ET 23-0003, License Amendment Request (LAR) for Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications

License Amendment Request (LAR) for Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications

ML23060A481
Person / Time
Site:	Wolf Creek
Issue date:	03/01/2023
From:	Boyce M Wolf Creek
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
ET 23-0003
Download: ML23060A481 (1)

Text

P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 Michael T. Boyce Vice President Engineering ET 23-0003 March 1, 2023 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: License Amendment Request (LAR) for Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications Commissioners and Staff:

Pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting a request for an amendment to the Technical Specifications (TS) for the Wolf Creek Generating Station (WCGS).

This license amendment request, if approved by the Commission, would revise the Technical Specifications by removing the Power Range Neutron Flux Rate - High Negative Rate Trip function. The contents of this amendment request is consistent with Westinghouse WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event, that was approved by the Nuclear Regulatory Commission (NRC) on October 23, 1989.

Attachment I provides the evaluation and justification for the proposed license amendment.

Attachment II provides the markup of the Technical Specification pages relevant to the change requested. Attachment III provides the markup of the Updated Safety Analysis Report (USAR).

Attachment IV provides the markup of the Renewed Facility Operating License. Attachment V provides the markup of the Technical Specification Bases for the proposed change for information only.

It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92, Issuance of amendment. The amendment application was reviewed by the WCNOC Plant Safety Review Committee. In accordance with 10 CFR 50.91, Notice for public comment; State consultation, a copy of this application is being provided to the designated Kansas State official.

WCNOC requests approval of this proposed amendment, following acceptance, by February 2024. Once approved, the amendment will be implemented in Refueling Outage 26, scheduled for Spring 2024.

W o If Cree k'-M~'~J"TJ>-

Nuclear Operating Corporation

ET 23-0003 Page 2 of 3 This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-8831 x8687, or Dustin Hamman at (620) 364-4204.

Sincerely, Michael T. Boyce MTB/jkt Attachment I:

Evaluation of Proposed Change Attachment II: Proposed Technical Specification Changes (Mark-up)

Attachment Ill: Proposed USAR Changes (Mark-up)

Attachment IV: Markup of the Renewed Facility Operating License Attachment V: Proppsed Technical Specification Bases Changes (Mark-up) for Information Only cc:

S. S. Lee (NRG), w/a, w/e R. J. Lewis, (NRG), w/a, w/e G. E. Werner (NRG), w/a, w/e

• Senior Resident Inspector (NRG), w/a, w/e J. Meinholdt (KDHE), w/a, w/e

ET 23-0003 Page 3 of 3 STATE OF KANSAS

) ) ss COUNTY OF COFFEY )

Michael T. Boyce, 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.

By

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MichaelT.Boyc Vice President Engineering

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Attachment I to ET 23-0003 Page 1 of 8 EVALUATION OF PROPOSED CHANGE

Subject:

License Amendment Request (LAR) for Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip function from Technical Specifications

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1. System Design and Operation 2.2. Current Technical Specifications Requirements 2.3. Reason for Proposed Change 2.4. Description of Proposed Change

3. TECHNICAL EVALUATION 3.1. Application of WCAP-11394-P-A 3.2. WCGS Cycle Specific Analysis

4. REGULATORY EVALUATION 4.1. Applicable Regulatory Requirements/Criteria 4.2. Precedent 4.3. No Significant Hazards Consideration 4.4. Conclusion

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

Attachment I to ET 23-0003 Page 2 of 8

1.

SUMMARY

DESCRIPTION The proposed license amendment would delete the Technical Specification (TS) requirement for the Power Range Neutron Flux Rate - High Negative Rate Trip function. This function, as specified in Table 3.3.1-1, Reactor Trip System Instrumentation, of the TS as Function 3.b, Power Range Neutron Flux Rate - High Negative Rate, would be deleted. The proposed change for the removal of the Power Range Neutron Flux Rate - High Negative Rate Trip function is aligned with the Nuclear Regulatory Commission (NRC) approved methodology depicted in the Westinghouse Topical Report WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event, dated October 23, 1989.

2.

DETAILED DESCRIPTION 2.1 System Design and Operation The Reactor Trip System (RTS) automatically keeps the reactor operating within a safe region by shutting down the reactor whenever the limits of the region are approached. Safe operating regions are defined by several considerations, such as mechanical/hydraulic limitations on equipment and heat transfer phenomena. These include mechanical limitations such as pressure and pressurizer water level, and also variables that directly affect the heat transfer capability of the reactor such as flow and reactor coolant temperatures. There are additional parameters that are calculated as a part of the RTS that, when established setpoints are reached, the reactor is shut down to protect against either damage to fuel cladding or the loss of system integrity; which could lead to the release of radioactive fission products into the containment. Additionally, the RTS assists the Engineered Safety Features (ESF) systems in mitigating accidents and aids in mitigation of Anticipated Operational Occurrences (AOOs).

The RTS consists of sensors that monitor the various plant parameters and are connected with analog circuitry which consists of two to four redundant channels, and digital circuitry, consisting of two redundant logic trains that receive inputs from the analog channels to complete the logic necessary to automatically open the reactor trip breakers. For the Power Range Neutron Flux Rate - High Negative Rate Trip function, the circuit trips the reactor when a sudden abnormal decrease in nuclear power occurs in two out of the four power range channels.

2.2 Current Technical Specifications Requirements The setpoints are established in Table 3.3.1-1 of Technical Specification 3.3.1, Reactor Trip System Instrumentation, and are directly monitored by the RTS. The Limiting Condition of Operation (LCO) for TS 3.3.1 require that the RTS instrumentation for each Function of Table 3.3.1-1 be OPERABLE in the illustrated MODE of applicability. The LCO for the Power Range Neutron Flux Rate - High Negative Rate Trip function requires all four channels to be operable in MODES 1 or 2, when there is the potential for a multiple rod drop accident to occur. MODES 3, 4, 5 and 6 do not require the Power Range Neutron Flux Rate - High Negative Rate Trip function channels to be operable as the reactor core is not critical and Departure from Nucleate Boiling (DNB) is not a concern. The Power Range Neutron Flux Rate - High Negative Rate Trip function is described as function 3.b of Table 3.3.1-1.

2.3 Reason for the Proposed Change With the documented conclusions contained within WCAP-11394-P-A and the Wolf Creek Generating Station (WCGS) Cycle specific analysis supported in Revision 2 of WCAP-17658-NP,

Attachment I to ET 23-0003 Page 3 of 8 the design Departure from Nucleate Boiling Ratio (DNBR) limits continue to be met with the new Rod Cluster Control Assembly (RCCA) methodology; independent of the Power Range Neutron Flux Rate - High Negative Rate Trip function. Thus, the proposed change in this amendment request would delete an unnecessary trip function which, in turn, would avoid unnecessary reactor trip initiating events.

2.4 Description of the Proposed Change The proposed change would delete the TS requirement for the Power Range Neutron Flux Rate

- High Negative Rate Trip function. This function, as specified in Table 3.3.1-1, Reactor Trip System Instrumentation, of the TS as Function 3.b, Power Range Neutron Flux Rate - High Negative Rate, would be deleted. The specific changes to Table 3.3.1-1 (i.e., the mark-up of the proposed change) are contained in Attachment II.

Currently, TS 3.3.1 requires that the Power Range Neutron Flux Rate - High Negative Rate Trip Function have four (4) of the channels operable while in MODES 1 and 2. Following application of the proposed change, the Power Range Neutron Flux Rate - High Negative Rate Trip Function would be deleted, meaning that the trip function would not be required for operation in any MODE.

Deletion of the High Negative Rate trip function from Table 3.3.1-1 would leave the High Positive Rate trip function as the only function for the Power Range Neutron Flux Rate Instrumentation.

The proposed change would remove the a listing for the High Positive Rate and relocate the associated description of the function to make a single row. This change seeks to simplify Table 3.3.1-1 following the removal of the High Negative Rate trip function and does not alter any of the descriptions associated with the High Positive Rate Function.

Additionally, the proposed changes would revise the Updated Safety Analysis Report (USAR)

Section 7.2 (Pages 5, 41, 46, and 49) and Section 15 Table 15.0-6 (Sheet 3, Page 32). The specific changes to the USAR are detailed in Attachment III. Section 7.2 of the USAR describes the RTS as configured at the WCGS, including the system description as well as the various trip functions. Currently, Section 7.2 describes the Power Range Neutron Flux Rate - High Negative Rate Trip function in its current configuration and the response given a sudden abnormal decrease in nuclear power. Table 15.0-6 of the USAR (Sheet 3) describes the reactor trip function response for an incident of RCCA mis-alignment. A response to the incident is the RTS Power Range Negative Flux Rate trip function. Following application of the proposed change, the descriptions of the Power Range Neutron Flux Rate - High Negative Rate Trip function in the aforementioned sections would be deleted and replaced with a statement indicating the functions removal per the License Amendment. Section 7.2 (Page 5) also relocates a description contained in the High Negative Rate trip function description to the description for the High Positive Rate trip function. The statement relocated is generic and does not specifically apply to the High Negative trip function.

The proposed changes would revise Paragraph 2.C.(2) of the renewed facility operating license as a result of the new amendment received should the proposed changes be approved. The new amendment number received will also be reflected in the applicable sections of the TS for which the amendment applies. The mark-up of the renewed facility operating license accounting for this is provided in Attachment IV.

Lastly, changes to the existing TS Bases are provided in Attachment V for information only. The removal of the Power Range Neutron Flux - High Negative Rate trip function is reflected in the markups. In a similar manner to the TS, the a listing for the High Positive Rate trip function will be removed as the function is the only Power Range Neutron Flux Rate trip function. The changes

Attachment I to ET 23-0003 Page 4 of 8 associated with the TS Bases will be implemented under the TS Bases Control Program, following approval of the proposed changes.

3.

TECHNICAL EVALUATION 3.1 Application of WCAP-11394-P-A The design basis for the Power Range Neutron Flux Rate - High Negative Rate Trip function ensures that protection is provided for multiple rod drop accidents. At high power levels, a multiple rod drop accident could cause local flux peaking that would result in an unconservative local departure from nucleate boiling ratio (DNBR). In an incident where one or more dropped Rod Cluster Control Assemblies (RCCAs) would occur, the RTS would detect the rapid decrease in neutron flux and would trip the reactor, thereby terminating the transient to ensure that the DNBR limits are maintained.

Historically, Westinghouse developed and submitted WCAP-10297 Dropped Rod Methodology for Negative Flux Rate Trip Plants. The document concluded that the Power Range Neutron Flux High Negative Flux Rate Trip Function was only required if the plant exceeded a threshold value of reactivity worth that is dependent on the plant design and fuel type being utilized. WCAP-11394-P, Methodology for the Analysis of the Dropped Rod Events, was later submitted and approved by the NRC. WCAP-11394-P concluded the following:

1. Due to the plant specific nature of the core physics characteristics and Dropped Rod Limit Lines, a generic safety analysis which bounds all plants is not feasible.

2. The methodology presented in WCAP-11394-P provides a means by which to evaluate DNB for all dropped rod events without taking credit for a direct trip. No direct automatic power reduction due to the dropped rods was assumed. The methodology also applies to plants which block automatic rod withdrawal due to dropped rods or are in manual rod control.

3. For all plant specific cases considered, the DNB design basis is met.

4. Successful plant/cycle specific application of this methodology will be sufficient to confirm that the DNB design basis is met for all dropped rod events initiated from full power.

The above conclusions presented in WCAP-11394-P demonstrates that the methodology does not require the use of a direct trip or automatic power reduction as a result of the dropped rods.

Successful plant/cycle specific application of the methodology presented in WCAP 11394-P, with consideration of the conclusions presented, ensures the DNB design basis is met.

The report, formally approved by the NRC on October 23, 1989, identified one condition of acceptance. This condition states that, A further review by the staff (for each cycle) is not necessary, given the utility assertion that the analysis described by Westinghouse has been performed and the required comparisons have been made with favorable results. WCGS has performed the required comparisons with the WCAP-11394-P-A methodology applied for a reference cycle.

Attachment I to ET 23-0003 Page 5 of 8 3.2 WCGS Cycle Specific Analysis WCAP-17658-NP Revision 2 applied the WCAP-11394-P-A generic methodology to a reference Wolf Creek cycle with satisfactory results. The reference cycle assumed in the WCGS Transitions of Method (TM) Program had the methodology applied and the required comparisons made.

These comparisons validated that the cycle specific analysis, independent of the Power Range Neutron Flux Rate - High Negative Rate Trip function, remained bounding per the requirements of the NRCs Safety Evaluation Report (SER). That is, the Power Range Neutron Flux Rate - High Negative Rate Trip function was not credited in the cycle specific dropped control rod analysis, with the analysis remaining within the acceptance criteria. Future fuel cycles will be assessed as part of the Reload Safety Evaluation (RSE) process for WCGS. WCAP-17658-NP was previously reviewed by the NRC during the review and issuance of Amendment 221 for WCNOC (Reference 6.1).

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(a)(1), Technical specifications, states, in part, Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section 10 CFR 50.36(c)(2)(ii), Technical specifications, states, 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.

10 CFR 50, Appendix A, Criterion 10 - Reactor Design. The reactor core and associated coolant, control, and protections 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.

Implementation of the proposed changes at WCGS will not challenge the regulatory requirements for technical specifications, nor does it alter the description of or challenge the 10 CFR 50, Appendix A, Criterion 10 requirements. The proposed change does not affect compliance with these regulations and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met. Section 3.1.4 of the WCGS USAR contains the description for compliance with General Design Criterion 10.

Attachment I to ET 23-0003 Page 6 of 8 4.2 Precedent Union Electric Company (Ameren Missouri) submitted a license amendment for Callaway Plant, Unit 1 to remove the Power Range Neutron Flux High Negative Rate Trip from their TS (Reference 6.2). The Nuclear Regulatory Commission (NRC) approved the request in Reference 6.3. The scope of the precedent is the same as the scope for this amendment request.

Tennessee Valley Authority (TVA) submitted a license amendment for Seqouyah Nuclear Plant, Units 1 and 2 to remove the Power Range Neutron Flux High Negative Rate Trip from their TS (Reference 6.4). The NRC approved the request in Reference 6.5. The scope of the precedent is the same as the scope for this amendment request.

4.3 No Significant Hazards Consideration Wolf Creek Nuclear Operating Corporation (WCNOC) is requesting Nuclear Regulatory Commission (NRC) approval to amend the Technical Specifications (TS) at the Wolf Creek Generating Station (WCGS). The proposed license amendment would delete the TS requirement for the Power Range Neutron Flux Rate - High Negative Rate Trip function. This function, as specified in Table 3.3.1-1, Reactor Trip System Instrumentation, of the TS as Function 3.b, Power Range Neutron Flux Rate - High Negative Rate, will be deleted. The proposed change for the removal of the Power Range Neutron Flux Rate - High Negative Rate Trip function is aligned with the NRC approved methodology depicted in the Westinghouse Topical Report WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event, dated October 23, 1989.

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, as discussed below:

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

Response: No.

Removing the Power Range Neutron Flux High Negative Rate Trip Function from the Wolf Creek Generating Station (WCGS) TS does not increase the probability or consequences of accidents resulting from dropped control rod events previously analyzed utilizing the Nuclear Regulatory Commission (NRC) approved WCAP-11394-P-A methodology. The associated accident analysis does not rely on the Power Range Neutron Flux High Negative Rate Trip Function to safely shut down the facility. Other Reactor Trip System protection functions unrelated to the aforementioned function are not impacted by the deletion of this function. Additionally, the safety analysis of the plant is unaffected by the change. Therefore, the radiological releases associated with the analysis are not affected.

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

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

Attachment I to ET 23-0003 Page 7 of 8 Response: No.

The proposed amendment does not create the possibility of or introduce any new accident scenarios. The proposed amendment does not introduce any new failure mechanisms, malfunctions, or accident initiators not previously considered in the licensing basis for WCGS. The proposed amendment does not challenge the performance of existing safety-related systems or components, nor does it impact the design function of other structures, systems and components (SSCs).

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The methodology presented in the NRC approved WCAP-11394-P-A has demonstrated that no direct trip or automatic power reduction is necessary to maintain the Departure from Nucleate Boiling (DNB) design basis when evaluated on a plant/cycle specific basis. WCGS has evaluated a reference cycle in WCAP 17658-NP Revision 2 with the methodology from WCAP-11394-P-A applied. The application of the methodology for WCGS achieved satisfactory results commensurate with the conclusions reached in WCAP-11394-P-A. As a result, removal of the Power Range Neutron Flux Rate - High Negative Rate Trip function would not significantly reduce the margin of safety for any of the fission product barriers. The application of the methodology in transient analysis for the referenced cycle has demonstrated that the trip function is not required for maintaining the DNB for dropped control rod events.

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

Based on the above, WCNOC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(C). Therefore, a finding of no significant hazards consideration is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (i) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (ii) such activities will be conducted in compliance with the Commissions regulations, and (iii) 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.

ENVIRONMENTAL CONSIDERATION WCNOC 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, Standards for protection against radiation, 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

Attachment I to ET 23-0003 Page 8 of 8 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 paragraph (c)(9) of 10 CFR 51.22, Criterion of categorical exclusion, identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring an environmental review. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.

REFERENCES 6.1 Letter from NRC dated May 31, 2019, to Wolf Creek Nuclear Operating Corporation (WCNOC), Issuance of Amendment No. 221 RE: Transition to Westinghouse Core Design and Safety Analyses Including Adoption of Alternative Source Term (CAC No.

MF9307; EPID L-2017-LLA-0211) (ML19100A122).

6.2 Union Electric Company (Ameren Missouri) Letter dated March 6, 1990, to NRC, Removal of Power Range, Neutron Flux, High Negative Rate Trip From Technical Specifications (ML20033G518).

6.3 Letter from NRC dated August 23, 1990, to Union Electric Company (Ameren Missouri),

Amendment No. 56 to Facility Operating License No. NPF-30 (ML021650565).

6.4 Tennessee Valley Authority Letter dated October 29, 2021, to NRC, Application to Modify the Seqouyah Nuclear Plant, Unit 1 and Unit 2, Technical Specification Table 3.3.1-1, Reactor Trip System Instrumentation (ML21302A238).

6.5 Letter from NRC dated July 12, 2022, to Tennessee Valley Authority, Issuance of Amendment Nos. 357 and 351 Regarding Revision to Technical Specification Table 3.3.1-1, Reactor Trip System Instrumentation (ML22165A105).

Attachment II to ET 23-0003 Page 1 of 2 Proposed Technical Specification Changes (Mark-up)

RTS Instrumentation 3.3.1 Wolf Creek - Unit 1 3.3-16 Amendment No. 123, 140, 221, 227 Table 3.3.1-1 (page 1 of 6)

Reactor Trip System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE (a)

1. Manual Reactor Trip 1,2 3(b), 4(b), 5(b) 2 2

B C

SR 3.3.1.14 SR 3.3.1.14 NA NA

2. Power Range Neutron Flux a.

High b.

Low 1,2 1(c), 2(f) 2(h), 3(i) 4 4

4 D

V W, X SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 d 112.3% RTP d 28.3% RTP d 28.3% RTP 3.

Power Range Neutron Flux Rate a.

High Positive Rate b.

High Negative Rate 1,2 1,2 4

4 E

E SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 d 6.3% RTP with time constant t 2 sec d6.3% RTP with time constant t 2 sec 4.

Intermediate Range Neutron Flux 1(c), 2(d) 2 F,G SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 d 35.3% RTP (continued)

(a)

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

(b)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(c)

Below the P-10 (Power Range Neutron Flux) interlock.

(d)

Above the P-6 (Intermediate Range Neutron Flux) interlock.

(f)

With keff 1.0.

(h)

With keff < 1.0, and all RCS cold leg temperatures 500o F, and RCS boron concentration the rods out (ARO) critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(i)

With all RCS cold leg temperatures 500o F, and RCS boron concentration the ARO critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

Attachment II to ET 23-0003 Page 2 of 2

Attachment III to ET 23-0003 Page 1 of 6 Proposed USAR Changes (Mark-up)

WOLF CREEK power range protection interlock (P-10). This trip function can also be reinstated below P-10 by an administrative action requiring manual actuation of two control board mounted switches. Each switch will reinstate the trip function in one out of the two protection logic trains. The source range trip point is set between the P-6 setpoint (source range cutoff power level) and the maximum source range power level. The channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing during plant shutdown or prior to startup.

This bypass action is annunciated on the control board.

4.

Power range high positive neutron flux rate trip This circuit trips the reactor when a sudden abnormal increase in nuclear power occurs in two out of the four power range channels. This trip provides DNB protection against certain rod ejection and rod withdrawal accidents (see Chapter 15.0).

5.

Power range high negative neutron flux rate trip This circuit trips the reactor when a sudden abnormal decrease in nuclear power occurs in two out of the four power range channels. This trip provides protection against two or more dropped rods and is always active. Protection against one dropped rod is not required to prevent occurrence of DNB per Section 15.4.3.

Figure 7.2-1 (Sheet 3) shows the logic for all of the nuclear overpower and rate trips.

b.

Core thermal overpower trips The specific trip functions generated are as follows:

1.

Overtemperature ' T trip This trip protects the core against low DNB and trips the reactor on coincidence, as listed in Table 7.2-1, with one set of temperature measurements per loop.

The setpoint for this trip is continuously calculated by analog circuitry for each loop by solving the following equation:

7.2-5 Rev. 25 Attachment III to ET 23-0003 Page 2 of 6 Figure 7.2-1 (Sheet

3) shows the logic for all of the nuclear overpower and rate trips.

WOLF CREEK TABLE 7.2-1 LIST OF REACTOR TRIPS Coincidence Protection Reactor Trip Logic Interlocks Comments 1.

Power range high neutron 2/4 Manual block of low High and low setting; manual flux setting permitted block and automatic reset by P-10 of low setting by P-10 2.

Intermediate range 1/2 Manual block per-Manual block and automatic high neutron flux mitted by P-10 reset 3.

Source range high neutron 1/2 Manual block per-Manual block and automatic flux mitted by P-6; reset; automatic block above interlocked with P-10 P-10 4.

Power range high positive 2/4 No interlocks neutron flux rate 5.

Power range high negative 2/4 No interlocks neutron flux rate 6.

Overtemperature

'T 2/4 No interlocks 7.

Overpower

'T 2/4 No interlocks 8.

Pressurizer low pressure 2/4 Interlocked with Blocked below P-7 P-7 Rev. 1 Attachment III to ET 23-0003 Page 3 of 6 Removed with License Amendment No. XXX

WOLF CREEK TABLE 7.2-3 REACTOR TRIP SYSTEM INSTRUMENTATION (Typical for Westinghouse Four Loop PWR)

Typical Trip Typical Time Reactor Trip Signal Typical Range Accuracy Response (sec)*

1.

Power range high neutron 1 to 120% of full power

+ 5.3% of full scale 0.2 flux 2.

Intermediate range high 8 decades of neutron

+ 12.3% of full scale; 0.2 neutron flux flux overlapping source range by 2 decades 3.

Source range high neutron 6 decades of neutron

+ 11.9% of full scale 0.2 flux flux (1 to 106 counts/sec) 4.

Power range high positive

+15% of full power

+ 2.3% of full scale 0.2 neutron flux rate 5.

Power range high negative

-15% of full power

+ 2.3% of full scale 0.2 neutron flux rate 6.

Overtemperature T TH 540 to 660°F

+ 6.1° F 5.0 TC 510 to 630°F TAV 530 to 630°F PPRZR 1,700 to 2,500 psig F() -50 to +50 Tsetpoint 0 to 150% power 7.

Overpower T TH 540 to 660°F

+ 3.6°F 5.0 TC 510 to 630°F TAV 530 to 630° F Tsetpoint 0 to 150% power 8.

Pressurizer low pressure 1,700 to 2,500 psig

+/- 18 psi (compensated 0.6 signal) 9.

Pressurizer high pressure 1,700 to 2,500 psig

+/- 18 psi (noncompensated 0.6 signal)

Rev. 33 Attachment III to ET 23-0003 Page 4 of 6 Removed with License Amendment No. XXX

WOLF CREEK TABLE 7.2-4 (Sheet 2)

(a)

(b)

Technical (c)

Trip Accident Specification Excessive Increase in Secondary Steam Flow (15.1.3)

Inadvertent Opening of a Steam Generator Atmospheric Relief or Safety Valve (15.1.4)

Spectrum of Steam System Piping Failures Inside and Outside of Containment in a PWR (15.1.5)

Spectrum of Rod Cluster Control Assembly Ejection Accidents (15.4.8)

See Note d, 3.

Intermediate Uncontrolled Rod Cluster Control 3.3.1, Table 3.3.1-1, range high Assembly Bank Withdrawal From a Function 4 neutron flux Subcritical or Low Power Startup trip Condition (15.4.1)

See Note d, 4.

Source range Uncontrolled Rod Cluster Control 3.3.1, Table 3.3.1-1, high neutron Assembly Bank Withdrawal From a Function 5 flux trip Subcritical or Low Power Startup Condition (15.4.1) 5.

Power range Spectrum of Rod Cluster Control 3.3.1, Table 3.3.1-1, high positive Assembly Ejection Accidents Function 3.a neutron flux (15.4.8) and Rod Withdrawal at rate trip Power Accidents (15.4.2) 6.

Power range Rod Cluster Control Assembly 3.3.1, Table 3.3.1-1, high negative Misalignment (15.4.3)

Function 3.b flux rate trip Rev. 25 Attachment III to ET 23-0003 Page 5 of 6 Removed with License Amendment No. XXX

WOLF CREEK TABLE 15.0-6 (Sheet 3)

Incident Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Equipment Uncontrolled Power range high flux, Pressurizer safety rod cluster overtemperature T, high valves,steam generator control pressurizer pressure, manual, safety valves assembly bank high pressurizer water level, withdrawal at overpower T, power range power high positive flux rate Rod cluster Power range negative control flux rate, overtemperature assebmly mis-T, manual alignment Startup of an Manual, power range high flux Low insertion limit inactive reac-above P8 setpoint with low annun-ciators for tor coolant flow boration loop at an in-correct temperature Chemical and Power range high flux, over-Pressurizer safety volume control, temperature T, manual, over-valves, steam generator system mal-power T high pressurizer safety valves rod function that pressure, high pressurizer insertion limit alarms, results in a water level position step counter, decrease in RT alarm, audible count boron rate from BF3 detector concentrattion source range high flux in the reactor alarm coolant Spectrum of Source range high flux, Low pressurizer pressure, Pressurizer safety ESFAS rod cluster con-intermediate range high flux, Hi-1 containment pressure, valves trol assembly power range high flux, Hi-3 containment pressure, (RCCA) ejection high positive flux rate, manual.

accidents manual 15.5 Increase in reactor coolant inventory Inadvertent Low pressurizer SIS, low-low SG level, manual Secondary system Auxiliary feedwater operation of pressure, manual, atmospheric relief valves, system the ECCS dur-safety injection trip steam generator safety ing power oper-valves, CVCS (letdown) tion CVCS malfunction that increases reactor coolant inventory High pressure level, manual Secondary system atmospheric relief valves, steam generator safety valves, CVCS (letdown)

Rev. 34 Attachment III to ET 23-0003 Page 6 of 6

Attachment IV to ET 23-0003 Page 1 of 2 Markup of the Renewed Facility Operating License

4 (5)

The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

The Operating Corporation, 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 the operation of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and 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 Operating Corporation is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100%

power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 233, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 229, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Antitrust Conditions Evergy Kansas South, Inc. and Evergy Metro, Inc. shall comply with the antitrust conditions delineated in Appendix C to this license.

(4)

Environmental Qualification (Section 3.11, SSER #4, Section 3.11, SSER #5)*

Deleted per Amendment No. 141.

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

Renewed License No. NPF-42 Amendment No. 233 Attachment IV to ET 23-0003 Page 2 of 2 Amendment No. XXX

Attachment V to ET 23-0003 Page 1 of 6 Proposed Technical Specification Bases Changes (Mark-up) for Information Only

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-11 Revision 84 BASES APPLICABLE b.

Power Range Neutron Flux - Low (continued)

SAFETY ANALYSES, LCO, and In addition, in MODE 3 (with any RCS cold leg temperature APPLICABILITY

< 500° F, or the RCS sufficiently borated, or the RCCA bank withdrawal event precluded per the specified conditions of footnote (i) in Table 3.3.1-1), 4, 5, or 6, the Power Range Neutron Flux - Low trip Function does not have to be OPERABLE because the reactor is shut down and the NIS power range detectors cannot accurately detect neutron levels in this range. Other RTS trip Functions and administrative controls provide protection against positive reactivity excursions in these MODES and specified conditions in the Applicability.

3.

Power Range Neutron Flux Rate - High Positive Rate The Power Range Neutron Flux Rate trips uses the same channels as discussed for Function 2 above.

a.

Power Range Neutron Flux - High Positive Rate The Power Range Neutron Flux - High Positive Rate trip Function ensures that protection is provided against rapid increases in neutron flux that are characteristic of an RCCA drive rod housing rupture and the accompanying ejection of the RCCA. This Function compliments the Power Range Neutron Flux - High and Low Setpoint trip Functions to ensure that the criteria are met for a rod ejection from the power range. This Function also provides protection for the rod withdrawal at power event.

The LCO requires all four of the Power Range Neutron Flux

- High Positive Rate channels to be OPERABLE.

In MODE 1 or 2, when there is a potential to add a large amount of positive reactivity from a rod ejection accident (REA), the Power Range Neutron Flux - High Positive Rate trip must be OPERABLE. In MODE 3, 4, 5, or 6, the Power Range Neutron Flux - High Positive Rate trip Function does not have to be OPERABLE because other RTS trip Functions and administrative controls will provide protection against positive reactivity excursions.

b.

Power Range Neutron Flux - High Negative Rate The Power Range Neutron Flux - High Negative Rate trip Function ensures that protection is provided for multiple rod drop accidents. At high power levels, a multiple rod drop accident could cause local flux peaking that would Attachment V to ET 23-0003 Page 2 of 6

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-12 Revision 84 BASES APPLICABLE b.

Power Range Neutron Flux - High Negative Rate SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY result in an unconservative local DNBR. DNBR is defined as the ratio of the heat flux required to cause a DNB at a particular location in the core to the local heat flux. The DNBR is indicative of the margin to DNB. No credit is taken for the operation of this Function for those rod drop accidents in which the local DNBRs will be greater than the limit.

The LCO requires all four Power Range Neutron Flux -

High Negative Rate channels to be OPERABLE. The Trip Setpoint is 4% RTP with a time constant 2 seconds.

In MODE 1 or 2, when there is potential for a multiple rod drop accident to occur, the Power Range Neutron Flux - High Negative Rate trip must be OPERABLE. In MODE 3, 4, 5, or 6, the Power Range Neutron Flux - High Negative Rate trip Function does not have to be OPERABLE because the core is not critical and DNB is not a concern.

4.

Intermediate Range Neutron Flux The Intermediate Range Neutron Flux trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition. This trip Function provides redundant protection to the Power Range Neutron Flux - Low Setpoint trip Function. The NIS intermediate range detectors are located external to the reactor vessel and measure neutrons leaking from the core. The NIS intermediate range detectors do not provide any input to control systems. Note that this Function also provides a signal to prevent automatic and manual rod withdrawal prior to initiating a reactor trip. Limiting further rod withdrawal may terminate the transient and eliminate the need to trip the reactor.

The LCO requires two channels of Intermediate Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function.

The Trip Setpoint is d 25% RTP.

Because this trip Function is important only during startup, there is generally no need to disable channels for testing while the Attachment V to ET 23-0003 Page 3 of 6

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-36 Revision 84 BASES ACTIONS E.1 and E.2 (continued) x

Overtemperature 'T; x

Overpower 'T; x

Power Range Neutron Flux - High Positive Rate; x

Power Range Neutron Flux - High Negative Rate; x

Pressurizer Pressure - High; and x

SG Water Level - Low Low.

A known inoperable channel must be placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Placing the channel in the tripped condition results in a partial trip condition requiring only one-out-of-three logic for actuation of the two-out-of-four trip logic. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed to place the inoperable channel in the tripped condition is justified in Reference 12.

If the inoperable channel cannot be placed in the trip condition within the specified Completion Time, the unit must be placed in a MODE where these Functions are not required OPERABLE. An additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to place the unit in MODE 3. Six hours is a reasonable time, based on operating experience, to place the unit in MODE 3 from full power in an orderly manner and without challenging unit systems.

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while performing routine surveillance testing of the other channels. The 12-hour time limit is justified in Reference 12.

F.1 and F.2 Condition F applies to the Intermediate Range Neutron Flux trip when THERMAL POWER is above the P-6 setpoint and below the P-10 setpoint and one channel is inoperable. Above the P-6 setpoint and below the P-10 setpoint, the NIS intermediate range detector performs the monitoring Functions. If THERMAL POWER is greater than the P-6 setpoint but less than the P-10 setpoint, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed to reduce THERMAL POWER below the P-6 setpoint or to increase THERMAL POWER above the P-10 setpoint. The NIS Intermediate Range Neutron Flux channels must be OPERABLE when the power level is above the capability of the source range, P-6, and below the capability of the power range, P-10. If THERMAL POWER is greater than the P-10 setpoint, the NIS power range detectors perform the monitoring and protection Attachment V to ET 23-0003 Page 4 of 6

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-58 Revision 89 TABLE B 3.3.1-1 (Page 1 of 2)

FUNCTION TRIP SETPOINT(a) 1.

Manual Reactor Trip 2.

Power Range Neutron Flux a.

High b.

Low 3.

Power Range Neutron Flux a.

High Positive Rate b.

High Negative Rate 4.

Intermediate Range Neutron Flux 5.

Source Range Neutron Flux 6.

Overtemperature T 7.

Overpower T 8.

Pressurizer Pressure a.

Low b.

High 9.

Pressurizer Water level - High 10.

Reactor Coolant Flow - Low 11.

Not Used 12.

Undervoltage RCPs 13.

Underfrequency RCPs 14.

Steam Generator (SG) Water Level Low - Low 15.

Not Used 16.

Turbine Trip a.

Low Fluid Oil Pressure b.

Turbine Stop Valve Closure NA 109% of RTP 25% of RTP 4% of RTP with a time constant 2 seconds 4% of RTP with a time constant 2 seconds 25% of RTP 105 cps See Table 3.3.1-1 Note 1 See Table 3.3.1-1 Note 2 1940 psig 2385 psig 92% of instrument span 89.9% of Normalized Flow 10578 Vac 57.15 Hz 23.5% of narrow range instrument span 590.00 psig 1% open Attachment V to ET 23-0003 Page 5 of 6

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-60 Revision 89 TABLE B 3.3.1-2 (Page 1 of 2)

FUNCTIONAL UNIT RESPONSE TIME 1.

Manual Reactor Trip 2.

Power Range Neutron Flux a.

High b.

Low 3.

Power Range Neutron Flux a.

High Positive Rate b.

High Negative Rate 4.

Intermediate Range Neutron Flux 5.

Source Range Neutron Flux 6.

Overtemperature T 7.

Overpower T 8.

Pressurizer Pressure a.

Low b.

High 9.

Pressurizer Water Level - High 10.

Reactor Coolant Flow - Low a.

Single Loop (Above P-8) b.

Two Loops (Above P-7 and below P-8) 11.

Not Used 12.

Undervoltage - Reactor Coolant Pumps 13.

Underfrequency - Reactor Coolant Pumps 14.

Steam Generator Water Level - Low-Low 15.

Not Used N.A.

0.5 second(1) 0.5 second(1) 0.5 second(1).

0.5 second(1)

N.A.

N.A.

6.0 seconds(1) 6.0 seconds(1) 2.0 seconds 1.0 second N.A.

1.0 second 1.0 second 1.5 seconds 0.6 second 2.0 seconds (1) Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

Attachment V to ET 23-0003 Page 6 of 6