ML20070G495
| ML20070G495 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/06/1991 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | GPU Nuclear Corp, Jersey Central Power & Light Co |
| Shared Package | |
| ML20070G497 | List: |
| References | |
| DPR-16-A-150 NUDOCS 9103120119 | |
| Download: ML20070G495 (6) | |
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'o UNITED STATES
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NUCLEAR REGULATORY COMMISSION g
p WASHINGTON, D. C. 20655 GPU NUCLEAR CORPORATION 2
AND JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.150 License No. DPR-16 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.,
(the licensee), dated December 18, 1989, as supplemented April 30, October 16, and November 16, 1990, 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 (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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
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. 2.
Accordingly, the-license is amended by changes to the Technical Specifications as indicated in the attachment sto this license amendment, and paragraph 2.C.(2) of Prnvisional Operating License No. DPR-16 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendrent No.
, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION s
ohn P. Stolz, Director r ' ct Directorate I-Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: March 6, 1991 1
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ATTACHMENT TO LICENSE AMENDMENT NO.1sn PROVISIONAL OPERATING LICENSE NO.' DPR-16 DOCKET NO. 50-219 4
Replace the following pages of the Appendix A Technical Specifications with the ~ enclosed ) ages as indicated. The revised pages are identified by amendment num>er and contain vertical lines indicating the areas of change..
Remove Insert 2.3-2 2.3-2 2.3-6' 2'.3-6 4.3-1
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FUNCTION LIMITING SAFETY SYSTEM SET *INGS B.
Neutron Flux, control Rod Block The Rod Block setting shall be ZEP S $ [(0.90 x 10-6) W + 53.1) (MFLPD) withamaxpuasetpointof108%forcoreflowequal to 61 x 10 lb/hr and greater.
The definitions of 5, W, FRP.and MFLPD used above for the APRM scram trip apply.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual operating value will be usec.
This adhstment may be acceeplished by increasing the
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APRM gain and thus reducing the flow referenced APRM rod block curve by the reciprocal of the APPM gain change.
C.
Reactor High,
$1060 psig Pressure, Scram D.
Reactor High Pressure, 2 4 1 1070 psig Relief Valves Initiation 3 9 1 1090 peig E.
Reactor High Prost, :
51060 poig with time delay Isclation Condenser 53 seconds Initiation F.
Reactor High Pressure, 4 4 1212.psig 112 poi safety Valve Initiation 4 9 1221 psig 112 poi 1 0 1230 peig.
112 psi G.
Low Pressure hain Steam 2825 peig'(initiated in IRM range 10)
'e Line, MSIV Closure H.
Main Steam Line Isolation 110% Valve Closure from Valve closure, scram full open I.
Reactor Low Water Level, 211'5" above the top of the active Scram fuel as indicated under normal operating conditions J.
Reactor Low-Low Water 27'2" above the top of the active Level, Main Steam Line fuel as indicated under normal Isolation Valve closure opera:ing conditions OYSTER CREEK 2.3-2 Amendment No. 73, 75, 111 150 a
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e7 The reactor coolant system safety valves offer yet another protective feature f or the reactor coolant system pressure safety limit since these valves are sized assuming no credit for other pressure relieving devices.
In coarpliance,
v;th Section I of the ASMI Boiler and Pressure Vessel Code, the safety valve must be set to open at a pressure no higher than 103n of design pressure, and they must limit the reactor pressure to no more than 110% of design pressure.
The safety valves are sized according to the Code for a condition of main steam isolation valve closure while operating at 1930 Mwt, followed by (1) a reactor scram on high neutron flu.x, (2) failure of recirculation pump trip on high pressure, [3] failure of the turnine typass valves to open, and (4) failure of the isolation condensers and relief valves to operate. Under these conditions, a total of 9 safety valves are required to turn the pressure transient. The ASKI BfPV Code allows a alt of working pressure (1250 petg) variation in the lift point of the valves.
This variation is recognized in Specification 4.3.
The low pressure isolation of the main stream lines at 825 peig was provided to give protwetion agatast fast reactor depressurization and the resulting rapid cool-down of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity saf ety limit.
Operation of the reactor at pressures lower than 825 psig requires that the reactor modo switch be in the STARTITP position and the IRMs be in the range 9, or lower, where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram. Thus, the commination of mais steam line low pressure isolation and isolation valves closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In additica the isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.
The low water level trip setting of 11'5* above the top of the active fuel has been established to assure that the reactor is not operated at a water level below that for which the fuel cladding integrity safety limit is applicable.
With the scram set at this point, the generatinn of steam, and thur the loss of inventory, is stopped.
For example, for a loss of foodwater flow a reactor scram at the value indicated and isolation valve closure at the low-low watet level set paint results t ' scre than 4 feet of water recaining above the cere after isolation (6).
During periods when the reactor is shut down, decay heat is present a.Sd adequate sater level must ce maintained to provide core cooling. Thus, the low-low level trip point of 7'2" above the core is provided to actuate the core soray system (when the core spray system is required as identified in section 3.4) to pte nde cooling water should the level drop to this point.'
The turbine etop valve (s) scram is provided to anticipate the pressure, neutron flux, and heat flux increase caused by the rapid closure of the turbine stop valve (s) and failure of the turbine bypass system.
The generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from f ast closure of the turbine cor. trol OYSTER CREEK 2.3-6 Amendment No.: 71, 75, 150
- correction: 11/30/87
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- E M ' 4.3 REACTCR COOLANT T
g iemhilitvt Applies to the surveillance requirements for,the reactor coolant system.
giiset ive t To determine the condition of the reactor coolant system and the operation of the safety devices related to it.
soeeifications A. Materials surveillance speciments and neutron flux monitors shall be installed in the reactor vessel adjacent to the wall at the midplane of the active core. Specimens and monitors shall be periodically removed, tested, and evaluated to determine the effects of neutron fluence on the fracture toughness of the vessel shell materials. The results of these evaluations shall be used to assess the adequacy of the P-T curves of Figures 3.3.l(a), (b) and (c).
New curves shall be generated as required.
B. Ineervice inspection of ASME Code class 1, Class 2 and Class 3 systems and components shall be performed in accordance with section II of the ASME Soiler and Pressure vessel Code and applicable Addenda as required by 10 CFR, section 50.55a(g),
except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(1).
s C. Innerrice testing of ASME Code Class 1, Class 2 and Class 3 pumpe and valves shall be performed in accordance with section II of the ASMR Roller and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, section 50.55a(g)(6)(1).
D. A visual examination for leaks shall be made with the reactor coolant system at pressure during each scheduled refueling outage or after major repairs have been made to the reactor coolant system in accordance with Article 5000, section II.
The requirassents of specification 3.3. A 9all be mat during the teat.
g E. Each replacement safety valve or valve that has been repaired shall be tested in accordance with subsection IWV-3510 of section II of the ASME Boiler and Pressure Vessel code.
Setpoints shall be as follows:
Mushar_of_Valvan Set Points fosic) 4 1212 1 12 4
1221 1 12 1
1230 1 12
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F. A sample of reactor coolant shall be analysed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the purpose of determining the content of chloride ion and to check the conductivity.
OYSTER CREEK 4.3-1 Amendswat.: 82, 90, 120, 150 1
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