ML20058N871
| ML20058N871 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/16/1993 |
| From: | Hopkins J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058N873 | List: |
| References | |
| NUDOCS 9312220314 | |
| Download: ML20058N871 (10) | |
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UNITED STATES ig<g[e,i 1
NUCLEAR REGULATORY COMMISSION
- j WASHINGTON. D.C. 20555 4001 TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY L
THE CLEVELAND ELECTRIC ILLUMINATIN(, COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.182 License No. NPF-3 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Toledo Edison Company, Centerior Service Company, and the Cleveland Electric Illuminating Company (the licensees) dated November 9,1992, as supplemented on November 22,_
1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:
9312220314 931216 PDR ADDCK 05000346" P
PDR.
(a) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.182, are hereby incorporated in the license..
The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance.
l FOR THE NUCLEAR REGULATORY COMMISSION i
I
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Jon B. Hopkins, Sr. Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical 1
Specifications Date of issuance:
December 16, 1993 t
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9 t-
e Mr. Louis F. Storz Davis-Besse Nuclear Power Station Toledo Edison Company Unit No. I cc:
Mary E. O'Reilly Robert E. Owen, Chief Centerior Energy Corporation Bureau of Radiological Health 300 Madison Avenue Services Toledo, Ohio 43652 Ohio Department of Health Post Office Box-118 Mr. William T. O'Connor, Jr.
Columbus, Ohio 43266-0118 Manager - Regulatory Affairs Toledo Edison Company Attorney General Davis-Besse Nuclear Power Station Department of Attorney General 5501 North State - Route 2 30 East Broad Street Oak Harbor, Ohio 43449 Columbus, Ohio 43216 Gerald Charnoff, Esq.
Mr. James W. Harris, Director Shaw, Pittman, Potts Division of Power Generation and Trowbridge Ohio Department of Industrial i
2300 N Street, N. W.
Regulations Washington, D. C. 20037 P. O. Box 825 Columbus, Ohio 43216 Regional Administrator, Region III U. S. Nuclear Regulatory Commission Ohio Environmental Protection Agency 801 Warrenville Road DERR--Compliance Unit F
Lisle, Illinois 60532-4351 ATTN:
Zack A. Clayton P. O. Box 1049 Mr. Robert B. Borsum Columbus, Ohio 43266-0149 Babcock & Wilcox Nuclear Power Generation Division State of Ohio 1700 Rockville Pike, Suite 525 Public Utilities Commission Rockville, Maryland 20852 180 East Broad Street Columbus, Ohio 43266-0573 Resident Inspector U. S. Nuclear Regulatory Commission Mr. James R. Williams 5503 N. State Route 2 State Liaison to the NRC Oak Harbor, Ohio 43449 Adjutant General's Department Office of Emergency Management Mr. Donald C. Shelton Agency Senior Vice President - Nuclear 2825 West Granville Road Centerior Service Company Columbus, Ohio 43235-2712 6200 Oak Tree Boulevard Independence, Ohio 44101 Mr. John K. Wood, Plant Manager Toledo Edison Company
-Davis-Besse Nuclear Power Station 5501 North State Route 2 Oak Harbor, Ohio 43449 r
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s ATTACHMENT TO LICENSE AMENDMENT NO. 1R7 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications w' the attached pages. The revised pages are identified by amendment number e..
contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
Remove Insert 3/4 5-3 3/4 5-3 3/4 5-4 3/4 5-4 B 3/4 5-2 B 3/4-5-2 B 3/4 6-3 B 3/4 6-3 t
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EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T.;; Jt280'F l
1 LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:
a.
One OPERABLE high pressure injection (HPI) pump, b.
One OPERABLE low pressure injection (LPI) puup, c.
One OPERABLE decay heat cooler, and d.
An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and manually transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
a.
With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next !? hours,
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b.
In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.
SURVE!LtANf! RE001REMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
At least once per 31 days by verifying that each valve (manugl.
a.
power operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position. is in its correct position.
DAVIS-BESSE, UNIT 1 3/4 5-3 Amendment No. 86,182
SURVEILLANCE REQUIREMENTS (Continued) 6.
At least once per 18 months, or prior to operation after ECCS piping has been drained by verifying that the ECCS piping is full of water by venting the ECCS pump casings and discharge piping high points.
c.
By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment snergency sump and cause restriction of the pump suction during LOCA conditions. This visual inspection shall be performed:
1.
For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY and 2.
Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.
d.
At least once per 18 months by:
1.
Verifying that the interlocks:
a) Close DH-11 and DH-12 and deenergize the pressurizer heaters, if either DH-11 or DH-12 is open and a simulated reactor coolant system pressure which is greater than the trip setpcint (<438 psig) is applied. The interlock to close DH-11 and/or DH-12 is not required if the valve is closed and 460 V AC power is' disconnected from its motor operators.
b) Prevent the opening of DH-ll and DH-12 when a simulated or actual reactor coolant system pressure which is greater than thetripsetpoint(<438psig)isapplied.
2.
a) A visual inspection of the containment emergency sump which verifies that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.
b) Verifying that on a Borated Water Storage Tank (BWST) Low-Low Level interlock trip, with the motor operators for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the BWST Outlet Valve HV-DH7A (HV-DH78) automatically close in 175 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (HV-DH98) which should be verified toopenin$J5 seconds.
3.
Verifying a total leak rate < 20 gallons per hour for the LpI system at:
a) Normal operating pressure or hydrostatic test pressure of
>150 psig for those parts of the system downstream of the pump suction isolation valve, and b) >45 psig for the piping from the containment emergency sump Tsolation valve to the pump suction isolation valve.
DAVIS-BESSE, UNIT 1 3/4 5-4 Amendment No. 3,25.ZE f 0,77/@5,182
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
BASES 3/4.5.1 CORE FLOODING TANKS The OPERABILITY of each core flooding tank ensures that a sufficient volume of borated water will be immediately forced into the reactor vessel
' in the event the RCS pressure falls below the pressure of the tanks.
This initial surge of water into the vessel provides the initial cooling mechanism during large pCS pipe ruptures.
The limits on volume, boron concentration and pressure ensure that the assumptions used for core flooding tank injection in the safety analysis are met.
The tank power operated isolation valves are considered to be
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.
In addition, as these tank isolation valves fail to meet single failure criteria, removal of power to the valves is required.
The limits for operation with a core flooding tank inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional tank which may result in unacceptable peak cladding tempera-tures.
If a closed isolation valve cannot be immediately opened, the full capability of one tank is not available and prompt action is required to place the reactor in a mode where this capability is not required.
3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems with RCS average temperature > 280*F ensures that sufficient emergency core cooling capability wT11-be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the core flooding tanks is capable of supplying sufficient core cooling to maintain the peak cladding tempera-tures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.
In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.
DAVIS-BESSE, UNIT 1 B 3/4 5-1 Aniendment No. 20
L EMERGENCY CORE COOLING SYSTEMS BASES With the RCS temperature below 280*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.
The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. The decay heat removal system leak rate surveillance requirements assure that the leakage rates assumed for the system during the recirculation phase of the low pressure injection will not be exceeded.
Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necesse'v to: (1) prevent total pump flow from exceeding runout conditions when the.ptem is in its minimum resistance configuration (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOM analyses.
Containment Emergency Sump Recirculation Valves DH-9A and DH-9B are de-energized during MODES 1, 2, 3 and 4 to precibde postulated inadvertent opening of the valves in the event of a Control Room fire, which could result in draining the Borated Water Storage Tank to the Containment Emergency Sump and the loss of this water source for normal plant shutdown. Re-energization of DH-9A and DH-9B is permitted on an intermittent basis during MODES 1, 2, 3 and 4 under administrative controls. Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.
Borated Water Storage Tank (BWST) outlet isolation valves DH-7A and DH-78 are de-energized during MODES 1, 2, 3, and 4 to preclude postulated inadvertent closure,f the valves in the event of a fire, which could result in a loss of the availability of the BWST.
Re-energization of valves DH-7A and DH-78 is permitted on an intermittent basis during H0 DES 1, 2, 3, and 4 under adminis-trative controls. Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.
3/4.5.4 BORATED WATER STORAGE TANK The OPERABILITY of the borated water storage tank (BWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on BWST minimum volume and boron concentration ensure that 1) sufficient water is available within contain-ment to permit recirculation cooling flow to the core, and 2) the reactor will remain suberitical in the cold condition following mixing of the BWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analysis.
The bottom 4 inches of the borated water storage tank are not available, and the instrumentation is calibrated to reflect the available volume. The limits on water volume, and boron concentration ensure a pH value of between 7.0 and 11.0 of the solution sprayed within the containment after a design basis accident.
The pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion cracking on mechanical systems and components.
DAVIS-BESSE, UNIT 1 B 3/4 S-2 Amendment No. 2D,723,182-
CONTAINMENT SYSTEMS BASES leakage rate are consistent with the assumptions used in the safety analyses.
The leak rate surveillance requirements assure that the leakage assumed for the system during the recirculation phase will not be exceeded.
Borated Water Storage Tank (BWST) outlet isolation valves DH-7A and DH-7B are ide-energized during MODES 1, 2, 3, and 4 to preclude postulated inadvertent
- closure of the valves in the event of a fire, which could result in a loss of lthe availability of the BWST. Re-energization of valves DH-7A and DH-7B is
' permitted on an intermittent basis during MODES 1, 2, 3 and 4 under admints-ltaken when re-energizing these valves under such controls. Station proce trative controls.
+
Containment Emergency Sump Recirculation Valves DH-9A and DH-9B are de-ener-
.gized during MODES 1, 2, 3, and 4 to preclude postulated inadvertent o)ening of the valves in the event of a fire, which could result in draining t1e Borated Water Storage Tank to the Containment Emergency Sump and the loss of this water source for normal plant shutdown. Re-energization of valves DH-9A and DH-9B is permitted on an intermittent basis during MODES 1, 2, 3, and 4 under administrative controls. Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.
3/4.6.2.2 CONTAINMENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature will be maintair,ed within limits during normal operation, and 2) adequate heat removal capacity is available when operated in conjunction with the containment spray systems during post-LOCA conditions.
3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.
Containment isolation within the required time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
Containment isolation valves and their required isolation times are addressed in the USAR. The opening of a closed inoperable containment isolation valve on an intermittent basis during plant operation is permitted under administrative control. Operating procedures identify those valves which may be opened under administrative control as well as the safety precautions which must be taken when opening valves under such controls, o
DAVIS-BESSE, UNIT 1 B 3/4 6-3 Amendment No. DE, f(7.182
i CONTAINMENT SYSTEMS BASES 3/4.6.4 COM3USTIBLE GAS CONTROL The OPEPABILITY of the Hydrogen Analyzers, Containment Hydrogen Dilution System, and Hydrogen Purge System ensures that this equipment will be available to raaintain the maximum hydrogen concentration within -
the containment vessel at or below three volume percent following a LOCA.
The two redundant Hydrogen Analyzers determine the content of hydro-i gen within the containment vessel. The Hydrogen Analyzers, although they have their OPERABILITY requirements in this Specification, are considered part of the post-accident monitoring instrumentation of Specification 3/4.3.3.6, Post-Accident Monitoring Instrumentation.
The Containment Hydrogen Dilution (CHD) System consists of two full capacity, redundant, rotary, positive displacement type blowers to supply air to the containment. The CHD System controls the hydrogen concentra-tion by the addition of air to the containment vessel, resulting in a pressurization of the containment and suppression of the hydrogen volume fraction.
l The Containment Hydrogen Purge System Filter Unit functions as a backup to the CHD System and is designed to release air from the con-tainment atmosphere through a HEPA filter and charcoal filter prior to discharge to the station vent.
i 3/4.6.5 SHIELD BUILDING 3/4.6.5.1 EMERGENCY VENTILATION SYSTEM The OPERABILITY of the emergency ventilation systems ensures that containment vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is neces-sary to meet the assumptions used in the safety analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions.
i t
1 DAVIS-BESSE, UNIT 1 B 3/4 6-4 Amendment No. 46,167