ML20199F455
| ML20199F455 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/04/1997 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Public Service Electric & Gas Co, Atlantic City Electric Co |
| Shared Package | |
| ML20199F461 | List: |
| References | |
| NPF-57-A-107 NUDOCS 9711240222 | |
| Download: ML20199F455 (7) | |
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NUCLEAR REGULATORY UOMMISSION m swiw at o u, o.c. sone onoi
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PUBLIC SERVICE ELECTRIC & GAS C'MPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.107 License No. NPF-57 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment filed by the Public Service Electric
- & Gas Company (PSE&G) dated March 31, 1997, as supplemented by letters dated July 16, August 26, and October 3,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.-
Accordingly, the license is amended by_ changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby amended to read as follows:
9711240222 971104 DR-ADOCK 0 34
! (2)_ Technical Specifications and Environmental Protection plan The Technical Specifications contained in Appendix A, as revised through Amendment No.107, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license.
PSE&G shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION i. Stolz, Direc r r ject Directorat
-2 D vision of Reactor Projects - I/II ffice of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: November 4, 1997
ATTACHMENT TO LICENSE AMENDMENT NO.197 FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following sages of the Appendix "A" Technical Specifications with the attached pages.
Tie revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 2-1 2-1 B 2-1 B 2-1 3/4 4-1 3/4 4-1 6-21 6-21
.____.,__m -
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low flow 2.1.1 THERMAL POWER shall not exceed 25% 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 25% of RATED THERMAL POWER and tha reactor vessel steam como 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.
EFRMAL POWER, High Pressure and High Flow t
2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not less than ) "O with two ree'.rculation loop operation and shall not be less than 1.12 with single recirculation loop operation, in both cased with the reactor vessel steam dome pressure greater than 785 psig and core ficw greater than 10% of rated flow.*
l APPLICABILITY:
OPERATIONAL CONDITIONS I and 2.
ACTION:
With MCPR 4ess than 1.10 with two recirculation loop operation or less than 1.12 with single recirculation loop operation and in both cases with the reactor vessel steam dome pressure greater than 785 psig and core flow greater 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.
REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as maasured 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 1225 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.
- Values applicable to Cycle 8 operation only.
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HOPE CREEK 2-1 Amendment No.107
i 2.1 SAFETY LIMITS BASES
2.0 INTRODUCTION
The fuel cladding, reactor pressure vessel and primary system piping are the principal berriers to the release of radioactive materials to the environs.
Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity safety Limit is set such that no fuel damage is calculated to occur if the lindt is not violated. Because fuel damage is not-directly observable, a step-back approach is used to establish a safety Limit _such that the MCPR is not less than 1.10 for two recirculation loop operation and 1.12 for single recirculation loop operation. MCPR greater than 1.10 for two recirculation loop operation and 1.12 for single recirculation loop operation represents 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 naterials from the environs. The integrity of this cladding barrier is related 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 ndgration from this source is incrementally cumulative and continuously measurable.
Puol cladding perforations, however, can result from thermal atresses which occur from reactor operation significant?y above design conditions and the Limiting Safety System Settings.
While fission product ndgration from cladding perforation ir 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.
2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the applicable NRC-approved critical power correlation is 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 wi31 clwags be greater than 4.5 psi.
Analyses show that with a bundle flow of 28 x 10 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 pai. Tgus, the bundle flow with a 4.5 pai driving head will be greater than 28 x 10 lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 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 50% of RATED THERMAL POWER.
- Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
HOPE CREEK B 2-1 Amendment No. 107
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM
_ RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION
....................n.......
...............................c........n 3.4.1.1 Two reactor coolant system recir'culation loops shal) be in operation with a.
Total core flow greater than or equal to 45% of rated core flo.1, or b.
THERMAL POWER less than or equal to the lindt specified in Figure 3.4.1.1-1.
APPLICABILITY:
OPERATIONAL CONDITIONS l' and 2*.
..CTI ON :
With one ruactor coolant system recirculation loep not in operation:
a.
- 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 THERMAL POWER to s 70% of RATED THERMAL POWER, and c)
Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit per Specification 2.1.2, and d)
Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value of 0.86 times the two recirculation loop lindt per Specification 3.2.1, and e)
DELETED.
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 is s 50% of lated 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
- See Special Test Exception 3.10.4.
HOPE CREEK 3/4 4-1 Amendment No.107
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ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-240ll-P-A (the latest approved revision)*, General Electric Standard Application for Reactor Fuel l
(GESTAR II).
The core operating limits shall be detcrmined so that all applicable lindts (e.g.,
fuel thermal-mechanical lindts, core thermal-hydraulic lindts, ECCS lidts, nuclear lindts 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 reloso cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be subndtted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, within the tin.e 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 affecced the ability to achieve and maintain safe shutdown in the event of a fire shall be submitted to the U.S. Nuclear kegulatory 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, Codo 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, Records of caanges made to the procedures required by e.
Specification 6.8.1.
f.
Records of radioactive shipments, g.
Records ol' sealed source and fission detector leak tests and results.
- For Cycle 8, as evaluated in the Saf ety Evaluation dated 11/ 4/97 to support License Amendment No.107.
HOPE CREEK 6-21 Amendment No.107
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