ML20064D709
| ML20064D709 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/08/1994 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20064D714 | List: |
| References | |
| NUDOCS 9403110377 | |
| Download: ML20064D709 (20) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
- f WASHINGTON, D.C. 20555 4 001 COMM0NWEALTli EDIS0N COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET NO. 50-254 00AD CITIES NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.144 License No. DPR-29 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated October 21, 1993, complies with the standards and requirements of the A b mic 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 Commission; 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.
The issuance of this amendment is in accordance with 10 CFR Part 51 of 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 3.B. of Facility Operating License No. DPR-29 is hereby amended to read as follows:
9403110377 940308 PDR ADOCK 05000254 P
1 B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 144, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION litV I ll (1 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:
March 8. 1994
ATTACHMENT TO LICENSE AMENDMENT NO. 144 FACILITY OPERATING LICENSE NO. OPR-29 DOCKET NO. 50-254 i
Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are i
identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3.5/4.5-2 3.5/4.5-2 3.5/4.5-3 3.5/4.5-3 3.5/4.5-4 3.5/4.5-4 3.5/4.5-5 3.5/4.5-5 3.5/4.5-15 3.5/4.5-15 3.5/4.5-16 3.5/4.5-16 3.5/4.5-17 3.5/4.5-17 3.5/4.5-23 3.5/4.5-23 l
QUAD-CITIES DPR-29 e.
Core spray header op instrumentation check Once/ day calibrate Once/3 months test Once/3 months f.
Logic system Once/Each functional refueling test ot:tage 2.
From and after the date that one 2.
LPCI mode of the RHR system of the core spray subsystems is testing shall be as specified in made or found to be inoperable Specifications 4.5.A.I.a b, c, for any reason, continued d, and f, except that each LPCI reactor operation is permissible division (two RHR pumps per only during the succeeding 7 division) shall deliver at least days unless such subsystem is 9000 gpm against a system head sooner made operable, provided corresponding to a reactor vessel that during such 7 days all pressure of 20 psig, with a active components of the other minimum flow valve open.
core spray subsystem and the LPCI mode of the RHR system and the diesel generators required for operation of such components if no external source of power were available shall be operable.
3.
The LPCI mode of the RHR system shall be operable whenever irradiated fuel is in the reactor vessel and prior to reactor startup from a cold condition.
4.
From and after the date that one of the RHR pumps is made or found to be inoperable for any reason, continued reactor opera-tion is permissible only during the succeeding 30 days unless such pump is sooner made oper-able, provided that during such 30 days the remaining active components of the LPCI mode of the RHR, containment cooling 3.5/4.5-2 Amendment No. 144
QUAD-CITIES DPR. mode of the RHR, all active components of both core spray subsystems, and the diesel generators required for operation of such components if no external source of power were available shall be operable.
5.
From and after the date that the LPCI mode of the RHR system is made or found to be inoperable for any reason, continued reac-tor operation is permissible only during the succeeding 7 days unless it is sooner made operable, provided that during such 7 days all active compo-nents of both core spray sub-systems, the containment cooling mode of the RHR (including two RHR pumps), and the diesel gen-erators required for operation of such components if no exter-nal source of power were avail-able shall be operable.
6.
If the requirements of Specifi-cation 3.5.A cannot be met, an orderly shutdown of the reactor shall be initiated, and the re-actor shall be in the cold shut-down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Containment Cooling Mode of the RHR B.
Containment Cooling Mode of the System RHR System Surveillance of the containment cooling mode of the RHR system shall be performed as follows:
1 1.
a.
Both loops of the 1.
RHR service water subsystem containment cooling mode testing:
of the RHR system, as defined in the bases for Item frequency Specification 3.5.8, shall be operable whenever a.
Pump and valve Once/3 irradiated fuel is in the operability months reactor vessel and prior to reactor startup from a cold condition.
3.5/4.5-3 Amendment No. 144 l
QUAD-CITIES DPR-29 1.
b.
From the effective date of this b.
Flow rate test - After pump amendment until November 1, each RHR service maintenance 1989, the "B" loop of the water pump shall and every containment cooling mode of the deliver at least 3 months RHR system for each reactor may 3500 gpm against share the Unit 1 "C"
and "D" RHR a pressure of 198 service water pumps using cross tie line 1/2-10509-16"-D.
c.
A logic system Each Consequently, the requirements functional test refueling of Specifications 3.5.B.2 and outage 3.5.B.3 will impose the corresponding surveillance testing of equipment associated with both reactors if the shared RHR service water pump or pumps, or the cross tie line, are made or found to be inoperable.
2.
From and after the date that one of the RHR service water pumps is made or found to be inoperable for any reason, continued reactor operation is permissible only during the succeeding 30 days unless such pump is sooner made operable, provided that during such 30 days all other active components of the containment cooling mode of the RHR system are operable.
3.
From and after the date that one loop of the containment cooling mode of the RHR system is made or found to be inoperable for any reason, continued reactor operation is permissible only during the succeeding 7 days unless such subsystem is sooner made operable, provided that all active components of the other loop of the containment cooling mode of the RHR system, both core spray subsystems, and both diesel generators required for operation of such components if no external source of power were available, shall be operable.
3.5/4.5-4 Amendment No.144
QUAD-CITIES DPR-29 4.
Containment cooling spray loo)s 2.
During each 5-year period, an air are required to be operable w1en test shall be performed on the the reactor water temperature is drywell spray headers and nozzles greater than 212*F and prior to and a water spray test performed reactor startup from a cold con-on the torus spray header and dition.
Continued reactor oper-nozzles.
ation is permitted provided that a maximum of one drywell spray loop may be inoperable for 30 days when the reactor water tem-perature is greater than 212*F.
5.
If the requirements of 3.5.B cannot be met, an orderly shutdown shall be initiated, and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C.
HPCI Subsystem C.
HPCI Subsystem 1.
The HPCI subsystem shall be Surveillance of HPCI subsystem operable whenever the reactor shall be performed as specified pressure is greater than 150 below with the following psig and fuel is in the reactor limitations.
For item 4.5.C.3, vessel.
the plant is allowed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in which to successfully complete the 2.
During startup following a test once reactor vessel pressure refuel outage or an outage in is adequate to perform each test.
which work was performed that In addition the testing required directly affects HPCI system by item 4.5.C.3.a shall be operability, if the testing completed prior to exceeding 325 requirements of 4.5.C.3.a cannot psig reactor vessel pressure.
If be met, continued reactor HPCI is made inoperable to perform startup is not permitted. The overspeed testing, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is HPCI subsystem shall be declared allowed to complete the tests inoperable, and the provisions before exceeding 325 psig.
of Specification 3.5.C.4 shall l
be implemented.
Item Freauency i
3.
Except for the limitation of 1.
Valve Position Every 31 days 3.5.C.2, if the HPCI subsystem is made or found to be 2.
Flow Rate Test-Every 92 days inoperable, continued reactor HPCI Pump shall
'l operation is permissible only deliver at least during the succeeding 14 days 5000 gpm against unless such subsystem is sooner a system head cor-made operable, provided that responding to a during such 14 days the reactor vessel automatic pressure relief pressure of 2 1150 j
subsystems, the core spray psig when steam is subsystems, LPCI mode of the RHR being supplied to system and the RCIC system are the turbine at 920 operable. Otherwise, the to 1005 psig.
provision of Specification i
3.5.C.4 shall be implemented.
3.5/4.5-5 Amendment No. 144
QUAD-CITIES DPR-29 3.5 LIMITING CONDITIONS FOR OPERATION BASES A.
Core Spray and LPCI Mode of the RHR System This specification assures that adequate emergency cooling capability is available whenever irradiated fuel is in the reactor vessel.
Based on the loss-of-coolant analytical methods described in General Electric Topical Report NEDC-31345P core cooling systems provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident, to limit calculated fuel cladding temperature to-less than 2200*F, to assure that core geometry remains intact, to limit cladding metal-water reaction to less than 1%, and to limit the calculated local metal-water reaction to less than 17%.
The limiting conditions of operation in Specifications 3.5.A.1 through 3.5.A.6 specify the combinations of operable subsystems to assure the availability of the minimum cooling systems noted above.
Core spray distribution has been show, in full-scale tests of systems similar in design to that of Quad-Cities 1 and 2, to exceed the minimum requirements by at least 25%.
In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additionally conservative in that no credit is taken for spray cooling of the reactor core before the internal pressure has fallen to 90 psig.
The LPCI mode of the RHR system is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident.
This system functions in combination with the core spray system to prevent excessive fuel cladding temperature.
The LPCI mode of the RHR system in combination with the core spray subsys cooling,forbreakareasofapproximately0.05ft}emprovidesadequate up to and including 4.26 ft, the latter being the double-ended recirculation line break with the equalizer line between the recirculation loops closed without assistance from the high-pressure emergency core cooling subsystems.
The allowable repair times are established so that the average risk rate for repair would be no greater than the basis risk rate.
The method and concept are described in Reference 3.
Using the results developed in this reference, the repair period is found to be less than 3.5/4.5-15 Amendment No. 144
i QUAD-CITIES DPR-29 1
half the test interval. This assumes that the core spray subsystems and LPCI constitute a one-out-of-two system; however, the combined effect of 1
the two systems to limit excessive cladding temperature must also be i
considered. The test interval specified in Specification 4.5 was 3 months.
Therefore, an allowable repair period which maintains the basic risk considering single failures should be less than 30 days, and this specification is within this period. Although it is recognized that the information given in Reference 1 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach 3revents complete acceptance of this method at this time. Therefore, tie times stated in the specific items were established with due regard to judgment.
Should one core spray subsystem become inoperable, the remaining core spray subsystem and the entire LPCI mode of the RHR system are available should the need for core cooling arise.
Based on judgments of the reliability of the remaining systems, i.e., the core spray and LPCI, a 7-day repair period was obtained.
Should the loss of one RHR pump occur, a nearly full complement of core and containment cooling equipment is available. Three RHR pumps in conjunction with the core spray subsystem will perform the core cooling function.
Because of the availability of the majority of the core cooling equipment, a 30-day repair period is justified.
If the LPCI mode of the RHR system is not available, at least two RHR pumps must be available to fulfill the containment cooling function.
The 7-day repair period is set on this basis.
B.
RHR Service Water The containment cooling mode of the RHR system is provided to remove heat energy from the containment in the event of a loss-of-coolant accident.
For the flow specified, the containment long-term pressure is limited to less than 8 psig and is therefore more than ample to provide the required heat-removal capability (reference SAR Section 5.2.3.2).
3.5/4.5-16 Amendment No.144
QUAD-CITIES DPR-29 The Containment Cooling mode of the RHR System consists of two loops.
Each loop consists of I Heat Exchanger, 2 RHR Pumps, and the associated valves, piping, electrical equipment, and instrumentation.
The "A" loop on each unit contains 2 RHR Service Water Pumas. Until November 1, 1989, the "B" loop on each unit may utilize tie "C" and "D"_RHR Service Water Pumps from Unit 1 via a cross-tie line. After November 1, 1989, each "B" loop will contain 2 RHR Service Water Pumps.
Either set of equipment is capable of performing the containment cooling function.
Loss of one RHR service water pump does not seriously jeopardize the containment cooling capability, as any one of the remaining three pumps can satisfy the cooling requirements.
Since there is some redundancy lef t, a 30-day repair period is adequate.
Loss of one loop of the containment cooling mode of the RHR system leaves one remaining system to perform the containment cooling function.
Based on the fact that when one loop of the containment cooling mode of the RHR system becomes inoperable, only one system remains, a 7-day repair period was specified.
C.
High-Pressure Coolant Injection The high-pressure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RHR system or core spray subsystems can protect the core.
The HPCI meets this requirement without the use of offsite electrical power.
For the pipe breaks for which the HPCI is intended to function, the core never uncovers and is continuously cooled, thus no cladding damage cccurs (reference SAR Section 6.2.5.3).
The repair times for the-limiting conditions of operation were set considering the use of the HPCI as part of the isolation cooling system.
D.
Automatic Pressure Relief The relief valves of the automatic pressure relief subsystem are a backup to the HPCI subsystem.
They enable the core spray subsystem and LPCI mode of the RHR system to provide protection against the small. pipe break in the event of HPCI failure by depressurizing the reactor vessel rapidly enough to actuate the core spray subsystem and the LPCI mode of the RHR system.
The core spray subsystem and the LPCI mode of the RHR system provide sufficient flow of coolant to limit fuel cladding temperatures to less than 2200*F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than 1%, and to limit the calculated local metal-water reaction to less than 17%.
3.5/4.5-17 Amendment No. 144
QUAD-CITIES DPR-29 4.5 SURVElt. LANCE REQUIREMENTS BASES The testing interval for the core and containment cooling systems is based on a quantitative reliability analysis, judgment, and practicality.
The core cooling systems have not been designed to be fully testable during operation.
For example, the core spray final admission valves do not open until reactor pressure has fallen to 350 psig.
Thus, during operation, even if high drywell pressure were simulated, the final valves would not open.
In the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable.
With a system, subsystem, loop, or equipment out-of-service, overall core and containment cooling reliability is maintained by verifying the operability of the remaining systems, subsystems, loops, or equipment. The verification of operability, as used in this context, for the remaining cooling systems means to administratively check by examining logs or other information to verify that the remaining systems are not out-of-service for maintenance or other reasons.
It does not mean to perform the surveillance requirements needed to demonstrate the operability of the remaining systems. However, if a failure, design deficiency, etc., causes the out-of-service period, then the verification of operability should be thorough enough to assure that a similar problem does not exist on the remaining systems.
For example, if an out-of-service period is caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test.
The surveillance requirements bases described in this paragraph apply to all core and containment cooling systems except HPCI and RCIC.
The systems can be automatically actuated during a refueling outage and this will be done.
To increase the availability of the individual components of the core and containment cooling systems, the components which make up the system, i.e., instrumentation, pumps, valve operators, etc., are tested more frequently. The instrumentation is functionally tested each month.
Likewise the pumps and motor-operated valves are also tested each month to assure their operability. The combination of a yearly simulated automatic actuation test and monthly tests of the pumps and valve operators is deemed to be adequate testing of these systems.
l The surveillance requirements bases described in this paragraph apply only to the RCIC and HPCI systems.
Following a refueling outage or an outage in which work was performed that directly affects system operability, the HPCI and RCIC pumps are flow rate tested prior to exceeding 325 psig and again at rated reactor steam pressure.
This combination of testing provides adequate assurance of pump performance throughout the range of reactor pressure at which it is i
3.5/4.5-23 Amendment No. 144
k k
UNITED STATES
( t,jg 7 )
NUCLEAR REGULATORY COMMISSION 5
e p W ASHINGTON, D C. 205 %-0001 COMMONWEALTH EDIS0N COMPANY MD IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET N0. 50-265 OVAD CITIES NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.140 License No. DPR-30 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated October 21, 1993, 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 f acility will 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 Commission's regulations; i
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 Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-30 is hereby amended to read as follows:
. B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 140, are hereby incorporated in the i
license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
-tin { b lt' j
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:
March 8, 1994 I
.. =..
. ~.
.. =. _...
4 S
ATTACHMENT TO LICENSE AMENDMENT NO. 140 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 50-265 F
Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3.5/4.5-2 3.5/4.5-2 3.5/4.5-2a 3.5/4.5-3 3.5/4.5-3 3.5/4.5-4 3.5/4.5-4 3.5/4.5-11 3.5/4.5-11 3.5/4.5-12 3.5/4.5-12 3.5/4.5-15 3.5/4.5-15
'r 5-i L
r
P QUAD CITIES DPR-30 e.
Core spray hesder
+
Ap instrumentation check Once/
day calibrate Once/3 i
months test Once/3-months f.
Logic Once/
system
.each functional refueling test outage 2.
From and after the date tnat one of the 2.
LPCI mode of the RHR system testing lt core spray subsystems is made or found shall be as specified.1n Specifications to be inoperable for any reason.
4.5.A.I.a. b, c. d. and f except that continued reactor operation is each LPCI division (two RHR pumps per permissible only during the succeeding 7 division) shall deliver at least 9000 days unless such subsystem is sooner gpm against a system head corresponding made operable, provided that during such to a reactor vessel pressure of-20 psig, 7 days all active components of the with a minimum flow valve open.
other core spray subsystem and the LPCI made of the RHR system a'id the diesel generators required for operation of such components if no external source of power were available shall be operable.
-3.
The LPCI mode of the RHR system shall be operable whenever 1rradiated fuel is in tne reactor vessel and crior to reactor j
startup from a cold condition.
4 from and after the cate that one of the RhR pumps is made or found to be i
inoperable for any reason, continued reactor operation is permissible only
-l during the succeeding 30 days unless such pump is sooner made operable.
provided that dJring such 30 days the remaining active components of the LPCI mode of the RHR. containment cooling mode of the RHR. all active components of both core spray subsystems..and the diesel generators required for operation of such components if no external source of power were available shall be operable.
5.
From and after the date that the LPCI mode of the RHR system is made or found to be inoperable for any reason'.
3.5/4.5-2
' Amendment No.140 -
l I
Q'JA0 CITIES DPR 30 continued reactor operation is permissible only during the succeeding 7 days unless it is sooner made operable, provioed that during such 7 days all active components of both core spray subsystems, the containment cooling mode of the RNR (including two RHR puilDs),
and the diesel generators required for operation of such components if no external source of power were available shall be operable.
6.
If the requirements of Speci-fication 3.5. A cannot be Fet, an orderly shutdown of the reactor shall be initiated, and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Containment Cooling Mode of the RHR B.
Containment Cooling Mode of the RHR System System Surveil (ance of the containment cooling mode of the RHR system shall be performed as follows:
1.
a.
Beth locps of the 1.
RNR service water subsystem containment cooling mode testing; of the RhR system, as oefined in the bases for item Frequency Speci fication 3.5.B, shall be operable a.
Pump and valve Once/3 whenever irradiated fuel operability months is in the reactor vessel and prior to reactor b.
Flow rate After pump startup f rom a cold test - each Faintenance condition.
RNR service and every water ptanp 3 months 1.
b.
From the effective date shall deliver of this anendment until at least 3500 hov. 1, 1989, t he "B"
gpm against a loop of the containment pressure of 198 cooling mode of the RHR psig system for each reactor may share the Unit 1 "C" c.
A logic Each and "D" RHR service water systen refue. ling pumps using cross tie functional outage line 1/2 10509 16" 0.
test Consequently, the requirements of Speci-fications 3.5.B.2 and 3.5 B.3 will impose the corresponding surveillence testing of equirrent associated with both reactors if the shared RNR service water-ptrip or pgs, or the cross tie line, are made or found to be inoperable.
2.
From and after the date that one of the RHR service water perps is nede or f ound to be inoperable f or any reason, continued reactor operation is permissible only during the succeeding 30 days unless such ptrip is sooner made operable, provided that during such 30 days all otner active components of the containment cooling mode of 3.5/4.5-3 Amendment ko.
14
QUAD CITIES-DPR-30 the RHR $ystem are operable.
3.
from and after the date that one loop of the containment cooling mode of the RHR system is made or found to be inoperable f or any reason, continued reactor operation is permissible only during the succeeding 7 days unless such subsystem is sooner made operable, provided that all active components of the other loop of _ the containment cooling mode of the RHR system, both core spray subsystems, and both diesel generators required for operation of such components if no external source of power were available, shall be i
During the time period from April 17, 1978 through April 30, 1978 while the 2A Containment Cooling Loop of the RHR System is made inoperable f or heat exchanger repair, continued reactor operation is permissible beyond the above 7 cay limitation, unless such loop is sooner made operable, provided that during the time the 7 day limit is exceeded, a visual inspection is performed daily to assure that proper valve alignment and system g
integrity is maintained in the 2.
During each 5 year period, an air test l
8'B" ReR loep, shall be performed on the drywett spray headers and nortles and a water spray 4.
Containment cooling spray test performed on ti.e torus spray header loops are required to be and nozzles.
operable when the reactor water temperature is greater than 212'F and prior to reactor startup f rom a cold condition.
Continued reactor operation is permitted provided that a maximum of one drywell spray loop may be inoperable for 30 days when the reactc* water temperature is greater than 2'2*F.
5.
If the requirements of 3.5.B cannot be net, an orderly shuidown shall be initiated, and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i l
3.5/4.5-4 Amendment No. 140
QUAD CITIES DPR-30 3.5 LIMIT!hG CONDITIONS FOR OPERATION BASES A.
Core Spray and LPCI Mode of the RHR System This specification assures that adequate emergency cooling capability is available.
Based on the loss of coolant analyses included in References 1 and 2 and in accordance with 10 CFR 50.46 and Amendix K, cere cooling systems provide suf ficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident, to limit the calculated fuel cladding tecperature to less than 2200cF, to assure that core geometry remains intact to limit the corewide cleading metal-water reaction to less than 1% and to limit the calculated local metal-water reaction to less than 17%.
The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in Reference 3.
Using the results developed in this reference, the repair period is found to be less than half the test interval. This asstanes that the core spray subsystems and LPCI constitute a one-out of-two system; however, the combined effect of the two systems to limit excessive cladding temerature must also be considered. The test interval specified in Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the basic risk considering single f ailures should be less than 30 days, and this specification is within this period.
Although it is recognized that the information given in Reference 3 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents complete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due regard to judgment.
Should one core spray subsystem become inoperable, the remaining core spray subsystem and the entire LPCI mooe of the RHR system are available should the need for core cooling arise. Based l
on judgments of the reliability of the remaining systems, i.e., the core spray and LPCI, a 7 day repair period was obtained.
Should the loss of one RHR pmp occur, a nearly full complement of core and containment cooling equipment is available. Three RHR purps in conjunction with the core spray subsystem will perf orm the core cooling f unction. Because of the availability of the majority of the core cooling equipment, a 30-day repair period is justified. If the LPCI mode of the RHR system is l
not available, at least two RhR ptrps must be available to fulfill the containment cooling function. The 7-day repair period is set on this basis.
B.
RHR Service Water The containment cooling mode of the RHR system is provided to remove heat energy from the containment in the event of a less-of coolant accident. For the flow specified, the containment long-term pressure is limited to less than 8 psig and is therefore more than ample to provide the required heat-removal capability (reference SAR Section 5.2.3.2).
The Containment Cooling mode of the RHR System consists of two loops. Each loop consists of 1 Heat Exchanger, 2 RHR P mps, and the associated valves, piping, electrical equipment, and instrumentation. The "A" loop on each unit contains 2 RHR Service Water Pmps. Until hov. 1, 1989, the "B" loop on each unit may utilize the "C" and "D" RHR service Water Ptrps f rom Unit 1 via a cross-tie line. After Nov. 1, 1989, each "B" loop will contain 2 RHR Service Water Pmps. Either set of equipment is capable of performing the containment cooling function.
Loss cf one RHR service water ptrp does not seriously jeopardite the containment cooling capability, as any one of the remaining three ptrps can satisfy the cooling requirements.
Since there is some redundancy left, a 30-day repair period is adequate. Loss of one loop of the containment cooling mode of the RHR system leaves one remaining system to perform the contairvnent cooling function. Based on the fact that when one system of the containment cooling mode of the RHR system becomes inoperable, only one system remains, a 7-day repair period was specified.
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QUAD CITIES DPR-30 C.
High-Pressure Coolant Injection The high pressure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RHR system or core spray subsystems can protect the core.
The HPCI meets this requirement without the use of offsite electrical power. For the pipa breaks f or which the HPC) is intended to function, the core never uncovers and is coC T r sty cooled, thus no cladding damage occurs (reference SAR Section 6.2.5.3).
The repal' t t
- e the limiting conditions of operation were set considering the use of the HPCI as 9 art - o 1
isolation cooling system.
D.
Automatic Pressure Relief The relief valves of the automatic pressure relief subsystems are a backup to the HPCI subsystem. They enable the core spray subsystem and LPCI mode of the RHR system to provide j
protection against the small pipe break in the event of HPCI f ailure by depressurizing the reactor vessel rapidly enough to actuate the core spray subsystems and LPCI mode of the RHR l
system. The core spray subsystem and/or the LPCI mode of the RHR system provide sufficient J
flow of coolant to limit fuel cladding terperatures to less than 2200'F, to assure that core I
geometry remains intact, to limit the core wide clad meta'.-water reaction to less than 1%, and I
to limit the calculated local metal-water reaction to less than 17%.
Analyses have shown that only f our of the five valves in the automatic depressurization system are required to operate. Loss of one of the relief valves does not significantly affect the pressure relieving capability, therefore continued operation is acceptable. Loss of two relief valves significantly reduces the pressure relief capability of the ADS: thus, a 7 day repair period is specified with the NPCI available, and a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> repair period with the HPCI unavailable.
E.
RCIC The RCIC system is provided to supply continuous makeup water to the reactor core when the reactor is isolated from the turbine and when the feedwater system is not available. Under these conditions the ptriping capacity of the RCIC system is suf ficient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vesset decreases to the RCIC initiation level, the system automatically starts. The system may also be manually initiated at any time.
The NPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable. Therfore, the specification calls for an operability check of the HPCI system should the RCIC system be found to be inoperable.
F.
Emergency Cooling Availability the purpose of Specification 3.5.F is to assure a minimtrn of core cooling equipment is available at all times.
If, for example, one core spray were out of service and the diesel which powered the opposite core spray were out of service, only two RHR pumps would be available. Likewise, if two RHR ptrips were out of service bnd two RHR service water ptrips.on the opposite side were also out of service no contairment cooling would be available. It is during the refueling outages that major maintenance is performed and during such time that all low pressure core cooling systems may be out of service. This specification provides that should this occur, no work will be performed on the primary system which could lead to draining the vessel. This work would include work on certain control rod drive comporwnts and recirculation systems. Thus, the specification precludes the events which could require core cooling. Specification 3.9 must also be consulted to determine other requirements for the diesel generators.
Quad Cities Units 1 and 2 share certain process systems such as the makeup demineralizers and the redwaste system and also some safety systems such as the standby gas treatment system, batteries, and 3.5/4.5-12 Amendment No. 140-
j QUA0 CITIES DPR 30 4.5 5URVE!LLAhCE REQUIREMEhi$ 8ASES
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The testing interval for the core and containment coeling systems is based on a quantitative reliability analysis, judgment, and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, the core spray final admission valves do not open until reactor i
pressure has fallen to 350 psig. Thus, during operation, even if high drywell pressure were simleted, the final valves would not open. In the case of the hPCI, automatic initiation during power operation would result in ptrping cold water into the reactor vessel which is not desirable.
With a system, subsystem, loop or equipenent out of service, overall core and containment cooling reliability is maintained by verifying the operab lity of the remaining systems, subsystems, loops or equipment. The verification of operability, as used in this context, for the remaining cooling systems means to adninistratively check by examining logs or other information to verify that the remaining systems are not out of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the remaining systems. However, if a f ailure, design deficiency, etc., causes the out of service period, then the verification of operability should be thorough enough to assure that a similar problem does not exist on the remaining systems. For exa cle, if an out-of service period is caused by f ailure of a ptro to deliver rated capacity due to a oesign deficiency, the other ptrps of this type might be subjected to a flow rate test.
The surveillance requirements bases in this paragraph apply to all core and containment cooling systems except RCIC and HPCI. The systems can be automatically actuated during a refueling outage and this will be done. To increase the availability of the individual corponents of the core and containment cooling systems, the components which make up the system, i.e.,
instrumentation, ptros, valve operators, etc.,
are tested more frequently. The instrumentation is functionally tested each month. Likewise the punes and motor operated valves are also tested each month to assure their operability. The combination of a yearly simlated automatic actuation test and monthly tests of the pumps and valve operators is oeemed to be acequate testing of these systems.
The surveillance requirements bases described in this paragraph apply only to the RCIC and HPCI systems, f ollowing a ref ueling outage or an outage in which work was performed that directly af fects system operability, the HPCI and RCIC piros are flow rate tested prior to exceeding 325 psig and again at rated reactor steam pressure. This combination of testing provides adequate assurance of pump performance througeout the range of reactor pressures at which it is required to operate. The low pressure limit is selected to allow testing at a point of stable plant operation and also to provide overlap with low pressure ECC systems. A time limit is provided in which to perform the required tests during startup.
This time limit is considered adequate to allow stable plant conditions to be achieved and the required tests to be performed. Flow rate testing of the NPCI and RCIC purps is also conducted every 92 days at rated reactor pressure to demonstrate system operability in accordance with the LCO provisions and to I
meet inservice testing requirements for the HPCI system. Applicable valves are tested in accordance with the provisions of the inservice testing program. In addition, monthly checks are made on the position of each manual, power operated or automatic valve installed in the direct flowpath of the suction or discharge of the ptrp or turbine that is not locked, sealed, or otherwise secured in position. At each refueling outage, a logic system functional test and a simulated automatic actuation test is performed on the HPCI and RCIC systems. The tests and checks described above are considered adequate to assure system operability.
The verification of the main steam relief valve operability during manual actuation surveillance testing must be made independent of terperatures indicated by thermocouples downstream of the relief valves, it has been found that a teeperature increase may result with the valve still closed. This is due to steam being vented through the pilot valves during the surveillance test. By 4
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