LR-N03-0511, Request for Change to Technical Specifications Regarding Fuel Vendor Change

Request for Change to Technical Specifications Regarding Fuel Vendor Change

ML040050272
Person / Time
Site:	Hope Creek
Issue date:	12/24/2003
From:	John Carlin Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR H03-08, LR-N03-0511, NEDC-33107P
Download: ML040050272 (27)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG DEC 2 4 2003 NuclearLLC LR-N03-0511 LCR H03-08 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS FUEL VENDOR CHANGE HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354

Reference:

Margaret E. Harding (Global Nuclear Fuel) letter to NRC, "Transmittal of GNF-A Proprietary Report, NEDC-33107P, 'GEXL80 Correlation for SVEA96+ Fuel,' dated September 2003," dated November 24, 2003 Pursuant to 10 CFR 50.90, PSEG Nuclear LLC (PSEG) hereby requests a revision to the Technical Specifications (TS) for the Hope Creek Generating Station. In accordance with 1 OCFR50.91 (b)(1), a copy of this submittal has been sent to the State of New Jersey.

The proposed changes support the use of General Electric Company (GE) fuel and reload analysis methods beginning with the upcoming Cycle 13. The proposed changes are consistent with NUREG-1433, "Standard Technical Specifications (STS) General Electric Plants, BWR/4," Revision 2.

PSEG has evaluated the proposed changes in accordance with 10CFR50.91(a)(1),

using the criteria in 1 OCFR50.92(c), and has determined this request involves no significant hazards considerations. An evaluation of the requested changes is provided in Attachment 1 to this letter. The marked up Technical Specification pages affected by the proposed changes are provided in Attachment 2.

PSEG plans to include GE14 fuel in the reload for Cycle 13, which is currently scheduled to begin in Fall 2004. PSEG therefore requests approval of the proposed License Amendment by September 16, 2004, to be implemented within 60 days of the completion of the Hope Creek Fall 2004 refueling outage.

The reference letter requests NRC review and approval of the GEXL80 correlation for modeling the Westinghouse SVEA96+ fuel design by March 31, 2004 to support reload analysis for Cycle 13.

rfm 95-2168 REV. 7/99

Document Control Desk DEC 2 4 2003 LR-N03-0511 PSEG proposes to meet with the staff at their earliest convenience to review the plans and schedule for transition to GE14 fuel at Hope Creek.

Should you have any questions or require additional information, please contact Mr.

Paul Duke at (856) 339-1466.

I declare under penalty of perjury that the foregoing is true 7d correct.

Executed on 2.qg(aal 2Oc)3 (date)

VP resident - Nuclear Assessments Attachments (2)

Document Control Desk DEC 2 4 2003

LR-N03-0511 C

Mr. H. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. J. Boska, Project Manager - Hope Creek U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, New Jersey 08625

Attachment I LR-N03-0511 LCR H03-08 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS FUEL VENDOR CHANGE Table of Contents

1.

DESCRIPTION......................................

.1

2.

PROPOSED CHANGE 1

3.

BACKGROUND.......................................

2

4.

TECHNICAL ANALYSIS...................................

3

5.

REGULATORY SAFETY ANALYSIS.....................................

4 5.1 No Significant Hazards Consideration.......................................

4 5.2 Applicable Regulatory Requirements/Criteria.......................................

6

6.

ENVIRONMENTAL CONSIDERATION

...................................... 6

7.

REFERENCES.......................................

6

Attachment I LR-N03-0511 LCR H03-08

1.

DESCRIPTION This letter is a request to amend Operating License NPF-57 for the Hope Creek Generating Station. The proposed changes are being made to support the introduction of GE14 fuel. To facilitate this new fuel introduction (NFI), NRC approved GE calculation methodologies will be used exclusively to determine fuel thermal limits and reload transient analysis results. The changes to the Hope Creek Technical Specifications: 1) reflect the exclusive use of GE methods by removing references to other methodologies, 2) modify and add Action statements to provide further thermal limit control during Single Loop Operation (SLO) consistent with GE methodology requirements, 3) revise TS Definitions and TS requirements for average planar linear heat generation rate (APLHGR) consistent with NUREG-1433, Standard Technical Specifications (STS) General Electric Plants, BWR/4," Revision 2 (Reference 1), and 4) correct an error in TS 6.9.1.9 introduced during implementation of a previous amendment. The references for TS Section 6.9.1.9 would be identified in the format prescribed in NUREG-1433, Rev. 2.

The TS Bases would also be revised to be consistent with GE methodology requirements. NRC approval for the GE methodologies and requirements was provided in Amendment 26 to GESTAR II, and included in GESTAR II Revision 14, June 2000.

The proposed changes are required to support the transition to General Electric Company (GE) fuel and reload analysis methods beginning with the upcoming Cycle 13 which will begin in Fall 2004.

2.

PROPOSED CHANGE The marked up pages for the proposed changes to the Technical Specifications are included in Attachment 2 of this submittal.

1.

One reference to ABB/CE calculational methodology would be deleted from the list of analytical methods that are used to determine the core operating limits in TS Section 6.9.1.9, "Core Operating Limits Report" (COLR). The references for TS Section 6.9.1.9 would be renumbered and identified in the format prescribed in NUREG-1433, Rev. 2.

2.

Limiting Condition for Operation (LCO) 3.4.1.1, "Recirculation Loops,"

would be revised as follows:

a.

Action a.1.d would be revised to require the Average Planar Linear Heat Generation Rate (APLHGR) limit to be reduced to a value specified in the Core Operating Limits Report for SLO.

Attachment I LR-N03-0511 LCR H03-08

b.

A new Action a.1.e would be added to require the LHGR limit to be reduced to a value specified in the COLR during SLO.

The associated TS Bases would also be revised to reflect the use of APLHGR and LHGR limits during SLO.

3.

Limiting Condition for Operation (LCO) 3.2.1, "Average Planar Linear Heat Generation Rate," would be revised consistent with NUREG-1433,

'Standard Technical Specifications (STS) General Electric Plants, BWR/4," Revision 2 (Reference 1). Specifically, the references to fuel type and average planar exposure would be deleted. TS 1.2, the definition for "Average Planar Exposure" would be deleted. The definition for "Average Planar Linear Heat Generation Rate" (APLHGR) in TS 1.3 would be revised consistent with NUREG-1433 to be applicable to the GE14 fuel design.

4.

A reference in TS 6.9.1.9 to CENPD-397-P-A, "Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology," which was inadvertently deleted in a previous amendment, would be restored.

Changes to the TS Bases would also be made to reflect the application of NRC approved GE methodologies. The marked up Bases pages are also included in of this submittal.

3.

BACKGROUND

1.

For the current operating cycle, the Hope Creek core contains a mixture of Westinghouse SVEA96+ and GE9B fuel. Core operating limits were determined using NRC approved Westinghouse methodology. PSEG plans to load GE14 fuel during the Hope Creek Fall 2004 refueling outage.

NRC approved GE calculation methodologies will be used exclusively to determine fuel thermal limits and reload transient analysis results.

2.

The current TS required action to reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value specified in the COLR for single loop operation is inconsistent with the approved GE methodology which establishes limits on APLHGR and LHGR for single loop operation.

3.

Limiting Condition for Operation (LCO) 3.2.1, "Average Planar Linear Heat Generation Rate," refers to APLHGR limits for each fuel type as a function of average planar exposure. The APLHGR limits are established in accordance with the approved analytical methods listed in TS 6.9.1.9.

The LCO 3.2.1 references to fuel type and average planar exposure are not needed since this information is located in the COLR.

Attachment I LR-N03-0511 LCR H03-08

4.

TS Amendment 131 revised TS 6.9.1.9 to add a reference to Topical Report CENPD-397-P-A, "Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology." The reference was inadvertently deleted during preparation of the retyped pages for HC TS Amendment 145.

4.

TECHNICAL ANALYSIS

1.

TS 6.9.1.9 identifies the previously reviewed and approved analytical methods used to determine the core operating limits. The proposed change deletes one reference to ABB/CE calculational methodology and retains the reference to NEDE-2401 1-P-A which will be used exclusively to determine core operating limits beginning with Cycle 13. The references for TS Section 6.9.1.9 will be identified in the format consistent with NUREG-1433, Rev. 2. TS 6.9.1.9 will be revised to state that the COLR will contain the complete identification of each of the TS referenced topical reports used to prepare the COLR.

2.

The proposed changes to LCO 3.4.1.1, Actions a.1.d and a.1.e are consistent with the approved GE methodology and ensure the appropriate adjustments are made to core operating limits for single loop operation.

3.

The proposed change to LCO 3.2.1 removes unnecessary detail from the TS while continuing to ensure fuel design limits are not exceeded. The APLHGR limits will continue to be established in accordance with the approved analytical methods listed in TS 6.9.1.9.

TS 1.2 is being deleted because the term "Average Planar Exposure" is being removed from LCO 3.2.1 and is not used elsewhere in the TS.

The proposed change to TS 1.3 is consistent with NUREG-1433 and makes the definition for "Average Planar Linear Heat Generation Rate" (APLHGR) applicable to the GE14 fuel design.

4.

The proposed change to TS 6.9.1.9 adding a reference to Topical Report CENPD-397-P-A is administrative in nature, correcting an error that was introduced during preparation of the retyped pages for HC TS Amendment 145.

The changes to the TS Bases are being made in support of the proposed TS changes and reflect the use of NRC reviewed and approved methods of evaluation.

Attachment I LR-N03-0511 LCR H03-08

5.

REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration PSEG Nuclear LLC (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 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 revised information and references relative to the fuel vendors calculation methodologies throughout the Technical Specifications are considered to be administrative in nature because they reflect the NRC approved methodologies to be used by PSEG Nuclear LLC and the fuel vendor to develop operating and safety limits for the fuel and core designs. The changes to the Recirculation System Action statements ensure the appropriate adjustments are made to core operating limits for single loop operation, and the Core Operating Limits Report (COLR) will still be developed in accordance with NRC approved methods. These proposed changes do not alter the method of operating the plant and have no effect on the probability of an accident initiating event or transient.

There are no significant increases in the radiological consequences of an accident previously evaluated. The basis of the COLR and the PSEG Nuclear LLC and fuel vendor calculation methodologies is to ensure that no mechanistic fuel damage is calculated to occur if the limits on plant operation are not violated. The COLR will continue to preserve the existing margin to fuel damage and the probability of fuel damage is not increased.

Therefore, the proposed change does not involve an increase in the probability or radiological 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.

These changes do not involve any new method for operating the facility,.any changes to setpoints, or any new facility modifications Attachment I LR-N03-0511 LCR H03-08 for the reload core operation. No new initiating events or transients result from these changes.

The revised information and references relative to the fuel vendor's calculation methodologies throughout the Technical Specifications are considered to be administrative in nature because they reflect the NRC approved methodologies to be used by PSEG Nuclear LLC and the fuel vendor to develop operating and safety limits for the fuel and core designs. The changes to the Recirculation System Action statements ensure the appropriate adjustments are made to core operating limits for single loop operation, and the COLR will still be developed in accordance with NRC approved methods.

Therefore, the proposed Technical Specification changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3.

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

Response

No.

The revised information and references relative to the fuel vendor's calculation methodologies throughout the Technical Specifications are considered to be administrative in nature because they reflect the NRC approved methodologies to be used by PSEG Nuclear LLC and the fuel vendor to develop operating and safety limits for the fuel and core designs. The changes to the Recirculation System Action statements ensure the appropriate adjustments are made to core operating limits for single loop operation, and the COLR will still be developed in accordance with NRC approved methods. The proposed changes will continue to ensure that the plant is operated within specified acceptable fuel design limits.

Therefore, the proposed Technical Specifications changes do not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed changes present 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.

LR-N03-0511 LCR H03-08 5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c)(2)(ii) Criterion 2 requires that TS LCOs include process variables, design features, and operating restrictions that are an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. LCO 3.4.1.1 requires adjustments to core operating limits for single loop operation. The proposed changes ensure the appropriate adjustments are made to core operating limits for single loop operation.

The proposed change to LCO 3.2.1 continues to ensure fuel design limits are not exceeded.

10 CFR 50.36(c)(5) requires that TS will include provisions relating to organization and management, procedures, record keeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. The proposed change to TS 6.9.1.9 lists the NRC-approved methods that will be used to determine core operating limits.

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.

6.

ENVIRONMENTAL CONSIDERATION PSEG has determined 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 a surveillance requirement. 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 criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

7.

REFERENCES

1.

NUREG-1433, "Standard Technical Specifications - General Electric Plants, BWRI4," Revision 2.

LR-N03-051 1 LCR H03-08 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. NPF-57 are affected by this change request:

Technical Specification Page Index i

1.2 1.3 3/4.2.1 3/4.4.1 6.9.1.9 References Bases 2.1.2 Bases 3/4.1.1 Bases 3/4.1.4 Bases 3/4.2.1 Bases 3/4.2.3 Bases 3/4.2.4 Bases 3/4.4.1 xxv 1-1 1-1 3/4 2-1 3/4 4-1 6-21 6-26 B 2-2 B 3/4 1-1 B 3/4 1-3 B 3/4 1-5 B 3/4 2-1 B 3/4 2-2 B 3/4 2-3 B 3/4 4-1 Insert A (TS 6.9.1.9)

LR-N03-051 1 LCR H03-08 I

1. NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-l I)"

2. CENPD-397-P-A, "Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology" The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e., report number title, revision, date, and any supplements).

Insert B (Bases 3/4.2.4)

The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. This specification assures that the Linear Heat Generation Rate (LHGR) in any fuel rod is less than the design linear heat generation even if fuel pellet densification is postulated. Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences (AOOs), and to ensure that the peak clad temperature (PCT) during postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46. Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to ensure that fuel system damage, fuel rod failure, or inability to cool the fuel does not occur during normal operation or the anticipated operational occurrences identified in Reference 1.

The analytical methods and assumptions used in evaluating the fuel system design limits are presented in Reference 1. The analytical methods and assumptions used in evaluating AOOs and normal operation that determine the LHGR limits are presented in Reference 1.

LHGR limits are developed as a function of exposure to ensure adherence to fuel design limits during the limiting AOOs. The exposure dependent LHGR limits are reduced by an LHGR multiplier (LHGRFAC) at various operating conditions to ensure that all fuel design criteria are met for normal operation and AOOs. A complete discussion of the analysis code is provided in Reference 2.

For single recirculation loop operation, the LHGRFAC multiplier is limited to a maximum value as given in the CORE OPERATING LIMITS REPORT. This maximum limit is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA.

DEFINITIONS -

SECTION 1.0 DEFINITIONS PAGE

1.

1 ACI O

N

1.

1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.1-1 1.4 CHANNEL CALIBRATION.l-1.5 CHANNEL CHECK.......................

1.6 CHANNEL FUNCTIONAL TEST.......................

.1-l 1.7 CORE ALTERATION....................................................

1-2 1.8 CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY......:............... 1-2 1.9 CORE OPERATING LIMITS REPORT..

1-2 1.10 CRlIICAL POWER RATIO................................................. 1-2 1.11 DOSE EQUIVALENT -131.......................

1-2 1.12 E-AVERAGE DISINTEGRATION ENERGY...

1-2 1.13 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME...

1-2 1.14 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.............

1-3 1.15 FRACTION OF LIMITING POWER DENSITY.

1-'

1.16 FRACTION OF RATED THERMAL POWER.

1-3 1.17 FREQUENCY NOTATION..................................................

1-3 1.18 IDENTIFIED LEAKAGE.1-3 1.19 ISOLATION SYSTEM RESPONSE TIME

.1-3 1.20 LIMITING CONTROL ROD PATTERN.........................................

1-3 1.21 LINEAR HEAT GENERATION RATE.1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST.......................................

1-4 1.Z3 MAXIMUM FRACTION OF LIMITING POWER DENSITY.1-4 1.24 MEMBER(S) OF TE PUBLIC...........................:..................

1-4 1.25 MINIMUM CRITICAL POWER RATIO.........................................

1-4 HOPE CREEK

1.

Amendment No.34 I

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.10 RECORD RETENTION.................................................... 6-21 6.11 RADIATION PROTECTION PROGRAM....................................... 6-23 6.12 HIGH RADIATION AREA.......................................

6-24 6.13 PROCESS CONTROL PROGRAM (PCP)....................................... 6-25 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)..

6-25 6.15 Deleted..

6-25 HOPE CREEK xxv Amendment No. 126

1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

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AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLEGR) shall be applicable to a specific planar height and in equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height dided by the of fuel rods in the fuel bundle CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel...

behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels -

the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

HOPE CREEK 1 -1 Amendment No. 90

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs/fc e h tid Cxe tue as /4ofunctvh of AERAGE P NAR VKWOS not eceeqrthe lim2its specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY:

OPERATIONAL CONDITION, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore APLEGR to within the required limits within 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 25S 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 4.2.2 All APLHGRs shall be verified to be equal to or less than the limits specified in the CORE OPERATING LIMITS REPORT:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 5 of RATED THERMAL POWER, and

c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

I HOPE CREEK 3/4 2-1 Amendment No. 126

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'EVe-L-CEA L AkfT 3/4.4 REACTOR COOLANT SYSTEM kc° a is is i

3/4.4.1 RECIRCULATION SYSTEM 0

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RECIRCULATION LOOPS

/at;v, LIMITING CONDITION R OPERATION 3.4.1.1 Two actor coolant system recirculation loops shall be in operation with:

a.

Total core flow greater than or equal to 45% of rated core flow, or

b.

THERMAL POWER less than or equal to the limit specified in Figure 3.4.1.1-1..

APPLIC ILITY:

OPERATIONAL CO 0

1 and ACTIO T

a.

With one reactor coo 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 r circulation flow control system in the Local

\\

~Manual mod, and b)

Reduce TH POWER to 70% of RATED THERMAL POWER, and C)

Increase he MINIMUM CRITICAL POWER RATIO (MCPR) Safety Lit pe pcification 2.1.2, and RM i

Aeage planar Line r H t Ge lerat ong

\\ t&&t7~>-PL{R)y=Fit o avalue specified ln the CORE

\\ ~RATINGr-LHITS REPORT for single loop operation, and e) f)

Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and I

40 g)

Perform surveillance requirement 4.4.1.1.2 if THERMAL POWER is 5 38% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is 5 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 Specifications 2.2.1 and 3.2.2; otherwise, with the Trip Setpoints and Allowable Values associated with one trip system not reduced to those applicable for single recirculation loop operation, place the affected trip system in the tripped condition and within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, reduce the Trip Setpoints and Allowable Values of the affected channels to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.2.2.

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 Specifications 3.2.2 and 3.3.6; otherwise, with the Trip Setpoint and Allowable Values associated with one trip function not reduced to those applicable for single recirculation loop operation, place at least one affected channel See Special Test Exception 3.10.4.

HOPE CREEK 3/4 4 -1 Amendment No. 126i

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT Continued) a The analytical methods used to determine the core operating 1 ts shall be revusly reviewed and approved by NRC as applicable in e

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

The CORE OPERATING LIMITS REPORT, 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.

SPECIAL REPORTS 6.9.2 Secial reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, within the time period specified for each report.

6.9.3 Violations of the requirements of the fire protection program described in the Final Safety Analysis Report which would have adversely affected the ability to achieve and maintain safe shutdown in the event of a fire shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, via the Licensee Event Report System within 30 days.

6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

SPECIAL REPORTS 6.10.2 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time interval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety.

c.

All REPORTABLE EVENTS submitted to the Commission.

d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

e.

Records of changes made to the procedures required by Specification 6.8.1.

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak tests and results.

HOPE CREEK 6-21 Amendment No. 131 l

ADMINISTRATIVE CONTROLS 6.15 TECHNICAL SPECIFICATION (S) BASES CONTROL PROGRAM This program provides a moans for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Eases of the TS shall be made under appropriate administrative controls and reviews.

b.

PSEG may make changes to the Bases without prior NRC approval provided the changes do not require either of the followings

1. A change in the TS incorporated in the License, or

2.

A change to the updated SAR or Bases that requires IRC approval pursuant to 10 CFR 50.59.

c.

Proposed changes to the Bases that require either condition of Specification 65.b above shall be reviewed and approved by the NRC prior to implementation.

d.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

e.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

Amendment No. 145 1 HOPE CREEK 6-26

BASES 2-:.2 L*-2MAL POWER, High Less re A^

.-e.uei cladoing integr::y 3ate:y s set s.n t damage s :a.: lated to occur i te _-:t s nt v:olate-.

Sin.te the parameters eh seult in fuel damage are n-.:

irectly bser-:acbe dur:-.:

reactor cperat:or.. the thermal and hydra_'.: cnditioas resultin; in a departure from nucleate boiling have een used to ark the beginning of :ne region where fuel damage could occur.

A.ithough it is recognized that 3 departure from nucleate boiling would ot necessarily result n damage tI E2R fuel rods, the critical power at which boiling transition is calculated t_

occur has been adopted as a convenient limit.

However, the uncertainties :n monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 9.9i of the fuel rds in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit CPR is determined using a statistical model that combines all of the uncertainties in operating parameters and in the procedures used to calculate criticaip er. Calculation of the Safe imit MCPR is defined in Reference 1 to GE f I and feren 2 for B u

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.)

2.

CENPD-

-P-A, efere i

Safety port fo Boiling ater Rea rs Reloa Fuel" (The roved rev aion at e time

.ae reload a ayses are erforme.

The pproved r isio n mber sha be identfied in/the CO OPERA NG LI GHS REPORT HOPE CREEK B 2-2 Amendment No.

126

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN SHUTDOWN MARGIN (SDM) requirements are specified to ensure:

a. The reactor can be made subcritical from all operating conditions, transients, and Design Bases Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

5DM can be demonstrated by using solely analytical method or pe orming a test.

SDM can be measured only by performing a test. A es involves collecting data with the reactor at a specified condition or series of conditions. The primary purpose of a SDM Demonstration is to ensure that SDM is equal to or greater than the DM Limit for a specific core exposure. The primary purpose of a SDM Measurement is to provide SD' in % delta k/k that can beused for:

1) ensuring that SDX is equal to or greater than the SDM Limit, for a range of core exposures, 2) determining the need for additional SDM Measurements during the cycle, 3) providing a benchmark for the core design (design vs. actual SDM),

and 4) providing a benchmark for potential future analysis of DM for such events as control rods incapable of full insertion.

This higher level of application requires that a SDM Measurement is determined from testing and not through solely analytical methods.

Since a SDM.

Measurement satisfies the primary purpose of a SD! Demonstration, it can considered a special type of SDM Demonstration.

All SDS Demonstrations involve some usage of analytical methods. The performance of tests lessens the usage of analytical methods, reduces'-

uncertainty in the results, and thus requires a smaller SDM Limit needed to show adequate SDM. At one.end of the spectrum is a series of local criticals where both SDM and the highest worth control rod are determined by test.

Although this technique has the minimum uncertainty and thus has the smallest SDM Limit, it still uses analytical methods to determine the worth of all the other control rods. At the other end of the spectrum is usage of solely analytical methods prior to core verification. This technique has the maximum uncertainty and 'thus has the largest SDM Limit.

The SDM Limit must be increased if the highest worth control rod is determined solely analytically versus a test using the reactor (requires a series of local criticals).

This higher limit accounts for uncertainties in the calculation of the highest worth control rod.

SD! is demonstrated to satisfy a variety of OPCON 5 surveillances at the beginning of each cycle and, if necessary, at any future entry to OPCON 5 during the cycle if the assumptions of the previous SDM Demonstration are no longer valid.

In most situations, the SDM Demonstration will be based solely on analytical methods and a test will not be performed. If SDM is demonstrated by using solely analytical methods. then SDM must be adjusted to account for Hope Creek B 3/4 1-1 Revised by NRC letter dated April 10, 2000

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in peak fuel enthalpy greater than 280 cal/gm in the event of a control rod drop accident.

The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal/gm. Thus requiring the RWM to be OPERABLE when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate control.

The R provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section of49 of the FSAR and the techniques of the analysis are presented in Reference 1.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation.

Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

Operability of a REM channel is assured for a given control rod when 50% of the LPRM inputs for each detector level are available for that rod.

When < 50% of the LPRM inputs on either detector level are available, a case-by-case evaluation of channel operability is required.

I HOPE CREEK B 3/4 1-3 Revised by NC letter dated anuray 8, 2001 I

Rv^AC7 7

oN7Ro

_.w*_

BASES rate, solution conaert:atr. or oron equivalent o ee. tne ATWS R e-.s:.c:

inva'.date he oginal system design basis.

?a:agraph '_: !4 ot 10 CR 50.62 states that:

Zach boi:in; ater reac:o:

s: have a Stant-iEy :izi -:ro:

Sys:e.

S:CS w

a

'low zazacit a.- bor.

_.::-: eo vale^.

co.nro.

apacity o 36 a2.cns per minuze o

3 weig-.:

per:en: sodiu3 er.:abora:e solution (natural bo:on enr,.hmenc!.

The described minimum system parameters C32.4 gp.n, 3.6 per:en:

concentration and ratural bordr. equivalent) will ensure an equivalent _n-e-_.n capability that exceeds the AWS Ruie requirement.

The stated minimum allowable pumping rate of 2.4 gallons per mnute: is et hrou;h the simultaneous oceration of both puns.

The standby liquid control system will also provide the capability o raise and maintain the long-term post-accident coolant nventory pH levels o cr above. This will prevent sgnificant fractions of the dissolved iodine from being converted to elemental iodine and then re-evolving to the containment atmosphere.

j.~~~~~~~~~~~~~

1. ~ A.

onro Rod Oro Aidnnay sMhdology on Wer actors Su; ry and Q alificac/on." Jy~.y 199.

/

(k'ak t 0 r;ec HOP C.EX B-o I/4 a-S me

)o 1

)

H~~~~~~~~~~-

A e d nt o 134-l V.

3:'4.2 POWvE? ZIS:_=RU710N IMITS BASES

.he secifications in this section help assure ha: the fuel an e operated safely and reliably during normal oera::on.

ai:ian, the s

specified n these specifications help ensure that the fel oes not ecee-specified safety and regulatory limits during anticpa:ed oerational occurrences and design basis accidents.

Specifically, these lImits:

1. Ensure that the limits specified in OCFR50.46 are not exceeded following the postulated design basis loss of coolant accident.

2. Ensure reactor operations remains within licensed, analyzed power/flow limits.

3. Ensure that the MCPR Safety Limit is not violated following any anticipated operational occurrence.

4. Ensure fuel centerline temperatur 9rac_

and peak cladding strain rena s below 1 during ste srati.

-t 0^

N SWKed coss sev sEit lt5 1

3/4.2.1 AVERAGE PLANARHE _____HE The AVERAGE P EAR HEAT GE Os TE the average Linear Heat Gen ration (LHGR) f all the fuel rod i

assembly at any axial locat on.

e Technic pecification APLHGR is the LHGR of the highest-powered rod divlded by cal peaking factor.

Limits on the APLHGR are specified to ensure that

.e fuel design limits are not exceeded.

The limiting value of the APLHGR limit is specified in the CORE OPERATING LIMITS REPORT.

The calculation procedure used to establish the APLHGR is based on a loss-of-coolant accident analysis.

The post LOCA peak cladding temperature PCT) is primarily a function of the APLHGR and is dependent only secondarily PGd to rod power distribution within an assembly.

The analytical mod used in evalu postulated loss-of-coolant accidents are esc d in Reference an These models are onsistent with the requirem &

ndix K tToCFR5O.

For plant operation with recirculation loop, a lower value for the APLHGR limit is specified in the CORE OPERATING LIMITS REPORT.

This lower value accounts for an earlier transition from nucleate boiling which occurs following a loss-of-coolant accident in the single loop operation compared to two loop operation.

3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The flow biased simulated thermal power-upscale scram setting and the flow biased neutron flux-upscale control rod block trip setpoints must be adjusted to ensure that the MCPR does not become less than the fuel cladding Safety Limit or that > 1% plastic strain does not occur in the degraded situation.

The scram setpoints and rod block setpoints are adjusted in accordance with the formula in Specification 3.2.2 whenever it is known that the existing power distribution would cause the design LHGR to be exceeded at RAT THERMAL POWER HOPE CREE exenodsre enecne t No.

g t ;Se S Zee re d A.e y

n Ak-k-k (,p xLsI t' er T_

I&t?

AC) 1~c AC~o 00' ev

'\\

0-N 6 t't, Os -, -

o em.o sve k.~at -as Aes5;9a HOPE CREEK B 3/4 1

Amendment No. 2 e-v e r zi eve beat oo r~

a oe r

at 2c

BASES 3/4.2.3 MINIM'UM CRITICAL POWER RATIO The required operating limit MCRs a: steady sate opera:t.ng conditio-=

as secified in Specification 3.2.3 are derived from the established ue; claddir.g

.tegri~y Safety Limit MCPR, and an analysis of abnormal opera::ona; transients.

For any abnormal operating transient analysis evaluation with he initial condition of the reactor being at he steady sate operating limit, is required that the resulting MCPR does not decrease below the Safety Lit MCPR at any tine during the transient assuming nstrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit 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 Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 s obtained.

sP Ce Bv uat a

enetrantie begins ththe

/ 2-Aedeial pa;K~mters~hown

/SSARTabf 15.0-3 2hat are *put oy ABC cre dca Weavio rase cmur ogam.

/he code use to aUat-transi ts The MCR operating lim ts dr ttransient anal v

i a

e deedn-n the operating core fowr

<25seftv lP to ensure adherence to fuel design lmits during he worst_

transient wth moderate Tecythat is postulated in Chap'ter Flow depencQ~en-t-Mre etermined by stea t BK methods using te/

1cloode (Reference curves are profVed based o edll rans-DOiet(e.

runout f both C t4<z T'F A*^l ?

~t e

imesi 1

t r

ne iansient\\

/ce Refer e 2) determi power dndent MC P i ts Mp) ue to the\\

/ senn tiv ty of >e transien '~sponse tinitial core lo ees 0~oe lev>Qbelow tho3 at which te urbine s~pvalve lsxeadtriscorl valve s clos re sram limit a

ypass high ad Iow Pp opea limits 3<%e provide d K peratio n b n 2 5

D THE POWERant

< ypass por levels.

\\

HOPE CREEK B 3/4 2-2 Amendment No.17A

F-WER tSTRIBJ7iON L:M:.s BASES MNIVIMUM CRITICAL POWER RATIO :ontinuedi At THERMAL POWER levels less than or eua; to 2i f RATE-TR' POWER, the reactor will be oerating at minimum ecirtulation pump-speed ana tne moderator void content will be ery,sma;'.

For al designated on:tr=

ron patterns which may be employed at this point, operating plant experienze indicates that the resulting :4CPR value 'is in excess of requirements by a considerable margin.

During initial start-up esting of the plant, a !CPR evaluation will be made at 25% of RATED THERMAL POWER level with m.nimum recirculation pump speed.

he MCPR margin will thus be demonstrated

d:

future MCPR evaluation below this power level will be shown to be unnecessa-v.

The daily requirement for calculating MCPR when THERMAL POWER is greater :nan or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control Zcd changes.

The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change n THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE l

8St 9T 5

^!hi spc gcatio assures a

h Li rHeat G eration R (H~

/

in 6

rod iofesth( the desi<< linear >&t eneraRvlon even i uelp>e w ~ ~ ~~~0 sfctoispslat

d.

////

References:

R - ;-en lect

/ ompany An t

ical Moda fo Loss-of-S o;a Anal is i

ccordan with 10 CFR 0, Appendi K(,

NEDE-20566 November 1 S;.

2.o/

eCENPD-30

-A.

Ref e Safety R rt for Boil ater s

\\ /

Reload uel" The aroed revisio at the time he reload alyes are/

+>(

per fo ~d. The appg~e revi

~umber shal ~ identifiti te C OPE REE tNG LIMIT E 3

e No.

2

>~~~~A S-2-as X4\\a~X 0

v1

(

I

} 2. g~o - a~t54-A Q~at~vc~to vo T

e t

Ose-B~~~etsxozt Cor~~e-TgA et S 8

Xeack~~~~ors) yiosks t i Y\\.

vt3

)-

ts L{ t~tS~o d } eA LE t00 HOPE CREEK B 3/4 2-3 Amendment No. 126

-~~~~

0 3/4.4 REACTOR COO NT SYSTEM S

BASES 3/4.4.1 RECIRCUL IONSYST C

OA The impact o s gle reci ulation loop operation up nt safety is assessed and shows at single loop operation is permitted if the MCPR fuel cladding Safety Li is increased as noted by Specification 2.1.2, APRM scram.

and control rod b o cnts are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2 r<;.a5 v 1 v PLHGR limits are decreased by the factor given in on ~t2>and MCPR o pet~~ri

(:~ iea tecon P. b

,e CouH.)

Additionally, surveillanc recirculation loop is imposed to exclude the possibility of excessive core internals vibration.

The surveillance on differential temperatures below 38%

THERMAL POWER or 50% rated recirc ion lo flow is to mitigate the undue thermal stress on vessel nozzles, er ap and vessel bottom head during the extended oper a

ulation loop mode.

An inoperable sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area-and reduce the capability of reflooding the core, thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation.

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a

-single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 506F of each other prior to startup of an idle loop.

The loop temperature must also be within 50°F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.

Sudden equalization of a temperature difference > 145F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

The objective of BWR plant and fuel design is to provide stable operation with margin over the normal operating domain.

However, at the high power/low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g., rod pattern, power shape).

To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise 1

n.

ed while operating in this region.

Stability tests at aie ewed to determine a generic region of the power/flow map in which rveillance of neutron flux noise levels should be performed.

a v decay ratio of 0.6 was chosen as the bases for determining the generic on for surveillance to account for the plant to plant variability of decay ratio with core and fuel-designs.

This generic region has been determined to correspond to a core flow of less than or equal to 45% of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.

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

126