ML20073M282
| ML20073M282 | |
| Person / Time | |
|---|---|
| Site: | River Bend  (NPF-47-A-075, NPF-47-A-75) |
| Issue date: | 10/07/1994 |
| From: | Beckner W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073M287 | List: |
| References | |
| NUDOCS 9410140255 | |
| Download: ML20073M282 (7) | |
Text
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S UNITED STATES I;,
NUCLEAR REGULATORY COMMISSION y.....,/
t WASHINGTON, D.C. 20E%4001 GULF STATES UTILITIES COMPANY **
CAJUN ELECTRIC POWER COOPERATIVE AND ENTERGY OPERATIONS. INC.
DOCKET NO. 50-458 1
RIVER BEND STATION. UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No.75 License No. NPF 47 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Gulf States Utilities * (the licensee) dated September 12, 1994, as supplemented by letter dated September 30, 1994, 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, as amended, 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 Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and
- E01 is authorized to act as agent for Gulf States Utilities Company, which has been authorized to act as agent for Cajun Electric Power Cooperative, and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
- Gulf States Utilities Company, which o:<ns a 70 percent undivided interest in River Bend, has merged with a wholly owned subsidiary of Entergy Corporation.
Gulf States Utilities Company was the surviving company in the merger.
9410140255 941007 PDR ' ADOCK 05000458 P
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-g-0 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-47 is hereby amended to read as follows:
(2) Technical Soecifications and Environmental Protection Plan I
The Technical Specifications contained in Appendix A, as revised through Amendment No. 75 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
E01 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.
FOR THE NUCLEAR REGULATORY COMMISSION W4 a /bA William D. Beckner, Director Project Directorate IV-1 Division of Reactor Projects III/IV l
Office of Nuclear Reactor Regulation l
Attachment:
Changes to the lechnical Specifications Date of Issuance: October 7,1994 l
l
ATTACHMENT TO LICENSE AMENDMENT NO. 75
[ACILITY OPERATING LICENSE N0. NPF-47 DOCKET NO. 50-458 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change.
The corresponding overleaf pages are also provided to maintain document completeness.
REMOVE INSERT B 2-7 B 2-7 3/4 2-7 3/4 2-7 3/4 2-7a 3/4 2-7a
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LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETp0lNTS (Continued)
Averaae Power Ranae Monitor (Continued)
The APRM trip system is calibrated using heat balance data taken during steady state conditions.
Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux-High setpoint; i.e.,
for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel.
For the Flow Biased Simulated Thermal Power-High setpoint, a time constant is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.
The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. The flow referenced trip setpoint must be adjusted by the specified formula in Specification 3.2.2 in order to maintain these margins.
l 3.
Reactor Vessel Steam Dome Pressure-Hiah High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products.
A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase.
The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine control valve fast closure and turbine stop valve closure trips are bypassed. For a load rejection or turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.
4.
Psactor vessel Water Level-tow The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease.
The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to, assure that there is adequate protection for the fuel and pressure limits.
e RIVER BEND - UNIT 1 B 2-7 Amendment No. 4h 75
1 LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 5.
Reactor vessel Water Level-Hich A reactor scram from high reactor water level, approximately two feet above normal operating level, is intended to offset the addition of reactivity effect associated with the introduction of a significant amount of relatively cold feedwater. An excess of feedwater entering the vessel would be detected by the level increase in a timely manner. This scram feature is only effective when the reactor mode switch is in the Run position because at THERMAL POWER levels below 10% to 15% of RATED THERMAL POWER, the approximate range of power level for changing to the Run position, the safety margins are more than adequate without a reactor scram.
6.
Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events.
The MSIV's are closed automatically from measured parameters such as high steam flow, high i
steam line radiation, low reactor water level, high steam tunnel temperature l
and low steam line pressure.
The MSIV's closure scram anticipates the pressure 4
and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.
7.
Main Steam Line Radiation-High The main steam line radiation detectors are provided to detect a gross l
failure of the fuel cladding.
When the high radiation is detected, a trip is initiated to reduce the continued failure of fuel cladding. At the same time the main steam line isolation valves are closed to limit the release of fission products.
The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures in the fuel cladding.
8.
Drywell Pressure-High High pressure in the drywell could indicate a break in the primary pressure boundary systems or a loss of drywell cooling. The reactor is tripped in order s
to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and to the primary containment. The trip setting was selected as low as possible without causing spurious trips.
9.
Scram Discharge Volume Water Level-High i
The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram.
Should this volume fill j
up to a point where there is insufficient volume to accept the displaced water, J
control rod insertion would be hindered.
The reactor is therefore tripped when j
l RIVER BEND - UNIT 1 B 2-8
POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETP0INTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-high scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (S,,) shall be established according to the following relationships:
a.
Two Recirculation Loop Operation Trio Setooint Allowable Value S s (0.66W + 48%)T S s (0.66W + 51%)T S,, s (0.66W + 42%)T S,, s (0.66W + 45%)T b.
Single Recirculation Loop Operation Trio Setooint Allowable Value S s (0.66W + 42.7%)T S s (0.66W + 45.7%)T S,, s (0.66W + 36.7%)T S,, s (0.66W + 39.7%)T where:
S and S, are in percent of RATED THERMAL POWER, W - Loop, recirculation flow as a percentage of the loop recircclation flow which produces a rated core flow of 84.5 million lbs/hr.
3xFRTP + 1 T
4xCMFLPD provided CMFLPD s 0.6 x FRTP + 0.4, otherwise FRTP T
CMFLPD T is applied only if less than or equal to 1.0.
FRTP is the FRACTION OF RATED THERMAL POWER.
CMFLPD is the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY.
APPLICABILITY:
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
With the APRM flow biased simulated thermal power-high scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conser-vative than the value shown in the Allowable Value column for S or S,5,and/or S as above determined, initiate corrective action within 15 minutes and adjust to be consistent with the Trip Setpoint value
- within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or reduce THERMAU POWER to less than 25% 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 />.
- With T < 1.0, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that the adjusted APRM readings result in a calculated T 21.0 when the APRM reading is substituted for FRTP, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.
RIVER BEND - UNIT 1 3/4 2-7 Amendment No. Bh.75
EQWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS 4.2.2 The FRTP and CMFLPD shall'be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-high scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:
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 15% 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 operat-ing with T s 1.0.
l d.
The provisions of Specification 4.0.4 are not applicable.
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.i RIVER BEND - UNIT 1 3/4 2-7a.
Amendment No. E 75 l