ML20045H824
| ML20045H824 | |
| Person / Time | |
|---|---|
| Site: | Clinton  (NPF-62-A-081, NPF-62-A-81) |
| Issue date: | 07/15/1993 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20045H825 | List: |
| References | |
| NUDOCS 9307210330 | |
| Download: ML20045H824 (15) | |
Text
. p*Mouq'o UNITED STATES
^g NUCLEAR REGULATORY COMMISSION c.
g p
WASHINGTON, D. C. 205$5
%...../
ILLINOIS POWER COMPANY S0YLAND POWER COOPERATIVE. INC.
DOCKET NO. 50-461 CLINTON POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 81 License No. NPF-62 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Illinois Power Company * (IP), and Snyland Power Cooperative, Inc. (the licensees) dated April 16, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's i
rules and regulations set forth in 10 CFR Chapter I; B.
The facil4ty will operate in conformity with the application, the provisions of the Art, 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 Comniission'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-62 is hereby amended to read as follows:
i
- 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.
9307210330 930715 PDR ADOCK 05000461 P
. (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 81
, 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. issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 1 tit $ I JY James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 15, 1993
ATTACHMENT TO LICENSE AMENDMENT NO. 81
{
FACILITY OPERATING LICENSE NO. NPF-62 DOCKET NO. 50-461 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, identified by an asterisk, are provided to maintain document completeness.
Remove Paoes insert Paces
- 3/4 1-7
- 3/4 1-7 3/4 1-8 3/4 1-8
- 3/4 4-9
- 3/4 4-9 i
3/4 4-10 3/4 4-10 3/4 5-4 3/4 5-4 3/4 5-5 3/4 5-5
^
3/4 5-6 3/4 5-6 B 3/4 4-3 B 3/4 4-3 B 3/4 4-4 B 3/4 4-4
- B 3/4 5-1
- B 3/4 5-1 B 3/4 5-2 B 3/4 5-2 i
B 3/4 5-3 B 3/4 5-3 j
i
REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION (Continued) 3.1.3.2 ACTION (Continued}:
3.
The sum of " fast" control rods with individual scram insertion times in excess of the limits of ACTION a.2 and of " slow" control rods does not exceed 5.
4.
No " slow" control rod, " fast" control rod with individual scram insertion time in excess of the limits of ACTION a.2, or otherwise inoperable control rod occupy adjacent locations in any direction, including the diagonal, to another such control rod.
Otherwise, 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 />.
b.
With a " slow" control rod (s) not satisfying ACTION a.1, above:
1.
Declare the " slow" control rod (s) inoperable and 2.
Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 60 days when operation is continued with three or more " slow" control rods declared inoperable.
Otherwise, 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 />.
With the maximum scram insertion time of one or more control rods exceed-c.
ing the maximum scram insertion time limits of Specification 3.1.3.2 as determined by Specification 4.1.3.2.c, operation may continue provided that:
1.
" Slow" control rods, i.e., those which exceed the limits of Specifi-cation 3.1.3.2, do not make up more than 20% of the 10% sample of control rods tested.
2.
Each of these " slow" control rods satisfies the limits of ACTION a.1.
i 3.
The eight adjacent control rods surrounding each " slow" control rod are:
a)
Demonstrated through measurement within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to satisfy the maximum scram insertion time limits of Specification 3.1.3.2 and b)
4.
The total number of " slow" control rods as determined by Specifica-tion 4.1.3.2.c, when added to the sum of ACTION a.3 as determined by Specification 4.1.3.2.a and b, does not exceed 5.
Otherwise, 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 />.
d.
The provisions of Specification 3.0.4 are not applicable.
CLINTON - UNIT 1 3/4 1-7 1
i
REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXIMUM SCRAM INSERTION TIMES SURVEILLANCE REOUIREMENTS 1
4.1.3.2 The maximum scram insertion time of the control rods shall be demon-strated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:
For all control rods prior to THERMAL POWER exceeding 40% of RATED a.
THERMAL POWER following CORE ALTERATIONS or after a-reactor shutdown that is greater than 120 days, b.
For at least 10% of the control rods, on a rotating basis, at least once l
per 120 days of POWER OPERATION.
4.1.3.3 The maximum scram insertion time for specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods shall be demonstrated through coolant pressure greater than or equal to 950 psig., measurement with reactor Alternatively, those specific control rods may be determined OPERABLE with reactor coolant pressure less than 950 psig by demonstrating an acceptable scram insertion time to notch position 13.
The scram time acceptance criteria for this alternate test shall be determined by linear interpolation between 0.95 seconds at a reactor coolant pressure of 0 psig and 1.40 seconds at 950 psig.
If this alternate test is utilized, the individual scram time shall also be measured with reactor coolant pressure greater than or equal to 950 psig prior to exceeding 40% of RATED THERMAL POWER.
For each of the above single control rod scram time tests,
- e control rod drive pumps shall be isolated from the accumulators 1
- The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 2 provided this surveillance requirement is completed prior to entry into OPERATIONAL CONDITION 1.
CLINTON - UNIT 1 3/4 1-8 Amendment No. 81 i
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2.1 The safety valve function of at least six of the following v.tives and the relief valve function of at least five additional valves, other than those satisfying the safety valve function requirement, shall be OPERABLE with the specified lift settings; and the acoustic monitor for each OPERABLE valve shall be OPERABLE.*
Number of Valves Function Setpoint** (psic) 7 Safety 1165 1 11.6 psi 5
Safety 1180 1 11.8 psi 4
Safety 1190 1 11.9 psi 1
Relief 1103 1 15.0 psi 8
Relief 1113 1 15.0 psi 7
Relief 1123 1 15.0 psi APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With the safety and/or relief valve function of one or more of the above a.
required safety / relief valves inoperable, be in at least HOT SHUTDOWN
.ithin 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 one or more safety / relief valves stuck open, provideo that suppres-sion pool average water temperatere is less than 110*F, close the stuck cpen s8fety/ relief valve (s); if suppression pool average water tempera-ture is 110 F or greater, place the reactor mode switch in the Shutdown position.
With one or more safety / relief valve acoustic monitor (s) inoperable, restore c.
the inoperable monitor (s) to OPERABLE status within 7 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 />.
d.
With either relief valve function pressure actuation trip system "A" or "B" inoperable, restore the inoperable trip system to OPERABLE status within 7 days; otherwise, 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 following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- One relief valve pressure actuation channel and/or one acoustic monitor channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the l
purpose of performing surveillance testing in accordance with Specifica-tions 4.4.2.1.1 and 4.4.2.1.2.
- The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
CLINTON - UNIT 1 3/4 4-9 Amendment No. 71 APR 9
$93
REACTOR COOLANT SYSTEM SAFETY / RELIEF VALVES j
SURVEILLANCE REQUIREMENTS 4.4.2.1.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE by performance of a:
a.
CHANNEL FUNCTIONAL TEST at least once per 31 days, and a b.
CHANNEL CALIBRATION at least once per 18 months.*
4.4.2.1.2 The relief valve function pressure actuation instrumentation shall be demonstrated OPERABLE by performance of a:
CHANNEL FUNCTIONAL TEST, including calibration of the trip unit, at least a.
i once per 92 days.
b.
CHANNEL CALIBRATION and LOGIC SYSTEM FUNCTIONAL TEST at least once per 18 months.
Each of the two trip systems or divisions of the relief valve function actuation logic associated with the Nuclear System Protection 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.
I k
t I
- The provisions of Specification 4.0.4 are not applicable provided the surveil-lance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
CLINTON - UNIT 1 3/4 4-10 Amendment No. 81
EMERGENCY CORE COOLING SYSTEMS LCCS - OPERATING SURVEILLANCE REQUIREMENTS (Continued) 4.5.1 (Continued) 2.
Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
b.
Verifying that when tested pursuant to Specification 4.0.5 each:
1.
LPCS pump develops a flow of at least 5010 gpm with a pump differential pressure greater than or equal to 276 psid.
2.
LPCI pump develops a flow of at least 5050 gpm with a pump differential pressure greater than or equal to 113 psid.
3.
HPCS pump develops a flow of at least.5010 gpm with a pump differential pressure greater than or equal to 363 psid.
c.
Fcr the LPCS, LPCI, and HPCS systems, at least once per 18 months performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.
d.
For the HPCS system, at least once per 18 months, verifying that the suction is automatically transferred from the RCIC storage tank to the suppression pool on a RCIC storage tank low water level signal and on a suppression pool high water level signal.
e.
For the ADS by:
1.
At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST of the accumulator low pressure alarm system.
2.
At least once per 18 months, performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
3.
At least once per 18 months, manually opening each ADS valve when-the r psig,eactor steam dome pressure is greater than or equal to 100 and observing that; a.
The control valve or bypass valve position responds accordingly, or
- Except that an automatic valve capable of automatic return to its ECCS
'I position when an ECCS signal is present may be in position for another mode of operation.
"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 reactor steam pressure and
~
flow are adequate to perform the test.
CLINTON - UNIT 1 3/4 5-4 Amendment No. 81
.~.~..
EMERGENCY CORE COOLING SYSTEMS a
3/4.5.2 ECCS - SHVTDOWN LIMITING CONDITION FOR OPERATION b.
There is a corresponding change in the measured steam flow, or c.
The acoustic tail-pipe monitor alarms.
4.
At least once per 18 months, performing a CHANNEL CALIBRATION of the accumulator low pressure alarm system and verifying an alarm setpoint of 1 140 psig on decreasing pressure.
3.5.2 At least two of the following shall_ be OPERABLE and capable of being powered from a diesel generator of Specification ~ 3.8.1.2.b.
a.
The low pressure core spray (LPCS) system with a flow path. capable of _
taking suction from the suppression pool and transferring the water i
through the spray sparger to the reactor vessel.
b.
Low pressure coolant injection (LPCI) subsystem "A" of the RHR system with a flow path capable of taking. suction from the suppression pool and transferring the water to the reactor vessel.
c.
Low pressure coolant injection (LPCI) subsystem "B" of the RHR system with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.
d.
Low pressure coolant injection (LPCI) subsystem "C" of the RHR system with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.
e.
The high pressure core spray-(HPCS) system with a flow path capable of-a taking suction from one of the following water sources and transferring the water through the spray sparger to the reactor vessel:
1.
From the suppression pool, or 2.
When the suppression pool level is less than the limit or is drained, from the RCIC storage tank containing at least 125,000 available gallons of water.
.l APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5*.
ACTION:
^
a.
~With one of the above required subsystems / systems inoperable, restore at--
least two subsystems / systems to OPERABLE status within 4-hours or suspend all operations that have a potential for draining the reactor vessel.
The provisions of Specification 3.0.4 are not applicable.
- .The' ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the reactor cavity to steam dryer pool gate is open and water level in these upper containment pools is 3
maintained within the limits of Specification 3.9.8 and 3.9.9.
1 CLINTON;- UNIT 1 3/4 5-5 Amendment No. 81
EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS - SHUTDOWN LIMITING CONDITION FOR OPERATION b.
With both of the above required subsystems / systems inoperable, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel. Restore at least one subsystem / system to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or establish PRIMARY CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
SURVEILLANCE RE0VIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1.
l 4.5.2.2 The HPCS system shall be determined OPERABLE at least once per i
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the RCIC storage tank required volume when the RCIC l
storage tank is required to be OPERABLE per Specification 3.5.2.e.
j l
l l
l i
CLINTON - UNIT 1 3/4 5-6 Amendment No. 81
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 one 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 stroking rate to 1011% per second in opening and closing directions on a control signal failure.
The analysis of the recirculation flow control failures on increasing and decreasing flow are presented in Sections 15.3 and 15.4 of the USAR respectively.
The required surveillance interval is adequate to ensure that the flow control valves remain OPERABLE and not so frequent as to cause excessive 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 mode 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 Specifica-tion 4.0.5.
The surveillance requirement for performing a CHANNEL CALIBRATION of the acoustic monitor (s) includes an exception to the provisions of Specification 4.0.4.
This exception allows reactor steam conditions to be established which are adequate to open the SRVs without resulting in unnecessary wear on the valves and to ensure that proper reactor pressure control can be maintained while opening and reclosing the valves.
Reactor steam conditions which are considered adequate to perform the test thus include the establishment of-sufficient reactor pressure as well as sufficient steam flow to ensure that the steam relieved by the SRVs can be compensated by the reactor pressure control system.
The low-low set system ensures that safety / relief valve discharges are minimized for a second opening of these valves, followirg 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 CLINTON - UNIT 1 B 3/4 4-3 Amendment No. 81
REACTOR COOLANT SYSTEM BASES 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 setpoint does not violate the design basis.
3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 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 recommendations 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 monitoring 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 IDENTIFIED 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 somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or 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 Position 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 values 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 isola-tion valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
CLINTON - UNIT 1 B 3/4 4-4 Amendment No. 81
3/4.5 EMERGENCY CORE COOLING SYSTEM i
BASES 3/4.5.1 AND 3/4.5.2 ECCS - OPERATING AND SHUTDOWN l
ECCS division 1 consists of the low pressure core spray system and low pressure coolant injection subsystem "A" of the RHR system and the automatic depressuriza-tion system (ADS) as actuated by ADS trip system "1".
ECCS division.2 consists-of low pressure coolant injection subsystems "B" and "C" of the RHR system and the automatic depressurization system as actuated by' ADS trip system "2".
The low pressure core spray (LPCS) system is provided to assure that the core is adequately cooled folowing a loss-of-coolant accident and, together with the LPCI system, provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller.
breaks following depressurization by the ADS.
The LPCS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental-draining.
The surveillance requirements provide adequate assurance that the LPCS system will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
t The low pressure coolant injection (LPCI) mode of the RHR system.is provided to assure that the core is adequately cooled following a loss-of-coolant accident.
The LPCI system, together with the LPCS system, provide adequate core flooding for all break sizes up to and including the double ended reactor recirculation line break, and for small breaks following depressurization by the ADS.
The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required.
Although all active components are testable.
and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. 'The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
ECCS division 3 consists of the high pressure core' spray system.
The high pres-sure core spray (HPCS) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a-small break-in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. 'The HPCS system permitsLthe reactor to be shut down while maintaining sufficient reactor vessel water level inventory untti the vessel is depressurized.
The HPCS system operates over a range of 1177 psid, differential pressure between reactor vessel and HPCS suction source, to O psid.
CLINTON - UNIT 1 B 3/4 5-1
EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING AND SHUTDOWN (Continued)
The capacity of the system is selected to provide the required core cooling.
The HPCS pump is designed to deliver greater than or equal to 467/1400/5010 gpm at differential pressures of 1177/1147/200 psid.
Initially, water from the reactor core isolation cooling-(RCIC) tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the RCIC tank water.
With the HPCS system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems.
In addition, the reactor core isolation cooling system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition.
The HPCS out-of-service period of 14 days, as specified in the corresponding ACTION statement, is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.
The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage.
Upon failure of the HPCS system to function properly after a small break loss-of-coolant accident, the automatic depressur'tation system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200*F. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig.
This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS.
ADS automatically controls seven selected safety-relief valves although the safety analysis only takes credit for six valves.
It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliability.
The surveillance requirements for the ADS include a requirement to manually open each ADS valve. This requirement includes an exception to the provisions of Specification 4.0.4.
This exception allows reactor steam conditions to be established which are adequate to open the ADS valves without resulting in unnecessary wear on the valves and to ensure that. proper reactor pressure control can be maintained while opening and reclosing the valves.
Reactor steam conditions which are considered adequate to perform the test thus include the establishment of sufficient reactor pressure as well as sufficient CLINTON - UNIT 1 B 3/4 5-2 Amendment No. 81
EMERGENCY CORE COOLING SYSTEM BASES steam flow to ensure that the steam relieved by the ADS valves can be compensated by the reactor pressure control system.
3/4.5.3 SUPPRESSION POOL The suppression pool is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA. This limit on suppression pool minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core. The OPERABILITY of the suppression pool in OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.3.1.
Repair work might require making the suppression pool inoperable. This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression pool must ha made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5.
In OPERATIONAL CONDITIONS 4 and 5 the suppression pool minimum required water volume is reduced because the reactor coolant is maintained at or below 200'f.
Since pressure suppression is not required below 212*F, the minimum required water volume is based on NPSH, recirculation volume and vortex prevention plus a safety margin of 2'4" for conservatism.
CLINTON - UNIT I B 3/4 5-3 Amendment No. 81
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