Text

Amends 125,119,145 & 141 to Licenses DPR-19,DPR-25,DPR-29 & DPR-30,respectively,consist of Changes to Facilities TS That Will Update Leakage Test Requirements of Drywell Airlock to Stds of 10CFR50,Appendix J & Section III.D.2
	ML20064H478
Person / Time
Site:	Dresden, Quad Cities
Issue date:	03/11/1994
From:	Dyer J; Office of Nuclear Reactor Regulation
To:	;
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ML20064H481	List: ML20064H478;
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	NUDOCS 9403180020
	Download: ML20064H478 (59)
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'o UNITED STATES 8'

NUCLE AR REGULATORY COMMISSION WASHING TON, D. C. 20555 B

%..... /

COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 125 License No. DPR-19 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated June 1, 1992, 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; 8.

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 (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; 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 1

requirements have been satisfied.

1 2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of facility Operating License No. DPR-19 is hereby amended to read as follows 9403180020 940311 PDR ADOCK 05000237 P

PDR

. (2)

Jechnical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

125, are hereby incorporated in the license. The licensee shall operate the facility in accordanc.'

with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION b IRCl. {.

tY 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 11, 1994

-l

\\

s/

ATTACHMENT TO LICENSE AMEN 0 MENT NO. 125 FACILITY OPERATING LICENSE NO. DPR-19 DOCKET NO. 50-237 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.

Pages indicated with an asterisk were inadvertently deleted in Amendment No. 122.

REMOVE INSERT iii lii v

v 1.0-3 1.0-3 3/4.7-7 3/4.7-7 3/4.7-9 3/4.7-9 3/4.7-19 3/4.7-19 3/4.7-19a 3/4.7-19b

3/4.7-31

3/4.7-32 8 3/4.7-33 8 3/4.7-33 B 3/4.7-34 8 3/4.7-34 8 3/4.7-43 8 3/4.7-43 B 3/4.7-43a

DRESDEN II DPR-19 Amendment No.

125 (Table of Contents. Cont'd.)

EASE 3.5.C HPCI Subsystem 3/4.5 - 7 3.5.0 Automatic Pressure Relief Subsystms 3/4.5 - 8 3.5.E

! solation Condenser System 3/4.5 - 9 3.5.F Hinimum Core and Containment Cooling System Availability 3/4.5 -11 3.5.G Deleted 3.5.H Maintenance of Filled Discharge Pipe 3/4.5 -13 3.5.I Average Planar LHGR 3/4.5 -15 3.5.J Local Steady State LHGR 3/4.5 -16 3.5.K Local Transient UlGR 3/4.5 -17 3.5.L Minimum Critical Pow r Ratio 3/4.5 -23 3.5.H Condensate Pump Room Flood Protection 3/4.5 -24 Limiting Conditions for Operation Bases (3.5)

B 3/4.5 -30 Surveillance Requirement Bases (4.5)

B 3/4.5 -39 3.6 Primary System Boundary 3/4.6 - 1 3.6.A Thermal Limitatiors 3/4.6 - 1 3.6.B Pressurization Temterature 3/4.6 - 2 3.6.C Coolant Chemistry 3/4,6 - 3 3.6.0 Coolant Leakage 3/4.6 - 5 3.6.E Safety and Relief Valves 3/4.6 - 6 3.6.F Structural Integrity 3/4.6 - 7 3.6.G Jet Pumps 3/4.6 -10 3.6.H Recirculation Pump Flow Mismatch 3/4.6 -12 3.6.I Shock Surpressors (Snubbers) 3/4.6 -16 Limiting Conditions for Operation Bases (3.6)

B 3/4.6 -25 Surveiilance Requirement Bases (4.6)

B 3/4.6 -39 3.7 Containment Systems 3/4.7 - 1 3.7.A Pri. nary Containment 3/4.7 - 1 3.7.B Standby Gas Treatment System 3/4.7 -19b l

3.7.C Secondary Containment 3/4.7 -26 3.7.D Primary Containemnt Isolation Valves 3/4.7 -27 Limiting Conditions for Operation Bases (3.7)

R 3/4.7 -33 Surveillance Requirement Bases (4.7)

B 3/4.7 -40 3.8 Radioactive Haterials 3/4.8 - 1 3.8.A Airborne Effluents 3/4.8 - 1 3.8.8 Liquid Effluents 3/4.8 - 9 3.8.C Mechancial Vacuum Pump 3/4.8 -14 3.8.D Radioactive Waste Storage 3/4.8 -15 3.8.E General Information 3/4.8 -15 3.8.F Solid Radioactive Waste 3/4.8 -19 3.8.G Miscellaneous Radioactive Materials Sources 3/4.8 -20 3.8.H Hiscellaneous LC0's 3/4.8 -21 Limiting Conditions for Operation Bases (3.8) 3/4.8 -32 Surveillance Requirement Bases (4.8) 3/4.8 -37 3.9 Auxiliary Electrical Systems 3/4.9 - 1 3.9.A Requirements 3/4.9 - 1 3.9.8 Availiability of Electric Power 3/4.9 - 2 111

DRESDEN II DPR-19 Amendment No.125 LT_able of Contents. Cont'd.)

PEla 4.3.D Control Rod Accumulators 3/4.3 -11 4.3.E Reactivity Anomalies 3/4.3 -12 4.3.F (N/A) 4.3.G Automatic Generation Control System 3/4.3 -13 4.4 Standby Liquid Control System 3/4.1 - 1 4.4.A Normal Operation 3/4.4 - 1 4.4.B Surveillance With Inoperable Components 3/4.4 - 2 4.4.C Boron Solution 3/4.4 - 3 4.5 Core and Containment Cooling Systems 3/4.5 - 1 4.5.A Core Spray and LPCI Subsystems 3/4.5 - 1 4.5.B Containment Cooling Subsystem 3/4.5 - 5 4.5.0 HPCI Subsystem 3/4.5 - 6 4.5.D Automatic Pressure Relief Subsystems 3/4.5 - 8 4.5.E Isolation Condenser System 3/4.5 - 9 4.5.F Core and Containment Cooling System 3/4.5 -11 4.5.G (Deleted) 4.5.H Haintenance of Filled Discharge Pipe 3/4.5 -13 4.5.I Average Planar Linear Heat Generation Rate 3/4.5 -15 4.5.J Linear Heat Generation Rate 3/4.5 -16 4.5.K Transient Linear' Heat Generation Rate 3/4.5 -17 4.5.L Minimum Critical Power Ratio 3/4.5 -23 4.5.H Condensate Pump Room Flood Protection 3/4.5 -24 4.6 Primary System Boundary 3/4.6 - 1 4.6.A Thermal Limitations 3/4.6 - 1 4.6.8 Pressurization Temperature 3/4.6 - 2 4.6.C Coolant Chemistry 3/4.6 - 3 4.6.0 Coolant Leakage 3/4.6 - 5 4.6.E Safety and Relief Valves 3/4.6 - 6 4.6.F Structural Integrity 3/4.6 - 7 4.6.G Jet Pumps 3/4.6 -10 4.6.H Recirculation Pump Flow Mismatch 3/4.6 -11 4.6.I Snubbers (Shack Suppressors) 3/4.6 -12 4.7 Containment System 3/4.7 - 1 4.7.A Primary Containment 3/4.7 - 1 4.7.B Standby Gas Treatment System 3/4.7 -19b l

4.7.C Secondary Containment 3/4.7 -25 4.7.0 Primary Containment Isolation Valves 3/4.7 -27 4.8 Radioactive Materials 3/4.8 - 1 4.8.A Gaseous Effluents 3/4.8 - 1 4.8.B Liquid Effluents 3/4.8 - S 4.8.C Mechanical Vacuum Pump 3/4.8 -10 4.8.D Radioactive Waste Storage 3/4.8 -11 4.8.E General 3/4.8 -12 4.8.F Miscellaneous Radioactive Materials Sources 3/4.8 -14 l

l y

DRESDEN II DPR-19 c

Amendment No.

125 1.0 DEFINITIONS (Cont'd)

R.

Primary Containment Inteority - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1.

All manual containment isolation valves on lines connecting to the reactor coolant system or containment which are not required to be open during accident conditions are closed, or comply with the requirements of Specification 3.7.D.

2.

Each primary containment air lock is in compliance with the requirements of Specification 3.7.A.8.

3.

All automatic containment isolation valves are operable or deactivated in the isolated position, or comply with the requirements of Specification 3.7.0.

4.

All blind flanges and manways are closed.

S.

Erotective Instrumentation Definitions 1.

Instrument Channel - An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.

2.

Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.

Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

3.

Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

4.

Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

T.

Rated Neutron Flux - Rated neutron flux is the neutron flux that corresponds to a steady-state power level of 2527 thermal megawatts.

U.

Rated Thermal Power - Rated thermal power means a steady-state power level of 2527 thermal megawatts.

1.0-3

l DRESDEN 11 DPR-19 d

Amendment No.

125 3.7 Litill1HG_CDNJ_ITION FOR _0PERATION 4.7 SVRVEILLANCE RE0VIREMENTS (Cont'd.)

(Cont'd.)

(b)

Deleted (c) 11.5 SCF per hour for any main steam isolation valvo at a test pressure of 25 psig.

c.

If two consecutive Type A tests fail to meet either 75 percent of L or 75percentofL,,

a Type A test shall be performed at each shutdown for refueling or approximately every 18 months until two consecutive Type A tests meet the above requirements, at which time the normal test schedule may be l

resumed.

d.

The accuracy of each Type A test shall be vertfled by a supplemental test which:

3/4.7-7 a

DRESDEN 11 DPR-19 m

Amendment No.:125 3.7-LIMITING CONDITION FOR OPERATION 4.7 SURVEILLANCE RE0VIREMENTS (Cont'd.)

.(Cont'd.)

(1)

Main steam line isolation valves-which shall be tested at a pressure of 25

_psig each operating cycle.

(2)

Bolted double-gasketed seals which shall be tested at a pressure of 48 psig whenever the seal is closed after being opened and each' operating cycle.

(3)

Air locks shall be tested per' Specification 4.7.A.8.

(4)

Deleted f.

Continuous-Leak Rate Monitor (1)

When the primary containment is inerted, the containment 3/4.7-9

DRESDEN II DPR-19 Amendment No.

125 e

3.7 LIMITING CONDITION FOR OPERATION 4.7 SVRVEILLANrE RE0VIREMENTS (cont'd)

(cont'd) shall be initiated and s

the reactor shall be in a cold shutdown condition in the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

8.

Primary Containment Air Locks 8.

Primary Containment Air Locks a.

Each primary containment a.

Each primary containment air lock shall be air lock shall be operable with:

demonstrated operable:

(1)

Both doors closed except when the (1)

By conducting an air lock is being overall air lock used for normal leakage test at transit entry and Pa, 48 psig and exit through the verifying that the containment, then overall air lock at least one air leakage rate is lock door shall be within its limit:

closed, and (a)

Within 72 (2)

An overall airlock hours of air leakage rate of lock opening less than or equal when to 0.05 La at Pa, containment 48 psig.

integrity is

required, b.

With one primary except when containment air lock the air lock door inoperable:

is being used for (1)

Maintain at least multiple the operable air entries.,

lock door closed

then at and either restore least once the inoperable air por 72 lock door to

hours, operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or (b)

At least lock the operable once per 6 air lock door months *, and closed.

(c)

Prior to establishing Primary Containment Integrity following air lock opening.

a Except during entry through an operable door to repair an inoperable door or to f acilitate the removal of persomet for o ctruletive time not to exceed one hour per year.

b the provisions of specification 1.0.cc are not ag>licable.

3/4.7-19

DRESDEN II DPR-19 Amendment No.

125 r

-3 7 LIMITING CONDITION EOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS (cont'd)

(cont'd)

(2)

Operation may then (2)

Concurrent with-continue until each overall air performance of the lock leakage test, next required conducted prior to overall air lock establishing leakage' test primary provided that the containment operable air lock integrity, by-door is verified verifying that to be locked only one door in closed

at least each air lock can.

once per 31 days, be opened at a time.

(3)

Otherwise, be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in cold shutdown within the following 24

hours, c.

With the primary containment air interlock mechanism inoperable:

(1)

Operations may continue provided the air lock is otherwise operable and entry and exit of the primary containment is administrative 1y controlled by a dedicated individual.

(2)

Otherwise, restore the air lock interlock mechanism to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the operable air lock door closed and verify that the operable air lock door is locked closed at least once per 31 days.

a Except during entry through an operable door to repair an inoperable door or to f acilitate the removal of personnet for a cumJlative time not to exceed one hour per year.

3/4.7-19 a

.~

DRESDEN II

-DPR-19 Amendment No.

125 3'.'7 L'IMITING_ CONDITION FOR OPERATION 4.7 SURVEILLANCE RE0VIREMENTS (cont'd)

(cont'd) d.

With the primary containment air lock inoperable, except as a result of an inoperable air lock door or air lock interlock mechanism:

(1)

Haintain at least one air lock door Closed.

(2)

Restore the inoperable air lock to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in at least cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B.

Standby Gas Treatment System B.

Standby Gas Treatment System 1.

Two separate and independent 1.

At least once per month, Otandby gas treatment system initiate from the control room sehsystems shall be operable 4000 cfm (plus or minus 10%)

at all times when secondary flow through each subsystem of containment integrity is standby gas treatment system required, except as specified for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the in sections 3.7.B.l(a) and subsystem heaters operating at (b).

rated power, a.

After one of the standby gas treatment system subsystems is made or found to be inoperable for any reason, reactor operation and fuel handling is permissible only during the succeeding.seven days, provided th t all active component is the other, standby gas treatment subsystem shall be operable.

3/4.7-19 b

DRESDEN II DPR-19 Amendment No.125

\\

INTENTIONALLY LEFT BLANK 3/4.7-31

DRESDEN II DPR-19 a

o Amendment No. 125 r

INTENTIONALLY 1. EFT BLANK 3/4.7-32

Y s

DRESDEN 11 DPR-19 Amendment No. 125 3.7 LIMITING CONDITION FOR OPERATION BASES A.

Primary Containment - The integrity of the primary containment and operation of the emergency core cooling system in combination, limit the off-site doses to values less than those suggested in 10 CFR 100 in the event of a break in the 3rimary system piping.

Thus, containment integrity is specified wienever the potential for violation of the primary reactor system integrity exists.

Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during initial core loading and while the low power test program is being conducted during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time which will greatly reduce the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring.

Procedures and the Rod Worth Minimizer would limit control worth to preclude a peak fuel enthalpy of 280 cal /gm.

In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offers a significant barrier to keep off-site doses well within 10 CFR 100.

The primary containment air lock's structural integrity and leak tightness are essential to the successful mitigation of a design basis accident event (DBA). The air lock is required to be operable whenever primary containment integrity is required.

For the air lock to be considered operable, the air lock interlock mechanism must be operable, the air lock must be in compliance with the 10 CFR 50, Appendix J, Type B air lock leakage test, and both air lock doors must be operable. The closure of a single door in an air lock will maintain primary containment operability since each door is designed to withstand the peak primary containment pressure calculated to occur following a DBA.

The action provisions have been modified to allow entry and exit to perform repairs on an affected air lock component or the removal of personnel should a component failure prevent exiting in the normal manner. The ability to open the operable door, even if it means the primary containment boundary is temporarily not intact, is acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which the operable door is expected to be open.

The operable door must be immediately closed after each entry and exit.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water B 3/4.7-33

,e

r DRESDEN 11 DPR-19 Amendment No. 125 3.7 LlHITING CONDITION FOR OPERATION BASES (Cont'd.)

volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1000 psig.

Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber design pressure. The design volume of the suppression chamber (water and air) was obtained by considering t1at the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

(Ref. Section 5.2.3 FSAR)

Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accidentisapproximately48psigwhichisbplowthedesignof 62 psig. Maximum water volume of 11, 655 f t results in a l

i 8 3/4.7-33 a

l DRESDEN 11 DPR-19 Amendment No. 125 4.7 SURVEILLANCE RE0VIREMENT BASES (Cont'd.)

4 The data reduction methods of the applicable ANSI standard will be applied for the integrated leak rate tests as specified in Appendix J of 10 CFR 50.

The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trer.ds. Whenever a double-gasketed penetration (primary ecntainment head equipment hatches and the t

suppression chamber access hatch) is broken and remade, the space between the gaskets is-pressurized to determine that the seals are performing properly. The test pressure of 48 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure.

It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other i

parts of the facility could occur.

Such leakage paths that may affect significantly the

' sequences of accidents are to be i

minimized.

Maintaining primary containment air locks cperable requires compliance with the leakage-rate test requirements of 10 CFR 50, Appendix J.

The periodic testing requirements verify that the air lock leakage does not exceed the specified allowed fraction of the overall primary containment leakage rate. The frequencies are required by 10 CFR 50, Appendix J.

Periodic testing of the interlock mechanism demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur.

The results of the loss-of-coolant accident analyses presented in Amendment No. 18 of the SAR indicates that fission products would not be released directly to the environs because of leakage from a

the main steam line isolation valves due to holdup in the steam system complex.

Although this effect would indicate that an adequate margin exists with regard to the release of fission products, a program will be undertaken to further reduce the potential for such leakage to bypass the standby gas treatment system.

Monitoring the nitrogen makeup requirements of the inerting system provides a method of observing leak rate trends and would detect gross leaks in a very short time. This equipment must be periodically removed from service for test and maintenance, but this out-of-service time will be kept to a practical minimum.

Surveillance of the reactor building-pressure suppression chamber vacuum breakers consists of operability checks and leakage tests B 3/4.7-43 I

s..

DRESDEN 11 DPR-19 Amendment No. 125 4.7 SURVEILLANCE RE0VIREMENT BASES (Cont'd.)

(conducted as part of the containment leak-tightness test). These vacuum breakers are normally in the closed position and open only during tests or a post accident condition. As a result, a testing frequency of 3 months for operability is considered justified for this equipment.

Inspections and calibrations are performed during refueling outages, this frequency being based on experience and judgement.

i B 3/4.7-43 a 4

a o.

A

[gDQ P E G g#'

o UNITED STATES l'

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 k.....,/

[_0MMONWEALTH EDIS0N COMPANY DOCKET N0. 50-249 DRESDEN NUCLEAR POWER STATION. UNIT 3 AMEN 0 MENT TO FACILITY OPERATING LICENE Amendment No. 119 License No. DPR-25 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated June 1, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility 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; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.B. of Facility Operating License No. DPR-25 is hereby amended to read as follows:

e 4

. B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.119, 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 and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION tillt f.

AV 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 11, 1994 1

r ATTACHMENT TO LICENSE AMENDMENT N0.

119 FACILITY OPERATING LICENSE NO. DPR-25 DOCKET N0. 50-249 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.

Pages indicated with an asterisk were inadvertently deleted in Amendment No. 117.

REMOVE INSERT iii iii y

v 1.0-3 1.0-3 3/4.7-7 3/4.7-7 3/4.7-9 3/4.7.9 3/4.7-19 3/4.7-19 3/4.7-19a 3/4.7-19b

3/4.7-31

3/4.7-32 B 3/4.7-33 8 3/4.7-33 8 3/4.7-34 8 3/4.7-34 B 3/4.7-43 8 3/4.7-43 B 3/4.7-43a

DRESDEN III DPR-25 Amendment No. 119 (Table of Contents. Cont'dd PElft 3.5.C HPCI Subsystem 3/4.5 - 7 3.5.0 Automatic Pressure Relief Subsystms 3/4.5 - 8 3.5.E Isolation Condenser System 3/4.5 - 9 3.5.F Minimum Core and Containment Cooling System Availability 3/4.5 -11 3.5.G Deleted 3.5.H Haintenance of Filled Discharge Pipe 3/4.5 -13 3.5.1 Average Planar LHGR.

3/4.5 -15 3.5.J Local Steady State LHGR 3/4.5 -16 3.5.K Local Transient LHGR 3/4.5 -l'/

3.5.L Minimum Critical Power Ratio 3/4.5 -23 2'

3.5.M Condensate Pump Room Flood Protection 3/4.F -24 Limiting Conditions for Operation Bases (3.5)

B 3/4.5 -30 Surveillaace Requirement Bases (4.5)

B 3/4.5 -39 3.6 Primary System Boundary 3/4.6 - 1 3.6.A Thermal Limitations 3/4.6 - 1 3.6.8 Pressurization Temperature 3/4.6 - 2 3.6.C Coolant Chemistry 3/4.6 - 3 3.6,0 Coolant Leakage 3/4.6 - 5 3.6.E Safety and Relief Valves 3/4.6 - 6 3.6.F Structural Integrity 3/4.6 - 7 3.6.G Jet Pumps 3/4.6 -10 3.6.H Recirculation Pump Flow limitations 3/4.6 -12 3.6.I Snubbers (Shock Suppressors) 3/4.6 -16 Limiting Conditions for Operation Bases (3.6)

B 3/4.6 -25 Surveillance Requirement Bases (4.6)

B 3/4.6.-39 3.7 Containment Systems 3/4.7 - 1 3.7.A Primary Containment 3/4.7 --I 3.7.B Standby Gas Treatment System 3/4.7 -19b-l 3.7.C Secondary Containment 3/4.7 -26 3.7.D Primary Containemnt Isolation Valves 3/4.7 -27 Limiting Conditions for Operation Bases (3.7)

B 3/4.7 -33 Surveillance Requirement Bases (4.7)

'B 3/4.7 -40 3.8 Radioactive Effluents 3/4.8 - l' 3.8.A Gaseous Effluents 3/4.5 --I 3.8.B Liquid Effluents 3/4.0 -.9 3.8.C Mechancial Vacuum Pump 3/4.8-3.8.0 Radioactive Waste Storage 3/4.8 -15 3.8.E Radiological Environmental Monitoring Program 3/4.8 -15 3.8.F Solid Radioactive Waste 3/4.8 -19 3.8.G Miscellaneous Radioactive Materials Sources

'3/4.8 -20 3.8.H Miscellaneous LC0 3/4.8 -21 Limiting Conditions for Operation Bases (3.8)

B 3/4.8 -32 Surveillance Requirement Bases (4.8)

B 3/4.8 -37_'

3.9 Auxiliary Electrical Systems 3/4.9 - 1 3.9.A Requirements 3/4.9 - 1 3.9.B Availiability of Electric Power 3/4.9 - 2 111

DRESDEN III DPR-25

+'

Amendment No. 119 (Table of Con 1ents. Cont'd.)

EA9a 4.3.0 Control Rod Accumulators 3/4.3 -11 4.3.E Reactivity Anomalies 3/4.3 -12 4.3.F (N/A) 4.3.G Automatic Generation Control System 3/4.3 -13 4.4 Standby Liquid Control System 3/4.1 - 1 4.4.A Normal Operation 3/4.4 - 1 4.4.B Surveillance With Inoperable Components 3/4.4 - 2 4.4.C Boron Solution 3/4.4 - 3 4.5 Core and Containment Cooling Systems 3/4.5 - 1 4.5.A Core Spray and LPCI Subsystems 3/4.5 - 1 4.5.B Contrinment Cooling Subsystem 3/4.5 - 5 4.5.C HPCI Subsystem 3/4.5 - 6 4.5.D Automatic Pressure Relief Subsystems 3/4.5 - 8 4.5.E Isolation Condenser System 3/4.5 - 9 4.5.F Core and Containment Cooling System 3/4.5 -11 4.5.G (Deleted) 4.5.H Maintenance of Filled Discharge Pipe 3/4.5 -13 4.5.I Average Planar Linear Heat Generation Rate 3/4.5 -15 4.5.J Linear Heat Generation Rate (LHGR) 3/4.5 -16 4.5.K Transient Linear Heat Generation Rate (LHGR) 3/4.5 -17 4.5.L Minimum Critical Power Ratio (MCPR) 3/4.5 -22 4.5.H Condensate Pump Room Flood Protection 3/4.5 -24 4.6 Primary System Boundary 3/4.6 - 1 4.6.A Thermal Limitations 3/4.6 - 1 4.6.8 Pressurization Temperature 3/4.6 - 2 4.6.C Coolant Chemistry 3/4.6 - 3 4.6.D Coolant Leakage 3/4.6 - 5 4.6.E Safety and Relief Valves 3/4.6 - 6 4.6.F Structural Integrity 3/4.6 - 7 4.6.G Jet Pumps 3/4.6 -10 4.6.H Recirculation Pump Flow limitations 3/4.6 -12 4.6.1 Snubbers (Shock Suppressors) 3/4.6 -16 4.7 Containment System 3/4.7 - 1 4.7.A Primary Containment 3/4.7 - 1 4.7.B Standby Gas Treatment System 3/4.7 -19b l

4.7.C Secondary Contrinment 3/4.7 -25 4.7.D Primary Containment Isolation Valves

'/4.7 -27 4.8 Rad',oactive Effluents 3/4.8 - 1 4.8.A Ga.eous Effluents 3/4.8 - 1 4.8.B Liquid Effluents 3/4.8 - 9 4.8.C Hechanical Vacuum Pump 3/4.8 -14 4.8.0 Radioactive Waste Storage 3/4.8 -15 4.8.E Radiological Environmental Monitoring Program 3/4.8 -15 4.8.F Solid Radioactive Waste 3/4.8 -19 4.8.G Miscellaneous Radioactive Materials Sources 3/4.8 -20 v

J

,j c

DRESDEN III DPR-25 Amendment No.

119 1.0 DEFINITIONS (Cont'd)

R.

Primary Containment Intearity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1.

All manual containment isolation valves on lines connecting to the reactor coolant system or containment which are not required to be open during accident conditions are closed, or comply with the requirements of Specification 3.7,0.

2.

Each primary containment air lock is in compliance with the requirements of Specification 3.7.A.8.

3.

All automatic containment isolation valves are operable or deactivated in the isolated position, or comply with the requirements of Specification 3.7,0.

4.

All blind flanges and manways are closed.

S.

Protective Instrumentation Definitions 1.

Instrument Channel - An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.

2.

Trip System - A trip system means an arrangement of instrument -hannel trip signals and auxiliary equipment required to !nitiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.

Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

3.

Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or_ system level.

4.

Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

T.

Rated Neutron Flux - Rated neutron flux is the neutron flux that corresponds to a steady-state power level of 2527 thermal megawatts.

t U.

Rated Thermal Power - Rated thermal power means a steady-state power level of 2527 thermal megawatts.

1.0-3

DRESDEN III DPR-25 Amendment No. 119 3.7 LIMITING CONDITION FOR OPERATION 4.7 SURVEILLANCE RE0VIREMENTS (Cont'd.)

(Cont'd.)

(b) deleted (c) 11.5 SCF per hour for any main steam isolation valve at a test pressure of 25 psig, c.

If two consecutive Type A tests fail to meet either 75 percent of L or 75percentofL,,

a Type A test shall be performed at each shutdown for refueling or approximately every 18 months until two consecutive Type A tests meet the above requirements, at which time the normal test schedule may be

resumed, d.

The accuracy of each Type A test shall be verified by a supplemental test which-1 (1)

Confirms the accuracy of the test by verifying that the difference 3/4.7-7 4

DRESDEN III DPR-25 Amendment No. 119 3.7 LIMITING CONDITION FOR OPERATION 4.7 SURVEILLANCE RE0VIREMENTS (Cont'd.)

(Cont'd.)

(2)

Bolted double-gasketed seals

'which shall be tested at a pressure of 48 psig whenever the.

seal is closed after.being opened and each operating cycle.

i.

(3)

Air locks shall be tested per Specification

'4.7.A.8.

(4)

Deleted f.

Continuous Leak Rate Monitor (1)

When the primary containment is inerted..the containment shall be continuously-monitored for gross leakage by review of the inerting system make-up requirements.

4 3/4.7-9

DRESDEN III DPR-25 Amendment No.

119

+

3.7 LIMITING CONDITION FOR OPERATION 4.7 SVRVElllANCE RE0UIREMENTS (cont'd)

(cont'd) shall be initiated and the reactor shall be in a cold shutdown condition in the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

8.

Primary Containment Air Locks 8.

Primary Containment Air Locks a.

Each primary containment a.

Each primary containment air lock shall be air lock shall be operable with:

demonstrated operable:

(1)

Both doors closed except when the (1)

By conducting an air lock is being overall air lock used for normal leakage test at transit entry and Pa, 48 psig and exit through the verifying that the containment, then overall air lock at least one air leakage rate is lock door shall be within its limit:

closed, and (a)

Within 72 (2)

An overall airlock hours of air leakage rate of lock opening less than or equal when to 0.05 La at Pa, containment 48 psig, integrity is

required, b.

With one primary except when containment air lock the air lock door inoperable:

is being used for (1)

Maintain at least multiple the operable air

entries, lock door closed' then at and either restore least once the inoperable air per 72 lock door to

hours, operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or (b)

At least lock the operable once per 6 air lock door months *, and closed.

(c)

Prior to establishing Primary Containment Integrity following air lock opening.

a Except during entry through an operable door to repair an inoperable door or to f acilitate the removal of personret for a cuwlative time not to exceed one hour per year.

b The provisions of specification 1.0.cc are not applicable, l

3/4.7-19

DRESDEN III DPR-25 Amendment No. 119 a

3.7 LIMITING CONDITION FOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS (cont'd)

(cont'd)

(2)

Operation may then (2)

Concurrent with continue until each overall air performance of the lock leakage test, next required conducted prior to overall air lock establishing leakage test primary provided that the containment operable air lock integrity, by door is verified verifying that to be locked only one door in 4

closed

at least each air lock can once per 31 days.

be opened at a time.

(3)

Otherwise, be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

With the primary containment air interlock mechanism inoperable:

(1)

Operations may continue provided the air lock is otherwise operable and entry and exit of the primary containment is administrative 1y controlled by a dedicated individual.

(2)

Otherwise, restore the air lock interlock mechanism to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the operable air lock door closed and verify that the operable air lock door is locked closed at least once per 31 days.

O Except during entry through an operable door to repair an inoperable door or to facilitate the removal of Mrsonnel for a cunulative time not to exceed one hour per year.

3/4.7-19 a

i DRESDEN III DPR-25 Amendment No. 119 3.7 LIMITING CONDITION FOR OPERATION 4.7 SVRVEILLANCE RE0VIREMENTS (cont'd)

(cont'd) d.

With the primary containment air lock inoperable, except as a result of an inoperable air lock door or air lock interlock mechanism:

(1)

Maintain at least one air lock door closed.

(2)

Restore the inoperable air lock to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in at least cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B.

Standby Gas Treatment System B.

Standby Gas Treatment System 1.

Two separate and independent 1.

At least once per month, standby gas treatment system initiate from the control room subsystems shall be operable 4000 cfm (plus or minus 10%)

at all times when secondary flow through each subsystem of containment integrity is standby gas treatment system required, except as specified for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the in sections 3.7.B.l(a) and subsystem heaters operating at (b).

rated power.

a.

After one of the standby gas treatment system subsystems is made or found to be inoperable for any reason, reactor operation and fuel handling is permissible

-only during the succeeding igyfdl days, provided that all active components in the other.

standby gas treatment subsystem shall be operable.

3,/4.7-19 b

DRESDEN !.11 DPR-25 Amendment he. 119 s

INTENTIONALLY LEFT BLANK

~

l 3/4.7-31

DRESDEN III DPR-25 Amendment No.

119 i

l INTENTIONALLY LEFT BLANK 3/4.7-32

t b

L 1

DRESDEN III DPR-25 Amendment No.119 3.7 LIMITING CONDITION FOR OPERATION BASES A.

Primary Containment - The integrity of the primary contair. ment and operation of the emergency core cooling system in combination, limit the off-site doses to values less than those suggested in 10 CFR 100 in the event of a break in the primary system piping.

Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists.

Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during initial core loading and while the low power test program is being conducted during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time which will greatly reduce the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring.

Procedures and the Rod Worth Minimizer would limit control worth to preclude a peak fuel enthalpy of 280 cal /gm.

In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offers a significant barrier to keep off-site doses well within 10 CFR 100.

The primary containment air lock's structural integrity and leak tightness are essential to the successful mitigation of a design basis accident event (DBA).

The air lock is required to be operable whenever primary containment integrity is required.

For the air lock to be considered operable, the air lock interlock mechanism must be operable, the air lock must be in compliance with the 10 CFR 50, Appendix J, Type B air lock leakage test, and both air lock doors must be operable. The closure of a single door in an air lock will maintain primary containment operability since each door is designed to withstand the peak primary containment pressure calculated to occur following a DBA. The action provisions have been modified to allow entry and exit to perform repairs on an affected air lock component or the removal of personnel should a component failure prevent exiting in the normal manner. The ability to open the operable door, even if it means the primary containment boundary is temporarily not intact, is acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which j

the operable door is expected to be open.

The operable door must a

be immediately closed after each entry and exit.

I The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water B 3/4.7-33 1

DRESDEN III DPR-25 Amendment No.119 3.7 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1000 psig.

Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber design pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

(Ref. Section 5.2.3 FSAR)

Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accidentisapproximately48psigwhichisbplowthedesignof 62 psig. Maximum water volume of 11, 655 ft results in a B 3/4.7-33 a

e DRESDEN III DPR-25 Amendment No.119 4.7 SVRVEILLANCE RE0VIREMENT BASES (Cont'd.)

The data reduction methods of the applicable ANSI standard will be apr ied for the integrated leak rate tests as specified in a mndix J of 10 CFR 50.

The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends. Whenever a double-gasketed penetration (primary containment head equipment hatches-and the suppression chamber access iatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly.

The test pressure of 48 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure.

It is expected that the majority of the leakage from valves, penetrations'and seals would be into the reactor building.

However, it is possible that leakage into other parts of the facility could occur.

Such leakage paths that may affect significantly the consequences of accidents are to be minimized.

Maintaining primary containment air locks operable requires compliance with the leakage-rate test requirements of 10 CFR 50, Appendix J.

The periodic testing requirements verify that the air lock leakage does not exceed the specified allowed. fraction of the overall primary containment leakage rate.

The frequencies are required by 10 CFR 50, Appendix J.

Periodic testing of the interlock mechanism demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur.

The results of the loss-of-coolant accident analyses presented in Amendment No.18 of the SAR indicates that' fission products would not be released directly to the environs because of leakage from the main steam line isolation valves due to holdup in the steam system complex. Although this effect would indicate that an adequate margin exists with regard to the release of fission products, a program will be undertaken l to further reduce the potential for such leakage to bypass the standby gas treatment system.

Monitoring the nitrogen makeup requirements of the inerting system provides a' method of observing leak rate trends and would detect gross leaks in a very short time.

This equipment must be periodically removed from service for test and maintenance, but this out-of-service time will be kept to a practical minimum.

Surveillance of the reactor building-pressure suppression chamber-vacuum breakers consists of operability checks and leakage tests B 3/4.7-43

DRESDEN III DPR-25 Amendment No. 119 4.7 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

(conducted as part of the containment leak-tightness test).

These vacuum breakers are normally in the closed position and open only during tests or a post accident condition. As a result, a testing frequency of 3 months for operability is considered justified for this equipment.

Inspections and calibrations are performed during refueling outages, this frequency being based on experience and Judgement.

B 3/4.7-43 a w

" [p ne u

'\\

UNITED STATES

-['

3

- )y, NUCLEAR REGULATORY COMMISSION WAsmNoTON, D. C. 20555

\\...../

COMMONWEALTH EDIS0N COMPANY MQ IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY

[ LOCKET NO 50-251

_0VAD CITIES NUCLEAR POWER STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENS1 Amendment No. 145 License No. DPR-29 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated June 1, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility 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:

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-29 is hereby amended to read as follows:

O 4 B.

Technical Specifications The Technical Specifications contatted in Appendices A and B, as revised through Amendment No.145, sre hereby i icorporated 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 and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION Mk T.

6V James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - lil/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 11, 1994 i

ATTACHMENT TO LICENSE AMENDMENT N0.

145 FACILITY OPERATING LICENSE N0. 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 number and contain marginal lines indicating the area of change.

REMOVE INSERT 1.0-3 1.0-3 3.7/4.7-5 3.7/4.7-5 3.7/4.7-Ila 3.7/4.7-Ilb 3.7/4.7-11c 3.7/4.7-Ild 3.7/4.7-12 3.7/4.7-12 3.7/4.7-21 3.7/4.7-21 3.7/4.7-21a 3.7/4.7-28 3.7/4.7-28 3

.s..

w r

QUAD CITIES DPR 29 M.

Operable A system, subsystem, train, component, or device shall be operable when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem train, component or device to perform its function (s) are also capable of performing their related support function (s).

N.

Operating Operating means that a system, subsystem, train, component or device is performing its intended functions in its required manner.

O.

Operating Cycle Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.

P.

Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:

1.

All manual containment isolation valves on lines connecting to the reactor coolant system or containment which are not required to be open during accident conditions are closed.

2.

Each primary containment air lock is in compliance with the requirements of Specification 3.7.A.7.

3.

All automatic containment isolation valves are operable or deactivated in the isolation position.

4.

All blind flanges and manways are closed.

O.

Protective Instrumentation Definitions 1.

Channel-A channelis an arrangement of a sensor and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its identity where individual channel outputs are combine in a logic.

2.

Trip System A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

1.0 3 Amendment No.145 l

l

t.

4

=

QUAD CITIES DPR 29

3) Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured

' leakage at P.,48 psig, or P,,

25 psig.

d.

Deleted d.

Type 8 and C tests shall be l'

conducted at P.,48 psig, at intervals no greater than 24 months except for tests involving:

1) Air locks shall be tested per Specification 4.7.A.7.

2) Main steam isolation valves which shall be leak tested at least once per operating' cycle, of a frequency not to exceed 24 months, at a pressure of 25 psig.

3) Bolted double gasketed seals which shall be tested at a pressure of 48 psig whenever the sealis closed after being -

opened and each operating cycle.

4) While valve M012001 is inoperable, valves M01220 2, M01-220 3, and.

M01220 4 shall be VERIFIED closed after each closure.

5) The pathways idpntified in Table 4.71, which will not be tested until the end of cycle 11 refueling outage.

3.7/4.7 5 Amendment No. 145 w

, ~

+-

m,

U-0 0

QUAD CITIES DPR 29 c.

This pressure differential may be decreased to less than 1.20 PSID for a maximum of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during required operability testing of the HPCI system pump, the RCIC system pump, the drywell.

pressure suppression chamber vacuum breakers, and reactor pressure relief valves, d.

If the Specifications of 3.7.A.6.c cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hours period, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.7/4.7 11 a Amendment No. 145

G A

4 QUAD CITIES DPR.29 7.

Primary Containment Air Locks 7.

Primary Containment Air Locks a.

Each primary containment air lock a.

Each primary containment air lock shall be Operable with:

shall be demonstrated Operable:

(1) Both doors closed except (1) By conducting an overall air when the air lock is being lock leakage test at P.,48 used for normal transit entry psig and verifying that the and exit through the overall air lock leakage rate is containment, then at least one within its limit:

d air lock door shall be closed, and (2) An overall air lock leakage (a) Within 7,2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of air rate of less than or equal to lock opening when 0.05 L, at P., 48 psig.

containment integrity is required, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , b.

With one primary containment air (b) At least once per 6 lock door inoperable:

months', and (1) Maintain at least the Operable (c) Prior to establishing air lock door closed' and Primary Containment either restore the inoperable Integrity following air lock air lock door to Operable opening.

status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the Operable air lock door closed.

]

1.

l Except during entry through an operable door to repair an inoperable door or to facilitate,

the removal of personnel for a cumulative time not to exceed one hour per year.

The provisions of Specification 1.0.00 are not applicable.

3.7/4.7 11 b Amendment No.

145-e

-)

v s

O QUAD CITIES DPR 29 (2) Operation may then continuo (2) Concurrent with each overall air unt;l performance of the next lock leakage test, conducted prior required overall air lock to establishing primary leakage test provided that the containment integrity, by verifying Operable air lock door is that only one door in each air lock verified to be locked closed' can be opened at a time, at least once per 31 days.

(3) Otherwise, be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

With the primary containment air lock interlock mechanism inoperable:

(1) Operations may continue provided the air lock is otherwise Operable and entry and exit of the primary containment is administratively controlled by a dedicated individual.

(2) Otherwise, restore the air lock interlock mechanism to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the Operable air lock door closed and verify that the Operable air lock door is locked closed at least once per 31 days, d.

With the primary containment air lock inoperable, except as a result of an inoperable air lock door or alt lock interlock mechanism:

(1) Maintain at least one air lock door closed.

Except during entry through an operable door to repair an inoperable door or to facilitate the removal of personnel for a cumulative time not to exceed one hour per year.

3.7/4.7 11 c Amendment No. 145

6 QUAD CITIES DPR 29 (2) Restore the inoperable air lock to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in at least cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 U

t 3.7/4.7 11 d Amendment No.145

= -.

u d

QUAD-CITIES DPR-29 I

B.

Standby Gas Treatment System B.

Standby Gas Treatment System 1.

Two separate and independent 1.

At least once per month, initiate from standby gas treatment circuits shall be the control room 4000 cfm (i 10%)

operable at all times when secondary flow through both circuits of the containment integrity is required, standby gas trr:atment system for at except as specified in sections least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the circuit heaters 3.7.B.1.(a) and (b).

operating at rated power.

a.

After one of the standby gas a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from the time that-treatment circuits is made or one standby gas treatment system found to be inoperable for any circuit is made or found to be.

reason, reactor operation and fuel inoperable for any reason and daily handling is permissible only during thereafter for the next succeeding the succeeding 7 days, provided 7 days, initiate from the control that all active components in the room 4000 cfm (210%) flow other standby gas treatment through the operable circuit of the system shall be demonstrated to standby treatment system for at be operable within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the circuit daily thereafter. Within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months heaters operating.

following the 7 days, the reactor shall be placed in a condition for which the standby gas treatment system is not required in accordance with Specification 3.7 C.1(a) through (d).

3.7/4.7 12 Amendment No. 145

6 s-a a

QUAD-CITIES DPR 29 3.7 LIMITING CONDITIONS FOR OPERATION BASES j

A.

Primary Containment The integrity of the primary containment and operation of the emergency core cooling system, in combination, limit the offsite doses to values less than those suggested in 10 CFR 100 in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concem about such violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vesselis required. There will be no pressure on the system at j

this time, which will greatly reduce the chances of a pipe break. The reactor may be l

taken critical during this period; however, restrictive operating procedure will be in i

effect again to minimize the probability of an accident occurring. Procedures and the rud worth minimizer would limit control rod worth to preclude a peak fuel enthalpy of 280 cal /gm. In addition, in the unlikely event that on excursion did occur, the reactor building standby gas treatwnt system, which will be operational during this time, offers a sufficient barrier to keep offsite doses well within 10 CFR 100 guidelines.

The primary containment air lock's structural integrity and leak tightness are essential to the successful mitigation of a design basis accident event (DBA). The air lock is required to be operable whenever primary containment integrity is required. For the air lock to be considered operable, the air lock interlock mechanism must be operable, the air lock must be in compliance with the 10 CFR 50, Appendix J, Type B air lock leakage test, end both air lock doors must be operable. The closure of a single door in an air lock will maintain primary containment operability since each door is designed to withstand th3 peak primary containme;it pressure calculated to occur following a DBA.

The action pavisions have been modified to allow entry and exit to perform repairs on an affected a r lock component or the removal of personnel should a component failure prevent exiting in the normal manner. The ability to open the operable door, even if it means the primary containment boundary is temporarily not intact,is acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which the operable door is expected to be open. The operable door must be immediately closed after each entry and exit.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1000 psig.

Since all of the gases in the drywell are purged into the pressure suppression chamber air space duriac a loss-of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 56 psig, the suppression chamber design pressure. The design volume of the suppression chamber 3.7/4.7 21 Amendment No. 145

QUAD CITIES DPR 29 (water and air) was chained by considering that the total voiume of teactor coolant to be condensed is dit;narged to the suppression chamber and that the drywell volume is purged to the suppmssion chamber. (Ref. Section 6.2.1 FSAR) l Using the minimum oi maximum water volume given in the specification, containment pressure during the design basis accident is approximately 48 psig, which is below the design value of 56 psig. Maximum water volume of 115,655 ft results in a 3

downcomer submergence of 4 feet; the minimum volume of 112,200 ft' results in a submergence approximately 4 inches less. The majority of the Bodega tests (Reference 1) were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to downcomer submergence, this specification is adequate.

3.7/4.7-21 a Amendment No.145

.c OUAD CITIES DPR 29 These doses are also based on the assumption of no holdup in the secondary containment resulting in direct release of fission products from the primary containment through the filters and stack to the environs. Therefore, the specified primary containment leak rate and filter efficiency are conservative and provide margin between expected offsite doses and 10 CFR 100 guidelines.

Although the dose calculations suggest that the accident leak rate could be allowed to increase to about 2.6%/ day before the guideline thyroid dose value given in 10 CFR 100 would be exceeded, establishing the test limit of 1.0%/ day provides an adequate margin of safety to assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak tightness capability of the structure over its service lifetime. Additional margin to maintain the containment in the as built condition is achieved by establishing the allowable operational leak rate.

The allowable operationalleak rate is derived by multiplying the maximum allowable leak rate by 0.75, thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.

The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification. Allowing the test intervals to be extended up to 8 months permits some flexibility needed to have the tests coincide with scheduled or unscheduled shutdown periods.

The data reduction methods of ANSI N45.41972 will be applied for integrated leak rate tests.

The penetration and air purge piping leakage test frequency, along with the containment leak rete tests, is adequate ito allow detection of leakage trends.

Whenever a doubis gasketed penetration (primary containment head equipment e

hatches and the suppression chamber access hatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of 48 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure. It is expected that the majority of the leakage from valves, penetrations, and seals would be irito the reactoc building.

However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of accidents are to be minimized.

Maintaining primary containment air locks operable requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J. The periodic testing requirements verify that the air lock leakage does not exceed the specified allowed fraction of the overall primary containment leakage rate. The frequencies are required by 10 CFR 50, Appendix J. Periodic testing of the interlock mechanism demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur.

3.7/4.7-28 Amendment No. 145

a.

, g >*, "t ouq'o s

UNITED STATES 8'

- 7, '

g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 g

g f

COMMONWEALTH EDIS0N COMPAN)

AND J_0WA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET N0. 50-265 OVAD CITIES NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 141 License No. DPR-30 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated June 1, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility 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; 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 Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 141, 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 and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION W {, h OV James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/lV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 11, 1994 D

s+

?

t ATTACHMENT TO LICENSE AMENDMENT NO 141 FACILITY OPERATING LICENSE N0. OPR-30 DOCKET N0. 50-265 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 1.0-3 1.0-3 3.7/4.7-3 3.7/4.7-3 3.7/4.7-6b 3.7/4.7-6c 3.7/4.7-6d 3.7/4.7-7 3.7/4.7-7 3.7/4.7-11 3.7/4.7-11 3.7/4.7-lla 3.7/4.7-16 3.7/4.7-16

OUAD CITIES v

DPR 30 y

2.

Each primary containment air lock is in compliance with the requirements of Specification 3.7.A.7.

3.

All automatic containment isolation valves are operable or deactivated in the isolation position.

4.

All bknd flanges and manways are closed.

O.

Protective instrumentation Definitions 1.

Channel A channelis an arrangement of a sensor and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channet terminates and loses its identity where individual channel outputs are combine in a logic.

2.

Trip System A trip system means an arrangement of instrument channel trip signals and auxihary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more plant parameters in order to initiate trip system action. initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

3.

Protective Action An action initiated by the protection system when a limit is reached. A protective action can be a the channel or system level.

4.

Protective Function - A system protective action which results from the protective action of the channels monitormg a particular plant condition.

R.

Rated Neutron Flux Rated neutron flux is the flux that corresponds to a stead-state powet level of 2511 thermal megawatts.

S.

Rated Thermal Power Rated thermal power means a steady state power level of 2511 thermal megawatts.

T.

Reactor Power Operation Reactor power operation is any operation with the mode switch in the Startup/ Hot Standby or Run position with the reactor Critical and above 1% rated thermal power.

U.

Reactor Vessel Pressure Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detector.

V.

Refueling Outage Refuehng outage is the period of time between the shutdown of the unit prior to a refuehng and the startup of the plant subsequent to that refuehng. For the purpose of designating frequency of testing surveillance, a refuehng outage shall mean a regularly scheduled refuehng outage; however, where such outages occur within 8 months of the completion of the previous refuehng outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

W.

Safety Umit. The safety limits are kmits below which the reasonable maintenance of the cladding and primary system are assured. Exceeding such a limit is cause for unit shutdown, and review by the NRC before resumption of the unit operation. Operation beyond such a hmit may not in itself result in serious consequences, but it indicates an operational deficiency subject to regulatory review.

X.

Secondary Containment integrity Secondary containment integrity means that the reactor building is intact and the following conditions are met:

1.

At least one door in each access opening is closed.

2.

The standby gas treatment system is operable.

3.

All reactor building automatic ventilation system isolation valves are operable or are secured in the isolated po itior*.

1.0 3 Amendment No. 141

ve, s

y u

OUAD CITIES

.j 1

DPR 30 i

b) s L,.1.0 percent by weight b.

If any periodic Type A test fails to of the containment air per meet either 0.75 L, or 0.75 L,c the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduced test sche'dule for subsequent Type'A pressure of P., 25 psig.

tests shall be reviewed and approved by the Commission, if two

2) A combined leakage rate of 5 consecutive Type A tests fall to meet 0.60 L, for all penetrations and either 0.75 L, or 0.75 L,i a Type A valves, except for main steam test shall be performed at each-isolation valves subject to Type B shutdown for refueling or and C tests when pressurized to approximately every 18 months, P.,

whichever occurs first, until two -

consecutive Type A tests meet either -

3) 11.5 scf per hour for any one 0.75 L, or 0.75 L,, at which time the main steam isolation valve when above test schedule may be resumed.

tested at 25 psig.-

b.

With the measured overallintegrated c.

The accuracy of each Type A test containment leakage rate exceeding

. shall be verified by a supplemental 0.75 L, or 0.75 L,, as applicable, test which:

restore the overall integrated leakage.

rate (s) to s 0.75 L, or s' O.75 L,, as

1) Confirms the accuracy of the applicable.

test by verifying that the i

difference between the supplemental data and the Type A test data is within 0.25 L, or-0.25 L,.

c.

With the measured combined leakage

2) Has a duration sufficient to i

rate for all penetrations and valves, establish accurately the ' change.

except for main steam isolation in leakage rate between the Type -

~

valves, subject to Type B and C tests A test and the supplemental test.

exceeding 0.60 L., restore the L

combine leakage rate for all

3) Requires the quantity of gas l

penetrations and valves, except for injected into the containment or main steam isolation valves, subject bled from the containment during to Type B and C tests to 0.60 L,.

the supplemental test to be equivalent to at least 25 percent of the total measured leakage at' P., 48 psig, or P., 25 psig.

d.

Deleted d.

Type B and C tests shall be conducted at P.,48 psig, at intervals -

no greater than 24 months except for -

l tests involving:

1) Air locka shall be tested per Specification 4.7.A.7.

3.7/4.7 3 Amendment No.141

r QUAD CITIES DPR 30 c.

This pressure differential may be decreased to less than 1.20 PSID for a maximum of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during required operability testing of the HPCI system pump, the RCIC system pump, the drywell pressure suppression chamber vacuum breakers, and reactor pressure relief valves.

d.

If the Specifications of 3.7 A.6.c cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hours period, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1 3.7/4.7 6b Amendment No. 141

s QUAD CITIES DPR 30 7.

Primary Containment Air Locks 7.

Primary Containment Air Locks a.

Each primary containment air lock a.

Each primary containment air lock.

shall be Operable with:

shall be demonstrated Operable:

(1) Both doors closed except when (1) By conducting an overall air lock the air lock is being used for leakage test at P.,- 48 psig and i

normal transit entry and exit verifying that the overall air lock through the containment, then at leakage rate is within its limit:

least one air lock door shall be closed, and (21 An overall air lock leakage rate of (a) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of air lock less than or equal to 0.05 L, at opening when containment P., 48 psig.

integrity is required, except when the air lock is being used for multiple entries, -

then at least once oer 72

hours, b.

With one primary containment air lock (b) At least once ps i door inoperable:

months', and (1) Maintain at least the Operable air (c) Prior to establish <ng Primary lock door closed

and either Containment integrity restore the inoperable air lock following air lock opening.

a door to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the Operable air lock door closed.

1 1

Except during entry through an operable door to repair an inoperable door of to facilitate

~

the removal of personnel for a cumulative time not :o exceed one hout per. year.

The provisions of Specification 1.0.00 are not applicable.

3.7/4.7 6c

' Amendment No.141-0

'.g-g 4

s OUAD CITIES DPR 30 (2) Operation may then continue (2) Concurrent with each overall air lock until performance of the next leakage test. conducted prior to required overall air lock leakage

. establishing primary containment test provided that the Operable

-integrity, by verifying that only one air lock door is verified to be door in each air lock can be opened at locked closed

at least once per a time.

31 days.

(3) Otherwise, be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , c.

With the primary containment air lock interlock mechanism inoperable:

(1) Operations may continue provided the air lock is otherwise Operable and entry and exit of the primary containment is administratively controlled by a dedicated individual.

(2) Otherwise, restore the air lock interlock mechanism to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock it the Operable air lock door closed and verify that the Operable air lock door is locked closed at least once per 31 days.

i d.

With the primary containment air lock inoperable, except as a result of an inoperable air lock door or air lock interlock mechanism:

+

(1) Maintain at least one air lock l

door closed.

(2) Restore the inoperable air lock to Operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least hot shutdown i

within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in at least cold shutdown within the following 24 ho'urs.

I Except during entry through an operable door to repair an inoperable door or to facilitate the removal of personnel for a cumulative time not to exceed one hour per year.

3.7/4.7 6d Amendment No.-

141

~

-~- -

,. ~, -,..,, - -

v b

QUAD vlTIES OPR 30 B.

Standby Gas Treatment System B.

Standby Gas Treatment System 1.

Two separate and independent standby 1.

At least once per month, initiate from the gas treatment circuits shall be operable at control room 4000 cfm (2 10%) flow all times when secondary containment through both circuits of the standby gas integrity is required, except as specified in treatment system for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months sections 3.7.8.1.(a) and (b).

with the circuit heaters operating at rated

power, a.

After one of the standby gas a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from the time that one treatment circuits is made or found to standby gas treatment system circuit be inoperable for any reason, reactor is made or found to be inoperable for operation and fuel handling is any reason and daily thereafter for permissible only during the the next succeeding 7 days, initiate succeeding 7 days, provided that all from the control room 4000 cfm (t active components in the other 10%) flow through the operable standby gas treatment system shall circuit of the standby treatment be demonstrated to be operable system for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and daily thereafter.

circuit heaters operating, Within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months following the 7 days, the reactor shall be placed in a condition for which the standby gas treatment system is not required in accordance with Specification 3.7.C.1(a) through (d).

b.

If both standby gas treatment system circuits are not operable, within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months the reactor shall be placed in a condition for which the standby gas treatment systern is not required in accordance with Specification 3.7.C.1.f a) through (d).

3.7/4.7-7 Amendment No.141 l

a

~

OUAD CITIES DPR 30 3.7 LIMITING CONDITIONS FOR OPERATION BASES A.

Primary Containment The integrity of the primary containment and operation of the emergency core cooling system, in combination, limit the offsite doses to values less than those suggested in 10 CFR 100 in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during initial core loading and while the low power test progrom is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time, which will greatly reduce the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedure will be in effect again to minimize the probability of an accident occurring. Procedures and the rod worth minimizer would limit control rod worth to preclude a peak fuel enthalpy of 280 cal /gm. In addition, in the unlikely event that an excursion did occur, the reactor building standby gas treatment system, which will be operational during this time, offers a sufficient barrier to keep offsite doses well within 10 CFR 100 guidelines.

t The primary containment air lock's structuralintegrity and leak tightness are essential to the successful l

mitigation of a design basis accident event (DBA). The air lock is required to be operable whenever primary containment integrity is required. For the air lock to be considered operable, the air lock interlock mechanism must be operable, the air lock must be in compliance with the 10 CFR 50, Appendix J. Type B sit lock leakage test, and both air lock doors must be operable. The closure of a single door in an air lock will maintain primary containment operability since each door is designed to withstand the peak primary containment pressure calculated to occur following a DBA. The action provisions have been modified to allow entry and exit to perform repairs on an affected air lock component or the removal of personnel should a component failure prevent exiting in the normal menner. The ability to open the operable door, even if it means the primary containment boundary is temporarily not intact, is acceptable due to the low probabuity of an event that could pressurize the primary containment during the short time in which the operable door is expected to be open. The operable door must be immediately closed after each entry and exit.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1000 psig.

Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 56 psig, the suppression chamber design pressure. The design volume of the suppression chamber (Water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber. (Ref. Section 6.2.1 FSAR) l Using the minimum or maximum water volume given in the specification, containment pressure during the design basis accident is approximately 48 psig, which is below the design value of 56 psig. Maximum water volume of 115,655 ft' resutts in a downcomer submergence of 4 feet; the minimum volume of 112,200 ft' results in a submergence approximately 4 inches less. The majority of the Bodega tests (Reference 1) were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to downcomer submergence, this specification is adequate.

3.7/4.7 11 Amendment No. 141-I

4..

b u.-

QUAD CITIES OPR 30 Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool.is maintained below 160*F during any period of rehef valve operation with sonic conditions at the discharge exit. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings, in addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include: (1) use of all available means to close the valve, (2) initiation of suppression pool water cooling heat exchangers, (3) initiation of reactor shutdown, and (4)If other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck open relief valve to assure mixing and uniformity of energy insertion to the pool.

The maximum temperatute at the end of the blowdown tested during the Humboldt Bay (Reference 2) and i

Bodega Bay tests was 170'F; this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170'F.

J 1

2 3.7/4,7 11 a Amendment No,141 em--

y

<.er,,w--g

-,,mr-e-.,

c,

+n,-

r r

~,, - - --,

QUAD CITIES DPR 30 multiplying the maximum allowable leak rate by 0.75, thereby providing a 25% margin to allow for leakage deterioration which may occur durmg the period between leak rate tests.

The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification. Allowing the test intervals to be extended up to 8 months permits some flexibihty needed to have the tests coincide with scheduled or unscheduled shutdown periods.

The data reduction methods of ANSI N45.41972 will be apphed for integrated leak rate tests.

The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends. Whenever a double-gasketed penetration (primary containment head equipment hatches and the suppression chamber access hatch)is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of 48 psig is consistent with the accident analysis and the maximum preoperational leak rate test pressure. It is expected that the majority of the leakage from valves, penetrations, and seals would be into the reactor building. However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of accidents are to be minimized.

Maintaining primary containment air locks operable requires compliance with the leakage-rate test requirements of 10 CFR 50, Appendix J. The periodic testing requirements verify that the air lock leakage does not exceed the specified allowed fraction of the overall primary containment leakage rate. The frequencies are required by 10 CFR 50, Appendix J. Periodic testing of the interlock mechanism demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur.

The results of the loss of coolant accident analysis referenced in Section 6.2.4.1 of the FSAR indicate that fission products would not be released directly to the environs because of leakage from the main steam line isolation valves due to holdup on the steam system complex. Although this effect would indicate that an adequate margin exists with regard to the release of fission products, a program will be undertaken to further reduce the potential for such leakage to bypass the standby gas treatment system.

Surveillance of the reactor building. pressure suppression chamber vacuum breakers consists of operability checks and leakage tests (conducted as part of the containment leak tightness tests). These vacuum breakers are normally in the closed position and open only during tests or an accident condition. As a resutt, a testing frequency of 3 months for operabikty is considered justified for this equipment. Inspections and cahbrations are performed during refueling outages, this frequency being based on experience and judgment.

Pressure suppression chamber drywell vacuum breakers monthly operability tests are performed to check the capability of the disks to open and close and to verify that the position indication and alarm circuits function properly. The disks must open during accident conditions and during transient additions of energy through rehef valves. This periodic operation of the disks and the quahty of equipment justify the freoJency of operability tests of this equipment.

Following each quarterly operability test, a differential pressure decay rate test is performer, to verify that leakage from the drywell to the suppression chamber is within specified limits.

Measurement of force to open, calibration of position switches, inspection of equipment, and functional testing are performed during each refueling outage. This frequency is based on equipment quality, experience, and judgement. Also, a more stringent differential pressure decay rate test is performed during refueling outages than is performed monthly. This test is performed to verify that totalleakage caths between the drywell and suppression chamber are not in excess of the equivalent to a 1 inch orifice.

This small leakage path is only a small fraction of the allowable, thus integrity of the containment system is assured prior to startup following each refuehng outage (Reference 1).

3.7/4.7 16 Amendment No. 141

ML20064H478

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