ML20044H407
| ML20044H407 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 05/28/1993 |
| From: | Robert Stransky Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044H408 | List: |
| References | |
| NUDOCS 9306080352 | |
| Download: ML20044H407 (11) | |
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UNITED STATES y
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. NUCLEAR REGULATORY COMMISSION g
WASHINGTON. D.C. 20066-0001
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THE CLEVELAND ELECTRIC ILLUMINATING COMPANY. ET AL.
l DOCKET NO. 50-440 i
PERRY NUCLEAR POWER PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE-Amendment No. 48-License No. NPF-58 l
1.
The Nuclear Regulatory Commission (the Commission) has found that:
t A.
The application for amendment by The Cleveland Electric Illuminating-Company, Centerior Service Company, Duquesne Light Company, Ohio-Edison Company, Pennsylvania Power Company, and Toledo' Edison Company (the licensees) dated June 30, 1992, 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; i
B.
The facility will operate in conformity with the application, the j
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 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 Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-58 is hereby.
amended to read as follows:
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9306080352 930528 DR ADOCK 0500 0
1 (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 48 are hereby incorporated into this license.
The Cleveland Electric Illuminating Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION y
i bert J. Stransk, Project Manager LProject Directorate III-3
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Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: May 28, 1993
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ATTACHMENT TO LICENSE AMENDMENT NO. 48 FACILITY OPERATING LICENSE NO. NPF-58 DOCKET NO. 50-440 Replace the following pages-of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.
Remove Insert B 2-7 B 2-7 3/4 3-6 3/4 3-6 3/4 3-8 3/4 3-8 6-21 6-21 l
LIMITING SAFETY SYSTEM SETTINGS 1
BASES i
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Average Power Ranoe Monitor (Continued) 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.
The 15% neutron flux trip remains active until the mode switch is placed in the Run position.
The APRM trip system is calibrated using heat balance data taken during steady state conditions.
Fission chambers provide the basic input to the sys-tem 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 trans-fer associated with the fuel.
For the Flow Biased Simulated Thermal Power-High setpoint, a time constant specified in the COLR 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 unneces-sary shutdown.
3.
Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products.
A pres-sure 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 set-ting 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.
PER W - UNIT 1 B 2-7 Amendment No. 48
LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 4.
Reactor Vessel Water Level-Low f
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.
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 i
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 steam line radiation, low reactor water level, high steam tunnel temperature and low steam line pressure.
The MSIV's closure scram anticipates the pressure 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 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.
I PERRY - UNIT 1 B 2-8
TABLE 3.3.1-2 hh REACTOR PROTECTION SYSTEM RESPONSE TIMES 4
I c:
RESPONSE TIME 25 FUNCTIONAL UNIT (Seconds) s
>d 1.
a.
Neutron Flux - High NA b.
Inoperative NA 2.
Average Power Range Monitor *:
a.
Neutron Flux - High, Setdown NA b.
Flow Biased Simulated Thermal Power - High
< 0.09**
c.
Neutron Flux - High 7 0.09 d.
Inoperative NA 3.
Reactor Vessel Steam Dome Pressure - High
< 0.35 us 4.
Reactor Vessel Water Level - Low, level 3 7 1.05 32 5.
Reactor Vessel Water Level - High, Level 8 7 1.05 u>
6.
Main Steam Line Isolation Valve - Closure 7 0.06 En 7.
Main Steam Line Radiation - High RA 8.
Drywell Pressure - High NA 9.
Scram Discharge Volume Water Level - High NA 10.
Turbine Stop Valve - Closure 11.
Turbine Control Valve Fast Closure, Valve Trip System
-< 0.06 Oil Pressure - Low
< 0.07#
EI 12.
Reactor Mode Switch Shutdown Position NA l
[
13.
Manual Scram NA 8
s a
- Neutron detectors are exempt from response time testing.
Response time shall be measured z
P from the detector output or from the input of the first electronic component in the channel.
m
- Not including the simulated thermal oower time constant specified in the COLR.
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- Measured from start of turbine control valve fast closure.
o TABLE 3.3.1-1 (Continued)
REACTOR PROTECTI0h SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(b) Unless adequate shutdown margin has been demonstrated per Specifica-tion 3.1.1 and the "one-rod-out" Refuel position interlock has been demonstrated OPERABLE per Specification 3.9.1, the shorting links shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn."
(c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.
(d) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(e) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
(f) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required.
(g) With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(h) This function is automatically bypassed when turbine first stage pressure is less than the value of turbine first stage pressure corresponding to 40%** of RATED THERMAL POWER.
cNot required for control rods removed per Specification 3.9.10.1 or 3,9.10.2.
- The Turbine First Stage Pressure Bypass Setpoints.and corresponding Allowable Values are adjusted based on Feedwater temperatures (see 3/4.2.2 for definition of AT). The Setpoints and Allowable Values for various ATs are as follows:
T(*F)
Setpoint (psig)
Allowable Value (psig) 0=T 1 212 1 218 0<
AT < 50 1 190 1 196 50 < AT < 100 1 1 68 1 174 100 < AT-170
< 146
< 152 1
PERRY - UNIT 1 3/4 3-5 Amendment No. 29
l TABLE 4.3.1.1-1 (Continued)
A REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS i3 CHANNEL OPERATIONAL l
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH l
!i FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED U
- 10. Turbine Stop Valve - Closure NA M
R 1
- 11. Turbine Control Valve fast l
Closure Valve Trip System 011 Pressure - Low NA M
R 1
12.
Reactor Mode Switch Shutdown Position NA R
NA 1,2,3,4,5
- 13. Manual Scram NA M
NA 1,2,3,4,5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for R*
at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.
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(c) Deleted Y
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.
The provisions of Specification 4.0.4 are not applicable provided the surveillance is. performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reaching 25% of RATED THERMAL POWER.
(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.
(f) The LPRMs shall be calibrated at least once per 1000 MWD /T using the TIP system.
(g) Calibrate trip unit setpoint at least once per 31 days.
p (h) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing loop flow (APRM % flow).
a g
(i) This calibration shall consist of verifying that the simulated thermal power time constant is within the limits specified in the COLR.
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a A
'(j)' This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
- =
P (k) With any control rod withdrawn.
Not applicable to control rods removed per 3
Specification 3.9.10.1 or 3.9.10.2.
(1) This function is not required to be OPERABLE when Drywell Integrity is not required.
(m) The CHANNEL CALIBRATION shall exclude the flow reference transmitters, these transmitters shall be calibrated at least once per 18 months.
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e TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL c
CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION (a)
SURVEILLANCE REQUIRED 1.
a.
Neutron Flux - High S/U,S,(b)
W R
2 l
5 W
R 3,4,5 b.
Inoperative NA W
NA 2,3,4,5 2.
Average Power Range Monitor:(f) a.
Neutron Flux - High, S/U.S,(b)
W SA 2
l Setdown S
W SA 3, 5 b.
Flow Biased Simulated ID)
Thermal Power - High 5,D W
W(d)(e), SA "), RII) f 1
D c.
Neutron Flux - High S
W W(d), 34 y
"4 d.
Inoperative NA W
NA 1,2,3,5 3.
Reactor Vessel Steam Dome Pressure - High S
M R(9) 1,2(3)*
4.
Reactor Vessel Water Leve.
I9)
Low, Level 3 S
M R
1, 2 5.
High, Level 8 S
M R(9) 1 6.
Main Steam Line Isolation
{
Valve - Closure NA M
R 1
R 7.
Main Steam Line Radiation -
High S
M R
1,2(3) 8.
Drywell Pressure - High S
M R(9) 1,2(I)
P 9.
Scram Discharge Volume Water S
Level High y
a.
Level Transmitter S
M R(9) 1, 2, 5(k) b.
Float Switches NA M
R 1,2,5(k)
.m A.
ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)
The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (0DCM), pursuant to Specifications 6.13 and 6.14, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.1S.
It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2.
The Semiannual Radioactive Effluent Release Reports shall also include the following:
an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.7.9 or 3.3.7.10, respectively; and description of the events leading to liquid holdup tanks exceeding the limit!, of Specification 3.11.1.4.
MONTHLY OPERATING REPORTS
- 6. 9.1. 8 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Resource Management, U. S. Nuclear Regulatory Cormission, Washington, D. C. 20555, with a copy to the Regional Administrator of the Regional Office no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
(1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.
(2) The Minimum Critical Power Ratio (MCPR) for Technical Specification 3.2.2.
(3)
The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.3.
(4) The Simulated Thermal Power Time Constant for Technical Specification 3.3.1.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel.
(The approved revision at the time reload analyses are performed shall be identified in the COLR.)
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 U.S. Nuclear Regulatory Commission Document Control Desk with copies to the Regional Administrator and Resident Inspector.
l PERRY - UNIT l' 6-21 Amendment No. A7,48
ADMINISTRATIVE CONTROLS SPECIAL REPORTS j
6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office within the time period specified' for each report.
6.9.3 Safety-relief valve failures will be reported to the Regional Administrator of the Regional Office of the NRC via the Licensee Event Report system within 30 days.
6.9.4 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 reported to the Regional Administrator of the Regional Office of the NRC 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.
6.10.2 The following records shall be retained for at least 5 years:
a.
Records and logs of unit operation covering the 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.
PERRY - UNIT 1 6-21a Amendment No. 33 l
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