ML17202G577

Amends 107 & 102 to Licenses DPR-19 & DPR-25,respectively, Revising Testing Requirements for Other Sys or Subsystems of ECCS or Standby Gas Treatment Sys When One Sys or Subsystem Inoperable

ML17202G577
Person / Time
Site:	Dresden
Issue date:	08/10/1989
From:	Shemanski P Office of Nuclear Reactor Regulation
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ML17202G578	List: ML17202G577 ML17202G580 ML20246D071
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NUDOCS 8908250251
Download: ML17202G577 (97)

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{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT NO. 2 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 107 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 December 21, 1988 and supplemental information provided dated May 4, 1989 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 tiave 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 Provisional Operating License No. DPR-19 is hereby amended to read as follows: B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 107, 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 to be implemented within 60 days. FOR THE NUCLEAR REGULATORY COMMISSION ?aJc.s~

Attachment:

Changes to the Technical Specifications Date of Issuance: August 10, 1989 Paul C. Shemanski, Acting Director Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects

ATTACHMENT TO LICENSE AMENDMENT NO. 107 PROVISIONAL OPERATING LICENSE 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. REMOVE INSERT iii iii iv

• iv vii vii 1.0-4 1.0-4 3/4.5-2 3/4.5-2 3/4.5-3 3/4.5-3 3/4.5-4 3/4.5-4 3/4.5-4a 3/4.5-5 3/4.5-5 3/4.5-6 3/4.5-6 3/4.5-7 3/4.5-7 3/4.5-7a 3/4.5-8 3/4.5-8 3/4.5-9 3/4.5-9 3/4.5-10

. 3/4.5-10 3/4. 5-11 3/4.5-11 3/4.5-12 3/4.5-12 3/4.5-12a 3/4.5-12a 3/4.5-13 3/4.5-13 3/4.5-29 B 3/4.5-29 B 3/4.5-30 B 3/4.5-30 B 3/4.5-31 B 3/4.5-32 B 3/4.5-31 B 3/4.5-33 B 3/4.5-32 B 3/4.5-34 B 3/4.5-33 B 3/4.5-35 B 3/4.5-34 B 3/4.5-36 B 3/4.5-35

• B 3/4.5-37 B 3/4.5-36 B 3/4.5-38 B 3/4.5-37 B 3/4.5-39 B 3/4.5-38 B 3/4.5-40 B 3/4.5-39 B 3/4.5-41 B 3/4.5-40 B 3/4.5-42 B 3/4.5-41 B 3/4.5-43 B 3/4.5-42 B 3/4.5-44

AMENDMENT NO. 107 CONT'D REMOVE 3/4.7-19 3/4.7-20 3/4.7-22 3/4.7-23 3/4.7-24 3/4.7-25 3/4.7-26 3/4.7-27 B 3/4.7-38 B 3/4.7-39 B 3/4.7-40 B 3/4.7-45 8 3/4.7-46 3/4.9-3 3/4.9-4 3/4.9-5 3/4.9-6 INSERT 3/4.7-19 3/4.7-20 3/4.7-22 3/4.7-23 3/4.7-24 3/4.7-25 3/4.7-26 3/4.7-27 B 3/4.7-38 B 3/4.7-39 B 3/4.7-40 B 3/4.7-45 B 3/4.7-46 . 3/4.9-3 3/4.9-4 3/4.9-5 3/4.9-6 3/4.9-6a

DRESDEN II DPR-19 Amendment No. 82,83,84,85,95,104,107 (Table of Contents, Cont'd.) 3.5.C 3.5.D 3.5.E 3.5.F 3.5.G 3.5.H

3. 5. I 3.5.J 3.5.K

3. 5. L 3.5.M 3.6 3.6.A 3.6.B 3.6.C 3.6.D 3.6.E 3.6.F 3.6.G 3.6.H

3. 6. I 3.7 3.7.A

3. 7. B 3.7.C 3.7.D 3.8 3.8.A 3.8.B 3.8.C 3.8.D 3.8.E 3.8.F 3.8.G 3.8.H 3.9 3.9.A 3.9.B HPCI Subsystem Automatic Pressure Relief Subsystems Isolation Condenser System Minimum Core and Containment Cooling System Availability Deleted Maintenance of Filled Discharge Pipe Average Planar LHGR Local Steady State LHGR Local Transient LHGR

~inimum Critical Power Ratio Condensate Pump Room Flood Protection Limiting Conditions for Operation Bases (3.5) Surveillance Requirement Bases (4.5) Primary System Boundary Thermal Limitations Pressurization Temperature Coolant Chemistry Coolant Leakage Safety and Relief Valves Structural Integrity Jet Pumps Recirculation Pump Flow Mismatch Shock Suppressors (Snubbers) Limiting Conditions for Operation Bases (3.6) Surveillance Requirement Bases (4.6) Containment Systems Primary Containment Standby Gas Treatment System Secondary Containment Primary Containment Isolation Valves Limiting Conditions for Operation Bases (3.7) Surveillance Requirement Bases (4.7) Radioactive Materials Airborne Effluents Liquid Effluents Mechanical Vacuum Pump Radioactive Waste Storage General Information Solid Radioactive Waste Miscellaneous Radioactive Materials Sources Miscellaneous LCO's Limiting Conditions for Operation Bases (3.8) Surveillance Requirement Bases (4.8) Auxiliary Electrical Systems Requirements Availability of Electric Power iii Page 3/4.5 - 7 3/4.5 - 8 3/4.5 - 9 3/4.5.:.11 3/4. 5 -13 3/4.5 -15 3/4.5 -16 3/4.5 -17 3/4. 5..;23... 3/4.5 -24 B 3/4.5 -30 B 3/4.5 -39 3/4.6 - 1 3/4.6 - 1 3/4.6 - 2 3/4.6 - 3 3/4.6 - 5 3/4.6 - 6 3/4.6 - 7 3/4.6 -10 3/4.6 -12 3/4.6 -16 B 3/4.6 -25 B 3/4.6 -39 3/4.7 - 1 3/4. 7 - 1 3/4.7 -19 3/4.7 -26 3/4. 7 -27 B 3/4.7 -33 B 3/4.7 -40 3/4.8 - 1 3/4.8 - 1 3/4.8 - 9 3/4.8 -14 3/4.8 -15 3/4.8 -15 3/4.8 -19 3/4.8 -20 3/4.8 -21 3/4.8 -32 3/4.8 -37 3/4.9 - 1 3/4.9 - 1 3/4.9 - 2

DRESDEN II DPR-19 Amendment No. 82, 83, 106, 107 (Table of Contents, Cont 1d.) 3.9.C 3.9.D 3.10 3.10.A 3.10. B 3.10. c 3.10. D 3.10.E 3.10.F 3.11 . 3.12 4.0 4.1 4.2 4.2.A 4.2.B 4.2.C 4.2.D 4.2.E 4.2.F 4.2.G 4.3 4.3.A 4.3.B 4.3.C Diesel Fuel Diesel Generator Operability Limiting Conditions for Operation Bases (3.9) Surveillance Requirement Bases (4.9) Refueling Refueling Interlocks Core Monitoring Fuel Storage Pool Water Level Control Rod and* Control Rod Drive Maintenance Extended Core Maintenance Spent Fuel Cask Handling Limiting Conditions for Operation Bases (3.10) Surveillance Requirement Bases (4.10) High Energy Piping Integrity (outside containment) Limiting Conditions for Operation Bases (3.11) Surveillance Requirement Bases (4.11) Fire Protection Systems - Sections 3.12.A through 3.12.H - Deleted per Generic Letters 86-10 and 88-12 (Amendment 106) SURVEILLANCE REgUIREMENTS Reactor Protection System Protective Instrumentation Primary Containment Isolation Functions Core and Containment Cooling Systems Initiation and Control Control Rod Block Actuation Refueling Floor Radiation Monitors Post Accident Instrumentation Radioactive Liquid Effluent Instrumentation Radioactive Gaseous Effluent Instrumentation Reactivity Control Reactivity Limitations Control Rods Scram Insertion Times iv Page 3/4.9 - 5 3/4.9 - 5 8 3/4.9 - 7 8 3/4.9 - 8 3/4.10- 1 3/4.10- 1 3/4.10- 1 3/4.10-.2 3/4.10-. 3 3/4.10- 4* 3/4.10- 5 8 3/4.10- 8 8 3/4.10-ll 3/4.11-1 8 3/4.ll-4 8 3/4.ll-4 3/4.l - 1 3/4.2 - 1 3/4.2 - 1 3/4.2 - 1 3/4.2 - 2 3/4.2 - 2 3/4.2 - 3 3/4.2 - 4 3/4.2 - 5 3/4.3 - 1 3/4.3 - 1 3/4.3 - 4 3/4.3 -10

Table 3.1.1 Table 4.1.1 Table 4.1.2 Table 3.2.l Table 3.2.2 Table 3.2.3 Table 3.2.4 Table 3.2.5 Table 3.2.6 Table 4.2.l Table 4.2.2 Table 4.2.3 Table 4.2.4 Table 4.5.l Table 4.6.2 Table 3. 7.1 Table 4.8.l Table 4.8.2 Table 4.8.3 Table 4.8.4 Table 4.8.5 Table 4.8.6 Table 4.11-1 Table 3.12-1 Table 3.12-2 Table 3.12-3 Table 3.12-4 Table 6.1. l Table 6.6.1 DRESDEN II DPR-19 Amendment No. 84,85,86,90,105,107 List of Tables Reactor Protection System (Scram) Instrumentation Requirements Scram Instrumentation Functional Tests Scram Instrumentation Calibration Instrumentation that Initiates Primary Containment Isolation Functions Instrumentation that Initiates or Controls the Core and Containment Cooling System Instrumentation that Initiates Rod Block Radioactive Liquid Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring Instrumentation Post Accident Monitoring Instrumentation Requirements Minimum Test and Calibration Frequency for Core and Containment Cooling Systems Instrumentation, Rod Blocks, and Isolations Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Post Accident Monitoring Instrumentation Surveillance Requirements. Surveillance of the HPCI Subsystem Neutron Flux and Sample Withdrawal Primary Containment Isolation Radioactive Gaseous Waste Sampling and Analysis Program Maximum Permissibl~ Concentration of Dissolved or Entrained Noble Gases Released From the Site to Unrestricted Areas in Liquid Waste Radioactive Liquid Waste Sampling and Analysis Program Radiological Environmental Monitoring Program Reporting Levels for Radioactivity Concentrations in Environmental Samples Practical Lower Limits of Detection (LLD) for Standard Radiological Environmental Monitoring Program Surveillance Requirements for High Energy Piping Outside Containment Fire Detection Instruments Sprinkler Systems CO Systems Fi~e Hose Stations Minimum Shift Manning Chart Special Reports vii Page 3/4.1 - 5 3/4.1 - 8 3/4.1 -10 3/4.2 - 8 3/4.2 -10 3/4.2 -12 3/4.2 -14 3/4.2 -15 3/4.2 -17 3/4.2 -19 3/4.2 -22 3/4.2 -24 3/4.2 -26 3/4.5 - la B 3/4.6 -26 3/4.7 -31 3/4.8-22 3/4.8-24 3/4.8-25 3/4.8-27 3/4.8-28 3/4.8-29 3/4.11-3 B 3/4.12-17 B 3/4.12-18 B 3/4.12-19 B 3/4.12-20 & 21 6-4 6-2.2

DRESDEN II DPR-19 Amendment No. 21, 82, 95, 107 1.0 Definitions (Continued) V. Reactor Power Operation - Reactor power operation is any operation with the mode switch in the 11 Startup/Hot Standbi 1 or 11 Run 11 position with the reactor critical and above 1% rated therma 1 power.

1.

Startup/Hot Standby Mode - In this mode the reactor protection scram trips, initiated by condenser low vacuum and main steamline isolation valve closure, are by-passed when reactor pressure is less than 600 psig; the low pres-sure main steamline isolation valve closure trip is bypassed, the reactor protection system is energized with IRM neutron-monitoring system trips and control rod with-drawal interlocks in service.

2.

Run Mode - In this mode the reactor protectiori ~ystem is energized with APRM protection and RBM interlocks in service. W. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detector. X. Refueling Outage - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant subsequent to that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage. Y. Safety Limit - The safety limits are limits 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 Nuclear Regulatory Commission (NRC) before resump-tion of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review. Z. 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 in compliance with the provisions of Specification 3.7.B.

3.

All automatic ventilation system isolation valves are operable or are secured in the isolated position.

1. 0-4

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

2.

From and after the date that one of the core spray subsystems is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless suc_h __ subsystem is sooner made operable, provided that during such seven days all active components of the other core spray subsystem and the LPCI subsystem and the diesel

• generators required for operation DRESDEN II DPR-19 Amendment No. 82, 98, 107 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.)

to a reactor vessel pressure of 90 psig

c. Pump Operability

d. Motor Operated Valve

e. Core Spray header delta p instru-mentation:

check calibrate test

f. Logic System Functional Test Once/month Once/month Once/day Once/3 months Once/3 months Each Refueling Outage

2.

Deleted 3/4.5-2

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.) of such components if no external source of power were available shall be operable.

3.

Except as specified in 3.5.A.4, 3.5.A.5 and 3.5.F.3 below, the 'LPCI subsystem shall be operable whenever irradiated fuel is in the reactor vessel.

4.

From and after the date that one of the LPCI pumps is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding thirty days unless such pump is sooner made operable, provided that during such thirty days the remaining active components of the LPCI and containment cooling subsystem and all active components of both core spray subsystems and the diesel generators required for operation of such components if no external source of power were available shall be operable.

5.

From and after the date that the LPCI subsystem is made or found to be inoperable DRESDEN II DPR-19 Amendment No. 82, 98, 107 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.)

3.

LPCI Subsystem Testing shall be as specified in 4.5.A.1.a, b, c, d, and f, except that three LPCI pilmps shall deliver at least 14,500 gpm against a system head corresponding to a reactor vessel pressure of 20 psig.

4.

Deleted

5.

Deleted 3/4.5-3 I ~

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.) for any reason, reactor operation is permissible only during the suc-ceeding seven days un-less it lS""SOoner made operable, provided that during such seven days all active com-- ponents of both core spray subsystems, the containment cooling subsys:tem (including 2 LPCI pumps) and the diesel generators required for operation of such components if no external source of power were avail~ble shall be operable.

6.

Containment cooling spray loops are required to be operable when the reactor water temperature is greater than 212°F except that a maximum of one dry-well spray loop may be inoperable for thirty days when the reactor water temperature is greater than 212°F.

7.

The doors of the core spray and LPCI pump compartments shall be closed at all times except during passage in order to consider the core spray and the LPCI subsystems operable. DRESDEN II DPR-19 Amendment No. 14,82,98,101,107 4.5 SURVEILLANCE RE UIREMENT Cont d. 3/4.5-4

6.

During each five year period an air test shall be performed on the drywell spray headers and nozzles.

7.

Whenever the LPCI and core spray subsystems are required to be operable, the doors of the core spray and LPCI pump compartment shall be verified to be closed weekly.

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

8.

If the requirements of 3.5.A cannot be met, an orderly shutdown of the reactor shall be initi-ated and the reactor shall be in the Cold Shutdown condition within 24 hours. Sub-sequently, the reactor mode switch may be placed in Refuel in accordance with 3.5.F.3 through 3.5.F.6. DRESDEN II Amendment No. 107 DPR-19 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-4a

DRESDEN II DPR-19 Amendment No. 34, 82, 101, 107 3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) (Cont 1d.) B. Containment Cooling Subsystem B. Surveillance of the Containment Cooling Subsystem shall be performed as follows:

1.

Except as specified in

1.

Containment Cooling 3.5.B.2, 3.5.B.3, and Service Water Subsystem 3.5.F.3 through 3.5.F.6 Testing: below, both containment cooling subsystem loops Item Freguenc_l'. shall be operable whenever irradiated fuel is in

a. Pump &

Once/3 the reactor vessel and Valve months reactor coolant Operability temperature is greater than 212°F.

b. Flow Rate After pump Test.

Each maintenance containment and every cooling 3 months water pump shall deliver at least 3500 gpm against a pressure of 180 psig.

c. Each manual, Every 31 days power operated or automatic valve, in the fl ow path that is not locked, sealed or other-wise secured in its posit ion, must be verified to be in its correct position.

2.

From and after the date that one of the containment cooling service water subsystem pumps is made or found to be inoperable for 3/4.5-5

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.) any reason, reactor operation is permissible only during the succeeding thirty days unless such pump is sooner made operable, provided that during such thirty days all other active components of the containment cooling subsystem are* operab 1 e.

3.

From and after the date that one containment cooling subsystem is made or found to be inoperable for any r!:!ason, reactor operation is permissible only during the succeeding seven days unless such subsystem is sooner made operable, provided that a 11 active components of the other containment cooling subsystem, both core spray subsystems and both diesel generators required for operation of such components if no external source of power were available, shall be operable.

4.

If the requirements of 3.5.B cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours. DRESDEN II DPR-19 Amendment No. 82, 98, 107 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-6

DRESDEN II DPR-19 Amendment No. 65, 73, 82, 107 3.5 LIMITING CONDITION FOR OPERATION (Cont'd.) 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) C. HPCI Subsystem

1.

Except as specified in 3.5.C.2 below, the HPCI subsystem shall be operable whenever the reactor pressure is greater than 150 psig and irradiated fuel is in the reactor vessel.

2.

a.

From and after the date . that the HPCI subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless such subsystem is sooner made operable, provided that during such seven days all active components of the Automatic Pressure Relief Subsystem, the core spray subsystems, LPCI subsystem, and isolation cooling system are operable.

b.

During reactor startup when HPCI surveillances are being performed, if the testing requirements of 4.5.C.3 or 4.5.C.4 cannot be met, continued reactor startup is not permitted. The HPCI system shall be declared inoperable and the provisions of 3.5.C.3 shall be implemented.

3.

If the requirements of 3.5.C cannot be met, an orderly shutdown shall 3/4.5-7 C. Surveillance of HPCI Subsystem shall be performed as indicated in Table 4. 5.1.

DRESDEN II DPR-19 Amendment No. 107 TABLE 4.5.1 SURVEILLANCE Of THE HPCI SUBSYSTEM Item

1.

Pump Operability

2.

Motor Operated Valve

3.

Flow Rate Test-HPCI pump shall deliver at least 5000 gpm against a system head corresponding tp a reacto~ vessel pr~ssure of 1150 psig when steam is being supplied to the turbine at (1000 +20, -80) psig.**

4.

Flow Rate Test-HPCI pump shall deliver at least 5000 gpm against a system head corresponding to a reactor vessel pressure of > 300 psig when steam is being-supplied to the turbine at 300 (+50, -100) psig.*

5.

Simulated Automatic Actuation test

6.

Logic System Functional Test ~reguency Every 31 days Every 31 days After pump maintenance and every 3 months Prior to exceeding 350 psi g fo 11 owing a refueling outage or an outage during which HPCI maintenance was performed. Each refueling outage Each refueling outage

• Entry into the Startup/Hot Standby Mode is permitted provided that the required testing is successfully completed within 12 hours after reactor steam pressure is adequate to perform the test.
  • Entry into the Run Mode is permitted provided that the required testing is successfully completed within 12 hours after reactor steam pressure is adequate to perform the. test.

3/4.5-7a

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.) be initiated and the reactor pressure shall be reduced to 150 psig within 24 hours. D. Automatic Pressure Relief Subsystems

1.

Except.as specified in 3.5.D.2 and 3.5.D.3 below, the Automatic Pressure Relief Subsystem shall be operable whenever the reactor pressure is greater than 150 psig and irradiated fuel is in the reactor vessel.

2.

From and after the date that one of the five relief valves of the automatic pressure relief subsystem is made or found to be inoperable, reactor DRESDEN II DPR-19 Amendment No. 75, 82, 104, 107 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) 3/4.5-8 D. Surveillance of the Automatic Pressure Relief Subsystem shall be performed as follows:

1.

During each operating cycle the following shall be performed:

a.

A simulated automatic initiation which opens all pilot valves, and

b.

With the reactor at pressure each relief valve shall be manually opened. Relief valve opening shall be verified by a compensating turbine bypass valve or control valve closure.

c.

A logic system functional test shall be performed each refueling outage.

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.) operation is permis-sible only during the succeeding seven(7) days provided that during such time the HPCI subsystem is operable. If the following MAPLHGR reduction factors (multi-pliers) are applied to Figure 3.5-1, the Auto-matic Pressure Relief Sub~ystem of ECCS shall be considered operable: (1) 0.89 for 8x8 fuel, or (2) 0.76 for 9x9 fuel.

3.

From and after the date that two relief valves are found or made to be inoperable, reac-tor operation is permis-sible only during the succeeding seven days provided that during such time the HPCI subsystem is operable and the multipliers specified in 3.5.D.2 are applied.

4.

If the requirements of 3.5.D.l cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to below 150 psig within 24 hours. E. Isolation Condenser System DRESDEN II DPR-19 Amendment No. 75, 82, 94, 104, 107 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) E. Surveillance of the Isolation Condenser System shall be performed as follows: 3/4.5-9

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

1.

Whenever the reactor pressure is greater than 150 psig and irradiated fuel is in the reactor vessel, the isolation condenser shall be operable except as specified in 3.5.E.2.

2.

From and after the date that the isolation condenser system is made or found to be inoperable for any reason, reactor operation is permissible only during th~ succeeding seven days unless such system is sooner made operable, provided that during such seven days all active components of the HPCI subsystem are operable. DRESDEN II DPR~l9 Amendment No. 19, 28, 82, 107 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-10

1.

I sol at ion Condenser System Testing:

a.

The shell side water level and temperature shall be checked daily.

b.

Simulated automatic actuation and func tional system test-ing shall be per-formed during each refueling outage or whenever major re-pairs are completed on the system.

c.

The system heat removal capability shall be determined once every five years.

d.

Calibrate vent line radiation monitors quarterly.

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

3.

If the requirements of 3.5.E cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 150 psig within 24 hours. F. Minimum Core and Containment Cooling System Availability

1.

Any combination of inoperable components in the core and containment cooling systems shall not defeat the capability of the remaining operable components to fulfill the core and containment cooling functions. DRESDEN II DPR-19 Amendment No. 19, 28, 82, 107 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-11 F. Surveillance of Core and Containment Cooling System

1.

Actions necessary to assure that the plant can be safely shut down and maintained in this condition in case of failure of the Dresden Dam shall be demonstrated to be adequate every third refue 1 i ng outage. If this Specification has been complied with for Dresden Unit 3, it shall not be required for Dresden Unit 2.

• • • ~
• r-t 3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

2.

When irradiated fuel is in the reactor vessel and the reactor is in cold shutdown or refuel conditions, at least two of the following pumps, with each having an operable flow path capable of taking suc-tion from the suppres-sion pool or the conden-sate storage tank and transferring the water to the reactor vessel, shall be operable except as specified in 3.5.F.4, 3.5.F.5 and 3.5.F.6 below:

a.

Two Core Spray pumps or,

b.

Two Low Pressure Cool-ant Injection pumps or,

c.

One Core Spray pump and one Low Pressure Cool-ant Injection pump.

3.

With one of the pumps and/ or associated flow paths required by 3.5.F.3. inoperable, restore at least two pumps and asso-ciated flow paths to oper-able status within 4 hours or suspend all operations with a potential for drain-ing the reactor vessel.

4.

With both of the pumps and/or associated flow paths required by 3.5.F.3 inoperable, suspend core alterations and all opera-tions with a potential for draining the reactor vessel. Restore at least one pump and associated flow path to operable status within 4 hours or estab*l i sh secondary con-tainment integrity within the next 8 hours. DRESDEN II DPR-19 Amendment No. 20, 58, 82, 101, 107 4.5 SURVEILLANCE RE UIREMENT Cont d. 3/4.5-12

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

5.

When irradiated fuel is in the reactor vessel and the reactor is in the cold shutdown or refuel condition, all low pres-sure core and containment cooling subsystems may be inoperable provided the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, the fuel pool water level is maintained above the low level alarm point, and the reactor cavity water temperature is below 140°F.

6.

When irradiated fuel is in the reactor vessel and the reactor is in the refuel condition, the torus may be drained com-pletely and control rod drive maintenance per~ formed provided that the spent fuel pool gates are open, the fuel pool water level is maintained above the low level alarm point, and the minimum total con-densate storage reserve is maintained at 230,000 gallons, and provided that not more than one control rod drive housing is open at one time, the control rod drive housing is blanked following removal of the control rod drive, no work is being performed in the reactor vessel while the housing is open and a special flange DRESDEN II DPR-19 Amendment No. 101, 107 4.5 SURVEILLANCE RE UIREMENT Cont d. 3/4.5-12a

)- 3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.) is available which can be used to blank an open housing in the event of a leak.

7.

When irradiated fuel is in the reactor and the vessel head is removed, ~ork that has the poten-tial for drain.ing the vessel *may be performed wi~h less than 112,000 ft of water in the suppression pool, pro-vided that: (1) the total volume of water in the suppression pool, dryer separator above the shield blocks, refuel-ing cavity, and the fuel storage pool above the bottom of the fuel pool gate is gr~ater than 112,000 ft ; (2) the fuel storage pool gate is removed; (3) the low pres-sure coolant injection and core spray systems are operable as specified in 3.5.F.3, 3.5.F.4 and 3.5.F.5; and (4) the automatic mode of the drywell sump pumps is disabled. G. Not Used H. Maintenance of Filled Discharge Pipe Whenever core spray, LPCI, or HPCI ECCS are required to be operable, the discharge piping from the pump discharge DRESDEN II DPR-19 Amendment No. 20, 58, 92, 101, 107 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) G. Not Used H. Maintenance of Filled Discharge Pipe The following surveil-lance requirements shall be adhered to, to assure that the discharge piping of the core spray, 3/4.5-13

DRESDEN II DPR-19 Amendment No. 107 THIS PAGE WAS LEFT INTENTIONALLY BLANK 3/4.5-29

DRESDEN II DPR-19 Amendment No. 82,94,95,104,107 3.5 LIMITING CONDITION FOR OPERATION BASES A. Core ~ and LPCI Mode of the RHR System - This specification assures that adequate emergency cooling capability is available. Based on the loss of coolant analyses included in References (1) and (2) in accordance with 10 CFR 50.46 and Appendix K, core cooling systems provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant acci-dent, to limit the calculated peak clad temperature to less than 2200°F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than 1%, and to limit the cdculated local metal-water reaction to less than 17%. The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in Reference (3). Using the results developed in this reference, the repair period is found to be less than 1/2 the test interval. This assumes that the core spray and LPCI subsystems constitute a 1 out of 3 system, however, the combined effect of the two systems to limit excessive clad temperatures must also be considered. The test interval specified in Specification 4.5 was 3 months. There-fore, an allowable repair period which maintains the basic risk* considering single failures should be less than 45 days and this specification is within this period. For multiple failures, a shorter interval is specified. ---j (1) "Loss of Coolant Accident Analyses Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations," NED0-24146A, Revisions 1, April 1979. (2) NED0-20566, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K. (3) APED-"Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards" - April 1969, I.M. Jacobs and P.W. Marriott. B 3/4.5-30

DRESDEN II DPR-19 Amendment No. 82,84,95,98,104,107 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) Although it is recognized that the information given in refer-ence 3 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analyt-ical approach prevents complete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due regard to judgment. Should one core spray subsystem become inoperable, the rema1n1ng. core spray and the entire LPCI system are available should the reactor core cooling arise. Based on judgments of the reli- -i ability of.the remaining systems; i.e. the core.spray and LP.CI, a 7-day repair period was obtained. Should the loss of one LPCI pump occur, a nearly full complement of core and containment cooling equipment is available. Three LPCI pumps in conjunction with the core spray subsystem will perform the core cooling function. Because of the availability of the majority of the core cooling equipment, a 30-day repair -f period is justified. If the LPCI subsystem is not available, at least 2 LPCI pumps must be available to fulfill the containment cooling function. The 7-day repair period is set on this basis.. When a core spray or LPCI subsystem (or pump) is inoperable, there is a requirement that other emergency core cooling systems be operable. The verification of operability, as used in this context, for these other systems means to administratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. B. Containment Cooling Service Water - The containment heat removal portion of the LPCI/containment cooling subsystem is provided to remove heat energy from the containment in the event of a loss of coolant accident. For the flow specified, the containment long-term pressure is limited to less than 8 psig and, there-fore, is more than ample to provide the required heat removal capability. (Ref. Section 5.2.3.2 SAR). I The containment cooling subsystem consists of two sets of cooling equipment. Each set contains 2 service water pumps, 1 heat exchanger and 2 LPCI pumps. Either set of equipment is capable of performing the containment cooling function. Loss of one containment cooling service water pump or one LPCI pump does not seriously jeopardize the containment cooling capability as any 2 of the remaining three pumps can satisfy the cooling B 3/4.5-31

DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) requirements. Since there is some redundancy left, a 30-day repair period is adequate. Loss of 1 containment cooling subsystem leaves one remaining system to perform the contain-ment cooling function. Based on the fact that when one containment cooling subsystem becomes inoperable, only one system remains, a 7-day repair period was specified. When a containment cooling subsystem (or pump) is inoperable, there is a requirement that other emergency core cooling systems be operable. The verification of operability, as used in this context, for these other systems means to administratively check* by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. C. ~iSh Pressure Coolant Injection - The high pressure coolant 1nJect1on subsystem is provided to adequately cool the ~ore for all pipe breaks smaller than those for which the LPCI and core spray subsystems can protect the core. The HPCI meets this requirement without the use of off-site electrical power. For the pipe breaks for which the HPCI is intended to function, the core never uncovers and is continuously cooled and thus no clad damage occurs (Ref. Section 6.2.5.3 SAR). The repair times for the limiting conditions of operation were set considering the use of the HPCI as part of the isola-tion cooling system. When the HPCI subsystem is inoperable, there is a requirement that other emergency core cooling systems be operable. The verification of operability, as used in this context, for these other systems means to administratively check by examining logs or other information to determine if certain components/ systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. D. Automatic Pressure Relief - The relief valves of the automatic pressure relief subsystem are a back-up to the HPCI subsystem. They enable the core spray and LPCI to provide protection against the small pipe break in the event of HPCI failure, by depressurizing the reactor vessel rapidly enough to actuate the core sprays and LPCI. The core spray and LPCI provide sufficient flow of coolant to adequately cool the core. B 3/4.5-32

3.5 DRESDEN II DPR-19 Amendment No. 82,94,95,104,107 LIMITING CONDITION FOR OPERATION BASES (Cont 1 d.) Analyses have shown that only four of the five valves in the automatic depressurization system.are required to operate. Loss of one of the relief valves does not significantly affect the pressure relieving capability, therefore, continued operation is acceptable provided the appropriate MAPLHGR reduction factor is applied to assure compliance with the 2200°F PCT limit. Loss of more than one relief valve significantly reduces the pressure relief capability of the ADS; thus, a 7-day repair period is specified with the HPCI available, and a 24-hour repair period otherwise. When one or more relief valves is inoperable, there is a require-ment that the HPCI subsystem be operable. The verification of operability, as used in this context, for HPCI means to admini-stratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform* the surveillance requirements needed to demonstrate the operability of the component/system. E. Isolation Coolins System - The turbine main condenser is normally available. The isolation condenser is provided for core decay heat removal following reactor isolation and scram. The isola-tion condenser has a heat removal capacity sufficient to handle the decay heat production at 300 seconds following a scram. Water will be lost from the reactor vessel through the relief valves in the 300 seconds following isolation and scram. This represents a minor loss relative to the vessel inventory. The system may be manually initiated at any time. The system is automatically initiated on high reactor pressure in excess of 1060 psig sustained for 15 seconds. The time delay is provided to prevent unnecessary actuation of the system during anticipated turbine trips. Automatic initiation is provided to minimize the coolant loss following isolation from the main condenser. To be considered operable, the shell side of the isolation condenser must contain at least 11~300 gallons of water. Make-up water to the shell side of the isolation condenser is provided by the condensate transfer pumps from the condensate storage tank. The condensate transfer pumps are operable from on-site power. The fire protection system is also available as make-up water. An alternate method of cooling the core upon isolation from the main condenser is by using the relief valves and HPCI subsystem in a feed and bleed manner. Therefore, the high pressure relief function and the HPCI must be available together to cope with an anticipated transient so the LCO for HPCI and relief valves is set upon this function rather than their function as depressuri-zation means for a small pipe break. B 3/4,5..:33

3.5 DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) When the isolation condenser system is inoperable, there is a requirement that the HPCI subsystem be operable. The verification of operability, as used in this context, for HPCI means to administratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. F. Emergency Coolins Availability - The purpose of Specification D is to assure a minimum of core cooling equipment* is available-at all times. If, for example, one core spray were out of service and the diesel which powered the opposite core spray were out of service, only 2 LPCI pumps would be available. Likewise, if 2 LPCI pumps were out of service and 2 containment service water pumps on the opposite side were also out of service no contain-ment cooling would be available. It is during refueling outages that major maintenance is performed and during such time that all low pressure core cooling systems may be out of service. This specification provides that should this occur, no work will be performed on the primary system which could lead to draining the vessel. This work would include work on certain control rod drive components and recirculation system. Thus, the specifi-cation precludes the events which could require core cooling. Specification 3.9 muit also be consulted to determine other requirements for the diesel generators. . When the unit or shared diesel generator is inoperable, there is a requirement that other emergency core cooling systems be operable. The verification of operability, as used in this context, for these other systems means to administratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. Dresden Units 2 and 3 share certain process systems such as the makeup demineralizers and the radwaste sistem and also some safety systems such as the standby gas treatment system, batteries, and diesel generators. All of these systems have been sized to perform their intended function considering the simultaneous operation of both units. For the safety related shared features of each plant, the Technical Specifications for that unit contain the operability and surveillance requirements for the shared feature; thus, the 8 3/4.5-34

3.5 DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) G. level of operability for one unit is maintained independently of the status of the other. For example, the shared diesel (2/3 diesel) would be mentioned in the specifications for both Units 2 and 3 and even if Unit 3 were in the Cold Shutdown Condition and needed no diesel power, readiness of the 2/3 diesel would be required for continuing Unit 2 operation. Specification 3.5.F.4 provides that should this occur, no work ~ will be performed which could preclude adequate emergency cooling capability being available. Work is prohibited unless it is in accordance with specified procedures which limit the period that the control. rod drive housing is open and assures that the worst possible loss of coolant resulting from the work will not result in uncovering the reactor core. Thus, this specification assures adequate core cooling. Specification 3.9 must be consulted to determine other requirements for the diesel generator. Specification 3.5.F.5 provides assurance that an adequate supply of coolant water is immedi~tely available to the low pressure core cooling systems and that the core will remain covered in the event of a loss of coolant accident while the reactor is depressurized with the head removed. Not Used. H. Maintenance of Filled Discharge Pipe - If the discharge piping of the core spray, LPCI, and HPCI are not filled, a water hammer can develop in this piping when the pump and/or pumps are started. I. Average Planar LHGR This specification assures that the peak cladding temperature following a postulated design basis loss-of-coolant accident will not exceed the 2200°F limit specified in 10 CFR 50 Appendix K considering the postulated affects of fuel pellet densification. The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average LHGR of all the rods in a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within

• a fuel assembly.

Since expected local variations in power dis-tribution within a fuel assembly affect the calculated peak clad temperature by less than plus or minus 20°F relative to the peak temperature for a typical fuel design, the limit on the average B 3/4.5-35

3.5 DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) J. planar LHGR is sufficient to assure that calculated temperatures are below the 10 CFR 50, Appendix K limit. The maximum average planar LHGRs shown in Figure 3.5.l are based on calculations employing the models described in Refer-ence (1), (2) and (3). Power operation with APLHGRs at or below those shown in Fig. 3.5.l assures that the peak cladding tempera-ture following a postulated loss-of-coolant accident will not exceed the 2200°F limit. ANF has analyzed the effects that Single Loop Operation has on LOCA events (Reference 4). For breaks in the idle loop, the above Dual Loop Operation resuits are conservative (Reference 1). For breaks in the active loop, the event is more severe primarily due to a more rapid loss of core flow. By applying a multiplica-tive 0.91 reduction factor to the results of the previous analyses, all applicable criteria are met. The maximum average planar LHGRs for G.E. fuel plotted in Figure 3.5.l at higher exposures result in a calculated peak clad temperature of less than 2200°F. However, the maximum average planar LHGRs are shown on Fig. 3.5.1 as limits because conform-ance calculations have not been performed to justify operation at LHGRs in excess of those shown. Local Steady State LHGR This specification assures that the maximum linear heat genera-tion rate in any fuel rod is less than the design linear heat generation rate even if fuel pellet densification is postulated. This provides assurance that the fuel end-of-life steady state criteria are met. (l)i'Loss of Coolant Accident Analyses Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations," NED0-24146A, Revi-sion l, April 1979. (2)XN-NF-82-88 "Dresden Unit 2 LOCA Analysis Using the ENC EXEM/BWR Evaluation Model MAPLHGR Results." (3)XN-NF-85-63 "Dresden Unit 3 LOCA-ECCS Analysis MAPLHGR results for 9x9 fuel," dated September 1985. (4)ANF-84-lll, "LOCA-ECCS Analysis for Dresden Units during Single Loop Operation with ANF Fuel," September 1987. B 3/4.5-36 r-

3.5 l* DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) K. Local Transient LHGR This specification provides assurance that th_e fuel wi 11 neither experience centerline melt nor exceed 1% plastic cladding strain for transient overpower events beginning at any power and termi-nating at 120% of rated thermal power. L. Minimum Critical Power Ratio (MCPR) The steady-state values for MCPR specified in the Specification were determined using the THERMEX thermal limits* methodology-described in XN-NF-80-19, Volume J. The safety limit implicit in the Operating limits is established so that during sustained operation at the MCPR safety limit, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition. The Limiting Transient delta CPR implicit in the operating limits was calculated such that the occurrence of the limiting transient from the operating limit will not result in violation of the MCPR safety limit in at least 95% of the random statistical combina-tions of uncertainties. Transient events of each type anticipated during operation of a BWR/3 were evaluated to determine which is most restrictive in terms of thermal margin requirements. The generator load rejection/turbine trip without bypass is typically the limiting event. The thermal margin effects of the event are evaluated with the THERMEX Methodology and appropriate MCPR limits con-sistent with the XN-3 critical power correlation are determined. Several factors influence which transient results in the largest reduction in critical power ratio, such as the cycle-specific fuel loading, exposure and fuel type. The current cycle's reload licensing analyses identifies the limiting transient for that cycle. For core flows less than rated, the MCPR Operating Limit estab-lished in the specification is adjusted to provide protection of the MCPR Safety Limit in the event of an uncontrolled recircula-tion flow increase to the physical limit of pump flow. This protection is provided for manual and automatic flow control by choosing the MCPR operating limit as the value from Figure 3.5-2 Sheet 1 or the rated core flow value, whichever is greater. For Automatic Flow Control, in addition to protecting the MCPR Safety Limit during the flow run-up event, protection is pro-vided against violating the rated flow MCPR Operating Limit during an automatic flow increase to rated core flow. This protection is provided by the reduced flow MCPR limit~ shown in Figure 3.5-2 Sheet 2 where the curve corresponding to the B 3/4.5-37

DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont 1d.) M. current rated flow MCPR limit is used (linear interpolation between the MCPR limit lines depicted is permissible). There-fore, for Automatic Flow Control, the MCPR Operating Limit is chosen as the value from Figure 3.5-2 Sheet 1, Sheet 2 or the rated flow value, whichever is greatest. Analyses have demonstrated that transient events in Single Loop Operation are bounded by those at rated conditions; however, due to the increase in the MCPR fuel cladding integrity safety limit in Single Loop Operation, an equivalent adder must be uniformly applied to all MCPR LCO to maintain the same.margins to the MCPR

• fuel cladding integrity safety limit.

Flood Protection Condensate pump room flood protection will assure the availabil-ity of the containment cooling service water system (CCSW) during a postulated incident of flooding in the turbine build-ing. The redundant level switches in the condenser pit will preclude any postulated flooding of the turbine building to an elevation above river water level. The level switches provide alarm and circulating water pump trip in the event a water level is detected in the condenser pit. B 3/4.5-38

DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 4.5 SURVEILLANCE REQUIREMENT BASES (A thru F) The testing interval for the core.and containment cooling systems is based on 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 stimulated, the final valves would not open. In the case of the HPCI, automatic initiation during power operation would result in pumping ~old water.into the reactor vessel ~hich is not desirable. The systems can be automatically aGtuated 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., instru-mentation, pumps, valve operators, etc., are tested more fre-quently. The instrumentation is functionally tested each month. Likewise the pumps and motor-operated valves are also tested each month to assure their operability. The combination of a yearly simulated automatic actuation test and monthly tests of the pumps and valve operators is deemed to be adequate testing of these systems. With components or subsystems out-of-service overall core and containment cooling reliability is maintained by verifying the operability of the remaining cooling equipment. The verification of operability, as used in this context, for these other systems means to administratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system.

However, if a failure, design deficiency, etc., caused 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 components.

For example, if an out-of-service period were caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test. The requirement of 180 psig at 3500 gpm at the containment cooling service water (CCSW) pump discharge provides adequate margin to ensure that the LPCI/CCSW system provides the design B 3/4.5-39

DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 4.5 SURVEILLANCE REQUIREMENT BASES (Cont 1d.) bases cooling water flow and maintains 20 psig differential pressure at the containment cooling heat exchanger. This differential pressure precludes reactor coolant from entering the river water side of the containment cooling heat exchangers. The verification of 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 m.ay result with the valve still closed. This is due to steam being vented through the.valve actuation mechanism during the-survei l lance 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 relief valve opening. This test method may be performed over a wide range of reactor pressure greater than 150 psig. Valve opera-tion below 150 psig is limited by the spring tension exhibited by the relief valves. G. Deleted H. Maintenance of Filled Discharge Pipe The surveillance requirements to assure that the discharge piping of the core spray, LPCI, and HPCI systems are filled provides for a visual observation that water flows from a high point vent. This ensures that the line is in a full condition. Between the monthly intervals at which the lines are vented, instrumentation has been provided to monitor the presence of water in the discharge piping. This instrumentation will be calibrated on the same frequency as the safety system instru-mentation. This period of periodic testing ensures that during the intervals between the monthly checks the status of the discharge piping is monitored on a continuous basis. I. Average Planar LHGR At core thermal power levels less than or equal to 25 per cent, operating plant experience and thermal hydraulic analyses indi-cate that the resulting average planar LHGR is below the maxi-mum average planar LHGR by a considerable margin; therefore, B 3/4.5-40

DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 4.5 SURVEILLANCE REQUIREMENT BASES (Cont 1d.) evaluation of the average planar LHGR below this power level is not necessary. The daily requirement for calculating average planar LHGR above 25 per cent rated thermal power is sufficient since power distribution shifts are slow when there have not been significant power or control rod changes. J. Local Steady State LHGR The LHGR for all fuel shall be checked daily during reactor operation at greater than or equal to 25 per cent power to determine if fuel burnup or control rod movement has caused changes in power distribution. A limiting LHGR value is pre-cluded by a considerable margin when employing a permissible cont~ol rod pattern below 25% rated thermal power. K. Local Transient LHGR The fuel design limiting ratio for centerline melt (FDLRC) 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 changes in power distribution. The FDLRC limit is designed to protect against centerline melt-ing of the fuel during anticipated operational occurrences. L. Minimum Critical Power Ratio (MCPR) M. At core thermal power levels less than or eq~al to 25 percent, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicates that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculating MCPR above 25 percent 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 reduced flow correction applied to the LCO provides margin for flow increase from low flows. Flood Protection The watertight bulkhead door and the penetration seals for pipes and cables penetrating the vault walls have been designed B 3/4.5-41

C* DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 4.5 SURVEILLANCE REQUIREMENT BASES (Cont 1d.) to withstand the maximum flood conditions. To assure that their installation is adequate for maximum flood conditions, a method of testing each seal has been devised. To test a pipe seal, another test seal is installed in the opposite side of the penetration creating a space between the two seals that can be pressurized. Compressed air is then supplied to a fitting on the test seal and the space inside the sleeve is pressurized to approximately 15 psi. The outer face of the permanent seal is then tested for leaks using a soap bubble solution. On completion of the test, the test seal is removed for use on other pipes and penetrations of the same size. In order to test the watertight bulkhead doors, a test frame must be installed around each door. At the time of the test, a reinforced steel box with rubber gasketing is clamped to the wall around the door. The fixture is then pressurized to approximately 15 psig to test for leak tightness. Floor drainage of each vault is accomplished through a carbon steel pipe which penetrates the vault. When open, this pipe will drain the vault floor to a floor drain sump in the conden-sate pump room. Equipment drainage from the vault coolers and the CCSW pump bedplates will also be routed to the vault floor drains. The old equipment drain pipes will be permanently capped to pre-clude the possibility of back-flooding the vault. As a means of preventing backflow from outside the vaults in the event of a flood, a check valve and an air operated valve are installed in the 2 11 vault floor drain line 6 10 11 above the floor of the condensate pump room. The check valve is a 2 11 swing check designed for 125 psig service. The air operated valve is a control valve designed for a 50 psi differential pressure. The control valve will be in the normally open position in the energized condition and will close upon any one of the following:

a.

Loss of air or power

b.

High level (5 10 11 ) in the condensate pump room B 3/4.5-42

DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.) Closure of the air operated valve on high water level in the condensate pump room is effected by use of a level switch set at a water level of 5 10 11

• Upon actuation, the switch will close the control valve and alarm in the control room.

The operator will also be aware of problems in the vaults/ condensate pump room if the high level alarm on the equipment drain sump is not terminated in a reasonable amount of time.

• 1t must be pointed out that these alarms provide information to the operator but that operator action upon the above alar~s is not a necessity for reactor safety since the* other proiii s i o*ns
• provide adequate protection.

A system of level switches has been installed in the condenser pit to indicate and control flooding of the condenser area. The following switches are installed: Level Function

a.

1 10 11 (1 switch) Al arm, Panel Hi-Water-Condenser Pit

b.

3 10 11 (1 switch) Alarm, Panel High-Circ. Water Condenser Pit

c.

5 10 11 (2 redundant Alarm and Ci re. Water Pump switch pairs) Trip Level (a) indicates water in the condenser ~it from either the hotwell or the circulating water system. Level (b) is above the hotwell capacity and indicates a probable circulating water failure. Should the switches at level (a) and (b) fail or the operator fail 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 switch pairs at level (c) are designed and installed to IEEE-279, "Criteria for Nuclear Power Plant Protection Systems. 11 As the circulating water pumps are tripped, either manually or automatically, at level (c) of 5 10 11 , the maximum water level reached in the condenser pit due to pumping will be at the 491 10 11 elevation (10 1 above condenser pit floor elevation 481 10 11

5 1 plus an additional 51 attributed to pump coastdown).

B 3/4.5-43

DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.) In order to prevent overheating of the CCSW pump motors, a vault cooler is supplied for each pump. Each vault cooler is designed to maintain the vault at a maximum 105°F temperature during operation of its respective pump. For example, if CCSW pump 28-1501 starts, its cooler will also start and compensate for the heat supplied to the vault by the 28 pump motor keeping the vault at less than 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 pum~ 1 ~ suction line. In this way, the vault coolers are supplied with cooling water totally inside the vault. The cooling water quantity needed for each cooler is approximately 1% to 5% of the design flow of the pumps so that the recirculation of this small amount of heated water will not affect pump or cooler operation. Operation of the fans and coolers is required during pump operability testing 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. B 3/4.5-44

3.7 LIMITING CONDITION FOR OPERATION (Cont'd.) shall be initiated and the reactor shall be in a cold shutdown condition in the following 24 hours.

8.

Standby Gas Treatment System

1. Two separate and.

independent standby gas treatment system subsystems shall be operable at all times when secondary containment integrity is required, except as specified in sections 3. 7.8.l(a) and (b).

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 that a 11 active components in the other standby gas treatment subsystem shall be operable. 3/4.7-19 DRESDEN II DPR-19 Amendment No. 17, 37, 82, 94, 107 4.7 SURVEILLANCE REQUIREMENTS (Cont'd.) B. Standby Gas Treatment System . l. At 1 east once per month, initiate.from* the control room 4000 cfm (plus or mi nus 10%) fl ow through each subsystem of the standby gas treatment system for at least 10 hours with the subsystem heaters operating at rated power. I J I I I l i

3. 7 LIMITING CONDITION FOR OPERATION (Cont'd.)

Within 36 hours 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.l.(a) through (d).

b.

If both standby gas treatment system subsystems are not operable, within 36 hours 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.l.(a) through (d).

2.

Performance Requirement

a.

Periodic Requirements: 3/4.7-20 DRESDEN II DPR-19 Amendment No. 17, 82, 107 4.7 SURVEILLANCE REQUIREMENTS (Cont'd.)

2.

Performance Requirement Tests

a.

At least once per 720 hours of system operation; or once per operating cycle, or every 18 months, whichever occurs first; or following painting, fire, or chemical

3.7 LIMITING CONDITION FOR OPERATION (Cont 1d.)

b.

The system shall be shown to operate.

3.

Post Maintenance Requirements

a.

After any maintenance or testing that could affect the HEPA filter or HEPA filter mounting frame leak tight integrity, the results of the inplace DOP tests at 4000 cfm (plus or minus 10%) on HEPA filters shall show less than or equal to 1% DOP penetration in 3/4.7-22 DRESDEN II DPR-19 Amendment No. 17, 82, 107 4.7 SURVEILLANCE REQUIREMENTS (Cont 1d.)

b.

At least once per operating cycle or every 18 months, whichever occurs first, the following conditions shall be demonstrated: (1) Pressure drop across the combined

• filters of each standby gas treat-ment system subsystem is less than 6 inches of water at 4000 cfm (plus or minus 10%) flow rate.

(2) Operability of inlet heater at rated power. (3) Automatic initiation of each standby gas treatment system subsystem. r

3.

Post Maintenance I Testing

a.

After any maintenance or testing that could affect the leak tight integrity of the HEPA filters, perform in-place DOP tests on the HEPA filters in accordance with Specification 3.7.8.2.a. (1).

3.7 LIMITING CONDITION FOR OPERATION (Cont 1d.) accordance with Specification

3. 7.8.2.a(l).

b.

After any maintenance or testing that could affect the charcoal adsorber leak tight integrity, the results of in-place halogenated hydrocarbon tests at 4000 cfm (plus or minus 10%) on charcoal adsorber banks shall show less than or equal to 1% penetration in accordance with Specification 3.7.8.2.a(2).

c.

The results of in-place air distribution tests shall show the air distribution is uni form within plus or minus 20% to each HEPA filter when tested initially and after any maintenance or testing that could affect the air distribution within the standby gas treatment system. 3/4.7-23 DRESDEN II DPR-19 Amendment No. 17, 82, 107 4.7 SURVEILLANCE REQUIREMENTS (Cont 1d.)

b.

After any maintenance or testing that could affect the leak tight integrity of the charcoal adsorber banks, perform halogenated hydrot~~bon tests

• on the charcoal adsorbers in accordance with Specification

3. 7.8.2.a.(2).

c.

Perform an air distribution test on the HEPA filter bank initially and after any maintenance or testing that could affect the air distribution within the standby gas treatment system. The test shall be performed at 4000 cfm (plus or minus 10%) flow rate.

DRESDEN II DPR-19 Amendment No. 17, 82, 107 This page has been deleted. 3/4.7-24 I I I -, I I I

DRESDEN II DPR-1 Amendment No. 107 This page has been deleted. 3/4.7-25 J I

3.7 LIMITING CONDITION FOR OPERATION (Cont'd.). C. Secondary Containment

1.

Secondary containment integrity shall be maintained during all modes of plant operation except when all of the following conditions are met.

a.

The reactor is subcritical and Specification 3.3.A is met.

b.

The reactor water temperature is below 212°F and the reactor coolant system is vented.

c.

No activity is being performed which can reduce the shutdown margin below that specified in Specification

3. 3. A.

3/4.7-26 DRESDEN II DPR-19 Amendment No. 17, 40, 82, 107 4.7 SURVEILLANCE REQUIREMENTS (Cont'd.) C. Secondary Containment

1.

Secondary containment surveillance shall be performed as indicated be low:

a.

Secondary. containment capability to maintain a 1/4 inch of water vacuum under calm wind (less than 5 mph) conditions with a filter train flow rate of not more than 4000 cfm, shall be demonstrated at each refueling outage prior to refueling. r .( ~ I.

• 3. 7 LIMITING CONDITION FOR OPERATION (Cont 1d.)

d.

The fuel cask or ir-radiated fuel is not being moved in the reactor building.

2.

If Specification 3. 7. C. l cannot be met, restore Secondary Containment Integrity within 4 hours or be in at least Hot Shutdown within the next 12 hours and in Cold Shutdown within the following 24 hours and establish the conditions listed in Specification 3.7.C.l.a through d. D. Primary Containment Isolation Valves

1.

During reactor power operating conditions, all isolation valves listed in Table 3.7.1 and all instrument line flow check valves shall

• be operable except as specified in 3.7.D.2.

3/4.7-27 DRESDEN II DPR-19 Amendment No. 40, 82, 107 4.7 SURVEILLANCE REQUIREMENTS (Cont 1d.) D. Primary Containment Isolation Valves

1.

The primary containment isolation valves surveillance shall be performed as follows:

a.

At least once per operating cycle the operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and closure t.imes.

DRESDEN II DPR-19 Amendment No. 37, 82, 107 3.7 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed (Reference 15) which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping. The maintenance of a drywell-suppression chamber differential pressure of 1.00 psid and a suppression chamber water level corresponding to a downcomer submergence range of 3.67 to 4.00 feet will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.

8.

Standby Gas Treatment System and C. Secondary Containment The secondary containment is designed to m1n1m1ze any ground level release of radioactive materials which might result from a serious accident. The reactor building provides secondary containment during reqctor operation, when the drywell is sealed and in service; the reactor building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the ~econdary containment is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required as well as during refueling. Only one of the two standby gas treatment system subsystems is needed to cleanup the reactor building atmosphere upon containment isolation. If one subsystem is found to be inoperable, there is no immediate threat to the containment system performance. Therefore, reactor operation or refueling operation may continue while repairs are being made. If neither subsystem is operable, the plant *is placed in a condition that does not require a standby gas treatment system. While only a small amount of particulates are released from the primary containment as a result of the loss of coolant accident, high-efficiency particulate filters before and after

• the charcoal filters are specified to minimize potential particulate release to the environment and to prevent clogging of the charcoal adsorbers.

The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system.leak tightness of less than 1% bypass leakage for the charcoal adsorbers using halogenated hydrocarbon and a HEPA filter efficiency of at least 99% removal of DOP particulates. B 3/4.7-38 -f

DRESDEN II DPR-19 Amendment No. 37, 82, 107 3.7 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) Laboratory carbon sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions. Operation of the standby gas treatment subsystems significantly different from the design flow will change the . removal efficiency of the HEPA filters and charcoal adsorbers. If the performance requirements are met as specified, the calculated doses would be less than the guidelines stated in 10 CFR 100 for the* accidents analyzed. The standby gas treatment system is designed to filter and exhaust the reactor building* atmosphere to the stack during secondary containment isolation conditions, with a minimum release of radioactive materials from the reactor building to the environs. One standby gas treatment fan is designed to automatically start upon containment isolation and to maintain the reactor building pressure to approximately a negative 1/4-inch water guage pressure; all leakage should be in-leakage. Should the fan fail to start, the redundant alternate fan and filter system is designed to start automatically. Each of the two fans has 200% capacity. (Ref. Section 5.3.2 SAR.) If one standby gas treatment system subsystem is inoperable, the other subsystem will be verified to be operable. The verification of operability, as used in this context, for the other subsystem means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonst.rate the operability of the component/subsystem. This substantiates the availability of the operable subsystem and results in no added risk; thus, reactor ope~ation or refueling operation can continue. If neither subsystem is operable, the plant is brought to a condition where the system is not required. While only a small amount of particulates are released from the pressure suppression chamber system as a result of the loss of coolant accident, high-efficiency particulate filters before and after the charcoal filters are specified to minimize potential particulate release to the environment and to prevent clogging of the charcoal adsorbers. The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The inplace test results should indicate a system leak tightness of less than 1% bypass leakage for the charcoal adsorbers using halogenated hydrocarbon and a HEPA filter efficiency of at least 99% removal of DOP particulates. Laboratory carbon sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions. Operation of the B 3/4.7-39

DRESDEN II DPR-19 Amendment No. 17, 36, 37, 82, 107 3.7 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) standby gas treatment subsystems significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. If the performance require-ments are met as specified, the calculated doses would be less than the guidelines stated in 10 CFR 100 for the accidents analyzed. D. Primary Containment Isolation Valves - Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is require.cl to minimize the potential leakage paths from the containment in the event of a loss of coolant accident. 4.7 SURVEILLANCE REQUIREMENT BASES A. Primary Containment Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress. The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspection of the paint during each major refueling outage, approximately once per year, assures the paint is intact. Experience with this type of paint at fossil fueled generating stations indicates that. the inspection interval is adequate. B 3/4.7-40

4. 7 SURVEILLANCE REQUIREMENT BASES (Cont 1d.)

DRESDEN II DPR-19 Amendment No. 17, 82, 107 occurred and additional testing is required immediately. The frequency of testing the alarms is based on experience and quality of the equipment. During each refueling outage, three drywell-suppression chamber vacuum breakers will be inspected to assure sealing surfaces and components have not deteriorated. Since valve internals are designed for a 40-year lifetime, an inspection program which cycles through all valves in 1/10 of the design lifetime is extremely conservative. The primary containment is normally slightly pressurized during ~eriods of reactor operation. Nitrogen used for inerting could leak o~t of the *containment but.air could not leak in to increase oxygen concentration. Once the containment is filled with nitrogen to the required concentration, no monitoring of oxygen concentration is necessary. However, at least once a week the oxygen concentration will be determined as added assurance. Recording N2 storage tank level weekly and after containment reinerting provides assurance of an adequate onsite supply. Weekly testing of the oxygen analyzer and monthly actuation of the nitrogen makeup and purge line valves provides assurance of operational readiness. B. Standby Gas Treatment System and C. Secondary Containment Initiating reactor building isolation and operation of the standby gas treatment system to maintain the design negative pressure within the secondary containment provides an adequate test of the reactor building isolation valves and the standby gas treatment system. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system operational capability. The frequency of tests and sample analysis is necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Standby gas treatment system in-place testing procedures will be established utilizing applicable sections of ANS1 N510-1975 standard as a procedural guideline only. Operation of the standby gas treatment system every month for 10 hours will reduce the moisture buildup on the adsorbent. If painting, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample B 3/4.7-45 -I

4.7 SURVEILLANCE REQUIREMENT BASES (Cont'd.) DRESDEN II DPR-19 Amendment No. 17, 82, 107 analysis should be performed as required for operational use. Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52, Revision 1 (June 1976). The charcoal adsorber efficiency test procedures will allow for the removal of one representative sample cartridge and testing in accordance with the guidelines of Table 3 of Regulatory Guide 1.52, Revision 1 (June 1976). The sample will be at least two inches in diameter and a length equal to the thickness of the bed. If the iodine removal efficiency test results are unacceptable, all adsorbent in the system will be replaced. High efficiency particulate filters are

• i nsta 11 ed before and after the charcoal filters to prevent cloggi~g of the carbon adsorbers and to minimize potential release of particulates to the environment.

An efficiency of 99% is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by in-place testing with DOP as the testing medium. Any HEPA filters found defective will be replaced with filters qualified pursuant to regulatory guide position C.3.d of Regulatory Guide 1.52, Revision 1 (June 1976). Once per operating cycle demonstration of HEPA filter press*ure drop, operability of inlet heaters at rated power, air distribution to each HEPA filter, and automatic initiation of each standby gas treatment system subsystem is necessary to assure system performance capability.

0.

Primary Containment Isolation Valves Those large pipes comprising a portion of the reactor coolant system, whose failure could result in uncovering the reactor core, are supplied with automatic isolation valves (except those lines needed for emergency.core cooling system operation or containment cooling). The closure times specified herein are adequate to prevent loss of more coolant from the circumferential rupture of any of these lines outside the containment than from a steam line rupture. Therefore, this isolation valve closure time is sufficient to prevent uncovering the core. B 3/4.7-46

3.9 LIMITING CONDITION FOR OPERATION (Cont'd.) permissible only during the succeeding seven days unless the second line is sooner placed in service providing both the Unit 2 and Unit 2/3 emergency diesel generators are operable. From and after the date that incoming power is not available from any line, reactor operation is permissible providing both the Unit 2 and Unit 2/3 emergency diesel generators are operating and all core and containment cooling systems are operable and the NRC is notified within 24 hours of the situation, the precautions to be taken during this situation, and the plans for prompt restoration of incoming power.

2.

a.

From and after the date that one of the diesel generators and/or its associated bus is made or found to be inoperable for any reason, except as specified in Specifi-cation 3.9.B.2.b below, reactor opera-tion is permissible according to Specifi-cation 3.9.B.2.c and 3.9.D only during the succeeding seven days unless such diesel generator and/or bus is sooner made DRESDEN II DPR-19 Amendment No. 72, 82, 96, 98, 107 4.9 SURVEILLANCE REQUIREMENT (Cont'd.) 3/4.9-3

3.9 LIMITING CONDITION FOR OPERATION (Cont'd.) operable, provided that during such seven days the operable diesel generator shall be demonstrated to be operable immediately and daily thereafter and two offsite lines as specified in 3.9.A. are available.

b.

Specification 3.9.B.2.a shall not apply when a diesel generator has been made inoperable for a period not to exceed 1-1/2 hours for the purpose of conducting preventative maintenance. Additionally, preventative maintenance shall not be undertaken unless two offsite lines are available and the alt.ernate diesel generator has been demonstrated to be operable.

c.

During any period when the unit or shared diesel genera-tor is inoperable, continued reactor operation is permis-sible only during the succeeding seven days provided that all of the low pres-sure core cooling and containment cooling subsystems shall be operable. If this requirement cannot be met, an DRESDEN II DPR-19 Amendment No. 82,87,92,96,98,107

4. 9 SURVEILLANCE REQUIREMENT (Cont'd.)

3/4.9-4

3.9 LIMITING CONDITION FOR OPERATION (Cont'd.) orderly shutdown shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours.

3.

From and after the date that one of the two 125 or 250V battery systems is made or found to be inoperable, except as specified in 3.9.B.4a or b, Unit shutdown shall be. initiated within 2 hours and the unit shall be in cold shutdown in 24 hours unless the failed battery can be sooner made operable.

4.

a.

Each 125 or 250 volt battery may be inoperable for a maximum of 7 days per operating cycle for maintenance and testing.

b.

If it is determined that a battery need be replaced as a result of maintenance or testing, a specific battery may be inoperable for an additional 7 days per operating cycle. C. Diesel Fuel There shall be a minimum of 10,000 gallons of diesel fuel supply on site for each diesel. D. Diesel Generator Operability Whenever the reactor is in the Cold Shutdown or DRESDEN II DPR-19 Amendment No. 82,87,98,107 4.9 SURVEILLANCE REQUIREMENT (Cont'd.) C. Diesel Fuel Once a month the quantity of diesel fuel available shall be logged. Once a month a sample of diesel fuel shall be checked for quality. D. Diesel Generator Operability

1.

Each diesel generator shall be manually 3/4.9-5

3.9 LIMITING CONDITION FOR OPERATION (Cont 1d.) Refueling modes, a minimum of one diesel generator (either the Dresden 2 diesel generator or the Unit 2/3 diesel generator) shall be operable whenever any work is being done which has the potential for draining the vessel,. secondary containment is required, or a core or containment cooling system is required. DRESDEN II DPR-19 Amendment No. 82,87,98,107 4.9 SURVEILLANCE REQUIREMENT (Cont 1d.)

2.

3.

4.

3/4.9-6 started and loaded once each month to demonstrate operational readiness. The test shall continue until both the diesel engine and the generator are at equilibrium conditions of

• temperature whil~ full

.1 oad output is maintained. During the monthly generator test, the diesel starting air compressor shall be checked for operation and its ability to recharge air receivers. During the monthly generator test, the diesel fuel oil transfer pumps shall be operated. Additionally, during each refueling outage, a simulated loss of offsite power in conjunction with an ECCS initiation signal test shall be performed on the 4160 volt emergency bus by: (a) Verifying de-energization of the emergency buses and load shedding from the emergency buses.

3.9 LIMITING CONDITION FOR OPERATION (Cont'd.) DRESDEN II DPR-19 Amendment No. 82,87,98,107 4.9 SURVEILLANCE REQUIREMENT (Cont'd.) 3/4.9-6a (b) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency buses with permanently connected loads, energizes the auto-connected emergency ioads through the load sequencer, and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.

)- <UNIT-ED'STAT,ES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 COMMONWEALTH EDISON COMPANY DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 102 License No. DPR-25

1.

The Nuclear Regulatory Commission (the Conunission) has found that: A. The application for amendment by the Commonwealth Edison Company (the licensee) dated December 21, 1988 and supplemental information provided dated May 4, 1989 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 Conunission; 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 Conunission'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-25 is hereby amended to read as follows: B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 102, 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 to be implemented within 60 days.

Attachment:

Changes to the Technical Specifications Date of Issuance: Auqust 10, 1989 FOR THE NUCLEAR REGULATORY COMMISSION Paul C. Shemanski, Acting Director Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects

ATTACHMENT TO LICENSE AMENDMENT NO. 102 FACILITY OPERATING LICENSE DPR-25 DOCKET NO. 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. REMOVE INSERT iii iii iv iv

• vii vii 1.0-4 i.0-4 3/4.5-2 3/4.5-2 3/4.5-3 3/4.5-3 3/4.5-4 3/4.5-4 3/4.5-4a 3/4.5-5 3/4.5-5 3/4.5-6 3/4.5-6 3/4.5-7 3/4.5-7 3/4.5-7a 3/4.5-8 3/4.5-8 3/4.5-9 3/4.5-9 3/4.5-10 3/4.5-10 3/4.5-11 3/4.5-11 3/4.5-12 3/4.5-12 3/4.5-12a 3/4.5-12a 3/4.5-13 3/4.5-13 B 3/4.5-30 B 3/4.5-30 8 3/4.5-31 B 3/4.5-31 B 3/4.5-32 B 3/4.5-32 B 3/4.5-33 B 3/4.5-33 B 3/4.5-34 B 3/4.5-34 B 3/4.5-35 B 3/4.5-35 B 3/4.5-36 B 3/4.5-36 B 3/4.5-37 B 3/4.5-37 B 3/4.5-38 B 3/4.5-38 B 3/4.5-39.

B 3/4.5-39 B 3/4.5-40 B 3/4.5-40 B 3/4.5-41 B 3/4.5-41 B 3/4.5-42 B 3/4.5-42 B 3/4.5-43 8 3/4.5-43 8 3/4.5-44

AMENDMENT NO. 102 CONT'D REMOVE 3/4.7-19 3/4.7-20 3/4.7-22 3/4.7-23 3/4.7-24 3/4.7-25 3/4.7-26 3/4.7-27 B 3/4.7-38 B 3/4.7-39 B 3/4.7-40 B 3/4.7-45 B 3/4.7-46 3/4.9-3 3/4.9-4 3/4.9-5 3/4.9-6 INSERT 3/4.7-19 3/4.7-20 3/4.7-22 3/4.7-23 3/4.7-24 3/4.7-25 3/4.7-26 3/4.7-27 B 3/4.7-38 B 3/4.7-39 B 3/4.7-40 B 3/4.7-45 B 3/4.7-46 3/4.9-3 3/4.9-4 3/4.9-5 3/4.9-6 3/4.9-6a

DRESDEN III DPR~25 Amendment No. 75,78,77,87,94,102 (Table of Contents, Cont'd.) 3.5.C 3.5.D 3.5.E 3.5.F 3.5.G 3.5.H

3. 5. I 3.5.J 3.5.K 3.5.L 3.5.M 3.6 3.6.A 3.6.B 3.6.C 3.6.D 3.6.E 3.6.F 3.6.G 3.6.H 3.6.I 3.7 3.7.A 3.7.B

3. 7.C 3.7.D 3.8 3.8.A 3.8.B 3.8.C 3.8.D 3.8.E 3.8.F 3.8.G 3.8.H 3.9 3.9.A 3.9.B HPCI Subsystem Automatic Pressure Relief Subsystems Isolation Condenser System Minimum Core and Containment Cooling System Availability Deleted Maintenance of Filled Discharge Pipe Average Planar LHGR Local Steady State LHGR Local Transient LHGR Minimum Critical Power Ratio (MCPR)

Conderisate Pump Room Flood Protection Limiting Conditions for Operation Bases (3.5) Surveillance Requirement Bases (4.5) Primary System Boundary Thermal Limitations Pressurization Temperature Coolant Chemistry Coolant Leakage Safety and Relief Valves Structural Integrity Jet Pumps Recirculation Pump Flow Limitations Snubbers (Shock Suppressors) Limiting Conditions for Operation Bases (3.6) Surveillance Requirement Bases (4.6) Containment Systems Primary Containment Standby Gas Treatment System Secondary Containment Primary Containment Isolation Valves Limiting Conditions for Operation Bases (3.7) Surveillance Requirement Bases (4.7) Radioactive Effluents Gaseous Effluents Liquid Effluents Mechanical Vacuum Pump Radioactive Waste Storage Radiological Environmental Monitoring Program Solid Radioactive Waste Miscellaneous Radioactive Material Sources Miscellaneous LCO Limiting Conditions for Operation Bases (3.8) Surveillance Requirement Bases (4.8) Auxiliary Electrical Systems Requirements Availability of Electric Power iii Page 3/4.5-7 3/4.5-8 3/4.5-9 3/4.5-11 3/4.5.:.13 3/4.5-15 3/4.5-16 3/4.5-17 3/4.5-22.. 3/4.5,.24 B 3/4.5-30 B 3/4.5-39 3/4.6-1 3/4.6-1 3/4.6-2 3/4.6-3 3/4.6-5 3/4.6-6 3/4.6-7 3/4. 6-10 3/4.6-12 3/4.6-16 B 3/4.6-25 B 3/4.6-39 3/4.7-1 3/4.7-1 3/4. 7-19 3/4.7-26 3/4.7-27 B 3/4. 7-33 B 3/4.7-40 3/4.8-1 3/4.8-1 3/4.8-9 3/4.8-14 3/4.8-15 3/4.8-15 3/4.8-19 3/4.8-20 3/4.8-21 B 3/4.8-32 B 3/4.8-37 3/4.9-1 3/4.9-1 3/4.9-2

DRESDEN III DPR-25 Amendment No. 75, 77, 101, 102 (Table of Contents, Cont'd.) 3.9.C 3.9.D

3. 10 3.10.A 3.10.B 3.10. c 3.10.D 3.10. E 3.10. F 3.11 3.12 4.0 4.1 4.2 4.2.A 4.2.B 4.2.C 4.2.D 4.2.E 4.2.F 4.2.G 4.3 4.3.A 4.3.B 4.3.C Diesel Fuel Diesel Generator Operability Limiting Conditions for Operation Bases (3.9)

Surveillance Requirement Bases (4.9) Refueling Refueling Interlocks Core Monitoring Fuel Storage Pool Water Level Control Rod and Control Rod Drive Maintenance Extended Core Maintenance Spent Fuel Cask Handling Limiting Conditions for Operation Bases (3.10) Surveillance Requirement Bases (4.10) High Energy Piping Integrity (outside containment) Limiting Conditions for Operation Bases (3.11) Surveillance Requirement Bases (4.11) Fire Protection Systems - Sections 3.12.A through 3.12.H - Deleted per Generic Letters 86-10 and 88-12 (Amendment 101) SURVEILLANCE REQUIREMENTS .Reactor Protection System Protective Instrumentation Primary Containment Isolation Functions Core and Containment Cooling Systems Initiation and Control Control Rod Block Actuation Refueling Floor Radiation Monitors Post Accident Instrumentation Radioactive Liquid Effluent Instrumentation Radioactive Gaseous Effluent Instrumentation Reactivity Control Reactivity Limitations Control Rods Scram Insertion Times iv Page 3/4.9-5 3/4.9-5 B 3/4.9-7 B 3/4.9-8 3/4.10-1 3/4.10-1 3/4.10-1 3/4. 10-2 3/4.10-3 3/4.10.;.4 3/4.10-5* B 3/4.10-8 B 3/4.10-ll 3/4.ll-l B 3/4.ll-4 B 3/4.ll-4 3/4.1-1 3/4.2-1 3/4.2-1 3/4.2-1 3/4.2-2 3/4.2-2 3/4.2-3 3/4.2-4 3/4.2-5 3/4.3-1 3/4.3-1 3/4.3-4 3/4.3-10

Table 3.1.1 Table 4.1.1 Table 4.1.2 Table 3.2.1 Table 3.2.2 Table 3.2.3 Table 3.2.4 Table 3.2.5 Table 3.2.6 Table 4.2.l Table 4.2.2 Table 4.2.3 Table 4.2.4 Table 4.5.l Table 4.6.2 Table 3.7.1 Table 4.8.1 Table 4.8.2 Table 4.8.3 Table 4.8.4 Table 4.8.5 Table 4.8.6 Table 4.11-1 Table 3.12-1 Table 3.12-2 Table 3.12-3 Table 3.12-4 Table 6.1.l Table 6.6.1 DRESDEN III DPR-25 Amendment No. 78, 79, 83, 87, 92, 100, 102 List of Tables Reactor Protection System (Scram) Instrumentation Requirements Scram Instrumentation Functional Tests Scram Instrumentation Calibration Instrumentation that Initiates Primary Containment Isolation Functions Instrumentation that Initiates or Controls the Core and Containment Cooling System Instrumentation that Initiates Rod Block Radioactive Liquid Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring Instrumentation Post Accident Monitoring Instrumentation Requirements Minimum Test and Calibration Frequency for Core and Containment Cooling Systems Instrumentation, Rod Blocks, and Isolations Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Post Accident Monitoring Instrumentation Surveillance Requirements Surveillance of HPCI Subsystem Neutron Flux and Sample Withdrawal Primary Containment Isolation Radioactive Gaseous Waste Sampling and Analysis Program Maximum Permissible Concentration of Dissolved or Entrained Noble Gases Released from the Site to Unrestricted Areas in Liquid Waste Radioactive Liquid Waste Sampling and Analysis Program Radioactive Environmental Monitoring Program Reporting Levels for Radioactivity Concentrations in Environmental Samples Practical Lower Limits of Detection (LLD) for Standard Radiological Environmental Monitoring Program Surveillance Requirements for High Energy Piping Outside Containment Fire Detection Instruments Sprinkler Systems co2 Systems Fire Hose Stations Minimum Shift Manning Chart Special Reports vii Page 3/4.1-5 3/4.1-8 3/4.1-10 3/4.2-8 3/4.2-10 3/4.2-12 3/4.2-14 3/4.2-15 3/4.2-17 3/4.2-19 3/4.2-22 3/4.2-24 3/4.2-26 3/4.5-7a B 3/4.6-30 3/4.7-31 3/4.8-22 3/4.8-24 3/4.8-25 3/4.8-27 3/4.8-28 3/4.8-29 3/4.11-3 B 3/4.12-17 B 3/4.12-18 B 3/4.12-19 B 3/4.12-20 & 21 6-4 6-22

1.0 DEFINITIONS (Cont'd.) DRESDEN III DPR-25 Amendment No. 75, 87, 102 V. Reactor Power Operation - Reactor power operation is any operation with the mode switch in the 11Startup/Hot Standby 11 or 11 Run 11 position with the reactor critical and above 1% rated thermal power.

1.

Startup/Hot Standby Mode - In this mode the reactor protection scram trips, initiated by condensor low vacuum and main steamline isolation valve closure, are by-passed when reactor pressure is less than 600 psig; the low pressure main steamline isolation valve closure trip is bypassed., the reactor protection system is energized with IRM neutron-monitoring system trips and control rod withdrawal interlocks in service.

2.

Run Mode - In this mode the reactor protection system is energized with APRM protection and RBM interlocks in service. W. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detector. X. Refueling-Outage - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant subsequent to that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of. the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage. Y. Safety Limit - The safety limits are limits 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 Nuclear Regulatory Commission (NRC) before re-sumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review. Z. 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 in compliance with the provisions of Specification 3.7.B.

3.

All automatic ventilation system isolation valves are operable or are secured in the isolated position.

1. 0-4

.\\ 1.\\. 3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

2.

From and after the date that one of the core spray subsystems is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless suc_h __ subsystem is sooner made operable, provided that during such seven days all active components of the other core spray subsystem and the LPCI subsystem and the diesel generators required for operation DRESDEN III DPR-25 Amendment No. 4, 75, 93, 102 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-2 to a reactor vessel pressure of 90 psig

c.

Pump Once/month Operability

d.

Motor Operated Valve

e.

Core Spray header delta p instru-mentation: check calibrate test

f.

Logic System Functional Test

2.

Deleted Once/month Once/day Once/3 months Once/3 month$ Each Refueling Outage

DRESDEN III DPR-25 Amendment No. 4, 75, 93, 102 3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) (Cont'd.) of such components if no external source of power were available shall be operable.

3.

Except as specified in

3.

LPCI Subsystem Testing 3.5.A.4, 3.5.A.5 and shall be as specified 3.5.F.3 below, the in 4.5.A.l.a, b, c, d, LPCI subsystem shall and f, except that be operable whenever three LPCI pumps shall irradiated fuel is in deliver at least 14,500 the reactor vesseJ. .gpm against a system head corresponding to a reactor vessel pressure of 20 psig.

4.

From and after the date

4.

Deleted that one of the LPCI pumps is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding thirty days unless such pump is sooner made operable, provided I that during such thirty days the remaining active components of the LPCI and containment cooling subsystem and all active components of both core spray subsystems and the diesel generators required for operation of such components if no external source of power were available shall be operable.

5.

From and after the

5.

Deleted I date that the LPCI J subsystem is made or found to be inoperable 3/4.5-3

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.) for any reason, reactor operation is permissible only during the succeeding seven days unless it TSSooner made operable, provided that during such seven days all active components of both core spray subsystems, the containment cooling subsystem (including 2 LPCI pumps) and the diesel generators required for operation of such components if no external source of power were available shall be operable.

6.

Containment cooling spray loops are required to be operable when the reactor water temperature is greater than 212°F except that a maximum of one drywell spray loop may be inoperable for thirty days when the reactor water temperature is greater than 212°F.

7.

The doors of the core spray and LPCI pump compartments shall be closed at all times except during passage in order to consider the core spray and the LPCI subsystems operable. DRESDEN III DPR-25 Amendment No. 4, 75, 90, 97, 102 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-4

6.

During *each five year period an air test shall be performed on the drywell spray headers and nozzles.

7.

Whenever the LPCI and core spray subsystems are required to be operable, the doors of the core spray and LPCI pump compartments shall be verified to be closed weekly.

-* -1,.. 3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

8.

If the requirements of 3.5.A cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown condition within 24 hours. Sub-sequently, the reactor mode switch may be placed in Refuel in accordance with 3.5.F.3 through 3.5.F.6. DRESDEN I II DPR-25 Amendment No. 4, 75, 90, 97, 102 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-4a

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.) B. Containment Cooling Subsystem

1.

Except as specified in 3.5.B.2, 3.5.B.3, and 3.5.F.3 through 3.5.F.6 below, both containment cooling subsystem loops shall be operable when~ ever irradiated fuel is in the reactor vessel and reactor coolant temperature is greater than 212°F.

2.

From and after the date that one of the containment cooling service water subsystem pumps is made or found to be inoperable for DRESDEN III DPR-25 Amendment No. 23, 75, 97, 102 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) B. Surveillance of the Containment Cooling Subsystem shall.be performed as follows: 3/4.5-5

1.

Containment Cooling Service Water Subsystem Testing: Item

a.

Pump & Valve Operability

b.

Flow Rate Test. Each containment cooling water pump shall deliver at least 3500 gpm against a pressure of 180 psig. Frequency Once/3 months After pump maintenance and every 3 months

c.

Each manual, Every 31 days power operated or automatic valve, in the fl ow path that is not 1 oc ked, sealed or other-wise secured in its position, must be verified to be in its correct position.

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.) any reason, reactor operation is permissible only during the succeeding thirty days unless such pump is sooner made operable, provided that during such thirty days all other active components of the containment cooling subsystem are operable.

3.

From and after the date that one containment cooling subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless such subsystem is sooner made operable, provided that all active components of the other containment cooling subsystem, both core spray subsystems and both diesel generators required for operation of such components if no external source of power were available; shall be operable.

4.

If the requirements of 3.5.B cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours. DRESDEN III DPR-25 Amendment No. 23,57,65,75,93,102 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) 3/4.5-6

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.) C. HPCI Subsystem

1.

Except as specified in 3.5.C.2 below, the HPCI subsystem shall be operable whenever the reactor pressure is greater than 150 psig and irradiated fuel is in the reactor vessel.

2.

a.

From and after the date that the HPCI subsystem is made or found to be inoperable for any reason, reactor operation is permissible DRESDEN III DPR-25 Amendment No. 57,65,75,102 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) C. Surveillance of HPCI Subsystem shall be performed as indicated in Table 4.5.1. only during the succeeding seven days unless such subsystem is sooner made operable, provided that during such seven days all active components of the Automatic Pressure Relief Subsystem, the core spray subsystems, LPCI subsystem, and isolation cooling system are operable.

b.

During reactor startup when HPCI Surveillances are being performed, if the testing requirements of 4.5.C.3 or 4.5.C.4 cannot be met, continued reactor startup is not permitted. The HPCI system shall be declared inoperable and the provisions of 3.5.C.3 shall be implemented.

3.

If the requirements of 3.5.C cannot be met, an orderly shutdown shall 3/4.5-7

DRESDEN III DPR-25 Amendment No. 102 TABLE 4.5.1 SURVEILLANCE OF THE HPCI SUBSYSTEM. Item

1.

Pump Operability

2.

Motor Operated Valve

3.

Flow Rate Test-HPCI pump shall deliver at least 5000 gpm against a system head corresponding to a reactor vessel pressure of 1150 psig when steam is being supplied to the turbine at (1000 +20, -BO) psig.**

4.

Flow Rate Test-HPCI pump shall deliver at least 5000 gpm against a system head corresponding to a reactor vessel pressure of > 300 psig when steam is being-supplied to the turbine at 300 (+50, -100) psig.*

5.

Simulated Automatic Actuation Test

6.

Logic System Functional Test Frequency Every 31 days Every 31 days After pump maintenance and every 3 months Prior to exceeding 350 psi g fo 11 owing a refue 1 i ng outage or an outage during which HPCI maintenance was performed. Each refueling outage Each refueling outage Entry into the Startup/Hot Standby Mode is permitted provided that the required testing is successfully completed within 12 hours after reactor steam pressure is adequate to perform the test.

• • Entry into the Run Mode is permitted provided that the required testing is successfully completed within 12 hours after reactor steam pressure is adequate to perform the test.

3/4.5-7a

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.) be initiated and the reactor pressure shall be reduced to 150 psig within 24 hours. D. Automatic Pressure Relief Subsystems 1.. Except as specified in 3.5.D.2 and 3.5.D.3 below, the Automatic Pressure Relief Subsystem shall be operable whenever the reactor pressure is greater than 150 psig and irradiated fuel is in the reactor vesse 1.

2.

From and after the date that one of the five relief valves of the automatic pressure relief subsystem is made or found to be inoperable, reactor DRESDEN III DPR-25 Amendment No. 28,40,63,73,94,102 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-8 D. Surveillance of the Automatic Pressure Relief Subsystem shall be performed as follows: l~

• During each operating*

cycle the following shall be performed:

a.

A simulated automatic initiation which opens all pilot valves, and

b.

With the reactor at pressure each relief valve shall be manually opened. Relief valve opening sha 11 be verified by a compensating turbine bypass valve or control valve closure.

c.

A logic system functional test shall be performed each refueling outage.

• 3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

operation is permis-sible only during the succeeding seven (7) days provided that during such time the HPCI subsystem is operable. If the follow-ing MAPLHGR reduction factors (multipliers) are applied to Figure 3.5-1, the Automatic Pressure Relief Subsystem of ECCS shall be considered operable: (1) 0.89 for 8x8 fuel, or (2) 0.76 for 9x9 fuel.

3.

From and after the date that two relief valves are found or made to be inoperable, reactor operation is permissible only during the succeed-ing seven days provided that during such time the HPCI subsystem is operable and the multipliers speci-fied in 3.5.D.2 are applied.

4.

If the requirements of 3.5.D.l cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to below 150 psig within 24 hours. E. Isolation Condenser System DRESDEN III DPR-25 Amendment No. 19,28,40,63,75,94,102 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-9 E. Surveillance of the Isolation Condenser System shall be performed as follows:

3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

1.

Whenever the reactor pressure is greater than 150 psig and irradiated fuel is in the reactor vessel, the isolation condenser shall be operable except as specified in 3.5.E.2.

2.

From and after the date that the isolation condenser system is made or found to be inoperable for any reason, reactor ope rat ion is permissible only during the succeeding seven days unless such system is sooner made operable, provided that during such seven days all active components of the HPCI subsystem are operable. DRESDEN III DPR-25 Amendment No. 19,28,75,102 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) 3/4. 5-10

1.

Isolation Condenser System Testing:

a.

The shel 1 side water level and temperature shall be checked daily.

b.

Simulated automatic actuation and functional* system testing shall be performed during each refueling outage or whenever major repairs are completed on the system.

c.

The system heat removal capability shall be determined once every five years.

d.

Calibrate vent line radiation monitors quarterly.

r-3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)

3.

If the requirements of 3.5.E cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to 150 psig within 24 hours. F. Minimum Core and Containment Cooling System Availability

1.

Any combination of inoperable components in the core and containment cooling systems shall not defeat the capability of the remaining operable components to fulfill the core and containment cooling functions. DRESDEN III DPR-25 Amendment No. 18, 28, 75, 102 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) 3/4. 5-11 F. Surveillance of Core and Containment Cooling System

1.

Actions necessary to assure that the plant can be safely shut down and maintained in this condition in case of failure of the Dresden Dam shall be demonstrated to be adequate every third refueling outage. If this Specification has been complied with for Dresden Unit 2, it shall not be required for Dresden Unit 3. --{ I

DRESDEN III DPR-25 Amendment No. 18, 28, 75, 97, 102 3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.) 4.5. SURVEILLANCE REQUIREMENT (Cont 1d.)

2.

When irradiated fuel is in the reactor vessel and the reactor is in cold shutdown or refuel condi-tions, at least two of the following pumps, with each having an operable flow path capable of taking suction from the suppres-sion pool or the condensate storage tank and.trans-ferring the water to the reactor vessel, shall be operable except as speci-fied in 3.5.F.4, 3.5.F.5 and 3.5.F.6 below:

a.

Two Core Spray pumps or,

b.

Two Low Pressure Coolant Injection pumps or,

c.

One Core Spray pump and one Low Pressure Coolant Injection pump.

3.

With one of the pumps and/or associated flow paths required by 3.5.F.3 inoperable, restore at least two pumps and associ-ated flow paths to operable status within 4 hours or suspend all operations with a potential for draining the reactor vessel.

4.

With both of the pumps and/or associated flow paths required by 3.5.F.3 inoperable, suspend core alterations and all operations with potential for draining the reactor vessel. Restore at least one pump and associated flow path to oper-able status within 4 hours or establish secondary contain-ment integrity within the next 8 hours. 3/4.5-12

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.)

5.

When irradiated fuel is in the reactor vessel and the reactor is in the cold shut-down or refuel condition, all low pressure core and containment cooling sub-systems may be inoperable provided the reactor vessel head is removed, the cavity is _flooded, the spent fuel pool gates are re~oved, the fuel pool water level is m~intained above the low level alarm point, and the reactor cavity water temperature is below 140°F.

6.

When irradiated fuel is in the reactor vessel and the reactor is in the refuel condition, the torus may be drained completely and control rod drive maintenance performed provided that the spent fuel pool gates are open, the fuel pool water level is maintained above the low level alarm point, and the minimum total condensate storage reserve is maintained at 230,000 gallons, and provided that not more than one control rod drive housing is open at one time, the control rod drive housing is blanked following removal of the control rod drive, no work is being performed in the reactor vessel while the housing is open and a special flange is DRESDEN III DPR-25 Amendment No. 97, 102 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-12a

3.5 LIMITING CONDITION FOR OPERATION (Cont 1d.) available which can be used to blank an open housing in the event of a leak.

7.

When irradiated fuel is in the reactor and the vessel head is removed, work that has the poten-tial for draining the . vessel may be performed wi§h less than 112,000 ft of water in the suppression pool, pro-vided that:

1) the total volume of water in the suppression pool, dryer separator above the shield blocks, refuel-ing cavity, and the fuel storage pool above the bottom of th~ fuel pool gate is gr~ater than 112,000 ft ; 2) the fuel storage pool gate is removed; 3) the low pres-sure coolant injection and core spray systems are operable as speci-fied in 3.5.F.3, 3.5.F.4 and 3.5.F.5; and 4) the automatic mode of the drywell sump pumps is disabled.

G. Not Used H. Maintenance of Filled Discharge Pipe Whenever core spray, LPCI, or HPCI ECCS are required to be operable, the discharge piping from the pump discharge DRESDEN III DPR-25 Amendment No. 18, 75, 97, 102 4.5 SURVEILLANCE REQUIREMENT (Cont 1d.) 3/4.5-13 G. Not Used H. Maintenance of Filled Discharge Pipe The following surveil-lance requirements shall be adhered to, to assure that the discharge piping of the core spray,

I ~ _\\~ DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES A. Core seray and LPCI Mode of the RHR System - This specification assures that adequate emergency cooling capability is available. Based on the loss of coolant analyses included in References (1) and (2) in accordance with 10 CFR 50.46 and Appendix K, core cooling systems provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident, to limit the calculated peak clad temperature to less than 2200°F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to les5 than 1%, and to limit the calculated local metal-water reaction to less than 17%. The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in Reference (3). Using the results developed in this reference, the repair period is found to be.less than 1/2 the test interval. This assumes that the core spray and LPCI subsystems constitute a 1 out of 3 system, however, the combined effect of the two systems to limit excessive clad temperatures must also be considered. The test interval specified in Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the basic risk considering single failures should be less than 45 days and this specification is within this period. For multiple failures, a shorter interval is specified. (1) "Loss of Coolant Accident Analyses Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations," NED0-24146A, Revision 1, April 1979. (2) NED0-20566, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K. (3) APED-11Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards" - April 1969, I.M. Jacobs and P.W. Marriott. B 3/4.5-30

• 3.5 DRESDEN III DPR-25 Amendment No. 75, 87, 93, 94, 102 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

Although it is recognized that the information given in reference 3 provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach

• prevents complete acceptance of this method at this time.

Therefore, the times stated in the specific items were established with due 1 regard to judgment. Should one core spray subsystem become inoperable, the remaining core spray and the entire LPCI system are available should t.he -t reactor core cooling arise. Based on judgments of the reli-ability of the remaining systems; i.e.,.the core spray and LPCI, a 7-day repair period was obtained. Should the loss of one LPCI pump occur, a nearly full complement of core and containment cooling equipment is available. Three LPCI pumps in conjunction with the core spray subsystem will perform the core cooling function. Because of the availability of the majority of the core cooling equipment a 30-day repair period is justified. If the LPCI subsystem is not available, at least 2 LPCI pumps must be available to fulfill the containment cooling function. The 7-day repair period is set on this basis. When a core spray or LPCI subsystem (or pump) is inoperable, there is a requirement that other emergency core cooling systems be operable. The verification of operability, as used in this context, for these other systems means to administratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. B. Containment Coolin Service Water - The containment heat removal portion of the L CI containment cooling subsystem is provided to remove heat energy from the containment in the event of a loss of coolant accident. For the flow specified, the containment long-term pressure is limited to less than 8 psig and, there-fore, is more than ample to provide the required heat removal capability. (Ref. Section 5.2.3.2 SAR).* The containment cooling subsystem consists of two sets of cooling equipment. Each set contains*2 service water pumps, 1 heat exchanger and 2 LPCI pumps. Either set of equipment is capable of performing the containment cooling function. Loss of one containment cooling service water pump or one LPCI pump does not seriously jeopardize the containment cooling capability as any 2 of the remaining three pumps can satisfy the cooling B 3/4.5-31

' l DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont 1d.) C. requirements. Since there is some redundancy left a 30-day repair period is adequate. Loss of 1 containment cooling subsystem leaves one remaining system to perform the contain-ment cooling function. Based on the facts that when one containment cooling subsystem becomes inoperable only one system remains, a 7-day repair period was specified. When a containment cooling subsystem (or pump) is inoperable, there is a requirement that other emergency core cooling systems be operable. The verification of operability, as used in this context, for these other systems means to administratively check by examining l-0gs or other information to determine if certain components/systems are out~of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. Hi~h Pressure Coolant Injection - The high pressure coolant inJection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI and core spray subsystems can protect the core. The HPCI meets this requirement without the use of off-site electrical power. For the pipe breaks for which the HPCI is -intended to function, the core never uncovers and is continuously cooled and thus no clad damage occurs. (Ref. Section 6.2.5.3 SAR). The repair times for the limiting conditions of operation were set considering the use of the HPCI as part of the isola-tion cooling system. When the HPCI subsystem is inope~able, there is a requirement that other emergency core cooling systems be operable. The verification of operability, as used in this context, for these other systems means to administratively check by examining logs or other information to determine if certain components/ systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. D. Automatic Pressure Relief - The relief valves of the automatic pressure relief subsystem are a back-up to the HPCI subsystem. They enable the core spray and LPCI to provide protection against the small pipe break in the event of HPCI failure, by depressuriztng the reactor vessel rapidly enough to actuate the core sprays and LPCI. The core spray and LPCI provide sufficient flow of coolant to adequately cool the core. B 3/4.5-32

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) E. Analyses have shown that only four of the five valves in the Automatic Depressurization System are required to operate. Loss of one of the relief valves does not significantly affect the pressure-relieving capability, therefore continued operation is acceptable provided the appropriate MAPLHGR reduction factor is . applied to assure compliance with the 2200°F PCT limit. Loss of more than one relief valve significantly reduces the pressure relief capability of the ADS; thus; a 7-day repair period is specified with the HPCI available, and a 24-hour repair period otherwise. When one or more relief valves is inoperable, there is a require~ ment that the HPCI subsystem be operable. The verification of operability, as used in this context, for HPCI means to admin-istratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. Isolation Coolina Srstem - The turbine main condenser is normally available. The 1so ation condenser is provided for core decay heat removal following reactor isolation and scram. The isola-tion condenser has a heat removal capacity sufficient to handle the decay heat production at 300 seconds following a scram. Water will be lost from the reactor vessel through the relief valves in the 300 seconds following isolation and scram. This represents a minor loss relative to the vessel inventory. The system may be manually initiated at any time. The system is automatically initiated on high reactor pressure in excess of 1060 psig su*stained for 15 seconds. The time delay is provided to prevent unnecessary actuation of the system during anticipated turbine trips. Automatic initiation is provided to minimize the coolant loss following isolation from the main condenser. To be considered operable the shell side of the isolation condenser must contain at least 11,300. gallons of water. Make-up water to the shell side of the isolation condenser is provided by the condensate transfer pumps from the condensate storage tank. The condensate transfer pumps are operable from on-site power. The fire protection system is also available as make-up water. An alternate method of cooling the core upon isolation from the main condenser is by using the relief valves and HPCI subsystem in a feed and bleed manner. Therefore, the high pressure relief function and the HPCI must be available together to cope with an anticipated transient so the LCO for HPCI and relief valves is set upon this function rather than their function as depressuriza-tion means for a small pipe break. B 3/4.5-33

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont 1d.) When the isolation condenser system is inoperable, there is a requirement that the HPCI subsystem be operable. The verification of operability, as used in this context, for HPCI means to administratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. F. Emergency Coolin~ Availability - The purpose of Specification D is to assure a minimum of core cooling equipment is available at all times.. If, for e.xample, one core spray were out of. service* and the diesel which powered the opposite core spray were out of service, only 2 LPCI pumps would be available. Likewise, if 2 LPCI pumps were out of service and 2 containment service water pumps on the opposite side were also out of service no contain-ment cooling would be available. It is during refueling outages that major maintenance is performed and during such time that all low pressure core cooling systems may be out of service. This specification provides that should this occur, no work will be performed on the primary system which could -lead to draining the vessel. This work would include work on certain control rod drive components and recirculation system. Thus, the specifica-tion precludes the events which could require core cooling. Specification 3.9 must also be consulted to determine other requirements for the diesel generators. When the unit or shared diesel generator is inoperable, there is a requirement that other emergency core cooling systems be operable. The verification of operability, as used in this context, for these other systems means to administratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system. Dresden Units 2 and 3 share certain process systems such as the makeup demineralizers and the radwaste system and also some safety systems such as the standby gas treatment system, batteries, and diesel generators. All of these systems have been sized to perform their intended function considering the simultaneous operation of both units. For the safety related shared features of each plant, the Technical Specifications for that unit contain the operability and surveillance requirements for the shared feature; thus, the B 3/4.5-34

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) level of operability for one unit is maintained independently of the status of the other. For example, the shared diesel (2/3 diesel) would be mentioned in the specifications for both Units 2 and 3 and even if Unit 3 were in the Cold Shutdown Condition and needed no diesel power, readiness of the 2/3 diesel would be required for continuing Unit 2 operation. Specification 3.5.F.4 provides that should this occur, no work will be performed which could preclude adequate emergency cooling capability being available. Work is prohibited unless it is i-n accordance with specified procedures which limit the period that the control rod drive housing is open and assures that the worst possible loss of coolant resulting from the work will not result in uncovering the reactor core. Thus, this specification assures adequate core cooling. Specification 3.9 must be consulted to determine other requirements for the diesel generator. Specification 3.5.F.5 provides assurance that an adequate supply of coolant water is immediately available to. the low pressure core cooling systems and that the core will remain covered in the event of a loss of coolant accident while the reactor is depressurized with the head removed. G. Not Used H. Maintenance of Filled Discharge Pipe - If the discharge piping of the core spray, LPCI, and HPCI are not filled, a water hammer can develop in this piping when the pump and/or pumps are started. I. Average Planar LHGR This specification assures that the peak cladding temperature following a postulated design basis loss-of-coolant accident will not exceed the 2200°F limit specified in 10 CFR 50 Appendix K considering the postulated affects of fuel pellet densification. The peak cladding temperature following a postulated loss-of-coolant accid~nt is primarily a function of the average LHGR of all the rods in a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within a fuel assembly. Since expected local variations in power dis-tribution within a fuel assembly affect the calculated peak clad temperature by less than plus or minus 20°F relative to the peak temperature for a typical fuel design, the limit on the average B 3/4.5-35 J

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont 1d.) planar LHGR is sufficient to assure that calculated temperatures are below the 10 CFR 50, Appendix K limit. The maximum average planar LHGRs shown in Figure 3.5.1 are based on calculations employing the models described in Refer-ences (1), (2) and (3). Power operation with APLHGRs at or below those shown in Figure 3.5.1 assures that the peak cladding temper-ature following a postulated loss-of-coolant accident will not exceed the 2200°F limit. ANF has analyzed the effects Single Loop Operation has on LOCA events (Reference 4). For breaks in the.idle loop, the above Dual Loop Operation results are conservative (Reference 1). For breaks in the active loop~ the event is more severe primarily due to a more rapid loss of core flow. By applying a multiplica-tive 0.91 reduction factor to the results of the previous analyses, all applicable criteria are met. J. Local Steady State LHGR This specification assures that the maximum linear heat genera-tion rate in any fuel rod is less than the design linear heat generation rate even if fuel pellet densification is postulated. This provides assurance that the fuel end-of-life steady state criteria are met. (1) 11 loss of Coolant Accident Analyses Report for Dresden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations, 11 NED0-24146A, Revision 1, April 1979. (2) XN-NF-81-75 11Dresden Unit 3 LOCA Model Using the ENC EXEM Evaluation Model MAPLHGR Results 11 (3) XN-NF-85-63 11Dresden Unit 3 LOCA-ECCS Analysis MAPLHGR results for 9x9 fuel 11 , dated Septebmer 1985. (4) ANF-84-111, 11 LOCA-ECCS Analysis for Dresden Units During Single Loop Operation with ANF Fuel, 11 September 1987. B 3/4.5-36

~- DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) K. Lo ca 1 Transient LHGR This specification provides assurance that the fuel will neither experience centerline melt nor exceed 1% plastic cladding strain for transient overpower events beginning at any power and termi-nating at 120% of rated thermal power. L. Minimum Critical Power Ratio (MCPR) The steady-state values for MCPR specified in the Specification were determined using the THERMEX thermal limits methodology described in XN-NF-80-19, Volume 3. The safety limit implicit in the Operating limits is established so that during sustained operation at the MCPR safety limit, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition. The Limiting Transient delta CPR implicit in the operating limits was calculated such that the occurrence of the limiting transient from the operating limit will not result in violation of the MCPR safety limit in at least 95% of the random statistical combina-tions of uncertainties. Transient events of each type anticipated during operation of a BWR/3 were evaluated to determine which is most restrictive in terms of thermal margin requirements. The generator load rejection/turbine trip without bypass is typically the limiting event. The thermal margin effects of the event are evaluated with the THERMEX Methodology and appropriate MCPR limits con-sistent with the XN-3 critical power correlation are determined. Several factors influence which transient results in the largest reduction in critical power ratio, such as the cycle-specific fuel loading, exposure and fuel type. The current cycle's reload licensing analyses identifies the limiting transient for that cycle. As described in ~pecification 4.3.C.3 and the associated Bases, observed plant data were used to determine the average scram performance used in the transient analyses for determin-ing the MCPR Operating Limit. If the current cycle scram time performance falls outside of the distribution assumed in the analyses, an adjustment of the MCPR limit may be required to maintain margin to the MCPR Safety Limit during transients. Compliance with the assumed distribution and adjustment of the MCPR Operating Limit will be performed as directed by the nuclear fuel vendor in accordance with station procedures. B 3/4.5-37

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont 1d.) For core flows less than rated, the MCPR Operating Limit established in the specification is adjusted to provide protection of the MCPR Safety Limit in the event of an uncontrolled recirculation flow increase to the physical limit of pump flow. This protection is provided for manual and automatic flow control by choosing the MCPR operating limit as the value from Figure 3.5-2 Sheet 1 or the rated core flow value, whichever is greater. For Automatic Flow Control, in addition to protecting the MCPR Safety Limit during the flow run-up event, protection is provided against violating the rated flow MCPR Operating Limit during an automatic flow increase to rated core flow. Thi~ protectiun is provided by the reduced.flow MCPR limits shown in Figure 3.5-2 Sheet 2 or 3 where the curve corresponding to the current rated flow MCPR limit is used (linear interpola-tion between the MCPR limit lines depicted is permissible). Therefore, for Automatic Flow Control, the MCPR Operating Limit is chosen as the value from Figure 3.5-2 Sheet 1, Sheet 2, Sheet 3 or the rated flow value, whichever is greatest. Analyses have demonstrated that transient events in Single Loop Operation are bounded by those at rated conditions; however, due to the increase in the MCPR fuel cladding integrity safety limit in Single Loop Operation, an equiva-lent adder must be uniformly applied to all MCPR LCO to maintain the same margins to the MCPR fuel cladding integrity safety limit. M. Flood Protection Condensate pump room flood protection will assure the availability of the containment cooling service water system (CCSW) during a postulated incident of flooding in the turbine building. The redundant level switches in the condenser pit will preclude any postulated flooding of the turbine building to an elevation above river water level. The level switches provide alarm and circulating water pump trip in the event a water level is detected in th~ condenser pit. B 3/4.5-38

4.5 SURVEILLANCE REQUIREMENT BASES (A thru F) DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 The testing interval for the core and containment cooling systems is based on 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 d~well pressure were stimulated, 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 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., instru-mentation, pumps, valve operators, etc.' are tested more fre-quently. The instrumentation is functionally tested each month. Likewise the pumps and motor-operated valves are also tested each month to assure their operability. The combination of a yearly simulated automatic actuation test and monthly tests of the pumps and valve operators is deemed to be adequate testing of these systems. With components or subsystems out-of-service overall core and containment cooling reliability is maintained by verifying the operability of the remaining cooling equipment. The verification of operability, as used in this context, for these other systems means to administratively check by examining logs or other information to determine if certain components/systems are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the operability of the component/system.

However, if a failure, design deficiency, etc., caused 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 components.

For example, if an out-of-service period were caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test. The requirement of 180 psig at 3500 gpm at the containment cooling service water (CCSW) pump discharge provides adequate margin to ensure that the LPCI/CCSW system provides the design B 3/4.5-39

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.) bases cooling water flow and maintains 20 psig differential pressure at the containment cooling heat exchanger. This differential pressure precludes reactor coolant from entering the river water side of the containment *cooling heat exchangers. The verification of 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 valve actuation mechanism 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 relief valve opening. This test method may be performed over a wide range of reactor pressure greater than 150 psig. Valve operation below 150 psig is limited by the spring tension exhibited by the relief valves. G. Deleted H. Maintenance of Filled Discharge Pipe The surveillance requirements to assure that the discharge piping of the core spray, LPCI, and HPCI systems are filled provides for a visual observation that water flows from a high point vent. This ensures that the line is in a full con-dition. Between the monthly intervals at which the lines are vented, instrumentation has been provided to monitor the presence of water in the discharge piping. This instrumenta-tion will be calibrated on the same frequency as the safety system instrumentation. This period of periodic testing ensures that during the intervals between the monthly checks the status of the discharge piping is monitored on a con-tinuous basis. I. Average Planar LHGR At core thermal power levels less than or equal to 25 percent, operating plant experience and thermal hydraulic analyses indi-cate that the resulting average planar LHGR is below the maxi-mum average planar LHGR by a considerable margin; therefore, B 3/4.5-40

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 4.5 SURVEILLANCE REQUIREMENT BASES (Cont 1d.) evaluation of the average planar LHGR below this power level is not necessary. The daily requirement for calculating average planar LHGR above 25 percent rated thermal power is sufficient since power distribution shifts are slow when there have not been significant power or control rod changes. J. Local Steady State LHGR The LHGR for all fuel shall be checked daily during reactor operation at greater than or equal to 25 percent power to determine if fuel burnup or control rod movement has caused changes in power distribution.. A limiting LHGR value is precluded by a considerable margin when employing a permis- . sible control rod pattern below 25% rated thermal power. K. Lo ca 1 Transient LHGR The fuel design limiting ratio for centerline melt (FDLRC) 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 changes in power distribution. The FDLRC limit is designed to protect against centerline melting of the fuel during anticipated operational occurrences. L. Minimum Critical Power Ratio (MCPR) At core thermal power levels less than or equal to 25 percent, the reactor will be operating at minimum recirculation pump speed and the moderator void content wi 11 be very sma 11. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicates that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow.when there have not been significant power or control ro*d changes. In addition, the reduced flow correction applied to the LCD provides margin for flow increase from low flows. M. F 1 ood Protection The watertight bulkhead door and the penetration seals for pipes and_ cables penetrating the vault walls have been designed B 3/4.5-41

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.) to withstand the maximum flood conditions. To assure that their installation is adequate for maximum flood conditions, a method of testing each seal has been devised. To test a pipe seal, another test seal is installed in the opposite side of the penetration creating a space between the two seals that can be pressurized. Compressed air is then supplied to a fitting on the test seal and the space inside the sleeve is pressurized to approximately 15 psi. The outer face of the permanent seal is then tested for leaks using a soap bubble solution~ On completion of the test, the test seal is removed for use on other pipes and penetrations of the same size. In order to test the watertight bulkhead doors, a test frame must be installed around each door. At the time of the test, a reinf~rced steel box with rubber gasketing is clamped to the wall around the door. The fixture is then pressurized to approximately 15 psig to test for leak tightness. Floor drainage of each vault is accomplished through a carbon steel pipe which penetrates the vault. When open, this pipe will drain the vault floor to a floor drain sump in the condensate pump room. Equipment drainage from the vault coolers and the CCSW pump bedplates will also be routed to the vault floor drains. The old equipment drain pipes will be permanently capped to pre-clude the possibility of back-flooding the vault. As a means of preventing backflow from outside the vaults in the event of a flood, a check valve and an air operated valve .are installed in the 2 11 vault floor drain line 6 10 11 above the floor. of the condensate pump room. The check valve is a 2 11 swing check designed for 125 psig service. The air operated valve is a control valve designed for a 50 psi differential pressure. The control valve will be in the normally open position in the energized condition and will close upon any one of the following:

a.

Loss of air or power

b.

High level (5 10 11 ) in the condensate pump room B 3/4.5-42

DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.) Closure of the air operated valve on high water level in the condensate pump room is effected by use of a level switch set at a water 1eve1 of 5 1 0 11 Upon actuation, the switch wi 11 close the control valve and alarm in the control room. The operator will also be aware of problems in the vaults/ condensate pump room if the high level alarm on the equipment drain sump is not terminated in a reasonable amount of time. It must be pointed out that these alarms provide information to the operator but that operator action upon the above alarms is not a necessity for reactor safety since the other provi-sions provide adequate protection. A system of level switches has been installed in the condenser pit to indicate and control flooding of the condenser area. The following switches are installed: Level Function

a.

l 1 011 (1 switch) Alarm, Panel Hi-Water-Condenser Pit

b.

3 10 11 (1 switch) Alarm, Panel High-Circ. Water Condenser Pit

c.

5 10 11 (2 redundant Alarm and Circ. Water Pump switch pairs) Trip 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 a probable circulating water failure. Should the switches at level (a) and (b) fail or the operator fail 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 switch pairs at level (c) are designed and installed to IEEE-279, "Criteria for Nuclear Power Plant Protection Systems. 11 As the circulating water pumps are tripped, either manually or automatically, at level (c) of 5 10 11, the maximum water level reached in the condenser pit due to pumping wi 11 be at the 491 1 0 11 e 1 evat ion (10 1 above condenser pit floor elevation 481 10 11 ; 5 1 plus an additional 5 1 attributed to pump coastdown). B 3/4.5-43

• 4.5 DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 SURVEILLANCE REQUIREMENT BASES (Cont'd.)

In order to prevent overheating of the CCSW pump motors, a vault cooler is supplied for each pump. Each vault cooler is designed to maintain the vault at a maximum 105°F temperature during operation of its respective pump. For example, if CCSW pump 28-1501 starts, its cooler will also start and compensate for the heat supplied to the vault by the 28 pump motor keeping the vault at less than 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. In this way, the vault coolers are supplied with cooling water totally inside the vault. The cooling water quantity needed for each cooler is approximately 1% to 5% of the design flow of the pumps so that the recirculation of this small amount of heated water will not affect pump or cooler operation. Operation of the fans and coolers is required during pump operability testing 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. B 3/4.5-44

\\-. ' I-3.7 LIMITING CONDITION FOR OPERATION (Cont 1d.) shall be initiated and the reactor shall be in a cold shutdown condition in the following 24 hours. B. Standby Gas Treatment System

1.

lwQseparate and independent standby gas treatment system subsystems shall be operable at all times when secondary containment integrity is required, except as specified in sections 3.7.B.l(a) and (b). 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, provi~ that all active components in the other standby gas treatment subsystem shall be operable. 3/4.7-19 'DRESDEN 111 DPR-25 Amendment No. 75, 77, 90, 102 4.7 SURVEILLANCE REQUIREMENTS (Cont 1d.) B. Standby Gas Treatment System

1.

At least once per month, initiate from .the control room 4000 cfm (plus or mi nus 10%) fl ow through each subsystem I of the standby gas treatment system for at least 10 hours with the subsystem heaters operating at rated power.

3.7 LIMITING CONDITION FOR OPERATION (Cont 1d.) Within 36 hours 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.l.(a) through (d).

b.

If both standby gas treatment system subsystems are not operable, within . 36 hours the reactor shall be placed in a condition for which the standby gas treatment system is not required in accordan~e with Specification 3.7.C.l.(a) through (d).

2.

Performance Requirement

a.

Periodic Requirements: 3/4. 7-20 DRESDEN Ill DPR-25 Amendment No. 15, 75, 102 4.7 SURVEILLANCE REQUIREMENTS (Cont'd.)

2.

Performance Requirement Tests

a.

At least once per 720 hours of system operation; or once per operating cycle, or every 18 months, whichever occurs first; or following painting, fire, or chemical}}