NLS2024025, License Amendment Request to Revise Technical Specifications Table 3.3.2.1-1, Control Rod Block Instrumentation

License Amendment Request to Revise Technical Specifications Table 3.3.2.1-1, Control Rod Block Instrumentation

ML24131A026
Person / Time
Site:	Cooper
Issue date:	05/09/2024
From:	Dia K Nebraska Public Power District (NPPD)
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NLS2024025
Download: ML24131A026 (1)

Text

H Nebraska Public Power District NLS2024025 May 9, 2024 Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 "Always there when you need us" 50.90

Subject:

License Amendment Request to Revise Technical Specifications Table 3.3.2.1-1, "Control Rod Block Instrumentation" Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, "Application for amendment oflicense, construction permit, or early site permit," Nebraska Public Power District (NPPD) is submitting a request for an amendment to Renewed License No. DPR-46 for Cooper Nuclear Station (CNS). The proposed change would relocate cycle specific Minimum Critical Power Ratio values from the CNS Technical Specifications (TS) Table 3.3.2.1-1 to the CNS Core Operating Limits Report. to this letter provides a description and assessment of the proposed change. provides the existing TS pages marked up to show the proposed change. provides revised (clean) TS pages. The changes to the TS Bases are provided for information only in Attachment 4 and will be incorporated upon implementation of the approved amendment.

Approval of the proposed amendment is requested by September 30, 2024. Once approved, the amendment shall be implemented by October 30, 2024, to allow plant operation after the completion of a refueling outage.

NPPD has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22( c )(9).

The proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," a copy of this application, with attachments, is being provided to the designated State of Nebraska Official. Copies to the Nuclear Regulatory Commission Region IV office and the CNS Resident Inspectors are also being provided in accordance with 10 CFR 50.4(b)(l).

COOPER NUCLEAR STATION 72676 648A Ave/ P.O. Box 98 / Brownville, NE 68321 http://www.nppd.com

NLS2024025 Page 2 of2 There are no regulatory commitments made in this submittal. If you should have any questions regarding this submittal, please contact Linda Dewhirst, Regulatory Affairs and Compliance Manager, at (402) 825-5416.

I declare under penalty of perjury that the foregoing is true and correct.

ExecutedOn: 5/~ ZnZ>{

ate Khalil Dia Site Vice President lbs Attachments: 1. Description and Assessment

2. Proposed Technical Specifications Change (Mark-up)

3. Revised Technical Specifications Pages

4. Proposed Technical Specifications Bases Changes (Mark-up) cc:

Regional Administrator w/ attachments USNRC - Region IV Cooper Project Manager w/ attachments USNRC - NRR Plant Licensing Branch IV Senior Resident Inspector w/ attachments USNRC-CNS Nebraska Health and Human Services w/ attachments Department of Regulation and Li censure NPG Distribution w/ attachments CNS Records w / attachments

NLS2024025 Page 1 of 6 Description and Assessment Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46 1.0 Summary Description 2.0 Detailed Description 3.0 Technical Evaluation 4.0 Regulatory Analysis 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusion 5.0 Environmental Evaluation

NLS2024025 Page 2 of6 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment oflicense, construction permit, or early site permit," Nebraska Public Power District (NPPD) is submitting a request for an amendment to Renewed License No. DPR-46 for Cooper Nuclear Station (CNS). The proposed change would amend the CNS Technical Specifications (TS) to modify TS Table 3.3.2.1-1, "Control Rod Block Instrumentation." The proposed change would relocate cycle specific Minimum Critical Power Ratio (MCPR) values to the CNS Core Operating Limits Report (COLR).

2.0 DETAILED DESCRIPTION NPPD proposes the following changes to the CNS TS:

1. Revise the notes associated with TS Table 3.3.2.1-1, "Control Rod Block Instrumentation," as shown below:

(a) THERMAL POWER 2: 27.5% and< 62.5% RTP and MCPR ~

less than the limit specified in the COLR and no peripheral control rod selected.

( d) THERMAL POWER 2: 62.5% and < 82.5% RTP and MCPR ~

less than the limit specified in the COLR and no peripheral control rod selected.

(e) THERMAL POWER 2: 82.5% and< 90% RTP and MCPR ~

less than the limit specified in the COLR and no peripheral control rod selected.

(t) THERMAL POWER 2: 90% RTP and MCPR < 1.40 less than the limit specified in the COLR and no peripheral control rod selected.

(g) THERMAL POWER 2: 27.5% and< 90% RTP and MCPR ~

less than the limit specified in the COLR and no peripheral control rod selected.

2. Revise TS 5.6.5, CORE OPERATING LIMITS REPORT (COLR) as shown below:

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.

2.

The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.

3.

The Linear Heat Generation Rates for Specifications 3.2.3 and 3. 7. 7.

4.

The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.

5.

The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.

6.

The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.

Relocation of the cycle specific MCPRs to the COLR, which is controlled by TS 5.6.5, would provide NPPD the flexibility to revise cycle specific MCPRs in accordance with Nuclear Regulatory Commission (NRC) approved methodologies without the need for a license

NLS2024025 Page 3 of 6 amendment. The COLR, including any mid-cycle revisions or supplements, is required to be submitted to the NRC for each reload cycle per TS 5.6.5.

3.0 TECHNICAL EVALUATION

NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," provides guidance to licensees for the removal of cycle dependent parameter limits from the TS provided these values are included in a COLR and are determined with NRC approved methodologies referenced in the TS. The specific values of these limits may be modified by licensees, without affecting nuclear safety, provided that these changes are determined using an NRC approved methodology and consistent with all applicable limits of the plant safety analysis that are addressed in the Final Safety Analysis Report. If any of the applicable limits of the safety analysis are not met, prior NRC approval of the change is required.

Control Rod Block Instrumentation MCPR values are calculated as part of the reload core design licensing analyses in accordance with NRC approved methods in NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel. Because the analyses are completed only two to three months prior to the start of the next refueling outage, a need to revise MCPR values for the next operating cycle would require submitting a license amendment request with a quick turnaround, placing an unnecessary burden on NPPD and NRC resources. Relocating the MCPR values to the COLR will allow NPPD to make cycle-specific changes that are consistent with NRC approved methodologies and within limits of the safety analysis without the burdensome process of amending the TS. The TS will continue to establish limits for MCPR for the Control Rod Block Instrumentation while the specific values for MCPR are relocated to the COLR.

Therefore, the requested changes are essentially administrative in nature, and the required level of safety will be maintained.

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria NRC GL 88-16 discusses that processing TS changes to update cycle-specific parameter limits each fuel cycle places an unnecessary burden on the licensee and the NRC if these limits are developed using an NRC approved methodology. The GL provides an alternative that relocates the specific parameter values to the COLR provided the values are determined using an NRC approved methodology and the TS require plant operation in accordance with the limits specified in the COLR.

10 CFR 50.36, Technical specifications - requires that the TS contain limiting conditions for operation, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

The proposed changes are consistent with the above regulatory guidance and regulation.

NLS2024025 Page 4 of 6 4.2 Precedent In March 2018, Duane Arnold Energy Center received approval to relocate the Control Rod Block Instrumentation MCPR values to the COLR (ML18011A059). Also in January 2014, Columbia Generating Station received approval to relocate the Control Rod Block Instrumentation MCPR values to the COLR as part of a change to implement a digital instrumentation system (ML133 l 7B623).

The TS for the plants below contain notes comparable to those in the CNS TS that establish limits on MCPR for the Control Rod Block Instrumentation. Similar to the proposed change to the CNS TS, the TS below do not provide the specific value for the Control Rod Block Instrumentation MCPR but specify the limit for MCPR as "less than the limit specified in the COLR."

• Brunswick Unit 1 (ML062900525) and Unit 2 (ML062900536)

• Browns Ferry Unit 2 (ML052780020)

• Peach Bottom Unit 2 (ML052720266)

• Susquehanna Unit 1 (ML052720300) and Unit 2 (ML052720301) 4.3 No Significant Hazards Consideration Determination Analysis Nebraska Public Power District (NPPD) requests an amendment to the Cooper Nuclear Station (CNS) Technical Specifications (TS) to modify TS Table 3.3.2.1-1, "Control Rod Block Instrumentation." The proposed change would relocate cycle specific Minimum Critical Power Ratio (MCPR) values to the CNS Core Operating Limits Report (COLR).

As required by 10 CFR 50.91(a), NPPD has evaluated the proposed change to the CNS TS using the criteria in 10 CFR 50.92 and determined that the proposed change does not involve a significant hazards consideration. An analysis of the issue of no significant hazards consideration is presented below.

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

Response: No The proposed change is an administrative change that does not affect any plant systems, structures, or components designed for the prevention or mitigation of previously evaluated accidents. No new equipment is added nor is installed equipment being changed or operated in a different manner. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on, nor does it contribute in any way to the probability or consequences of transients or accidents. The COLR will continue to be controlled by the CNS programs and procedures that comply with TS 5.6.5.

Transient analyses addressed in the Updated Safety Analysis Report will continue to be performed in the same manner with respect to changes in the cycle dependent parameters obtained from the use of Nuclear Regulatory Commission (NRC) approved reload design

NLS2024025 Page 5 of 6 methodologies, which ensures that the transient evaluation of new reloads are bounded by previously accepted analyses.

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

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

Response: No The proposed administrative change does not involve any changes to the operation, testing, or maintenance of any safety-related, or otherwise important to safety systems.

Systems important to safety will continue to be operated and maintained within their design bases. Relocation of the Control Rod Block Instrumentation MCPR values to the COLR has no influence or impact on new or different kind of accidents.

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

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

Response: No The margin of safety is not affected by the relocation of cycle-specific Control Rod Block Instrumentation MCPR values from the TS to the COLR. Appropriate measures exist to control the values of these cycle-specific limits since it is required by TS that only NRC approved methods be used to determine the limits. The proposed change continues to require operation within the core thermal limits as obtained from NRC approved reload design methodologies and the actions to be taken if a limit is exceeded remain unchanged in accordance with existing TS.

Therefore, the proposed change has no impact to the margin of safety.

Based on the above, NPPD concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion Based on the considerations discussed above, (1) there is 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.

NLS2024025 Page 6 of6 5.0 ENVIRONMENTAL EVALUATION NPPD 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 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.

NLS2024025 Page 1 of3 Proposed Technical Specifications Changes (Mark-up)

Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46 Marked Up Pages 3.3-19 5.0-21

Table3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation Control Rod Block Instrumentation 3.3.2.1 FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1.

Rod Block Monitor

a.

Low Power Range - Upscale (a)

b.

Intermediate Power Range - Upscale (d)

c.

High Power Range - Upscale

( e),(f)

d.

lnop (f),(g)

e.

Downscale (f),(g)

2.

Rod Worth Minimizer 1(h),2(h)

3.

Reactor Mode Switch - Shutdown Position (i) 2 2

2 2

2 2

SR 3.3.2.1.1 SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

SR 3.3.2.1.1 SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

SR 3.3.2.1.1 SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

SR 3.3.2.1.1 SR 3.3.2.1.1 SR 3.3.2.1.5 SR 3.3.2.1.2 SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8 SR 3.3.2.1. 7 U)

U)

U)

NA

~ 92/125 divisions of full scale NA NA (a) THERMAL POWER~ 27.5% and < 62.5% RTP and MCPR -<-4--:+G less than the limit specified in the COLR and no peripheral control rod selected.

(b) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

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

(d) THERMAL POWER~ 62.5% and < 82.5% RTP and MCPR <-4:-+G-less than the limit specified in the COLR and no peripheral control rod selected.

(e) THERMAL POWER~ 82.5% and< 90% RTP and MCPR ~

less than the limit specified in the COLR and no peripheral control rod selected.

(f) THERMAL POWER~ 90% RTP and MCPR -<-4-A-G-less than the limit specified in the COLR and no peripheral control rod selected.

(g) THERMAL POWER~ 27.5% and < 90% RTP and MCPR ~

less than the limit specified in the COLR and no peripheral control rod selected.

(h) With THERMAL POWER :5 9.85 RTP.

(i) Reactor mode switch in the shutdown position.

U)

Less than or equal to the Allowable Value specified in the COLR.

3.3-19 Amendment No. 242

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV. 8.1.

5.6.4 (Deleted) 5.6.5 Core Operating Limits Report (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.

2.

The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.

3.

The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.7.

4.

The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.

5.

The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.

6.

The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

Cooper

1.

NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).

( continued) 5.0-21 Amendment No.

NLS2024025 Page 1 of3 Revised Technical Specifications Pages Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46 Revised Pages 3.3-19 5.0-21

Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation Control Rod Block Instrumentation 3.3.2.1 FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1.

Rod Block Monitor

a.

Low Power Range - Upscale (a)

b.

Intermediate Power Range - Upscale (d)

c.

High Power Range - Upscale (e),(f)

d.

lnop (f),(g)

e.

Downscale (f),(g)

2.

Rod Worth Minimizer

3.

Reactor Mode Switch - Shutdown Position (i) 2 2

2 2

2 2

SR 3.3.2.1.1 SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

SR 3.3.2.1.1 SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

SR 3.3.2.1.1 SR 3.3.2.1.4 SR 3.3.2.1.5(b)(c)

SR 3.3.2.1.1 SR 3.3.2.1.1 SR 3.3.2.1.5 SR 3.3.2.1.2 SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8 SR 3.3.2.1. 7 (j)

(j)

0)

NA

~ 92/125 divisions of full scale NA NA (a) THERMAL POWER~ 27.5% and < 62.5% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(b) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

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

(d) THERMAL POWER~ 62.5% and < 82.5% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(e) THERMAL POWER~ 82.5% and< 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(f) THERMAL POWER~ 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(g) THERMAL POWER ~ 27.5% and < 90% RTP and MCPR less than the limit specified in the COLR and no peripheral control rod selected.

(h) With THERMAL POWER s 9.85 RTP.

(i) Reactor mode switch in the shutdown position.

(j)

Less than or equal to the Allowable Value specified in the COLR.

Cooper 3.3-19 Amendment No.

5.6 Reporting Requirements ( continued) 5.6.3 Radioactive Effluent Release Report Reporting Requirements 5.6 The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.8.1.

5.6.4 (Deleted) 5.6.5 Core Operating Limits Report (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3.7.7.

2.

The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7, and MCPR99.9% for Specification 3.2.2.

3.

The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.7.

4.

The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.

5.

The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.

6.

The Minimum Critical Power Ratios in Table 3.3.2.1-1 for Specification 3.3.2.1.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

Cooper

1.

NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).

( continued) 5.0-21 Amendment No.

NLS2024025 Page 1 of2 Proposed Technical Specifications Bases Changes (Mark-up)

Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DPR-46 Marked Up Page B 3.3-45

BASES Corih*ol Rod Block lnslruman!alion B 3,3.2,1 APPLICABLE SAFETY ANALYSESr LCO, and APPLICABl:UTY {OOfltinued)

Cooper

1.

Rod Block Monitor The RBM is dasfgned to prevent violation of lh@ MCPR SL and tM cladding. 1 % plastjo strain fuel design limit thitt may resutt from a sjllgle control rod WB,hdrawat error (RWE) event. TM anal*ytk:eJ methodiS and assumptions used in evaluating the RV!JE event are summarfzed in R~

4. A statistical analysis of RWE events was performed ro determine the RSM resp(lfflse for bolh channel& for aech event. Fron, these responses. lhe fuel thermal performance as a function of RBM

.AJlowabte Vatue was determined. The Allowab4e Values are* chOGen as a function of P(IWer revel. Based on lhe $pecifted.Alov,able Val~.

operating limits are establf,shed, The RBM Function satisfies Criterion 3 of' 10 CFR. 50.36(c)(2)(ij) (Ref. 5).

T\\'/0 channers of lhe RSM are required to be OPERABLE~ 'Nith U,eir setpoints within lhe app.ropriate Allowable Values, to ensure that no singre instrument failure can preclude a rod block from this FundiOn. The actual setpolnts are callbrated oonslstent wUh appficable setpolnt m-ethodofogy.

The RBM is assumed to mitigate the oonseqoonces of an RWE event when operating ~ 30% RT P (anatyticai limit), and a peripheral control rod is ool selected. B&low this PO'Net' -,vel or If a peripheral control rod is seleci:ed1 the oonsequences of an RWE event wil not exceed the MCPR SL and, therefore, lhe RSM Is not ra~ulred to be OPERABLE (Ref, 4 ). When operating < 90% RTF\\ analyses (Ref. 4) have sh0\\\\l11i that with a.n.ini.tia.. *.. M

.. cPRa****

.... 1.*.n.**.o. RW.E.' eve.ntW1. * *u res*u.tt in.exceed.*~

... *.*... in. g*t*h*e.MC. *p.*,R Sl~ Abio1 the analyse. de,rnonstrate lhat Wtien operating al~ 00% RTP with MCPR ~.

oo. * *. E event will refiutt in exceeding the MCPR SL (Ref, 4 ). Thntfoft~.., Ulese oonditions1 the RSM is also not required to be OPERABLE.

~*

The RWM is a. backup to openatOf" oonlrol of the rod sequences. The RWM eriforces the bankedl poailion wilhd rawal eeque<iee (BPVIJS) by alerting tha ~tor when Ute rod.paUam is not jn ao::;ordanct, wHh BPWS. Compliallee with BP\\IVS ensures that the initial conditions of the CRDA amdysh.i are not violated.

The analytical methods arJd Hsumptioos used in evaluating the CRDA are* summarized in References 6 and 7. The BPWS requires that conlml