ML20141J013
| ML20141J013 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/22/1997 |
| From: | Pulsifer R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20141J017 | List: |
| References | |
| NUDOCS 9705270208 | |
| Download: ML20141J013 (7) | |
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UNITED STATES g
j NUCLEAR REGULATORY COMMISSION I t WASHINGTON, D.C. so66tM001
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COMONWEALTH EDISON COMPANY A!Q MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 0UAD CITIES NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 174 License No. DPR-30 1
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated April 21, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 Comission; C.
There is reasonable assurance (i) 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; D.
The issuance of this amendment will not be inimical to the comon 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 3.B. of Facility Operating License No. DPR-30 is hereby amended to read as follows:
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9705270208 970522 PDR ADOCK 05000265 P
i B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through. Amendment No.174, are hereby incorporated in the i
license. The licensee shall operate the facility in accordance
-with the Technical Specifications.
l 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION Robert M. Pulsifer, Project Manager.
Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 22, 1997 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 174 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 50-265 Revise the Appendix A Technical Specifiestions by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 2-la B 2-3a 5-5a 5-Sa 6-16a 6-16a 1
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l SAFETY LIMITS 2.1 I
1, 2.0.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 7
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2J.
$AFETY LIMITS
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THERMAL POWER. Low Pressure or Low Flow j
1 2.1.A THERMAL POWER shall not exceed 25% of RATIED THERMAL POWER with the reactor j
vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
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APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.
ACTION:
' With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel j
steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least j
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.
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THERMAL POWER, Hiah Pressure and Hiah Flow a
2.1.B The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 for Unit 1 and 1.10* for Unit 2 with the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow. During single recirculation loop operation, 1
this MCPR limit shall be increased by 0.01.
i APPLICABillTY: OPERATIONAL MODE (s) 1 and 2.
p ACTION:
With MCPR less than the above applicable limit and the reactor vessel steam dome pressure i
greater than'or equal to 785 psig and core flow greater than or equal to 10% of rated flow, be in j
)
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.
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- Applicable to Unit 2 for cycle 15 only.
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QUAD CITIES - UNIT 2 2-1a Amendment No.174 i
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SAFETY LIMITS B 2.1 I
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BASES i
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approach. Much of the data indicates that BWR fuel can survive for an extended period in an I
environment of transition boiling.
i The Unit 1 MCPR Safety Limit is 1.07, based on General Electric methods for calculating the MCPR i
l Safety Limit. The Unit 2 MCPR Safety Limit is 1.10*, based on Siemens Power Corporation (SPC) l l
methods for calculating the MCPR Safety Limit.
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2d&
Reactor Coolant System Pressure The Safety Limit for the reactor coolant system pressure has been selected such that it is at a i
pressure below which it can be shown that the integrity of the system is not endangered. The i
. reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of i
fission products it is essential that the integrity of this system be protected by establishing a i
pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in j
the reactor vessel.
i The reactor coolant system pressure Safety Limit of 1345 psig, as measured by the vessel steam l
space pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor vessel.
The 1375 psig value is derived from the design pressures of the reactor pressure vessel and coolant system piping. The respective design pressures are 1250 psig at 575aF and 1175 psig at 560'F. The pressure Safety Limit was chosen as the lower of the pressure transients permitted by the applicable design codes, ASME Boiler and Pressure Vessel Code Section ill for the pressure vessel, and USASI B31.1 Code for the reactor coolant system piping. The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% x 1250
= 1375 psig), and the USASI Code permits pressure transients up to 20% over design pressure (120% x 1175 = 1410 psig). The Safety Limit pressure of 1375 psig is referenced to the lowest elevation'of the reactor vessel. The design pressure for the recirculation suction line piping (1175 psig) was chosen relative to the reactor vessel design pressure. Demonstrating compliance of peak vessel pressure with the ASME overpressure protection limit (1375 psig) assures compliance of the suction piping with the USASI limit (1410 psig). Evaluation methodology to assure that this Safety Limit pressure is not exceeded for any reload is documented by the specific fuel vendor. The design basis for the reactor pressure vessel makes evident the substaritial margin of protection i
against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is a factor of 1.5 below the yield strength of 40,100 psi at 575'F. At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yield strength.
The relationships of stress levels to yield strength are comparable for the primary system piping and provides similar margin of protection at the established pressure Safety Limit.
The normal operating pressure of the reactor coolant system is nominally 1000 psig. Both pressure relief and safety relief valves have been installed to keep the reactor vessel peak pressure below 1375 psig. However no credit is taken for relief valves during the postulated full closure of all MSIVs without a direct (valve position switch) scram. Credit, however, is taken for the neutron flux scram. The indirect flux scram and safety valve actuation provide adequate margin below the allowable peak vessel pressure of 1375 psig.
- Applicable to Unit 2 cycle 15 only.
QUAD ClTIES - UNIT 2 B 2-3 a Amendment No.174 w
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. REACTOR CORE 5.3 5.0 DESIGN FEATURES El REACTOR CORE 1
Fuel Assemblies 4
4 5.3.A The reactor core shall contain 724 fuel assemblies. Each assembly consists of a l
matrix of Zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. The assemblies may contain water rods or water boxes. Limited substitutions of Zircoloy or ZlRLO filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.
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Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.'
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Control Rod Assemblies
'5.3.B The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B C) and/or hafnium metal. The control rod assembly shall have a nominal axial absorber length of 143 inches.
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QUAD CITIES - UNIT 2 5-5 a Amendment No. 174
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m_._._
R3 porting Requiramants 6.9
. ADMINISTRATIVE CONTROLS.
(3)
Commonwealth Edison Topical Report NFSR-0085, Supplement 1, " Benchmark-of BWR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons,"
1 (latest approved revision).
(4) of BWR Nuclear Design Methods - Neutronic Licens,Supnalyses," (latest Commonwealth Edison Topical Report NFSR 0085, ment 2, Benchmark mg approved revision).
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(5)
Advanced Nuclear Fuels Methodology for Boiling Water Reactom, XN-NF l j
19(P)(A), Volume 1 Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.
t CASMO/MICROBURN BWR Nuclear Design Methods", Revision Commonwealth Edison Topical Report NFSR-0091, " Benchmark of (6) 1 and 2, December 1991, March 1992, and May 1992, respectively; SER letter dated March 22,1993.
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i (7)
Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A) l Revision 1, and Revision 1 Supplement 1 Advanced Nuclear Fuels Corporation,
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May 1995.
i (8)
Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced l l
Nuclear Fuels 9X9-lX and 9X9-9X BWR Reload Fuel, ANF-89-014(P)(A),
Revision 1 and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, 3
October 1991.
(9)*
Comed letter,
- Comed Response to NRC Staff Request for Additional.
j information (RAI) Regarding the Application of Siemens Power Corporation i
ANFB Critical Power Correlation to Coresident General Electric Fuel for LaSalle j
Unit 2 Cycle 8 and Quad Cities Unit 2 Cycle 15, NRC Docket No.'s 50-373/374 i
and 50-254/265", J.B. Hosmer to U.S. NRC, July 2,1996, transmitting the -
l ort, Application of the ANFB Critical Power Correlation to Coresident topical refor Quad Cities Unit 2 Cycle 15, EMF-96-051(P), Siemens Power GE Fuel i
Corporation - Nuclear Division, May 1996, and related information.
i c.
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 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.
6.9.B Special Reports Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
- Applicable to Unit 2 for cycle 15 only.
QUAD CITIES - UNIT 2 6-16 a Amendment No.174