ML20070J239
| ML20070J239 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/08/1991 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070J242 | List: |
| References | |
| NUDOCS 9103150312 | |
| Download: ML20070J239 (22) | |
Text
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t VW ASHINGTON, D C. 20555 s
COMMONWEALTH EDISON COMPANY A,t,D N
IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET NO. 50-265 00AD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 130 License No. DPR-30 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amer.dment by Comonwealth Edison Company (the licensee) dated May 25, 1989, as supplemented January 25, 1991, complies with the standards and re Act of 1954, as amended (the Act) quirements of the Atomic Energy 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.
Thereisreasonableassurance(1)thattheactivitiesauthorized 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 common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordence 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:
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2 B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.- 130, 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 C0l1 MISSION St-s Ric Barrett,' Director Proas... Directorate 111 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: March 8, 1991 f
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ATTACHMENT TO LICENSE AMENDMENT NO.130 FACILITY OPERATING LICENSE NO. DPR-29
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DOCKET NO. 50-254 t
i 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. 5-5 3.F'4.5-5 3.5/4.5-6 3.5/4.5-6 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8 3.5/4.5-10 3.5/4.5-10 3.5/4.5-23
'3.5/4.5-23 3.5/4.5-24 3.5/4.5-24 3.5/4.5-25
-3.5/4.5-25 3.5/4.5-26 3.5/4.5-26 4
3.5/4.5-27 I
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2
___. _ _ _ _ _ _ ~
j QUAD-CITIES' I
DPR-29 4.
Containment cooling spray loops 4.
During each 5 year period, an are required to be operable when air test shall be performed on
{
the reactor water temperature is the drywell spray headers and greater than 212 F and prior to nozzles and a water spray test reactor startup from a cold con-performed on the torus spray dition.
Continued reactor oper-header and nozzles.
ation is permitted provided that a maximum of one drywell spray loop may be-inoperable for 30 i
days when the reactor water tem-l perature is greater than 212'F.
1 5.
If the requirements of 3.5.B cannot be met, an orderly shut-down shall be initiated, and the reactor shall be in a 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 />.
C.
HPCI Subsystem C.
HPCI Subsystem 1.
The HPCI subsystem shall be Surveillance of HPCI subsystem shall operable whenever the reactor be performed as: specified below with -
pressure is greater than.150 the-following limitations.
Fnr item psig and fuel is in the-4.5.C.3, the plant.is allowed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> reactor vessel.
in which to successfully-complete the test once reactor vessel pressure is' l
Y-2.
During startup following a refuel adequate to perform each test.
In outage or an outage in which work addition, the' testing required by item-was performed that directly affects 4.5.C.3.a shall be completed prior to HPCI system operability if the exceeding-325 psig reactor vessel-testing requirements of 4.5.C.3 pressure.- If HPCI is made inoperable cannot be met, continued reactor
-to perform overspeed testing, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' startup is not permitted.
The is allowed to complete the tests before HPCI subsystem shall be declared exceeding 325-psig.
inoperable, and the provisions of-Specification 3.5.C.4 shall be Item Frecuency implemented.
1.
-Valve Position Every 31 days-3.
Except for the limitations of 3.5.C.2, if the HPCI subsystem 2.
Flow Rate-Tes't-Every. 92 days is made or found to be inoperable, HPCI Pump shall j
continued reactor operation is deliver at least-l-
permissible only during the suc-5000 gpm against ceeding 14 days unless such sub-I a system head cor-system is sooner made operable, responding to e i
provided that during such 14 days
-I reactor vessel the automatic pressure relief._
pressure af > 1150 L
subsystems, the core spray sub-
-psig when steam is systems, LPCI mode of the RHR teing supplied to
-system, and the RCIC system.are-the turbine at-920 operable.
Otherwise, the pro-
-to 1005 psig.
visions of Specification 3.5.C.4 l
shall be implemented.-
1 3.5/4.5. Amendment No. 130
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QUAD-CITIES OPR-29 3.
G ow Rate Test-During startup HPCI pump shall following a deliver at least refuel outage 5000 gpm against or an outage in a system head which work was i
corresponding to performed that a reactor vessel
-directly affects pressure of:
HPCI system operability.
a.
1 300 psig when steam is being supplied to the turbine at 250 to 325 psig, and b.
1 1150 psig when steam is being sup-plied to the turbine at 920 to 1005 psig.
4.
If the requirements of Specifica-4.
Simulated Auto-Each refueling tion 3.5.C.1, 3.5.C.2 or 3.5.C.3 matic Actuation outage cannot be met, an orderly shutdown Test shall be initiated, and the reactor pressure shall be reduced to < 150 l
5.
Logic System Each refueling psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Functional Test outage.
D.
Automatic Pressure Relief Subsystems D.
Automatic Pressure Helief Subsystems Surveillance of the automatic pressure relief subsystem shall be performed as follows:
1.
The automatic pressure relief-1.
The following surveillance shall subsystem shall be operable be carried out on a six-month whenever the reactor pressure is surveillance interval:
greater than 90 psig, irradiated fuel is in the reactor vessel a.
With the reactor at pressure and prior to reactor startup each relief valve shall be from a cold condition, manually opened.
Relief valve opening shall be verified by a compensating turbine bypass valve or control valve closure.
2.
From and after the date that two 2.
A logic system functional test shall of the five relief valves of the be performed eacn refueling outags automatic pressure relief subsystem are made or found to be inoperable 3.5/4.5-6 Amendment No. 130-N iN
r QUAD-CITIES'-
DPR-29 L
when the reactor is pressurized above 90 psig with irradiated fuel in the reactor vessel, reactor operation is permissible only during the succeeding-7 days unless repairs are made and provided that-during such time the HPCI subsystem is operable.
3.
If the requirements of Specifi-3.
A! simulated automatic-initiation cation 3.5.D cannot be met, an-which opens all pilot _ valves-orderly shutdown shall-be-initi -
shall_be performed each re -
ated and the reactor pressure fueling outage.-
shall-be reduced to 90 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.
. When it is determined that two valves of the autonatic pressure relief. subsystem are inoperable.
'the HPCI-shall be demonstrated.
to be operable immediately.'
i E.
Reactor Core Isolation Cooling System E.
Reactor Core Isolation Cooling System-
- 1.
The RCIC system will be operable-Surveillance of the RCIC-system shall-whenever the reactor pressure is be performed as specified below'with:
greater than 150 psig and fuel is-the following limitations.1 For item in the reactor vessel 4.5.E.3, the plant is allowedL12 hours--
in-which to.successfully' complete the-2.
During startup following a refuel; test once' reactor vessel: pressure is-outage-or an outage in which work.
adequate to perform.each test; In.
i was performed that directly affects; addition,.the testing required by item the RCIC system operability, if the-4.5.E.3.alshall be completed prior to 3
testing requirements of-4.5.E.3 exceeding 325 psig reactor vessel u
cannot be met, continued reactor startup-is not permitted..The' '
pressure.
If RCIC is made inoperable to perform overspeed testing,D24 hours
' RCIC' system _sna11 be declared-
. is allowed to complete the tests before inoperable,-and the provisions of_
exceeding 325 psig.
Specification '3.5. E.4 'shall-be implemented.-
Item
- Frequency--
3.
_Except for the limitations of 1.
- Valve Position:
Every 31 days I
- 3.5.E.2, if-the_RCIC' system is mada or founa to be inoperable, 2L Flow' Rate Test -
Every 92 days continued reactor operation is permissible only during the:suc.
- RCIC Pump shall-
_ deliver at:1 east' ceeding 14 days-unless-such sys-l 400'gpm-against tem is sooner,made' operable, a system head' provided that during-'such 14 days l.
- corresponding to l
- the HPCI_ system'is-operable.
a reactor vessel
- Otherwise, the provisions of pressure of=,> 1150 -
Specification: 3.5.E.4 shall= e psig when steam'is
. implemented.
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1 QUAD-CITIES DPR-29 being supplied to the turbine at 920 to 1005 psig.
3.
Flow Rate Test-RCIC During startup pump shall deliver following a at least 400 gpm refuel outage against a system or an outage head corresponding in which work to a reactor vessel was performed pressure of:
that directly affects RCIC system operability, a.
> 300 psig when stcam is being supplied to the turbine at 250 to 325 psig, and b.
> 1150 psig when steam is being supplied to the turbine at 920 4.
If the requirements of Specification to 1005 psig.
3.5.E.1, 3.5.E.2, or 3.5.E.3 cannot l
be met, an orderly shutdown shall be 4.
Simulated Automatic Each refueling initiated and the reactor pressure shall Actuation Test outage be reduced to < 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.
Logic System Each refueling F.
Minimum Core and Containment Cooling Functional Test outage System Availability F.
Minimum Core and Containment Cooling 1.
Any combination of inoperable System Availability components in the core and containment cooling systems Surveillance requirements to assure shall not defeat the capability that minimum core and containment of the remaining operable cooling systems are available have components to fulfill the core been specified in Specification 4.2.B.
and containment cooling functions.
2.
When irradiated fuel is in the reactor vessel and the reector is in the cced shutdown condition, all low pressure core and contain-ment cooling systems may be in-operable provided no work is being done which has the potential for draining the reactor vessel.
j 3.5/4.5-8 Amendment No.130
QUAD-CITIES DPR-29 G.
Maintenance of Filled Discharge Pipe G.
Maintenance of Filled Discharge Pipe The following surveillance require-ments shall be adhered to to assure that the discharge piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC are filled:
1.
Whenever core spray, LPCI mode 1.
Every month the LPCI mode of the of the RHR, HPCI, or RCIC are RHR, core spray ECCS, HPCI and RCIC required to be operable, the discharge piping shall be vented discharge piping from the pump from the high point and water flow discharge of these systems to
- observed, the last check valves shall be filled.
2.
Following any period where HPCI, 2.
The discharge pipe pressure for RCIC, LPCI mode of the RHR or Core Spray and LPCI mode of RHR core spray have been out of shall be maintained at greater service end drained for than 40 psig and less than 90 maintenance, the discharge psig.
If pressure in any of piping of the inoperable system these systems is less than 40 shall be vented from the high psig or greater than 90 psig, point prior to the return of the this condition shall be alarmed system to service.
in the control room and immediate corrective action i
taken.
If the discharge pipe pressure is not within these limits in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the i
occurrence, an orderly shutdown l
shall be initiated, and the i
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 /> after initiation.
3.
Whenever the HPCI or RCIC system 3.
Filled discharge piping for HPCI is lined up to take suction from and RCIC systems is ensured by the torus, the discharge piping maintaining the leval in the of the HPCI and RCIC shall be Contaminated Condensate Storage vented from the high point of Tanks (CCST's) at or above 9.5 the system and water flow ob-feet.
If the CCST level falls served every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, below 9.5 feet, restore the 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 both HPCI and RCIC to take a suction from the torus per 4.5.G.3.
3.5/4.5-10 Amendment No. W4',130
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QUAD-CITIES l
DPR-29 4.5 SURVEILLANCE 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.
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.
Likewisn 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.
With components or subsystems out of service, overall core and containment cooling reliability is maintained by demonstrating the operability of the remaining cooling equipment.
The degree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment.
For routine out-of-service periods caused by preventative maintenance, etc., the pump and valve operability checks will be performed to demonstrate operability of the remaining components.
However, if a failure, design deficiency, etc., causes the out-of-service period, then the demonstration of operability should be thorough enough to assure that a similar problem does not exist on the remaining components.
For example, if an out-of-service period is caused by failure of a pump to deliver rated capacity due to a 1
design deficiency, the other pumps of this type might be subjected to a flow rate test in addition _to the operabil:ty checks.
The surveillance requirements bases described in this paragraph apply only-to the RCIC and HPCI systems.
With a cooling system out of service, overall core and containment cooling reliability is maintained by verifying the operability of the remaining cooling systems.
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 l
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.
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 pressures at which it is 3.S/4.5-23 Amendment No. 130
QUAD-CITIES DPR-29 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 E J t
systems.
A time limit is provided in which to perform the required tests during start-up.
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 HPCI and RCIC pumps is also conducted every 92 days at rated reactor pressure to demonstrate system operability in accordance with the LC0 provisions and to 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,..er operated or automatic valve installed in the direct flowpa*' of the suction or discharge of the pump or turbine that is not locked, sealed.,. otherwise secured in position.
At each refueling outage, a logic system functional test and a simulated automatic actua-tion test is performed on the HPCI and RCIC systems.
The tests and checks described above are considered adequate to ass re system operability.
The verification of the main steam relief valve operability during manual actuation surveillance testing must be made independent of temperatures indicated by thermocouples downstream of the relief valves.
It has been found that a temperature increase may result with the valve still closed.
This is due to steam 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 made for the relief valve opening and passing steam flow._ Closure response of the-turbine control valves during relief valve manual actuation would likewise serve as an adequate verification for the relief valve opening.
This test method may be performed over a wide range of reactor pressures greater than 150 psig.
Valve operation below 150 psig is limited by the spring tension exhibited by the relief valves.
The surveillance requirements to ensure that the discharge piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC systems is filled provides for a visual observation that water flows from a high point vent.
This ensures that the line is in a full condition.
Instrumentation has been provided on core spray ano :.PCI mode of RHR to monitor t
l the pressure of water in the discharge piping between the monthly intervals at l
which the lines are vented and alarm the control room if the pressure is inade-ruate.
This instrumentation will De calibrated on the same frequency as the
- afety system instrumentation 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 monitored on a continuous basis.
F alarm point of 40 psig for the low pressure of the fill system has been chose-.ecause, due to elevations of piping within the plant, 39 psig is required-to keep the lines full.
The shutoff head of the fill system pumps is less than 90 psig and therefore will not defeat the iow pressure cooling pump discharge pressure interlock of 100 psig as shown in Table 3.2-2.
A margin of 10 psig is provided by the high pressure alarm point of 90 psig.
HPCI and RCIC systems normally take a suction from the Contaminated Condensate Storage Tanks (CCSTs).
The level in the CCST's is maintained at or above a
I 3.5/4.5-24 Amendment No. 130 u
xm
l QUAD-CITIES DPR-29 9.5 feet.
This level corresponds to an elevation which is greater than the elevation of the last check valves in the discharge pipes of either the HPCI or RCIC systems.
Therefore, filled discharge piping of HPCI or RCIC systems is ensured when lined up to the CCST and tank level is at or above 9.5 feet.
The watertight bulkhead and submarine doors and the penetration seals for pipes and cables penetrating the vault walls and ceilings have been designed to withstand the maximum flood conditions.
To as:ure that their installation is adequate for maximum flood conditions, a method of testing each seal has been devised.
In order to test an electrical penetration or pipe seal, compressed air is supplied to 6 test connection and the space between the fittings is pressurized to approximately 15 psig.
The outer faces are then tested for leaks using a soap bubble solution.
In order to test the submarine doors, a test frame must be installed around each door.
The frame is then pumped to a pressure of approximately 15 psig and held to test for leaktightness.
The watertight bulkhead doors are tested by pressurizing the volume between the double gasket seals to approximately 15 psig.
The gasket seal area is inspected using a soap bubble solution.
Each RHR service water vault contains a sump, which will collect any floor or equipment leakage inside the vault.
A sump pump will automatically start on high level in the sump, and will purrp the water out of the vault, via 2 discharge check valves outside the vault to the service water discharge nipe.
A composite samplar is located on the sump discharge line.
A radiation..,nitor is also located on the service water discharge.
The sump discharge watst is not expected to be l
i contaminated, and any in-leakage to the vault is prevented by 2 check valves, Surveillance of these check valves is performed each operating cycle to assure j
i their integrity.
The previously installed bedplate drains to the turbine building equipment drain sump have been canped off permanently, j
A level switch set at a water level of 6 inches is located inside each vault.
Upon actuation, the switch alarms in the control room to notify the operator of trouble in the vault.
The operator will also be aware of problems in the vaults / condensate pump room if the high-level a'larm on the equipment drain sump is not terminated in a reasonable amount of time.
A system of level switches has been installed in the condenser pit to indicate and control flooding of the condenser area.
The followi;g swit nes are installed:
Level Function a.
1 foot (one alarm, low water switch) level b.
3 feet (one alarm, high water switch) level c.
5 feet (two alarm and cir-redundant culating water switch pairs) pump trip 3.5/4.5-25 Amendment No. 130
%MheoLL
l QUAD-CfTIES DPR-29 i
Level (a) indicates water in the condenser pit from either the hotwell or the circulating water system.
Level (b) is above the hotwell capacity and indicates l
a probable circulating water failure.
Should the switches at levels _(a) and (b) fail or the rpetotor fails to trip the circulating water pumps on alarm at level (b), the actuation of either level switch pair at level (c) shall trip the circulating water pumps automatically and alarm in the control room.
These redundant level switth pairs at level (c) are designed and installed to IEEE-279, "Critaria for Nuclear Power Plant Protection Systems." As the circulating water pumps are tripped, eitMr manually or automatically at level (c) of 5 feet, the maximum w;:ter level reached in-the condenser pit due to pumping will be at elevation 568 feet 9 inches elevation (10 l
feet above condenser pit floor elevation 558 feet 6 inches; 5 feet plus an additional 5 feet attributed to pump coastdown).
In order to prevent the RHR service water pump motors and diesel generator cooling water pump motors from overheating a vault cooler is supplied for each pump.
Each vault cooler is designed to inaintain the vault at a maximum 105 F temperature during operation of its. respective pump.
For example, if diesel generator cooling water pump 1/2-3903 starts, its cooler also starts and maintains the vault at 105*F by removing heat supplied to the vault-by the motor of pump 1/2-3903.
If, at the same time that pump 1/2-3903 is in operation, RHR service water pump IC starts, its cooler will also start and compensate for the added heat supplied to the vault by the-IC pump motor keeping the vault at 105 F.
Each of the coolers is supplied with cooling water from its respective pump's discharge line.
After the water has been passed through the cooler it returns to its respective pump's suction line.
The cooling water quantity _needed for each cooler is approximately 1% to 5% of the design flow of the pumps so that tne recirculation of this small amount of heated water wi_11 not affect pump or cooler operation.
Operation of the fans and coolers is required during shutdown and thus-additional surveillance is not required.
Verification that access doors to each vault are closed following entrance by personnel is covered by station-operating procedures.
The LHGR shall be checked daily to determine if fuel burnup or control rod move-ment has caused changes in power distribution.
Since changes due to Nrnup are slow and only a few control rods are moved daily,-a daily check of pNeo distri-i bution is adequate.-
Average-Planar LHGR At core thermal power levels less than or equal to 25%, operating plant-experience and thermal hydraulic analyses indicate that the resulting average planar LHGR is below the maximum average planar LHGR by a considerable margin;_
therefore, evaluation of the average planar LHGR below this-power level is not necessary.
The daily requirement for calculating average planar LHGR above 25%
rated thermal power is sufficient, since power distribution shifts are slow when there have not been significant power or control rod changes.
3.5/4.5-26 Amendment No. 130
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j QUAD-CITIES DPR-29 Local LHGR The LHGR as a function of_ core height shall be checked daily during reactor operation at greater than or equal-to 25% power to determine if fuel burnup or control rod movement has caused cl.anges in power distribution.
A limiting-LHGR value is precluded by considerable margin when employing any permissible control rod pattern below 25% rated thermal power.
Minimum Critical Power Ratio (MCPR)
At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will i
be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicate.
that the resulting MCFR value is in excess of requirements by a considerable margin. With this low void content, any inativertent core flow increase would only place operation in a more conservative mode relative to MCPR.
The daily requirement for calculating MCPR above 25% rated thermal power is sufficient,. since power distribution shifts are very slow when there have not been significant-power or control rod changes.
In addition, the K correction, as specified in the CORE OPERATING LIMITS REPORT, applied to tbc LCO p,rovides margin for flow increases from' low flows.
3.5/4.5-27 Amendment No. 130 j
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COMMONWEALTH EDISON COMPANY AND 10WA-ILLINOIS GAS AND ELECTRIC COMPANY DOCKET NO. 50-254 QUAD CITIES NUC_ LEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE
.=
Amendment No. 124 License No. DPR-29 e
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Comonwealth Edison Company (thelicensee)datedMay 25, 1989, as supplemented January 25, 1991, complies with the standards and re Act of 1954, as amended (the Act) quirements of the Atomic Energy and the Commission'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; i
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 Commission's regulations;-
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of~this amendment is in accordance with 10 CFR Part bl 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-29 is hereby amended to read as-follows:
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2-B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 124, 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 Richard J.:Barrett, Director Project Directorate III-2 Division of Reactor _ Projects 111/IV/V Office of Nuclear Reactor Regulation l
Attachment:
Changes to the Technical i
Specifications Date of Issuance:
March 8,1991 i
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I ATTACHMENT TO LICENSE AMENDMENT NO. 124 FACILITY OPERATING LICENSE NO. OPR-30 DOCKET NO. 50-265 l
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 j
indicating the area of change.
REMOVE INSERT 3.5/4.5-4a 3.5/4.5-4a l
3.5/4.5-5 3.5/4.5-5 3.5/4.5-6 3.5/4.5-6 3.5/4.5-6a 3.5/4.5-6a 3.5/4.5-7 3.5/4.5-7 3.5/4.5-15 3.5/4.5-15 3.5/4.5-15a 3.5/4.5-15a l
OUAD R372t5 OPR-30 f
C.
HPCI Subsystem turveillance of the HPCI subsysts shall be performed as specified below with the following limitations. For item 4.5.C.3, 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 test once reactor pressure is adequate to perform each test. In addition. the testing I
required by item 4.5.C.3.a shall be I
completed prior to exceedin 325 psig I
reactor vessel pressure. ! HPCI ts made inoperable to perform overspesd testing. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed to complete the tests before exceeding 325 psig.
h Freauency 1.
The NPCI subsystem shall be 1.
Valve Position Every 31 days operable whenever the reactor pressure is greater than 150 psig and fuel is in the reactor vessel.
2.
During startup following a 2.
Flow Rate fest -
tvery g2 days refuel outage or an outage in HPCI pury shall unich work was performed that deliver at least directly affects HPCI system 5000 gpm against operability, if the testing a system head requirerrents of 4.5.C.3 cannot be met. continued reactor corresponding to a reactor startup is not permitted. The vessel pressure HPCI subsystem shall be declared of 11150 psig when inoperable, and the provisions of Specification 3.5.C 4 shall steam is being surplied to the be implemented.
turbine at 920 to 1005 psig.
3.
Except for the limitations of 3.
Flow Rate fest -
During 3 5.C.2. if the HPCI subsystem HPCI pump shall startup it made or found to be deliver at least following tu,perable, continued reactor 5000 gpm against a refuel operation is permissible only a system head outage during the succetding 14 days unless such subsystem is sooner corresponding to or an putage a reactor vessel in which work made operable provided that pressure of:
was performed curing such 14 days the automatic pressure relief 4 1 300 psig that directly when steam is affects MPCI subsystem, the core spray being supplied system subsystems. LPCI mode of the RHR to the turbine operability.
system, and the RCIC system are at 250 to 325 operable. Otherwise, the esig, and provisions of Specification 3 5.C.4 shall be implemented.
D. 1 1150 psig when Steam ts being supplied to the turbine at 920 to 1005 pstg.
e.
If the reautrements of 4
Simulated tach refueling Speciftestion 3.5.C.). 3.5.C.2 Automatic outage or 3.5.C.3 cannot be met, an Actuation Test orderly shutdown shall be initiated. and the reactor 5.
tngic System tach refueling pressusa sh411 be reduced to Functional outage
<150 psig within 2e hours.
Test 3.5/4.5-da Amendment No. 124
QUAO-CIT 2t$
opt.30 0.
Automatic Pressure Relief Subsystems p.
Aut:matic Pressure Relief Subsystems Surveillance of the automatic pressure r'elief subsystem shall be performed as follows:
1.
The automatic pressure relief 1
The following surveillance shall subsystem shall be operable whenever the reactor pressure is be carried out on a sta-month surveillance interval greater than 90 psig irradiated toal is in the reactor vessel 4.
With the tcactor at pressure and prior to reactor startup from a cold conettion.
each relief valve shall be manually opened. Relief valve opening shall be verifled by a consensating turbine bypass valve or control valve closure.
2.
From and after the date that two 2.
A logic s/ stem functional test of the five relief valves of the shall be performed each i
automatic pressure relief subsystem is made or found to be refueling outage.
inoperable when the reactor is
^
pressurized above 90 psig with trradlated fuel in the reactor vessel, reactor operation is permissible only during the succeeding 7 days unless repairs are mace and provided that curing such time the NPCI subsystes is operable.
3.
If the reoutrements of Specif t-3.
A sisolated automatic initiation cation 3.5.0 cannot be met, an creerly shutdown shall be initi-W11ch opens all pilot valves sted and the reactor pressure shall be parformed each re-shall be reduced to 90 psig fueling outage, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 M en it is determined that t=0 relief valves of the automatic pressure relief subsystem are inoperable, the NPCI shall be I
demonstrated to be operable inned t ately.
3.5/4.5-5 Amendment No. 124
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DPR 30 E.
Reacter Core Isolation Cooling System E.
Reactor Core Isolation Cooling System e
Surveillance of the RCIC system shall be performed as specified below with the following limitations. For iten 4.5.t.3. 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 teich to successfully conglete the test once reactor vessel pressure is adequate to perform each test. In addition, the testing required by itse 4.i t.3.a shall be completed prist to asseding 321 psig reactor vessel pressure. If RCIC is made inoperable to perform overspeed testing. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed to complete the tests before exceeding 325 psis.
ILM Fraouanev 1.
The RCIC system will be operable 1.
Valve Position Every 31 days unenever the reactor pressure is greater than 150 psig (nd fuel is in the reactor vessel.
2.
During startup following a 2.
Flow Rate Test -
Every g2 days refuel outage or an outage in RCIC pung shall l
wnich work was performed that deliver at least directly affects RCIC system 400 gem against I
operability. If the testing a system need requirements of 4.5.t.3 cannot be met, continued reactor corresponding to a reactor startup is not permitted. The vessel pressure RCIC systee shall be declared of 11150 psig when ineperable. and the prov's tons 9 5pecification 3.5.t.4 shall steam is being ve inclemented.
supplied to the turbine at 920 to 1005 psig.
3.
Except for the limitations of 3.
Flow Rate Test -
During 3.5.t.2. if the RCIC system is RCIC pump shall startup made or found to be inoperable.
celiver at least following continued reactor operation is 400 gpe against a refuel permitted only during the a systere head outage succeccing 14 days unless such system is sooner made operable.
corresponding to or en outage a reactor vessel in which work provided that during such 14 pressure oft was performed days the HPCI system is operable. Otherwise, the 4 1 300 psig that directly ween steam is affects RCIC provisions of Specification 3.5 t.4 shall be implemented.
being supplied system to the turbine operability.
at 250 to 325 ps19. and B. 1 1150 psig when steam is being supplied to the -
turnine at 920 to 1005 psig.
l 4
If tre requirements of 4
Sista14ted tach refue;ing i
Speciftestion 3.5.t.l. 3.5.E,2 automatic outage or 3.5.E.3 cannot be met. An Actuation Test orderly SPutdown shall be initiated and the reactor 5.
Logic System tach refueling pressure SM11 be reduced to Functional outage
<l50 psig witnin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Test 3.5/4.5-6 Amendment No. 124
QUAD.C37!CS OPR-30 o
I
' F.
M1nteum Core and Containment Cooling F.
System Availability Hinteue Core and Containment Cooling System Ayallability Surve111ance requirements to assure that minisua core and containment cooling systems are available have been spectfted in $pecification 4.2.0.
1.
Any conetnation of inoperable components in the core and containment cooling systems shall not defeat the capability of the remaining operable ccaponents to fulfill the core and contairunent cooling functions.
2.
men irradiated foal is in the reactor vessel and the reactor is in the cold shutdow) condition, all low-pressure core and containment cooling systems may be inoperable provided no work 3.5/4.5-6a Amendment No. 124 1
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QUAD-C8T2ES DPR-30 is being done which has the potential
(
for draining the reactor vessel.
3.
When irradiated fuel is in the reactor an the vessel head is removed, the s ypression chamber may be drained completely and no more than one control rod drive housing opened at any one time provided that the spent fuel pool gate is open and the fuel pool water level is maintained at R level of greater than 33 feet above the bottom of the pool.
Additionally, a minimum condensate storage reserve of 230.000 gallons shall be maintained, no work
}:
shall be performed in the reactor vessel while a control rod drive housing is blanked following removal of the control rod drive, and a special flange shall be available which can be used to blank an open housing in the event of a leak.
4 When irradiated fuel is in the reactor and the vessel head is removed, work that has the potential for draining the_ vessel may be carried on with less than-ll2,200 ft3 of water in the suppression pool, provided that: (1) the total volume of water in the suppression pool, refueling cavity, and the fuel storage pool above the bottom of the fuel pool gate is greater than 112,200 ftJ: (2) the fuel storage pool gate is removed; (3) the low-pressure core and containment cooling systems are operable; and (4) -
the automatic mode of the drywell sump pumps is disabled.
G.
Maintenance of Filled Discharge Pipe G.
Maintenance of Filled-Discharge Pipe The following surveillance requirements shall be adhered to to assure that the discharge piping of the core spray, LPCI mode of the RHR, 1.
Whenever core spray, LPCI mode of the HPCI, and RCIC are filled:
RHR, HPCI, or RCIC are required to be operable, the discharge piping from 1.
Every month the LPCI mode of the the pump discharge of these systems to the last check valves shall be RHR, Core Spray ECCS, HPCI, and filled.
RCIC discharre piping shall be vented from the high point and water flow observed.
3.5/4.5-7 Amendment No. 124
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T guAO-CIT!!$
DPG-30 g
4.8 SuevtitLANCE At0VIREMENTS BA$t$
The testing interval for the core and contairunent cooling systems is bassa on a wantitative reliability analysis. Judgment, and practicality. The core caoling systems have not been designed to be fully testable during operation. For example, the core spray final tenission valves do not open until reactor pressure has fallen to 3$0 psig.
Thus, dr ing operation, even if high drywell pressure were siseJ1sted, the final valves would not open. In the case of the HPCI.
automatic initiation during power operation would result in punging cold water into the reactor vessel umich is not destrable.
The surveillance recuirenents bases in this paragraph apply to all core and containment cooling systems sucept RCIC and HPCI. The systems can be-automatically actuated during a refueling outage and this will be alone. To increase the av4114tility of the individual components of the core and contattunent cooling systems, the components which make up the system. i.e..
Instrumentation. pugs, valve operators. etc.. are tested more frequently.
The instrumentation is functionally tested each month, tikewise the pumps and motor-operated valves are also tested each month to assure their operett11ty.
The connination of a yearly sim leted automatic actuation test and monthly tests of the puses and valve operators 18 deemed to be adequate testing of these systems. With components or subsystems out of service overall core and contatnment cooling reltattitty is m intained by demonstrating the operabtitty of the renatning cooling equipment. The degree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment. For routine out-of-service periods caused by preventative maintenance, etc.. 'the puso and valve operabtitty checks will be performed to demonstrate operability of the remaining congonents.
However. if a failure, design deficiency. etc.. causes the out-of-service period, then the depenstratton of oper40111ty should be thorough enough to assure that a stattar problem does not exist on the remaining c onponent s.
For example, if an out-of-service period caused by failure of a pump to deliver rated capacity due to a design deficiency. the other puses of this type might be subjected to a flow rate test in addition to the operability checks.
The survet11ance requirements bases described in this paragraph apply only to the eCIC and HPCI systems.
With a cooling systen out of service. Overall core and containment cooling reliability is maintained by verifying the operabiltty of the rematntng cooling systems. Tne verification of operability,as used-in this-contest. for the renaining cooling systems means to sentnistrativelyiheck by esamining leg *, or other information to verify that the remaining systems are not out-of.servire for maintenance or other reasons.
It does not mean to perform the surveillance requirements needed to demonstrate the operability of the remaining sys t ems. However, if a failure, design deficte3cy. etc.. causes the-out-of-service period, then the vertfication of operability should be thorough For example, if an out-of-service period is caused by failure of a pump toenough to assure deliver rated capacity due to a design-deficiency, the other pumps of this type might be subjected to a flow rate test.
Following a refueling outage or en outage in which work was performed that directly af fects systen operability, the HPCI and eCIC pumps are flow rate tested prior to exceeding 32s psig and again at rated reactor steam pressure. This comeination of testing provides adequate ass:Jrance of pumo performance throughout the range of reactor pressures at-which it is-required to operate.
at a point of stable plant operation and also to provide overlap with lowThe low pressure limit is selec pressure (CC systems.
tests during startup. A time Itmit-is provided in which to perform the required plant conditions to be achieved and.the required tests to be performed.This time iteit is considered Flow rate testing of the HPCI and eCIC pumps is also conducted every 92 days at rated reactor pressure to demonstrate system operability in accordance with the LCO provisions and to 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 autenatic valve installed -in the direct flowpath of the suction or discharge of the pump or turbine that is not locked, sealed, or otherwise secured in posttion.
At each refueling outage, a logic system functional test and a sitmJ14ted automatic actuation test is performed on the HPCI and eCIC systems.
to assure system operability.The tests and checks described above are considered adequate The vertfication of the main steam relief valve operability during manual actuation surveillance testing mst be made independent of temperatures indicated by thermocouples downstream of the reitef valves. It has been found that a temperature increase may result with the valve still closed.
This is due to steam being vented through the pilot valves during the surveillance tes,1.
By 3.s/4.5-is Anenenent no. 124 1
QUAD-CIT!ES DPR-30 first opening a turbine bypass valve, and then observing its closure response during relief valve actuation, positive verification can be made for the relief valve opening and passing steam flow.
Closure response of the turbine control valves during relief valve manual actuation would likewise serve as an adequate verification for the relief valve opening.
This test method may be performed over a wide range of reactor pressures greater than 150 psig.
Valve operation below 150 psig is limited by the spring tension exhibited by the relief valves.
The surveillance requirements to ensure that the cischarge piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC systems is filled provides for a visual observation that water flows from a high point vent, This ensures that the line is in a full condition.
3 Instrumentation has been providea on core spray and LPCI mode of RHR to monitor the pressure of water in the discharge piping between the monthly intervals at which the lines are vented and alarm the control room if the pressure is inadequate.
This instrumentation will be calibrated on the same frequency as the safety system instrumentation 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 monitored on a continuous basis. An alarm point of 2 0 psig for the low pressure of the 4
fill system has been chosen because, due to elavations of piping within the plant, 39 psig is required to keep the lines full.
The shutoff head of the fill system pumps is less than 90 psig and therefore will not defeat the low-pressure cooling pump discharge press interlock 100 psig as shown ir. Table 3.2-2.
A magin of 10 psig is provided by the high pressure alarm point of 90 psig.
HPCI and RCIC systems normally take a suction from the Contaminated Condensate Storage Tanks (CCST'sh The level in the CCST's is maintained at or above 9.5 feet.
This level corresponds to an elevation which is greater than the elevation of the last check valves in the discharge pires of either the HPCI or RCIC systems.
Therefore, filled discharge piping of HPCI or NCIC systems is ensured when lined up to the CCST and tank level is at or above 9.5 fee'.
The watectight bulkheaj and submarine doors and the penetration seals for pipes and cables penetrating the vault walls and ceilings have been designed to withstand the maximum f'.ood conditions.
To assure that their installation is adequate for maximum flood conditions, a method of-testing each seal has been devised.
In order to test an electrical penetration or pipe seal, compressed air is supplied to a test connection and the space between the fittings is pressurized to approximately 15 psig.
The outer faces are then tested for leaks using a soap bubble solution.
3.5/4.5-15a Amendment No. 124