ML20114D029
| ML20114D029 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 08/21/1992 |
| From: | Hannon J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20114D032 | List: |
| References | |
| GL-88-01, GL-88-1, NUDOCS 9209040098 | |
| Download: ML20114D029 (12) | |
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UNITED STATES S
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NUCLEAR REGULATORY COMMISSION
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DOMGT NO. 50-461 CLINTON POWER STATION. VIOT NO. i AMEN 0 MENT 10 FACILITY OPERATING LICENSE Amendment No. 65 License No. NPr-62 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by lilinois Power Company * (IP),
ar.J Soyland Power Cooperative, Inc. (the licensees) dated July 11,1990, as supplemented May 7, 1991, 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 5:ill 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 l
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 sith 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 Speci-fications as indicated in the attachment to th s license amendment, anc paragraph 2.C.(2) of facility Operating Licenss No. NPF-62 is hereby amended to read as follows:
- lllinois Power Company is authorized to act as agent for Soyland Power Cooperative, Inc. and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
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2 (2) 12.chnical Sntcifica11ons and Environmental Protection Plaq The Technical Specifications contained in f.ppendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 65, are hereby incorporated into this license.
Illinois Power 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 issuanta to be implemented within 60 days of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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1 John H. Hannon, Director Project Directorate 111-3 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 21, 1992 e
._..._____..________.q I
4 All A(HME!41 19 Ljjilj5LAfMDMEt41 f10, 65 S
fAlil1IY OPERAllfdG LltEftSE tjp. f4PF 62 k'.E E1 f40.59-461 J
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.
The corresponding overleaf pages are provided to maintain document completeness.
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Remove insert _
3/4 0-3 3/4 0-3 3/4 4-12 3/4 4-12 3/4 4-12a 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4-14 0 3/4 4-3 6 3/4 4-3 B 3/4 4-4 8 3/4 4-4 B 3/4 4-4a D
i e.
F hf3AntilLITY Ob iMF
'LLANCE RE0V.lREMENTS (Continu3sil g
1 0 '.
(Continued) 1 orovisions of Specification 4.0.2 are applicable the above required
.Jencies fnr performing irservice insoection and t-sting activities.
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~ormance of the above insarvice inspection and testing activities shall o.
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in adattio. *o other specified Surveillance Requirements.
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' thing in the ASME Boiler and Pressure Vessel Code shall be construed to I
t arsede the requirements of any Technical Specific =" n.
f.
r
'nser'eice Inspection Program ror piping
. NRC Generic j
L 88-01 shall be performed in accordt TC Staff positions w
1 on
.edule, methods and personnel, and sa included in the gentcic letter.
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CLINTON - UNIT 1 3/4 0-3 A:rendment No. 65 2
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REACTOR COOLANT SYSTEM SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION LIMITING CONDITION FOR OPERATION 3.4.2.2 The low-low set function of the following reactor coolant system
-safety / relief valves shall be OPERABLE with the following settings *:
Low-Low Set Function Setpoint* (psia) i 15 psi Valve No.
Open Close F0510 1033 926 F051C 1073 936 F047F 1113 946 F051B 1113 946 F051G 1113 946 A,PPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, ard 3.
ACTION:
a.
With the low-low set function of one of the above required reacter coolant system safety / relief valves inoperable, restore the inoperable low-low set C
function to OPERABLE status within 14 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 />, b.
With the low-low set function of more than one of the above reouired reactor coolant system safety / relief valves inoperable, be in a'. least HOT SHUTOOWN 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 il hours-c.
With either low-low set function pressure actuation trip system "A" or
' 1" inoperable, restore the inoperable trip system to OPERABLE status within 7 days; otherwise, be in at least HDT 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 following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVCILLS"C; REQUIREHENTS 4.4.2.2 The low-low set function pressure actuation instrumentation sh&1. be demonstrated OPERABLE by performance of a:
a.
CHANNEL FUNCTIONAL TEST, including calibration of the trip unit, at least once per 37 days.
b.
CHANNEL CALIBRATICN and LOGIC SYSTEM FUNCTIONAL TEST at least once per 18 months.
Each of the two trip systems or divisions of t5e low-low set function actuation logic associated with the Nuclear S); tem Protecticn System shall be manually tested independent of the SELF TEST SYSTEM during separate refueling outages such that both divisions and all channel trips are tested at least once every four fuel cycles.t
- 0ne 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 the pur-pose of performing surveillance testirn in accordance with Specification 4.4.2.2.
- The lift setting pressure shall correi.Jond to ambient conditions of the valves at nominal operating temperatures and pressures.
tManual testing for the purpose of satisfying Specification 4.4.2.2.b. is not re-quired until af ter shutdown during the first regularly scheduled refueling outage.
CLINTON - UNIT 1 3/4 4-11 Amendment No. 5
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4 REACTOR COOLANT SYSTEM 144.4.3 REACTOR COOLANT SYSTEM LEAKAGE
' tEAKAGE DETECTION SYSTEMS LIMITl',;G CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage dete ' ton systems shall be OPERABLE:
a.
The drywell atmosphere particulate radioactivity monitoring system, b.
The drywell sump flow monitoring system, and c.
Either the drywell atmosphere 93:eous radioactivity monitoring system or the drywell air coolcrs condensate flow rate monitoring system.
APPLICABilill: OPERA 110NAL CONDITIONS 1, 2, and 3.
a 110R:
With only two of the above required leakage detection systems OPERABLE, a..
operation msy continue fer up to 30 days when the drywell atmosphere particulate radioactivity monitoring s> item it inoperable provided grab samples.of the drywe atmosphere are cotained and 4lalyzed 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.
-operations may continue:
1, with the drywell equipmant drain sump flow monitoring subsystem inoperable provided the drywell equipment drain sump flow rate is monitored and determined by alternate means at-least once per 12 hou rs,-
2.
for up to 30. days with the drywell floor drain sump flow monitoring L
. subsystem inoperable provided the drywell floor drain sump flow rate is monitored' and determined by alternate mear.s at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,.
l L
- c.
operation may continue for up to 30 days when the drywell atmosphere l
gaseous radioactivity monitoring system and the drywell air coolers p
condensate flow rato monitoring system are inoperable provided grab samples L
of the drywe111 atmosphere are obtained and analyzed at least once per 24' L
hours.
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 />.
CLli4 TON - UNIT 1 3/4 4-12 Amendment No. 65 l
L
d REACTOR COOLANT-SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAG[
LEAKAGE DETCCTION SYSTEMS SVRVEILLANCE RE0VIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demon-strated OPERABLE by:
a.
Drywell atmosphere particulate and gaseous-monitoring systems-performance of a CHANNEL CHECK at least once per 12 nours, a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.
b.
Drywell-sump flow monitoring system-performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CAllBRATION TEST at least once per 18 months.
c.
Drywell air cooler condensate flow rate monitoring system performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CAllBRATION at least once per 18 months.
d.
Flcw testing the drywell floor drain sump inlet piping for blockage at least once every 18 months during shutdown.
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'CL.'NTON - UNIT l-3/4 4-12a Amendment No. 65 l
l EACTOR COOLANT SYST[M OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATI@j 3.4.3.2 Reactor coolant system leakage shall be limited to:
a.
b.
5 gpm UNIDENTIFIED LEAKAGE.
c.
25 Spm IDENTiflED LEAKAGE (averaged over any 24-hour period).
d.
0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm I
from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1, at rated recctor pressure.
e.
No greater than a 2 gpm increase in UNIDENTiflED LEAKAGE within a 24-hour period or less during OPERATIONAL CONDITION 1.
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 SHUIDOWN 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 hourt, 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 />.
c.
With any reactor coolant syst ssure isolation valve leakage greater than the abot limit, isolate tu high eressure porticn 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 two other closed manual or deactivated autamatic valves, 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 />.
d.
With any reactor coolant system UNIDENTIFIED LEAKAGE increase greater than 2 gpm within any 24-hour period or less (during OPERATIONAL CONDITION 1),
within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from the time of discovery isolate the scurce of increased leakage or verify that the source of increased lea' age is not associated with service sensitive Type 304 or 316 austenitic siainless steel; otherwise be in at leas? HOT SHUTDOWN within the nex. 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 b; 'rs.
CLINTON - UNIT )
3/4 4-13 Amendment f:o.33,65 l
SURVEILLANCE RE0VIREMENTS 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 drywell atmospheric par ticulate and gaseous radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (not a means of quantifying leakage),
b.
Monitoring the drywell floor drain sump flow rate at least once per 8 l
- hours, c.
Monitoring the drywell equipment drain sump flow rate at least once per l
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, d.
Monitoring the drywell air coolers condensate flow rate at least once per l
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and e.
Monitoring the reactor vessel head flange leak detection system at least l
once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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.
I b.
Prior to returning toe valve to service following maintenance,- repair, or replacement work on the valve or its associated actuator, c.
As outlined in ASME Code,Section XI, paragraph IWV-3427(b).
The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CCNDITION 3.
CLINTON - UNIT I 3/4 4-14 Amendment No.jy h5
REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)
The recirculation flow control valves provide regulation of individual recir-culation loop drive flows; which, in turn, will vary the flow rate of coolant through the reactor core over a range consistent with the rod pattern and re-circulation pump speed.
The recirculation flow control system consists of the electronic-and hydraulic components necessary for the positioning of the two hydraulically actuated flow control valves.
Solid state control logic will generate a flow control valve " motion inhibit" signal in response to any sne of several-hydraulic power unit or analog control circuit failure signals.
The
" motion inhibit" signal causes hydraulic power unit shutdown and hydraulic isolation such that the flow control valve fails "as is."
This design feature insures that the flow control valves do not respond to potentially erroneous control signals.
Electronic limiters exist in the position control loop of each flow control valve to limit the flow control valve stioking rate to 1011% per second in opening and closing directions on a control signal failure.
The analysis of the recirculation flw control failures on increasing and decreasing flow are prese '.ed in Secuons 15.3 and 15.4 of the USAR respectively.
The required surveillance interval is adequate to ensure that the flow control valves remain OPERABLE ard not so frequent as to cause exce nive wear on the system components.
3/4.4.2 SAFETY / RELIEF VALVES The safety. valve function of the safety / relief valves (SRV) operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1375 psig in accordance with the ASME Code. A total of 11 OPERABLE safety-relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.
Any combination of 5 SRVs operating in the relief c.sde and 6 SRVs operating in the safety mode is acceptable.
Demonstration of the safety-relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Spwifica-tion 4.0.5.
- The low-low set system ensures that safety / relief valve discharges are minimized
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for a second opening of these valves, following any overpressure transient.
This is achieved by automatically lowering the closing setpoint of 5 valves and lowering the opening setpoint of 2 valves following the initial opening.
In this way, the frequency and magnitude of the containment blowdown duty cycle is substantially reduced.
Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced set-point-does not violate the design basis.
I CLINTON - UNIT 1 B 3/4 4-3 Amendment No.73,33,65
REACTOR COOLANT SYSTE_M iLAji[3 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 114.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. With certain exceptions as noted in the Clinton Power Station Updated Safety Analysis Report, these detection systems are consistent with the recommenda?ns of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary leakage Detection Systems," May 1973.
Except for the drywell particulate and gaseous radioactivity monitors, the systems provide the ability to measure leakage from fluid systems in the drywell.
The drywell sump flow monitor.ng system consists of the drywell floor drain sump flow monitoring subsystem and the drywell equipment drain sump flow monitoring subsystem. OPERABILITY of each of these subsystems requires that the applicable portion of the monitoring subsystem associated with the v-notched weir box be OPERABLE. Other portions of the subsystem, including the sump pump control circuit and the associated timer, cycle counter and level switches, may be utilized as appropriate to provide an alternate means of monitoring and determining UNIDENTIFIED or l'1ENTIFIED leakage under the provisions of the associated ACTION statements for the respective subsystem.
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 con-sidered.
The evidence obtained from experiments suggests that for leakage some-4 what greater than that specified for UNIDENTiflED LEAKAGE the probability is small that the imperfection er crack associated with such leakage would grow rapidly. With respect to IGSCC-related cracks in service sensitive austenitic stainless steel piping however, an additional limit on the allowed increase in UNIDENTIFIED LEAKAGE (within a 24-hour period or less) is imposed in accordance with Generic Letter 88-01, "NRC Postion on IGSCC in BWR Austenitic Stainless Steel Piping." since an abrupt increase in the UNIDENTIFIED LEAKAGE could be indicative of leakage from such a_ source.
In all cases, if the leakage rates exceed the valui.s specified or the leakage -is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action. The reactor will also be -shut down if an-increase in UNIDENTIFIED LEAKAGE exceeds the specified limit and the source of increased leakage cannot be isolated or it cannot be determined within a short period of time that the source of increased leakage is not associated with austenitic stainless steel.
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 De considered as a portion of tne allowed limit.
j CLINTON - UNIT 1 B 3/4 4-4 Amendment No.33,65
i REACTOR COOLANT SYSTEM BASES 3/4.4.4 CHEMISTRY I
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 stainless steel The effect of chloride
't as great when the oxygen concentration in the coolant is low, thus t ppm limit on chlorides is permitted during POWER OPERATION. During shu.uowr, end 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.
Witt, the conductivity meter inoper-able, 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.
CLINTON - UNIT 1 B 3/4 4-5 Amendment No.65 1
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