ML20149C612
| ML20149C612 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/03/1988 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Iowa-Illinois Gas & Electric Co |
| Shared Package | |
| ML20149C616 | List: |
| References | |
| DPR-29-A-104, DPR-30-A-100 NUDOCS 8802090215 | |
| Download: ML20149C612 (10) | |
Text
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o UNITED STATES
g
[
g NUCLEAR REGULATORY COMMISSION g
a WASHINGTON, D C. 20555
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COMMONWEALTH EDISON COMPANY f!LD IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET N0. STN 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 104 License No. DPR-29 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated October 6, 1987 as supplemented by l
November 24, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; l
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health ano safety of the public, and (ii) that such activities will be conoucted in compliance with the Comission's regulations i
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 LFR 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 folicws:
G802090215 880203 DR ADOCK 050 4
.- B.
Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No.104, are hereby incorpurated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendirent is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Daniel R. Muller Director Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects Attachnent:
Changes to the Technical Specifications Date of Issuance: February 3, 1988 i
i
ATTACHMENT TO LICENSE AMENDMENT NO.104 FACILITY OPERATING LICENSE NO. DPR-29 DOCKET NO. 50-254 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 nunber and contain marginal lines indicating the area of change.
REMOVE INSERT l
3.5/4.5-8 3.5/4.5-8 3.5/4.5-16 3.5/4.5-16 t
l
...........~.._
- 4 l
QUAD-CITIES OPR-29 2.
The discharge pipe pressure for 2.
Following any period where HPCI, Core Spray and LPCI mode of RHR RCIC, LPCI mode of W RHR or shall be maintained at greater core spray have been out of W n 40 psig and less than 90 service and deelned for moln-psig, if pressure in any of tenance, the discharge piping of Nse systems is less than 40 the Inoperable system shall be psig or greater than 90 psig, vented from the high point prior this condition shall be alarmed to the return of N system to in the control roca and service.
Iss'ediate corrective action taken, if N discharge pipe 3.
Whenever the HPCI or RCIC system pressure is not within these is lined up to take suction from limits in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> u ter W the torus. the discharge piping occurrence, en orderly shutdown of the HPCI and RCIC shall be shall be initiated, and N re-vented from the high point of actor shall be in a cold shut-the system and water flow ob-down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> served every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
efter initiation.
4.
The pressure switches which non-3.
Filled discharge piping for HPCI Itor N discharge lines and N and RCIC systems is ensured by discharge of the fill system maintaining N level in the pw'c to ensure that N y are Cont elnated Condensate Storage full shall be functionally Tanks (CCST's) at or above 9.5 tested every month and call-feet, if the CCST level falls brated every 3 months. The below 9.5 feet, restore N pressure switches shall be set level within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or line up to alern at a decreasing pres-both HPCI and RCIC to take a sure cf > 40 psig and an in-suction from W torus per creasing pressure of i 90 psig.
4.5.G.3.
H.
Condensete Pung Roca Flood Protection H.
Condensate Pw'p Room Flood Protection 1.
The following surveillance re-1.
The systems installed to prevent qulraments shall be observed to or mitigate W consequences of assure that the condensate pwip flooding of the condensate pwie room flood protection is oper-roce shall be operable prior to
- able, startup of W roector, s.
The piping and electrical 2.
The condenser pit water level penetrations, bulkheed switches shall trip the condon-doors, and sutaarine doors ser circulating water pwips and for N vaults containing alern in the control room if wa-N RHR service water puips ter level In W condenser pit and diesel generator cooling exceeds a level of 5 feet above pwips shall be checked W pit floor. If a failure oc-during each operating cycle cuts in one of Wse trip and by pressurlaing to 15 + 2 alors circuits, N failed cir-psig and checking for Teaks cult shall be Innediately placed using a soap bubble in a trip condition and reactor solution. The criterla for operation shall be permissible ecceptance shall be no for W following 7 days unless visible leakage through the W circuit is sooner made oper-soap bubble solution.
able.
3669K 3.5/4.5-8 Amen *ent No.
104
QUAD-CITIES DPH-29 4.5 SURVEILLANCE REQUIREMENTS BASES The testing interval for the core and cont'airvient cooling systems is based on a quantitative The core cooling systems have not been designed reliability analysis, judpent, and practicality.For emanple, W core spray final adelssion valves do not to be fully testable during operation.
open until reactor pressure has fallen to 350 psig. Thus, during operation, even if high drywell In the case of the HPCI, automatic pressure were simulated, the final valves would not open.initletion during power operation would resu is not desirable.
To The systems can be auton.atically actuated during a refueling outage and this will be done.
Increase N avallebility of N individual cmponents of the core and containment cooling systems, N cwponents which make up the system, f.e., instrwnentation, pwps, valve operators, i
l The instrwentation is functionally tested each nonth. Likewise The
)
etc., are tasted more frequently.
N peps and motor-operated valves are also tested each month to assure Wir operability.
1 conbination of a yearly simulated automatic actuation test and monthly tests of W punps and valve operators is deemed to be adequate testing of these systems.
With emponents or subsystems out of service, overall core and containment cooling reliability is j
maintained by denonstrating N operability of the remaining cooling equipment. The degree of
)
operability to be demonstrated depends on W nature of N reason for the out-of-service For routine out-of-service periods caused by preventative maintenance, etc., the pung equipnent.
and valve operability checks will be performed to demonstrate operability of N remaining if a failure, design deficiency, etc., causes W out-of-servlee period, caponents. However, then W demonstration of operability should be thorough enough to assure that a similar problem j
For sample, if an out-of-service period caused by does not exist on N remaining conponents.
i f ailure of a pep to deliver rated capacity due to a design deficiency, W other pumps of tnis type alght to subheted to a flow rate test in addition to the operability checks.
l The verification of N main steam relief valve operability durIng manual actuation surveillance testing must be made indep. dent of tenperatures indicated by thermocouples downstream of N relief valves, it Pas been found N t a tecperature increase may result with the valve still This is di.e to steam being vented through the pilot valves during N surveillance test.
closed.
By first opening a turbine bypass valve, and then observing its closure response during relief valva actuation, positive verification can be made for the relief valve opening and possing stown Closure response of W turbine control valves during rollef valve manual actuation would flow.
This test nothod may be likewise serve as an adequate verification for the relief valve opening. Valve operation below 150 perforvied over a wide range of reactor pressures greater than 150 psig.
psig is limited by the spring tension exhibited by the relief valves.
The surveillance reautrenants to ensure that the discharge piping of the core spray, LPCI mode of HF<l and RCIC systems is filled provides for a visual observation that water flows frca a W RHR highpolntvent. Tnis ensures Nt tte line is in a full condition.
Instrumentation has been provided on core spray and LPCI mode of RHR to monitor N pressure of I
]
water in the discharge piping between W monthly Intervals at which the lines are vented and This instrumentation will be callbrated on i
alarm ttie control r,xn if ttie pressure is inadequate.
This N same f requency.ns the safety system instreientation and W alerte system tested nonthly.
testing ensures that, during the interval between the sonthly venting checks, the status of the An alors point of 40 psig for N Icre discharge piping is monitored on a continuous basis.
pressure of the fil! system has been chosen because, due to elevations of piping within the plant, 39 psig is required to keep the lines full. The shutoff head of N fill system peps is less N n 90 psIg and W refore will not defeat the bw-pressure cooling pep discharge pressure interlock 100 psig as shown in Table 3.2-2.
A margin of to psig is provided by the high pressure alarm point of 90 prig.
IfCl and RCIC systens normally take a suction from W Contaminated Condensate Storege Tanks (CCST's). The level in the CCST's is maintained at or above 9.5 feet. This level corresponds to an elevation which is greater than W elevation of the last check valves in the discharge pipes of el W r the PCI or RCIC systems. N refore, filled discharge piping of IFCI or RCIC systems is ensured when lined up to N CCSI and tank level is at or above 9.5 feet.
3669K 3.5/4.5-16 Mendeent No.
104 i
/
'o UNI TED STATES 8
~g NUCLEAR REGULATORY COMMISSION n
i WASHING TO N, D. C. 20555
- %...../
COMMONWEALTH EDISON COMPANY AND IOWA-ILLINDIS GAS AND ELECTRIC COMPANY DOCKET NO. STN 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2, AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 100 License No. DPR-30 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated October 6, 1987 as supplemented by November 24, 1987, complies with the standards and requirements of the Atonic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; l
and E.
The issuance of this anendment 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 changer, 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:
i
1 2-B.
Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No.100, are hereby incorporated in this 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 m k h?g Daniel R. Huller, Director Project Directorate III-2 Division of Reactor Projects - III, IV, Y anc Special Projects
Attachment:
Changes to the Technical l
Specifications Date of Issuance:
February 3, 1988
)
i i
i
ATTACHPENT TO LICENSE AMENDMENT NO. 100 FACILITY OPERATING LICENSE N0. DPR-30 DOCKET h0. STN-50-265 Revise the Aprendix A technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned anendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3.5/4.5-8 3.5/4.5-8 3.5/4.5-16 3.5/4.5-16 I
O QUAD-CITIES Offt-30 2.
The discharge pipe pressure for 2.
Following any period where HPCI, Core Spray and LPCI mode of RHR RCIC, LPCI mode of W fMR or shall be maintained at greater core spray have been out of ser.
W n 40 psis and less N n 90 vice and drained for moln-psig, if pressure in any of tenance, W discharge piping of W se systems is less than 40 the Inoperable system snell se vented frca the high point prior psig or greater than 90 psig, d to W return of W system to this condition shall be alors.
in W control room and
- service, isnediate corrective ection taken. If W discharge pipe 3.
Whenever the HPCI or RCIC system pressure is not within W se is lined up to take suction from limits in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after W the torus, the discharge pitting cccurrence, an orderly shutduwn of the HPCI and RCIC small te shall be initiated, and the re-vented from the high point of actor shall be in a cold shut-the system and water flow ob-down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> served every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, after initiation.
4.
The pressure switches which mon-3.
Filled discharge piping for HPCI Itor the discharge lines and W and RCIC systems is ensured by discharge of W fill system maintaining the level in the p g to ensure that N y are Contaminated Condensate Storsos full shall be functionally Tanks (CCST's) at or above 9.5 tested every month and cali-feet. If ttw, CCST level f alls brated every 3 wths. The below 9.5 foot, restore the pressure s=liches shall tm set level within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or line up to alarin at a decreasing pres-both HPCI and RCIC to take a sure of > 40 psig and an in-suction frem the torus per creasing ~ pressure of 5 90 psig.
4.5.G.3.
H.
Condensate P g Room Flood Protection H.
Condensate Pug Room Flood Protection 1.
The systems installed to prevent 1.
The following surveillance re-or mitigate the consequences of quirements shall be observed to flooding of the condensate pg assure that the condensate pug room shall be operable prior to rom flood protection is oper-startup of the reactor.
able.
2.
N condenser pit water level s.
The piping and electrical switches shall trip W condon-penetrations, bulkhead ser circulating water p g s and doors, and submar!ne doors alarin in the control rom if wa-for the vaults containing ter level in W condenser pit W RHR service water p g s exceeds a level of 5 feet above end diesel generator cooling W pit floor, if a failure oc-p g s shall be checked j
curs in one of W se trip and during each operating cycle j
alarm circuits, the failed cir-by pressurl:Ing to 15 + 2 cult shall be inmediately placed pstg and checking for Teoks in a trip condition end reactor using a soap bubble operation shall be permissible solution. The criterla for for W following 7 days unless acceptance shall be no W circuit is sooner mode oper-visible leekage through h able.
soap bubble solution.
3669K 3.5/4.5-a Mane ent no. 100 l
j l
~
QUAD-CITIES DPR-)O 4.5 $URyElLLANCE REQUIREMENTS BASES The testing interval for N core and contairvient cooling systeres is based on a quantitative reliability analysis, judront, and practicality. The core cooling systems have not been designed to be fully testable during operation. For eseg le, W core spray final se lssion valves do not open until reactor pressure has fallen to 350 psig. Thus, during operation, even if high drywell I
pressure were simulated, the final valves wovfd not open. In N case of the HPCl, automatic initiation during power operation would result in p g ing cold water into N reactor vessel which is not desirable.
The systems can be autmatically actuated during a refueling outage and this will be done. To increase the availability of the Individual cm ponents of the core and containment cooling systems, the ccrqponents which make up N system, i.e., instreentation, p g s, valve operators,
)
etc., are tested more f requently. The Instrumentation is functionally tested each month.
Likewise the pugs and motor-operated valves are also tested each month to assure their I
operability. The caelnation of a yearly simleted autmatic actuation test and northly tests of N pgs and valve operators is deemed to be adequate testing of Wse systems.
With cceiponents or subsystems out of service, overall core and contaltvoont cooling reliability is maintained by denenstrating the operability of the remaining cooling equipeent. W degree of operability to be demonstrated depends on the nature of the reason for N out-of-service e qu i p'ent. For routine out.cf-service periods caused by preventative maintenance, etc., W pwp and valve operability checks will te performsd to demonstrate operability of the remaining cavonents. However, if a f ailure, design deficiency, etc., causes the out-of-service period, then the der'onstration of operability should be thorough enough to assure that a similar problem For exariple if an out-of-service period caused by does not exist on N remaining ccroonents.
f ailure of a pwp to de!iver rated capacity due to a des lgn deficiency, W other pumps of this type might be subjected to a flow rate test in addition to the operability checks.
The verification of the main ste m rollef valve operability during manual actuation surveillance testing must be made independen* of tenceratures indicated by thermocouples &wnstream of the relief valves. It has been found that a terverature increase may result with N valve still closed. This is due to stem being vented through the pilot valves during the surveillance test.
By first opening a turbine bypass valve, and then observing its closure response during relief valve actuation, positive verification can be u de for the relief valve opening and passing ste m flow. Closure response of the turbine control valves during relief valve manuel actuation would likewise serve as an adequate verification for the relief valve opening. This test method may be perfcced over a wide range of reactor pressures greater than 150 psig. Valve operation below 150 psig is limited by N spring tension exhibited by the relief valves.
The surveillance requirements to ensure that the discharge piptog of the core spray, LPCI mode of N RHR, HPCI, and RCIC systes is filled provides for a visual observation that water flows f rm a high point vent. This ensures that N line is in a full condition.
instrumentation tas toen provided on Core Spray and LPCI node of RHR to monitor the pressure of water in N discharge piping between N ponthly intervals at which the lines are vented and etern the control rom if the pressure is inadequate. This instrwentation will be calibrated on the sane f requency as W safety system Instreontation and the alarm system tested monthly. This testing ensures that, during the Interval between the monthly venting checks, the status of the discharge piping is sonitored on a continuous basis. An alarm point of > 40 psig for the low pressure of the fill system has been chosen because, due to elevations of piping within the plant.
39 psic is required to keep the lines full. N shutof f head of the fill system pumps is less than 90 psig and Wrefore will not defeat the low-pressure cooling puno discharge press interlock 100 psig as shown in Table 3.2-2.
A margin of 10 psig is provided by N high pressure alarm point of 90 psig.
HPCI and RCIC systems normally take a suction from the Contaminated Condensate Storage Tanks (CCST's). N tevel in N CCST's is maintained at or above 9.5 feet. This level corresponds to en elevation which is greater than the elevation of N last check velves in W discharge pipes of either the WCl or RCic systems. Therefors, filled discharge piping of HPCI or MCIC systems is ensured when li ed up to the CCST and tank level is at or above 9.5 feet.
36f.9K 3.5/4.5-3 Amen hent Ns. 100
_ - _