NLS2019020, Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit Mcpr.

Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit Mcpr.

ML19171A266
Person / Time
Site:	Cooper
Issue date:	05/23/2019
From:	Dent J Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2019020
Download: ML19171A266 (21)

Text

H Nebraska Public Power District Always there when you need us 50.90 NLS2019020

. May 23, 2019 Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. *20555-0001

Subject:

Application to Revise Technical Specifications to Adopt TSTF-564, "Safety Limit MCPR" Cooper Nuclear Station, Docket No.. 50-298, License No. DPR-46

Dear Sir or Madrun:

Pursuant to 10. CFR 50.90, Nebraska Public Power District (NPPD) is submitting a request for an amendment to the Technical Specifications (TS) for Cooper Nuclear Station (CNS).

NPPD requests adoption ofTSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications, into the CNS TS. The proposed amendment revises the TS safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL.

Attachment 1 provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides revised ( clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only.

No regulatory commitments are made in this submittal.

Approval of the proposed amendment is requested by May 31, 2020. Once approved, the amendment shall be implemented within 60 days.

In accordance with 10 CFR 50.91(b)(l), "Notice for Public Comment; State Consultation," a copy of this application, with attachments, is being provided to the designated State of Nebraska Official.

If you should have any questions regarding this submittal please contact Jim Shaw, Licensing Manager, at (402) 825-2788.

COOPER NUCLEAR STATION P.O. Box 98 / Brownvilie_'NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com

NLS2019020 Page 2 of2 I declare under penalty of perjury that the foregoing is true and correct.

Executed On: S/J3b~OJ~

• *

• 1 ate Dent, Jr.

Vice President - Nuclear and Chief Nuclear Officer

/dv Attachments: 1. Description and Assessment

2. Proposed Technical Specifications Changes (Mark..:up)

3. Revised Technical Specifications Pages

4. Proposed Technicat Specifications Bases Changes (Mark-up) -

Information Only

• 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 Licensure NPG Distribution w/o attachments CNS Records w/ attachments

NLS2019020 Page 1 of 4 Attachment 1 Application to Revise Technical Specifications to Adopt TSTF-564, "Safety Limit MCPR" Description and Assessment Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 1.0 Description 2.0 Assessment 2.1 Applicability of Safety Evaluation 2.2 Variations 3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Analysis 3 .2 Conclusion 4.0 Environmental Evaluation

NLS2019020 Attachment 1 Page 2 of 4

1.0 DESCRIPTION

Nebraska Public Power District (NPPD) requests adoption of TSTF-564, "Safety Limit MCPR,"

Revision 2, which is an approved change to the Improved Standard Technical Specifications, into the Cooper Nuclear Station (CNS) Technical Specifications (TS). The proposed amendment revises the TS safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation NPPD has reviewed the safety evaluation for TSTF-564 provided to the Technical Specifications Task Force in a letter dated November 16, 2018. This review included a review of the Nuclear Regulatory Commission (NRC) staffs evaluation, as well as the information provided in TSTF-564. NPPD has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRC staff are applicable to CNS and justify this amendment for the incorporation of the changes to the CNS TS.

The CNS reactor is currently fueled with GNF2 fuel bundles. The proposed safety limit in SL 2.1.1.2 is 1.07, consistent with Table 1 ofTSTF-564.

The MCPR value, calculated as the point at which 99.9% of the fuel rods would not be susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences, is referred to as MCPR99 .9%* Technical Specification 5.6.5, "Core Operating Limits Report (COLR)," is revised to require the MCPR99.9% value to be included in the cycle-specific COLR.

2.2 Variations NPPD is proposing the following variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs safety evaluation.

The CNS TS utilize different numbering than the Standard Technical Specifications on which TSTF-564 was based. Specifically, Section 5.6.5, "Core Operating Limits Report (COLR)," of the CNS TS is numbered 5.6.3 in TSTF-564. This difference is administrative and does not affect the applicability of TSTF-564 to the CNS TS.

The traveler and safety evaluation discuss the applicable regulatory requirement and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). CNS was not licensed to the 10 CFR 50, Appendix A, GDC. The CNS equivalent of the referenced GDC is located in Appendix F of the CNS Updated Safety Analysis Report. Specifically, the traveler and safety evaluation refer to GDC 10. The ,equjvalent draft GDC dated Jµly 11, 1967, that CNS is licensed to, is Criterion 6. This difference does not alter the conclusion that the proposed change is applicable to CNS.

NLS2019020 Page 3 of 4

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis Nebraska Public Power District (NPPD) requests adoption ofTSTF-564, "Safety Limit MCPR,"

which is an approved change to the Improved Standard Technical Specifications, into the Cooper Nuclear Station Technical Specifications (TS). The proposed change revises the TS safety limit on minimum critical power ratio (SLMCPR). The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition boiling. A single SLMCPR value will be used instead of two values applicable when one or two recirculation loops are in operation. TS 5.6.5, "Core Operating Limits Report (COLR)," is revised to require the current SLMCPR value to be included in the COLR.

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

Response: No.

The proposed amendment revises the TS SL1\1CPR and the list of core operating limits to be included in the COLR. The SLMCPR is not an initiator of any accident previously evaluated. The revised safety limit values continue to ensure for all accidents previously evaluated that the fuel cladding will be prqtected. from failure due to transition boiling..

The proposed change does not affect plant operation or any procedural or administrative controls on plant operation that affect the functior:is of preventing or mitigating any accidents previously evaluated.

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

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

Response: No.

The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems qr components. No new.equipment will be installed.

As a result, the proposed change will not, create any credible new failure mechanisms; malfunctions, or accident initiators not considered in the design and licensing bases ..

NLS2019020 Page 4 of 4 Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. This will result in a change to a safety limit, but will not result in a significant reduction in the margin of safety provided by the safety limit. As discussed in the application, changing the SLM CPR methodology to one based on a 95%

probability at a 95% confidence level that no fuel rods experience transition boiling during an anticipated transient, instead of the current limit based on ensuring that 99.9%

of the fuel rods are not susceptible to boiling transition, does not have significant effect on the plant response to any analyzed accident. The SLM CPR and the TS Limiting Condition for Operation on MCPR continue to provide the same level of assurance as the current limits and do not reduce a margin of safety.

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

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

3 .2 Conclusion In 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.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed change 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 change 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 change.

NLS2019020 Page 1 of 5 Attachment 2 Proposed Technical Specifications Changes (Mark-up)

Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 Revised Pages 2.0-1 5.0-21 Pages Included for Reference 3.2-2 3.2-3

Sls 2.0 2.0 SAFETY LIMITS (Sls) 2.1 Sls 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall bes; 25% RTP.

2.1.1.2 With the reactor steam dome pressure c: 785 psig and core flow ~ 10%

rated core flow: . _.----[1.07 .~

MCPR shall be c: 1.1~eiFSul4ion loop operation or~ 1.Hi for single msiFSulatisn lssp sper:ation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

• 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 1337 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

}

2.2.1 Restore compliance with all Sls; and 2.2.2 Insert all insertable control rods.

Cooper 2.0-1 Amendment No. -zs+

MCPR 3.2.2 No Changes. Included for Reference.

3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY: THERMAL POWER~ 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within limits. A.1 Restore MCPR(s) to within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits.

B. Required Action and B.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion Time <25% RTP.

not met SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within 12 limits specified in the COLR.

• hours after ~ 25%

RTP In accordance with the Surveillance Frequency Control Program (continued)

Cooper 3.2-2 Amendment No. 258

MCPR

.-. 3.2.2 No Changes. Included for Reference.

I SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.2.2.2 Determine the MCPR limits. Once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each completion of SR 3.1.4.1 Once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each completion of SR 3.1.4.2 Cooper 3.2-3 Amendment No. 178

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 confonnance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

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

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.

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

~---'-----L---~

3.3.2.1. , and MCPR99 _9%

for Specification

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

3.2.2.

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

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

(continued)

Cooper 5.0-21 Amendment N o . ~

NLS2019020 Page 1 of 3 Attachment 3 Revised Technical Specifications Pages Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 Revised Pages 2.0-1 5.0-21

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall bes 25% RTP.

2.1.1.2 With the reactor steam dome pressure .:: 785 psig and core flow .:: 10%

rated core flow:

MCPR shall be.:: 1.07. .I 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

• 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 1337 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

Cooper 2.0-1 Amendment No.

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.

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:

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

(continued)

Cooper 5.0-21 Amendment No.

NLS2019020 Page 1 of7 Attachment 4 Proposed Technical Specifications Bases Changes (Mark-up) -

Information Only Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 Revised Pages B 2.0-1 B 2.0-2 B 2.0-3 B 2.0-4 B 3.2-4 B 3.2-5

Reactor Core Sls B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND USAR, Appendix F (Ref. 1) establishes, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and abnormal operational transients.

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2 for General Electric Company (GE) fuel. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is r.elated to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,

MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. Hie MGPR fuel eladdiflg i1;tcgrity SL cAsures tl,st duriAg AOFffisl 019erstioA SAd durifl§ abROr1T1al open~tiorial trar,siel"lts, ~t least 99.9% of the fuel rods in the oore do not e>cperionoo traRsitien beiling.

This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95195 , which corresponds to a 95% probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.

(continued)

Cooper B 2.0-1 12118103

Reactor Core SLs B 2.1.1 BASES BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water {zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

The reactor vessel water level SL ensures that adequate core cooling capability is maintained during all MODES of reactor operation. Sufficient reactor vessel water level ensures adequate margin for effective action in the event of a level drop.

APPLICABLE SAFETY ANALYSES The Tech Spee SL is setr---l::ft1e-i'tt@-t-efe5-'tefl---e-tP"H:-e'l"it-eft-at-&tr-a--Mt;t"tt-+t111ttt;"""'T5'-'l;-6--8e-generically on a fuel product MCPR correlation basis as the MCPR which The Reactor Protection System setpoints {LCO 3.3.1.1, corresponds to a 95% 11 Reactor Protection System (RPS) Instrumentation"), in probability at a 95% combination with the other LCOs, are designed to prevent any confidence level that anticipated combination of transient conditions for Reactor transition boiling will not Coolant System water level, pressure, and THERMAL POWER occur, referred to as level that would result in reaching the MCPR limit.

SLMCPR95195.

2.1.1.1 Fuel Cladding Integrity General Electric Company

{GE) Fuel GE critical power correlations are applicable for all critical power calculations at pressures~ 785 psig and core flows~ 10% of rated flow. For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevatio~ head, the core pressure

{continued)

Cooper B 2.0-2 RevisieA g L _ _ _ _____ _

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fyel Cladding Integrity General Electric Company SAFETY ANALYSES (GEl Fue 1 (continued)

The Technical drop at low power and flows will always be> 4.5 psi.

Specifications SL Analyses3 (Ref. 2) show that with a bundle flow of value is dependent 28 x 10 lb/hr, bundle pressure drop is nearly on the fuel product independent of bundle power and has a value of line and the 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be> 28 x 103 lb/hr. Full scale ATLAS test corresponding data taken at pressures from 14.7 psia to 800 psia MCPR correlation, indicate that the fuel assembly c:--itical power at this which is cycle flow is approximately 3.35 MWt. With the design independent. The peaking factors, this corresponds to a THERMAL POWER value is based on > 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure< 785 psig is conservative.

the Critical Power Ratio (CPR) data statistics and a 2.1.1.2 MCPR GE Fuel 95% probability with 95% The fuel cladding integrity SL is set such that no confidence that significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel rods are not damage are not directly observable during reactor operation, susceptible to the thermal and hydraulic conditions that result in the boiling transition, onset of transition boiling have been used to mark the referred to as beginning of the region in which fuel damage could occur.

Although it is recognized that the onset of transition SLMCPRes,es- boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to The SL is based on occur h~s ~een adopted a~ a convenient limi~. He~1e 1,el", t~e GNF2 fuel. For cores loaded with a single fuel product line, the SLMCPR95, 95 is the MCPRes,es for the fuel type. For cores loaded with a mix of The MGPR SL is eetermiRea ijSiR9 a statistieal meael that applicable fuel GembiRes all t~e YRG9rtaiRties iR eperatiRg parametgrs aRd types, the the f)receeijres \:1Sed te cal m~la.te criti eal f)awer. TR-e-SLMCPRes,es is f)rahaei l ity af the ecc\:lrreRGe ef heiliRg traRsitian is aeterminee ijSiRg the appreved GeReral Electric Critical based on the largest Pewel" eel"l"elatieRs. getails af the fijel claaeiRg iRte§l"ity (i.e., most limiting) of the MCPR values for the fuel product (continued) lines that are fresh or once-burnt at the start of the cycle. B 2.0-3 Revision a

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR GE fuel (continued)

SAFETY ANALYSES SL esleulatiefl are giveA iA Refereflee 2. Refereflee 2 else iflelueles a tabulatieR ef tho uRsoFtaiRtioe used iR tho aeterlTliR!*ieR ef tl'ie MGPR SL aAEI ef tl"lc flOA'liAal values ef tl"le parameters useEI iA tl"lc MGPR SL statistieal aflalysis.

2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active irradiated fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the care height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action. Fuel zone zero (FZZ) is used as a reference point which corresponds to at or above top of the active irradiated fuel.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to prevent the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core*operates-;:

within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

• 6 (continued)

Cooper B 2.0-4 June 10, 1999

.,;;.,,.1*.-.

MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. Tl ,e MCPR Safety Lir11it (SL) is set such that 99.9% of the fuel rods aooid boiling tlansition if H,e lit 11it is I aot oiolated {refer to ti ,e Bases for SL 2.1. 1.2}. The operating limit MCPR is established to ensure that no fuel damage results during abnormal operational trans* . Although fuel damage does not

, and that 99.9% of n

• r I a fuel rod actually experienced boiling transition (Ref.

1), the critical power at which boiling transition is calculated to occur has the fuel rods are been adopted as a fuel design criterion.

not susceptible to boiling transition if The onset of transition boiling is a phenomenon that is readily detected the limit is not during the testing of various fuel bundle designs. Based on these violated. experimental data, correlations have been developed to predict critical bundle power {i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters {e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

• APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the abnormal operational transients to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 6, and 7. To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields

~~ largest change in CPR (LlCPR). When the largest LlCPR is added to

~ the"MCPR SL, the required operating limit MCPR is obtained. b' d 'th com me w1 MCPR99 _9 % is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical Power correlations. Details of the MCPR99 _9 % calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPRg 9 _9 % statistical analysis.

Cooper B 3.2-4 09/11/15

MCPR B 3.2.2 BASES APPLICABLE SAFETY ANALYSIS (continued) ~ the MCPR99 _9o/o value and The MCPR operating limits aerived from e transient analysis~e r-=:.-1 dependent on the operating core flow and power state (MCPRt a~

MCPRp, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 6 and 7).

Flow dependent MCPR limits are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 8) to analyze slow flow runout transients. The operating limit is dependent on the ma~x1_*m_u_m_co_r_e_ _ _ _~

flow limiter setting in the Recirculation Flow Control System. by approved transient analysis Power dependent MCPR limits (MCPRp) are determined ~liAli[L:oiit'.ffiie=""'----~

eAe diFReAsieAal ti:aAsieAt sede (Ref. 9). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure (MCPR99.9% value, scrams are bypassed, high and low flow MCPRp operating limits are MCPRt values, and rovided for operating between 25% RTP and the previously mentioned MCPRp values) b ss power level.

satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 10).

LCO limits {for each type cf fuel at rated power and flow) specified in the COL are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPRr and MCPRp l i m i t s ~ , which are based on the MCPR

- - - - - - - - - - - - - - .-----------199.9% limit specified in the COLR.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25%

RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels Cooper B 3.2-5 oe11111e I