ML20044H119
| ML20044H119 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/26/1993 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044H120 | List: |
| References | |
| GL-88-01, GL-88-1, NUDOCS 9306070433 | |
| Download: ML20044H119 (11) | |
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7 UNITED STATES
(('ikZ.[4j NUCLEAR REGULATORY COMMISSION Q ',j f
WASHINGTON, D.C. 20555-0001 y~
QETROIT EDISON COMPANY DOCKET NO. 50-341 FERMI-2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 89 License No. NPF-43 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Detroit Edison Company (the licensee) dated September 30, 1992, 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 (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 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.
Theissuanceofthisamendmentisinaccordancewith10CFit'Part51of the Comission'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-43 is hereby amended to read as follows:
9306070433 930526 PDR ADOCK 05000341 P
. )
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 89
, and the Environmental Protection Plan i
contained in Appendix B, are hereby incorporated in the license. DECO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance with full implementation within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION hl Ledyard B. Marsh, Director Ir Project Directorate III-l' f
Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 26, 1993 t
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1 ATTACHMENT TO LICENSE AMENDMENT NO. 89 j
FACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341 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.
REMOVE INSERT 3/4 4-9*
3/4 4-9*
3/4 4-10 3/4 4-10 3/4 4-11 3/4 4-11 3/4 4-11a 3/4 4-llb*
3/4 4-12*
3/4 4-12*
B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-2a
- 0verleaf pages provided to maintain document completeness.
No changes contained in these pages.
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REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage detection systems shall-be OPERABLE:
a.
The primary containment atmosphere gaseous radioactivity monitoring system channel.
b.
The primary containment sump flow monitoring system consisting of:
1.
The drywell floor drain sump level, flow and pump-run-time system, and i
2.
The drywell equipment drain sump level, flow and pump-run-time system.
c.
The drywell floor drain sump level monitoring system.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With only two of the above required leakage detection systems OPERABLE, restore the inoperable detection system to OPERABLE status within 30 days; when the required gaseous radioactive monitoring system is ino'perable,.
operation may continue for up to 30 days provided grab samples of the _
containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be i
demonstrated OPERABLE by:
i a.
Primary containment atmosphere gaseous monitoring systems-performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.
b.
Primary containment sump flow and drywell floor drain sump level monitoring systems-performance of a CHANNEL FUNCTIONAL TEST at least once per 31_ days and a CHANNEL CALIBRATION TEST at'least' once per 18 months.
i FERMI - UNIT 2 3/4 4-9 Amendment No. 89
2 i
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE l
LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be _ limited to:
a.
No PRESSURE B0UNDARY LEAKAGE.
b.
5 gpm UNIDENTIFIED LEAKAGE.
c.
25 gpm total leakage averaged over any 24-hour period.
d.
I gpm leakage at a reactor coolant system pressure of 1045 i 10 psig_
from any reactor coolant system pressure isolation valve specified in Tabl e 3.4.3.2-1.
e.
2 gpm increase in UNIDENTIFIED LEAKAGE within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during OPERATIONAL CONDITION 1.
f.
2 gpm increase in UNIDENTIFIED LEAKAGE within any 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period during.
OPERATIONAL CONDITIONS 2 and 3.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits _within 4 l
hours or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUIDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
With any reactor coolant system pressure isolation valve ' leakage greater than the above limit, isolate the high pressure portion of _the affected.
system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at -least one other closed manual, deactivated automatic, or check *~ valve,.or be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
With one or more of the high/ low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-2 inoperable, restore'the.
inoperable monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; a
restore the inoperable monitor (s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- Which has been verified not to exceed the allowable leakage limit at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.
FERMI - UNIT 2 3/4 4-10 Amendment No. 57, 89 t
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[IMITING CONDITION FOR OPERATION (Continued)
ACTION:
(Continued) e.
In OPERATIONAL CONDITION 1, with any reactor coolant system UNIDENTIFIED LEAXAGE increase greater than 2 gpm within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, identify the source of leakage increase as not service sensitive Type 304 or 316 austenitic stainless steel within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I f.
In OPERATIONAL CONDITIONS 2 and 3, with any reactor coolant system I
UNIDENTIFIED LEAKAGE increase greater than 2 gpm within any 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, identify the source of leakage increase as not service sensitive Type 304 or 316 austenitic stainless steel within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i SURVEILLANCE REOUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to.be within each of the above limits by:
a.
Monitoring the primary containment atmospheric gaseous radioactivity at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,*
b.
Monitoring the primary containment sump flow rate at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in OPERATIONAL CONDITION 1** and at least once per 4 l
hours in OPERATIONAL CONDITIONS 2 and 3, c.
Monitoring the drywell floor drain sump level at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> **, in OPERATIONAL CONDITION 1** and at least once per 4 l
r hours in OPERATIONAL CONDITIONS 2 and 3, and d.
Monitoring the reactor vessel head flange leak detection system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.*
8
- Not a means of quantifying leakage.
- The provisions of Specification 4.0.2 are not applicable to the l
surveillance requirement in OPERATIONAL CONDITION 1.
i FERMI - UNIT 2 3/4 4-11 Amendment No. 89 I
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- l 4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit:
a.
At least once per 18 months, and b.
Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.
The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.
4.4.3.2.3 The high/ low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with alarm setpoints per Table 3.4.3.2-2 by performance of a:
a.
CHANNEL FUNCTIONAL TEST at least once per 31 days, and b.
CHANNEL CALIBRATION at least once per 18 months.
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FERM1 - UNIT 2 3/4 4-11a Amendment No.89 t
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i FERMI - UNIT 2 3/4 4-llb Amendment No. 89.
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TABLE 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER VALVE DESCRIPTION i
1.
RHR System E11-F015A LPCI Loop A Injection Isolation Valve Ell-F015B LPCI Loop B Injection Isolation Valve Ell-F050A LPCI loop A Injection Line Testable Check Valve Ell-F050B LPCI loop B Injection Line Testable Check Valve i
Ell-F008 Shutdown Cooling RPV Suction Outboard Isolation Valve Ell-F009 Shutdown Cooling RPV Suction Inboard Isolation Valve Ell-F608 Shutdown Cooling Suction Isolation Valve 2.
Core Spray System E21-F005A Loop A Inboard Isolation Valve E21-F005B Loop B Inboard Isolation Valve E21-F006A Loop A Containment Check Valve i
E21-F006B Loop B Containment Check Valve 3.
High Pressure Coolant Injection System E41-F007 Pump Discharge Outboard Isolation Valve E41-F006 Pump Discharge Inboard Isolation Valve 4.
Reactor Core Isolation Cooling System E51-F012 Pump Discharge Isolation Valve,
E51-F013 Pump Discharge to Feedwater Header Isolation Valve TABLE 3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE PRESSURE MONITORS ALARM SETPOINT VALVE NUMBER SYSTEM fosia)
Ell-F015A & B, Ell-F050A & B RHR LPCI s 449 l
Ell-F008, F009, F608 RHR Shutdown Cooling s 135 E21-F005A & B, E21-F006A & B Core Spray s 452 E41-F006, F007 HPCI s 71 E51-F012, F013 RCIC s 71 FERMI - UNIT 2 3/4 4-12 Amendment No. Jf, 85, 89
I REACTOR COOLANT SYSTEM BASES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSIEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.
3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered.
The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.
However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action. Service sensitive reactor coolant system Type 304 and 316 austenitic stainless steel piping; i.e., those that are subject to high stress or that contain relatively stagnant, intermittent, or low flow fluids, requires additional surveillance and leakage limits.
The additional limit placed upon the rate of increase in i
UNIDENTIFIED LEAKAGE in OPERATIONAL CONDITION I meets the NRC Staff guidance in Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping".
The applicability of the Generic Letter 88-01 limit to OPERATIONAL I
I CONDITION 1 only ensures that the expected increases in UNIDENTIFIED LEAKAGE I
experienced during reactor vessel heatup and pressurization during startup do not cause unwarranted entries into the applicable ACTION statement.
The rate of I
increase in UNIDENTIFIED LEAKAGE limit in OPERATIONAL CONDITIONS 2 and 3 ensures I
that the above service sensitive reactor coolant system Type 304 and 316 I
austenitic stainless steel piping is monitored during reactor startup prior to I
reactor vessel heatup and pressurization.
The surveillance interval for I
t determination of UNIDENTIFIED LEAKAGE in OPERATIONAL CONDITION 1 meets the I
guidance in Supplement I to Generic Letter 88-01.
l Tu purpose of the RCS interface valves leakage pressure monitors (LPMs) is to provide assurance of the integrity of the Reactor Coolant System pressure isolation valves which form a high/ low pressure boundary.
The LPM is designed to alarm on increasing pressure on the low pressure side 'of the high/ low pressure interface to provide indication to the operator of abnormal interface valve leakage.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the i
j FERMI - UNIT 2 B 3/4 4-2 Amendment No. U, gg
f REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued) i stainless steel.
The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION. During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.
Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.
The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
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i FERMI - UNIT 2 B 3/4 4-2a Amendment No.14, 89