LR-N19-0038, Application to Revise Technical Specifications to Adopt TSTF-564, Safety Limit MCPR

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

ML19112A214
Person / Time
Site:	Hope Creek
Issue date:	04/22/2019
From:	Carr E Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR H19-02, LR-N19-0038
Download: ML19112A214 (21)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 PSEG Nuclear LLC 1 0 CFR 50.90 LR-N1 9-0038 LAR H1 9-02 APR 2 2 2019 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555-0001 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 NRC DOCKET NO. 50-354

Subject:

Application to Revise Technical Specifications to Adopt TSTF-564, "Safety Limit MCPR" Pursuant to 1 0 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) for Renewed Facility Operating License No.

NPF-57 for Hope Creek Generating Station.

PSEG requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Hope Creek Generating Station 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 a SL. provides a description and assessment of the proposed changes. Enclosure 1 provides the existing TS pages marked to show the proposed changes. Enclosure 2 provides existing TS Bases pages marked to show the proposed changes for information only.

Approval of the proposed amendment is requested by September 20, 201 9, in order to support the Hope Creek Refueling Outage H1 R22. Once approved, the amendment shall be implemented prior to restart following the Refueling Outage.

There are no regulatory commitments contained in this letter.

95-2168 REV. 7/99

APR 2 2 2019 Page 2 1 0 CFR 50.90 LR-N 1 9-0038 In accordance with 1 0 CFR 50.91 , " Notice for public comment; State consultation,"

paragraph (b), PSEG is notifying the State of New Jersey of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

If you have any questions or require additional information, please contact Mr. Lee Marabella at (856) 339-1 208.

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

Executed on t-tb:J--:/1 CJ (Date)

Eric Carr Site Vice President Hope Creek Generating Station Attachment

1. Description and Assessment

Enclosures:

1. Proposed Technical Specifications Changes (Mark-Up) For Hope Creek

2. Proposed Technical Specifications Bases Changes (Mark-Up) For Information Only For Hope Creek cc: Administrator, Region I, NRC Project Manager, NRC NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. T. Cachaza, Station Commitment Tracking Coordinator

LR- N19-0038 Attachment 1 Description and Assessment

LR-N1 9-0038 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 License Amendment Request to Revise Technical Specifications to Adopt TSTF-564, "Safety Limit MCPR" Table of Contents 1 .0 DESCRIPTION.......................................................................................... 2 2.0 ASSESSMENT.......................................................................................... 2 2.1 Applicability of Safety Evaluation............................................................... 2 2.2 Variations............................................................................................. 2

3.0 REGULATORY ANALYSIS

........................................................................... 3 3.1 No Significant Hazards Consideration Analysis............................................ 3 3.2 Conclusion........................................................................................... 4 4.0 ENVIRONMENTA L EVALUATION .................................................................. 4 1

LR-N1 9-0038 1 .0 DESCRIPTION PSEG Nuclear (PSEG) requests adoption of TSTF-564, "Safety Limit MCPR," Revision 2, which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Hope Creek Generating Station (Hope Creek) 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 a S L.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation PSEG has reviewed the safety evaluation for TSTF-564 provided to the Technical Specifications Task Force in a letter dated November 1 6, 201 8. This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-564. PSEG has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRC staff are applicable to Hope Creek and justify this amendment for the incorporation of the changes to the Hope Creek TS.

The Hope Creek reactor is currently fueled with GNF2 and GE1 4 fuel bundles. For cores loaded with a mix of applicable fuel types the SLMCPR95195 is based on the most limiting (i.e.

largest) MCPR95195 value for all fresh and once burnt fuel types in the core. Current fresh and once burnt fuel in the Hope Creek reactor is GNF2. The proposed Safety Limit in SL 2.1 .2 is 1 .07, consistent with Table 1 of TSTF-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_90fo. Technical Specification 6.9.1 .9, "Core Operating Limits Report," is revised to require the MCPR99_9% value to be included in the cycle-specific Core Operating Limits Report (COLR).

2.2 Variations PSEG is proposing the following variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs safety evaluation dated November 1 6, 201 8.

The Hope Creek TS are based on NUREG-01 23, Standard Technical Specifications for General Electric Boiling Water Reactors, and, therefore, the wording, numbering and format vary slightly from NUREG-1 433, Standard Technical Specifications General Electric BWR/4 Plants, shown in TSTF-564, Revision 2, and the applicable parts of the NRC staff's safety evaluation.

Specifically, section 2.1 .2 THERMAL POWER, High Pressure and High Flow of the Hope Creek TS is numbered 2.1 .1 .2 in TSTF-564. Section 3/4.2.3 MINIMUM CRITICAL POWER RATIO of the Hope Creek TS is numbered 3.2.2 in TSTF-564. Section 6.9.1 .9 CORE OPERATING LIMITS REPORT of the Hope Creek TS is numbered 5.6.3 in TSTF-564. In addition, Hope Creek TS 6.9.1 .9 currently lists the individual specifications that address core operating limits documented in the COLR. Reference to the MCPR99_9% value is being added as a note to be consistent with the existing TS format. These administrative differences and editorial changes made in reference to the Core Operating Limits Report do not affect the applicability of TSTF-564 to the Hope Creek TS.

2

LR-N1 9-0038 The Hope Creek TS contain requirements that differ from the Standard Technical Specifications on which TSTF-564 was based, such as, TS 3/4.4.1 , Recirculation Loops. LCO 3.4.1 .1 ACTION a.1 .c) requires revision as a result of the change to TS 2.1 .2. Reference to Specification 2.1 .2 is being deleted, as the MCPR95195 is not dependent on the number of recirculation loops in operation, and replaced with a reference to the MCPR limit listed in the COLR for single loop operation. No similar change was made in the TSTF-564 model application because the associated existing Standard Technical Specification LCO 3.4.1 .b.

already references the COLR. This change would align the Hope Creek TS with the current Standard Technical Specification wording and does not affect the applicability of the TSTF-564 Revision 2 justification.

PSEG is not providing Revised Technical Specification Pages with this submittal. The pages will be provided in support of final NRC approval.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis PSEG Nuclear (PSEG) requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Hope Creek Generating Station ( Hope Creek) 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 6.9.1 .9, "Core Operating Limits Report," is revised to require the current SLMCPR value to be included in the Core Operating Limits Report (COLR).

PSEG has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 1 0 C FR 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 The proposed amendment revises the TS SLMCPR and the list of individual specifications that address core operating limits to be included in the Core Operating Limits Report (COLR). The S LMCPR 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 protected 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 functions 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.

3

LR- N1 9-0038

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

Response: No The proposed amendment revises the TS SLMCPR and the list of individual specifications that address core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems or components (SSCs). 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.

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

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

Response: No The proposed amendment revises the TS SLMCPR and the list of specifications that address 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 SLMCPR methodology to one based on a 95% probability with 95% confidence 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 a significant effect on plant response to any analyzed accident. The SLMCPR and the TS Limiting Condition for Operation (LCO) on MCPR continue to provide the same level of assurance as the cur rent 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, PSEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 1 0 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 1 0 CFR 20, or 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 4

LR-N1 9-0038 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 1 0 CFR 51 .22( c)(9).

Therefore, pursuant to 1 0 CFR 51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

5

LR-N19-0038 Enclosure 1 Proposed Technical Specifications Changes (Mark-Up)

For Hope Creek

LR-N1 9-0038 LAR H1 9-02 PROPOSED TECHNICAL SPECIFICATIONS CHANGES (MARK-UP)

The following Technical Specifications for Renewed Facility Operating License NPF-57 are affected by this change request Technical Specification

2. 1 . 2 2-1

3. 2. 3* 3/4 2-3

3. 4. 1 . 1 3/4 4-1

6. 9. 1 . 9 6-20

• Information Only TS Page

2.0 SAFETY LIMIT S AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Lo w Pressure or Low Flow 2.i.l THERMAL POWER shall not exceed 24% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 24% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 W i th reactor steam dome pressure greater th an 785 psig and core flow greater tha n 10% of rated flow:

r-41 07 I The MINIMUM CRITICAL POWER RAT IO ( MCPR ) shall be 2 \t&g .fer t*.m reeireulatieE:

leep eperatien and shall be 1.12 fer sinle reeirelatien leep eperatien.

APPLICABILITY: OPERATIONAL C O N DIT I O N S 1 and 2.

ACT ION :

With reactor steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow and the MCPR below the values for the fuel stated in LCO 2.1.2, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

HOPE CREEK 2-1 Amendment No. 2-H

• • NO CHANGES THIS PAGE**

(INFORMATION ONLY)

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL PO WER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit sp ecified in the CORE OPERATING LIMITS REPORT.

APP LICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 24% of RATED THERMA L POWER.

ACTION:

a. With the end -of-c ycle recirculation pump trip system inoperable per Specification 3.3.4.2, op erat i o n may continue p rovided that, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, MCPR is determined to be greater than or eq ual to the EOC-RPT i nop e ra ble limit specified in the CORE OPERATING LIMITS REPORT.

• b. With MCPR less than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT, init iate corrective action within 15 minutes and restore MCPR to within the required limit with i n 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 24% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, shall be determined to be equal to or g reater than the appl i ca ble MCPR limit specified in the CORE OPERATING LIMITS REPORT:

a. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is greater than or equal to 24% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Progra m thereafter .

b. Initially and in accordance with the Surveillance Frequency C ontrol Program when the reactor is operating with a LIMITI NG CONTROL ROD PATTERN for MCPR.

HOPE CREEK Amendment No. 187

3/4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4. 1. 1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION: limit to a value specified in the CORE OPERATING LIMITS REPORT for single loop operation

a. With one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a) Place the recirculation flow control system in the Local Manual mode, and b) Reduce T HERMAL POWER tos 59.89% of RATED THERMAL POWER, and c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR)

Safety Limit per Specification 2.1.2, and d) . Reduce the AVERAGE PLANAR LINEAR HEAT GENERATION RATE{APLHGR) limit to a value specified in the CORE OPERATING LIMITS REPORT for single loop operation, and e) Reduce the LINEAR HEAT GENERATION RATE (LHGR) limit to a value specified in the CORE OPERATING LIMITS REPORT for single loop operation, and f) Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and g) Perform surveillance requirement 4.4.1.1.2 if THERMAL POWER is s 38% of RATED THERMAL POWER or the recirculation loop flow in the operating loop iss 50% of rated loop flow.

2. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the Average Power Range Monitor (APRM) Scram Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specification 2.2.1; otherwise, declare the APRM channel INOPERABLE and take the action of RPS Instrumentation TS 3.3. 1 ACTION a.

3. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the APRM Control Rod Block Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specification 3.3.6; otherwise declare the APRM channel INOPERABLE and take the action of Control Rod Block Instrumentation TS 3.3.6 ACTION a and b.

See Special Test Exception 3.10.4.

HOPE CREEK 3/4 4-1 Amendment No. 242

ADMINISTRATIVE CONTROLS 6.9.1.8 Deleted CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the PSEG Nuclear LLC generated CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following Technical Specifications:

2.2 Reactor Protection System Instrumentation Setpoints 3/4.1.4.3 Rod Block Monitor 3/4.2.1 Average Planar Linear Heat Generation Rate 3/4.2.3 Minimum Critical Power Ratio 3/4.2.4 Linear Heat Generation Rate 3/4.3.1 Reactor Protection System Instrumentation 3/4.3.6 Control Rod Block Instrumentation The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC as applicable in the following document

1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-11)"

The CORE OPER/1.TING LIMITS REPORT COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT COLR (i.e., report number title, revision, date, and any supplements).

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPER.A.TING LIMITS REPORT COLR, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.10

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Limiting Condition for Operation Section 3.4.6, "RCS Pressure/Temperature Limits"

2. Surveillance Requirement Section 4.4.6, "RCS Pressure/Temperature Limits"

- The MCPR99.9% value, for both Two RecirculationLoop Operation and Single Recirculation Loop Operation, used to calculate the LCO 3.2.3, Minimum Critical Power Ratio, limit shall be specified in the COLR.

HOPE CREEK 6-20 Amendment No.

LR- N1 9-0038 Enclosure 2 Proposed Technical Specifications Bases Changes (Mark-Up) For Information Only For Hope Creek

LAR H1 9-02 LR-N1 9-0038 PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES (MARK-UP)

FOR INFORMATION ONLY The following Technical Specification Bases for Renewed Facility Operating License NPF-57 are affected by this change request:

Technical Specification Bases Page B 2-1 B 2-2 B 3/4 2-2 B 3/4 4-1

2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

the limit specified in Specification 2.1 .2 ing, reactor pressure vessel and primary system piping are the principal barr'ers to the release of radioactive materials to the environs. Safety imits are established to protect the integrity of these barriers during rmal plant operations and anticipated transients. The fuel cladding integri y Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a tep-back approach is used to establish a Safety Limit such that the MCPR is recirculation operation en4 gingle recirculation operation. These MCPR values represent a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cr9cking.

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 and the Limiting Safety System Settings. 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 Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition intended by design for planned operation.This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95;95, which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur.

2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the applicable NRC-approved critical power correlations are not valid for all critical power calculations performed at reduced pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to BOO psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 5 0% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 24% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

Amendment No.

HOPE CREEK B 2-1 (PSEG Issued)

SAFETY LIMITS BASES the on set of tran sition 2.1.2 THERMAL The fuel cladding integrity Sa ety Limit is set such that no fu damage is calculated to occur if the 'mit is not violated. Since t parameters which result in fuel damage re not directly observable d reactor operation, the thermal and hydra lie conditions resulting in a deparEttre frem nttcleate boiling have been used to mark the beginning region where fuel damage could occur. Alt ough it is recognized that departttre frem nueleate boiling would not n cessarily result in damage to BWR fuel rods, the critical power at which boili transition is calculated to occur has been adopted as a convenient limit. Heuever, the uaeertainties in meniterin§ tee cere eperatin§ state and in the procedures used to calculate the critical pemer :t:esttH: in an ttf\certainty in the "--altte ef the critical peHer. '3?herefere, the fuel ela:deiin§ integ:t:"ity Safety Limit is defined as the CPR in the limitiA§ fuel assemely feF Hhieh mere than 99.9% ef Ehe fuel reds in the cere are expected te aveia beilin§ transition eensiderin§ the ewer distribution Hithin the cere afld all uneertahrties.

'3?he SafeEy Limit !!CPR is determined ttsia a stEistiel model that eem:biaes all ef the uaeertaiflties in eeratia§ arameters and in the procedures used to calculate critical po\#er. Caleulatiea of the Safety Limit HCPR is defined ifl Refereace 1.

Reference:

1. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A (The approved revision at the time the reload analyses are performed.

The approved revision number shall be identified in the CORE OPERATING LIMITS REPORT. )

HOPE CREEK B 2-2 Amendment No. +/--§-4

INSERT A The Technical Specification Safety Limit value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95;gs.

For cores with a single fuel product line, the SLMCPR95;95 is the MCPR95;95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95;95 is based on the largest (i.e. most limiting) of the MCPR values for the fuel product lines that are fresh or once burnt at the start of the cycle.

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 M I NIMUM CRITICAL POWER RATIO

• The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established aladdin interity afety Limit MCPR99.9%, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the afety Limit MCPR99.9% at any time during the transient assuming instrument trip setting given in Specification 2.2.

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 (Reference 1). The MCPR99.9% is expected to always be greater than the MCPR95;95 because the MCPR99.9% includes uncertainties not factored into the MCPR95;95 and the 99.9 percent probability basis associated with the MCPR99.9% is more conservative than the 95 percent probability at a 95 percent confidence level basis for the MCPR95;95.

To assure that the ftael olaeein interity afety Limit MCPR99*9% is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR) . The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the afety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPR(f) and MCPR(p),

respectively) to ensure adherence to fuel design limits during the worst transient with moderate frequency that is postulated in Chapter 15 of the UFSAR.

Flow dependent MCPR limits (MCPR(f)) are determined by steady state methods using a core thermal hydraulic code (Reference 1). MCPR(f) curves are provided based on the maximum credible flow runout transient (i.e.,

runout of both loops) 1 The methods described in Reference 1 are used to determine the power dependent MCPR limits (MCPR(p)). 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 scram limits are bypassed, high and low MCPR(p) operating limits are provided for operation between 24t of RATED THERMAL POWER and the bypass power levels.

HOPE CREEK B 3/4 2-2 Amendment No . l-+4 (PSEG Issued)

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single loop operation is permitted if the MCPR fuel cladding Safety Limit is increased as noted by Speoification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2 respectively. APLHGR limits are decreased by the factor given in the CORE OPERATING LIMITS REPORT (COLR), LHGR limits are decreased by the factor given in the COLR, and MCPR operating limits are adjusted as specified in the COLR.

The Average Power Range Monitor Scram and rod block functions vary as a function of recirculation loop drive flow (w). The effective drive flow correction term (b.w) is defined as the difference in indicated drive flow (in percent of drive flow which produces rated core flow) between two loop operation (TLO) and single loop operation (S LO) at the same core flow. b.w is based on a physical phenomenon and represents the amount of drive flow from the active loop that flows backwards through the inactive loop's jet pumps during SLO. The flow input to the APRM STP Scram function Allowable Value (AV) and Nominal Trip Set Point (NTSP) is adjusted by b.w during S LO to account for this phenomenon.

The form of the function equation is: Slope x (Flow [w] - Flow Offset [b.w]) + Power Offset.

GEH's setpoint methodology is described in NEDC-33864P Appendix P, P1 and P2 (VTD 432598). The methodology also accounts for increased uncertainty in the idle recirculation loop flow signal, which requires the NTSP to be further from the AV under S LO than it is under T LO.

This is accomplished by reducing the power offset term for the APRM STP-Upscale RPS Trip (Table 2.2.1-1 Function 2.b):

TLO AV: 0.56w + 60%

TLO NTSP: 0.56w + 58%

S LO AV: 0.56(w- b.w) + 60%

S LO NTSP: 0.56(w- b.w) + 57%

When the S LO mode is manually enabled the NUMAC APRM instrument applies an offset term to the flow signal. To avoid an additional action to manually adjust the power offset (from 58% to 57 %), the S LO NTSP equation is solved for the same power offset as the TLO NTSP equation.

Using b.w = 9% yields a flow offset of 10.8% to maintain the power offset at 58%:

0.56(w- 9%) + 57%:::: 0.56(w- 10.8%) + 58%

This 10.8% flow offset term is defined as the "S LO Setting Adjustment" (the actual value is 10.79 but it is rounded up to one decimal place for conservatism since the S LO Setting Adjustment is programmed to one decimal place in the NUMAC equipment). This term is applied to the NTSP during S LO by the NUMAC APRM to both account for the 9% b.w flow offset and the increased margin required to the AV. The b.w and S LO Setting Adjustment values have been inserted into the APRM STP-Upscale equations in Table 2.2.1-1.

This same methodology is also applied to the APRM STP-Upscale Rod Block Trip (Table 3.3.6-2 Function 2.a).

HOPE CREEK B 3/4 4-1 Amendment No.

(PSEG Issued)