter 15 analyses that are analyzed at hot zero power. The safety analyses that are applicable at these conditions are:

15.1.1 Feedwater System Malfunctions That Result in a Decrease in Feedwater Temperature 15.1.2 Feedwater System Malfunctions That Result in a Increase in Feedwater Flow 15.1.4 Inadvertent Opening of a Steam Generator Atmospheric Relief or Safety Valve 15.1.5 Steam System Piping Failure 15.4.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low Power Startup Condition 15.4.8 Spectrum of Rod Cluster Control Assembly Ejection Accidents Accident analyses not listed here have been shown to be limiting assuming the reactor is critical (i.e., power level is greater than zero percent) or because the temperature throughout the duration of the event remains above 564 OF. Therefore, only the analyses listed above may be affected by defeating the feedwater isolation on low Tavg and turbine trip on reactor trip function.

Defeating the feedwater isolation on low Tavg and the turbine trip on reactor trip function of the P-4 interlock in MODE 3 will have no impact on any accidents previously evaluated in the USAR since the signal to be blocked is not credited in these USAR Chapter 15 accident analyses.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commission's regulatory requirements related to the content of the TSs are contained in Title 10, Code of Federal Regulations (10 CFR), Part 50, Section 50.36, "Technical Specifications." The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety systems settings and control settings, (2) limiting conditions for operation, (3) surveillance requirements (SRs), (4) design features, and (5) administrative controls.

Attachment I to ET 10-0014 Page 9 of 12 0

10 CFR 50.36(c)(2) Limiting condition for operation. (ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4.

A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

0 Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, General Design Criteria (GDC) -13, Instrumentation and control, requires that instrumentation be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

• Appendix A to 10 CFR Part 50, GDC-20, Protection system functions, requires that the protection system be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

• Appendix A to 10 CFR Part 50, GDC-21, Protection system reliability and testing, requires that the protection system be designed and tested for high functional reliability.

Appendix A to 10 CFR Part 50, GDC-22 through 25 and GDC-29 require various design attributes for the protection system, including independence, safe failure modes, separation from control systems, requirements for reactivity control malfunctions, and protection against anticipated operational occurrences.

The proposed changes to include the enabled functions of the Reactor Trip, P-4 interlock in TS Table 3.3.2-1 is consistent with the WCGS design and analysis and ensures proper actuation to satisfy the anticipatory trip function.

Therefore, requirements and the recommendations of these regulations and guidance continue to be met with the proposed change.

Attachment I to ET 10-0014 Page 10 of 12 4.2 Precedent, License Amendment No. 126, dated April 23, 1998, approved changes to the Callaway Plant Technical Specification Bases that allows blocking the feedwater isolation on low Tavg coincident with P-4 function in MODE 3. The Safety Evaluation for the license amendment states:

The Bases for Functional Unit 11 b, Reactor Trip P-4, in Table 3 3-3 would be revised to add a note allowing the feedwater isolation function on P-4 (reactor trip and bypass breakers open) coincident with low Tavg (Tayg < 564 OF) to be blocked. The reason for the change is to decrease unnecessary cycling of the MFIVs and AFW system which adversely impacts startup and shutdown evolutions. This feedwater isolation function provides backup protection for excessive cooldown events and is not credited in any FSAR analyses. The licensee has proposed to install a bypass switch to block this signal during startup and shutdown evolutions with Tavg - 564 OF just prior to opening the reactor trip breakers. The feedwater isolation function would be restored by manually defeating the bypass prior to entering MODE 2. This change is acceptable.

This amendment approved other changes to the Callaway Plant Technical Specifications that are not applicable to the changes being proposed by WCNOC. The difference between the Callaway-Plant amendment and the changes proposed by WCNOC included changing TS Table 3.3.2-1, Function 8.a. to specify that the feedwater isolation on low Tvg coincident with P-4 function is not required in MODE 3. Additionally, WCNOC is proposing additional changes to address the additional enabled function of the P-4 interlock.

4.3 Significant Hazards Consideration The amendment application proposes changes to the Wolf Creek Generating Station (WCGS)

Technical Specification (TS) Table 3.3.2-1, Function 8.a., (ESFAS Interlocks, Reactor Trip, P-4) of TS 3.3.2, "Engineered Safety Feature Actuation System Instrumentation."

Wolf Creek Nuclear Operating Corporation (WCNOC) is proposing to add footnote (m) to Function 8.a. to identify the enabled functions and the applicable MODES for that function.

WCNOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment:

1.

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

Response: No Overall protection system performance will remain within the bounds of the previously performed accident analyses. Defeating the feedwater isolation low Tavg coincident with P-4 function will not impact any accidents previously evaluated in the Updated Safety Analysis Report (USAR) since feedwater isolation on low Tavg coincident with P-4 is not credited. Bypassing the turbine trip on reactor trip function will not impact any accidents previously evaluated in the USAR since the turbine trip on reactor trip function of P-4 is not credited.

Attachment I to ET 10-0014 Page 11 of 12 The instrumentation utilized to initiate transfer to the plant trip steam dump controller does not serve a primary protective function so as to warrant inclusion in the TS. The instrumentation does not serve to ensure that the plant is operated within the bounds of initial conditions assumed in design basis accident and transient analyses. Likewise, the transfer to the plant trip steam dump controller instrumentation does not serve as part of the primary success path of a safety sequence analysis used to demonstrate that the consequence of these events are within the appropriate acceptance criteria.

The ESFAS will continue to function in a manner consistent with the accident analysis assumptions and the plant design basis. As such, there will be no degradation in the performance of, nor an increase in, the number of challenges to equipment assumed to function during an accident situation. The proposed changes to the TSs do not affect the probability of any event initiators. There will be no change to normal plant operating parameters or accident mitigation capabilities.

Therefore, this change will not increase 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 previously evaluated?

Response: No There are no changes in the method by which any safety related plant system performs its safety function and the normal manner of plant operation is unaffected, other than the proposed allowance to defeat feedwater isolation on low Tavg coincident with P-4 and the proposed allowance to defeat the turbine trip on reactor trip function of P-4.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change. There will be no adverse effect or challenges imposed on any safety related system as a result of this change. Therefore, the possibility of a new or different type of accident is not created. The proposed change does not affect the steam generator high-high level trip ESFAS function which initiates feedwater isolation and trips the turbine and main feedwater pumps.

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

3.

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

Response: No There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions.

There will be no impact on departure from nucleate boiling ratio (DNBR) limits, heat flux hot channel factor (FQ(Z))

limits, nuclear enthalpy rise hot channel factor (FAH) limits, peak centerline temperature (PCT) limits, peak local power density or any other margin of safety.

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

Attachment I to ET 10-0014 Page 12 of 12 4.4 Conclusion Based on the considerations discussed above, 1) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, 2) 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.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment II to ET 10-0014 Page 1 of 2 ATTACHMENT II PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

Attachment II to ET 10-0014 Page 2 of 2 ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 5 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE(a)

7.

Automatic Switchover to Containment Sump

a.

Automatic Actuation 1.2,3,4 2 trains C

SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays SR 3.3.2.6

b.

Refueling Water 1.2,3,4 4

K SR 3.3.2.1 a 35.5% of Storage Tank (RWST)

SR 3.3.2.5 instrument span Level - Low Low SR 3.3.2.9 SR 3.3.2.10 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection

8.

ESFAS Interlocks

a.

Reactor Trip, P-1 1,2,3 2 per train, F

SR 3.3.2.11 NA 2 trains

b.

Pressurizer Pressure, 1,2,3 3

L SR 3.3.2.5

< 1979 psig P-11 SR 3.3.2.9 (a)

The Allowable Value defines the UmItIng Safety System Settings. See the Bases for the Trip Setpoints.

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Wolf Creek - Unit 1 3.3-35 Amendment No. 123, 126.2, 32, 183

Attachment III to ET 10-0014 Page 1 of 2 ATTACHMENT III REVISED TECHNICAL SPECIFICATION PAGES

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 5 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE(a)

7.

Automatic Switchover to Containment Sump

a.

Automatic Actuation 1,2,3,4 2 trains C

SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays SR 3.3.2.6

b.

Refueling Water 1,2,3.4 4

K SR 3.3.2.1 a 35.5% of Storage Tank (RWST)

SR 3.3.2.5 instrument span Level - Low Low SR 3.3.2.9 SR 3.3.2.10 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection

8.

ESFAS Interlocks

a.

Reactor Trip, P-4(

1,2,3 2 par train, F

SR 3.3.2.11 NA 2 trains

b.

Pressurizer Pressure, 1,2,3 3

L SR 3.3.2.5

< 1979 psig P-11 SR 3.3.2.9 (a)

The Allowable Value defines the Limiting Safety System Settings. See the Bases for the Trip Setpoints.

(in)

The functions of the Reactor Trip, P-4 Interlock required to meet the LCO are:

" Trip the main turbine - MODES 1 and 2 Isolates MFW with coincident low T., - MODES I and 2 Allows manual block of the automatic reactuation of SI after a manual rest of SI - MODES 1, 2, and 3 Prevent opening of MFIVs if closed on SI or SG Water Level-High High-MODES 1, 2, and 3 I

Wolf Creek - Unit 1 3.3-35 Amendment No. 123, 126, 4132,183

Attachment IV to ET 10-0014 Page 1 of 5 ATTACHMENT IV PROPOSED TS BASES CHANGES (for information only)

Attachment IV to ET 10-0014 Page 2 of 5 ESFAS Instrumentation B 3.3.2 BASES APPLICABLE

b.

Automatic Switchover to Containment Sump - Refueling SAFETY ANALYSES, Water Storaae Tank (RWST) Level - Low Low Coincident LCO, and With Safety Injection (continued)

APPLICABILITY The transmitters are located in an area not affected by HELBs or post accident high radiation. Thus, they will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties. The Trip Setpoint is;> 36% of span.

Automatic switchover occurs only if the RWST low low-1 level signal is coincident with SI. This prevents accidental switchover during normal operation. Accidental switchover could damage ECCS pumps if they are attempting to take suction from an empty sump. This is one of the few functions that requires the bistable output to energize to perform its required action. The automatic switchover Function requires the SI Function for OPERABILITY.

Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating Functions and requirements.

These Functions must be OPERABLE in MODES 1, 2, 3, and 4 when there is a potential for a LOCA to occur, to ensure a continued supply of water for the ECCS pumps. These Functions are not required to be OPERABLE in MODES 5 and 6 because there is adequate time for the operator to evaluate unit conditions and respond by manually starting systems, pumps, and other equipment to mitigate the consequences of an abnormal condition or accident. System pressure and temperature are very low and many ESF components are administratively locked out or otherwise prevented from actuating to prevent inadvertent overpressurization of unit systems.

8.

Engineered Safety Feature Actuation System Interlocks To allow some flexibility in unit operations, several interlocks are included as part of the ESFAS. These interlocks permit the operator to block some signals, automatically enable other signals, prevent some actions from occurring, and cause other actions to occur. The interlock Functions back up manual actions to ensure bypassable functions are in operation under the conditions assumed in the safety analyses.

Wolf Creek - Unit 1 B 3.3.2-30 Revision 0

-~ - r Attachment IV to ET 10-0014 Page 3 of 5 ESFAS Instrumentation B 3.3.2 BASES APPLICABLE SAFETY ANALYSES LCO, and APPLICABILITY (continued)

a.

Engineered Safety Feature Actuation System Interlocks - Reactor Trip, P-4 The P-4 interlock is enabled when a reactor trip breaker (RTB) and its associated bypass breaker is open. Manual reset of SI following a 60 second time delay, in conjunction with P-4, generates an automatic SI block. This Function allows operators to take manual control of SI systems after the initial phase of injection is complete. Once SI is blocked, automatic actuation of SI cannot occur until the RTBs have been manually closed.

T*he fun*s of the P-4dock a re: **

STrips them i turb ine; J

o Isola MFW with coine*nt low T..g; S

Ilows manual bi of the automatic rea ation of S1 after a ma I reset of SI; and S

Allows ing of the steam du valves and tran rs the steam dump the load rejection T,,controller to the pl trip controller, and Prevents openin the MFW isolation ves if theywerecs onSIor SG Water vel - High High.

• I,

'9--w M Each of the above/unctions is interlocked with P-4 to

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2. g_.*gi

,-e..uI4

• avert or reduce the contnued cooldown of the RCS following a reactor trip.-.ý excessive cooldown of the RCS 64j*6+

tcould cause an insertion of positive reactivity with a subsequent incregise in core power. To avoid such a situation, the notedfunctions have been interlocked with P-4 as part of the design of the unit control and protection system.

I TSERT 1 3.3a -316 None of noted Fy ions serve mitigation fution in the licensin sis safety a yses. Only t iturbine

t.

uncio plicitly ass ed since it I immediate consequ of the rea r trip Functio. Neither tu ine Wolf Creek - Unit 1 B 3.3.2-31 Revision 0

Attachment IV to ET 10-0014 Page 4 of 5 INSERT B 3.3.2-31a The functions of the P-4 interlock are:

Function Required MODE Trips the main turbine 1, 2 Isolates MFW with 1,2 coincident low Tavg o Allows manual block of 1,2, 3 the automatic reactuation of SI after a manual reset of SI; and Allows arming of the None steam dump valves and transfers the steam dump from the load rejection Tavg controller to the plant trip controller; and o

Prevents opening of the 1,2,3 MFW isolation valves if they were closed on SI or SG Water Level -

High High.

INSERT B 3.3.2-31b The turbine trip function, MFW isolation coincident with low Tavg function, arming of the steam dump valves function do not serve a mitigation function in the licensing basis safety analyses.

Block of the SI signals is required to support long-term ECCS operation in the post-LOCA recirculation mode. Block of the opening of the MFW isolation valves on SI or SG Water Level -

High High prevents reopening the valves for mitigation of a high water level in the SGs, which could result in carryover of water into the steam lines and excessive cooldown of the primary system.

Attachment IV to ET 10-0014 Page 5 of 5 ESFAS Instrumentation B 3.3.2 BASES APPLICABLE

a.

Engineered Safety Feature Actuation System SAFETY ANALYSES, Interlocks - Reactor Trip, P- (continued)

LCO, and APPLICABILITY t pnorui of th therf r Fun *ns ass iated wit he) actrip sied, is iredt how th the unit

)

Iising is sa tanal

• accept ce crite *ar The RTB position switches that provide input to the P4 interlock (P4 generated when one train's RTB and the alternate train's Bypass Breaker are both open) only function to energize or de-energize or open or close contacts. Therefore, this Function has no adjustable trip setpoint with which to associate a Trip Setpoint and Allowable Value.

This Function does not have o e 0 A in E 4, 5, or 6 because the main turbine, the MFW System, and the Steam Dump System are not in operation.

b.

Engineered Safety Feature Actuation System Interlocks -

Pressurizer Pressure. P-11 The P-11 interlock permits a normal unit cooldown and depressurization without actuation of SI or main steam line isolation. With two-out-of-three pressurizer pressure channels (discussed previously) less than the P-11 setpoint, the operator can manually block the Pressurizer Pressure - Low and Steam Line Pressure - Low SI signals and the Steam Line Pressure - Low steam line isolation signal (previously discussed). When the Steam Line Pressure - Low steam line isolation signal is manually blocked, a main steam isolation signal on Steam Line Pressure - Negative Rate - High is automatically enabled.

This provides protection for an SLB by closure of the MSIVs. With two-out-of-three pressurizer pressure channels above the P-1I setpoint, the Pressurizer Pressure - Low and Steam Line Pressure - Low SI signals and the Steam Line Pressure - Low steam line isolation signal are automatically enabled. The operator can also Wolf Creek - Unit I B 3.3.2-32 Revision 0

Attachment V to ET 10-0014 Page 1 of 1 ATTACHMENT V REGULATORY COMMITMENTS The following table identifies those actions committed to by WCNOC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr.

Richard Flannigan at (620) 364-4117.

Regulatory Commitments Due Date / Event The proposed changes to the WCGS Technical Specifications will Within 90 days of NRC be implemented within 90 days of NRC approval, approval.