ML20091K982
| ML20091K982 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/17/1992 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20091K985 | List: |
| References | |
| NUDOCS 9201270185 | |
| Download: ML20091K982 (13) | |
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o UNITED STATES
!" 3 m ~,h NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555 c
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BALTIPORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS HUCLEAR POWER PLAtlT l' NIT NO.1 AMENDPENT TO FACILITY OPERATIMG LICENSE Amendment tio.166 License No. DPR-53 L
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Baltimore Gas and Electric Company (the licensee) dated July ?,1991, as supplemented November 15, 1991, complies with the standards and requirements of the Atomic Energy Act of 195a, as amended (the Act) and the Comission's rules end regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confortr.ity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendmert can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendrent will not be inimical to the common defense and security or to the health and safety of the public;
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and E.
The issurnce of this amendment is in accordance with 10 CFR Part 51 of tb Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPP,-53 is hereby amended to read as follows:
9201270185 920117 PDR ADOCK 05000317 P
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.c (2)--:TechnicalSpecifications The-Technical-Specifications contained in Appendices A and B, as revised-through Amendnent No.166, are hereby incorporated in the 11cc.se. The licensee shall operate-the facility in accordance with the Technical Specifications.
Y 3.
This license atendment is effective as of the date of its issuance and shall be implemented when the spent fuel cask handling crane modifications are complete prior to July 31, 1992.
FOR THE NUCLEAR REGULATORY COMMISSION ga.s a. Cp Robert A. Capra, Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to-the" Technical Specifications l
Date of Issuance:
January 17, 1992 L
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'o UNITED STATES NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D, C,20555 a
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sb BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 146 License No -DPR-69 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Baltimore Gas and Electric Company (the licensee) dated July 2, 1991, as supplemented November 15, 1991, complies with the standards and requirements of the Atomic Energy Act of-1954, as amended (the Act) and the Comission's
- rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There'is reasonab'le 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the-public; and-E.
The issuance of. this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements
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-have been satisfied.
l 2.
Accordingly. the-license is amended by changes to the Technical-L Specifications as indicated in the attachment to this license l
atendment, and paragraph 2.C.2 of Facility Operating License No.
DPR-69 is hereby amended to read as follows:-
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2-(2) Tec}nicalSpecifications
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.The Technical Specifications contained'in Appendices
" A and B, as revised' through Amendment No.- 146. are hereby incorporated in the_ license. The licensee shall operate the facility in accordance with the Technical Specifications..-
.3.-
This license amendment is effective as of the date of its issuance and shall be-implemented when the spent fuel cask handling crane modifications are complete ~ prior.to July 31, 1992.
F0!! THE NUCLEAR REGULATORY CONNISSION.
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Robert A. Capra, Director Project Directorate I_1-Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attach' ment:
Changes to the Technical Specifications Date of-Issuance: -January 17, 1992 i
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t ATTAC HME.NT T.O.L..IC. ENS..E.A. !..!EN.D.M..EN.TS.
AMENDMENT NO.,166 FACIL1TY_0PERATING, LICENSE,NO,DPR-53 AMENDMENT NO.146-FACIL11Y OPERATING LICENSE NO. DPR-69 DOCKET NOS. 50-317 AND 50-318 Revise Appendix A as follows:
ggmove_Paaes Insert _Pages 3/4 9-7 3/4 9-7 3/4 9-16 3/4 9-16 B3/4 9-2 B3/4 9-2 B3/4 9-3 B3/4 9-3 r
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REFUELING OPERATIONS
[RANE TRAVEL - SPENT FUEL STORAGE POOL _ BUILDING LIMITING CQND1110R FOR OPERATION 3.9.7 Loads in excess of 1600 pounds shall be prohibited from travel over fuel assemblies in the storage pool unless such loads are handled by the single-failure-proof Spent fuel Cask Handling Crane, i
APPLICABillly: With fuel assemblies in the storage pool.
ACTION:
With the requirements of the above specification not satisfied, place the crane loao in a safe condition.
The provisions of Specification 3.0.3 are not applicable.
SEPYEILLANff_RE0VIREMENTS 4.9.7.1 The weight of each load, other than a fuel asse..bly and CEA, shall be verified to be 51600 pounds prior to moving it over fuel assemblies unless such loads are handled by the single-failure-proof Spent Fuel Cask Handling Crane.
4.9.7.2 Slings and special lifting devices shall be visually inspected and verified operable within 7 days prior to and at least once per 7 days thereafter during Spent fuel Cask Handling Crane operation over the spent fuel storage pool.
4.9.7.3 In addition to the requirements of Section 4.9.7.2, pre-operational and periodic tests and preventive maintenance sha'll be performed per plant procedures.
CALVERT CLIFFS - UNIT 1 3/4 9-7 Amendment No. 166
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., p IN CALVERT CLIFFS - UNIT 1 3/4 9-16 Amendment No. Jff, !66.
i BEFUELING OPERATIONS BMES
}LLS.6 REFUELING MACHINE OPERABILITY The OPERABILITY requirements for the refueling machine ensure that:
(1) the refueling machine will be used for movement of CEAs and fuel assemblies, (2) the refueling machine has sufficient load capacity to lift a CEA or fuel assembly, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL - SP.DjJ,.10EL STORAGE BUILDitiQ lhe restriction on movement of loads in excess of the nominal weight of a fuel assembly and CEA over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited 'o that c.ontained in a single fuel assembly, and (2) any possible distorth n of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses. The Spent Fuel Cask Handling Crane, which has a critical load capacity of 125/15 ton, meets the
" single failure-proof" criteria of NUREG-0554 and NUREU-0612.
3/4.9.8 COOLANT CIRCULATION The requirement that at least one shutdown cooling loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel 0
below 140 F as rt fired during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident.and prevent boron stratification.
The requirement to have two shutdown cooling loops OPERABLE when there is less than 23 feet of water above the core ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling, thus in the event of a failure of thb operating shutdown cooling loop, adequate time is provided to initiate emergency procedures to cool the core.
In MODE 6, shutdown cooling flow must provide sufficient heat removal to match. core decay heat generation rates and maintain the core exit temperature within the MODE limit. Thus, as decay heat production is reduced with time, shutdown cooling flow may be proportionally reduced.
Pursuant to NRC Generic Letter 88-17, flow reduction is necessary for operations near the mid point of the hot leg piping to prevent vortex formation at the shutdown cooling suction nozzle.
Prevention of vortex formation reduces the potential for a loss of shutdown cooling due to air binding of the low pressure safety injection (LPSI) pump (s) operating to provide shutdown cooling flow.
In accordance with the recommendations of NRC Bulletin 88-04, " Safety Related Pump Loss," a minimum flow rate CALVERT CLIFFS - UNIT 1 B 3/4 9-2 Amendment No. JU)f#,166
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'A'i REFUELING OPERATIQtlS BASES requirement;of 1500 gpm is imposed. This protects the vendor recommended minimum continuous duty flow rate of 1340 gpm for-the LPSI pumps. - The 1500 gpm minimum flow rate is also more than adequate to-preclude a-boron dilution event in MODE 6 operation.and in no way restricts the ability to increase flow as necessary to remove decay heat.
3/4.9.9 CONTAINMENT' PURGE VALVE 1.5.0LATIQN SYSTEM The OPERABILITY of this system ensures that 1 4 contalnuant. purge valves will:be automatically isolated upon. detection of high radiation levels within the containment.
The-OPERABILITY of this system is
. required to restrict the release of radioactive material from the containment atmosphere to the_ environment.
7 3/4.9.10 and 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND SPENT-FQFL POOL WATER LEVEL-
.The restrictions on minimum wa'ter level ensure that sufficient water-depth is available to remove 99% of the assumed 10% iodine gap activity released from_the rupture of an. irradiated fuel assembly.
The minimum
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. water d(pth is consistent with-the assumptions of the accident analysis, e
3/4.9.12 SPENT FUEL p00L VENTILATION SYSTEM
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Thu limitations on the spent fuel pool ventilation system c.=re that all ~ radioactive material released from an irradiated fuel assembly will-:be filtered through_ the HEPA~ filters and charcoal adsorber prior:to
. discharge to the atmosphere. -The OPERABILITY of:this: system and the resulting iodine removal capacity aro consistent with the_ assumptions-of-
-the accident 1 analyses.
'3/4.9.14-CONTAINMENT VENT-ISOLATION VALVES:
..The OPERABILITY 1and closure-restrictions;on the containment vent isolation valves are sufficient to restrict radioactive material' release from-a' fuel element rupture based uaon the-lack of containment:
pressurization ' potential while-in t1e REFUELING MDDE.
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Lh CALVERT CLIFFS - UNIT 1.
B 3/4 9-3 Amendment No. EE/JEE/JEE,166 u
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3 REFUEllNG OPERATIONS CRANE TRAVEL - SPENT FUEL STORACE P0OL BQllJllRG LIMITING CONDITION FOR OPERATION Loads in excess of .1 pounds shall be prohibited from travel i
3.9.7 over fuel assemblies in tc; storage pool unless such loads are handled by the single-failure-proof Spent Fuel Cask Handling Crane.
APPLICABILITY: With fuel assemblies in the storage pool, ACILQH:
With the requirements of the above specification not satisfied, place the crane load in a safe condition.
The provisions of Specification 3.0.3 are not applicable.
SMRVEILLANCE RE0VIREMENTS 4.9.7.1 The weight of each load, other than a fuel assembly and CEA, shall be verified to be s 1600 pounds prior to moving it over fuel assemblies unless such loads are handled by the single-failure-proof Spent Fuel Cask Handling Crane.
4.9.7.2 Slings and special lifting devices shall be visually inspected and verified operable within 7 days prior to and at least once per 7 days thereafter during Spent Fuel Cask Handling Crane operation over the spent fuel storage pool.
4.9.7.3 In addition to the requirements of Section 4.9.7.2, pre-operational and periodic tests and preventive maintenance shall be performed per plant procedures.
CALVERT CLIFFS - UNIT 2 3/4 9-7 Amendment No.146
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R(IVilnNG OPERATIONS BASES 3/4.9.6 REFVELING MACHINE OPERABILITY The OPERABILITY requirements for the refueling machine ensure that:
(1) the refueling machine will be used for movement of CEAs and fuel assemblies, (2) the refueling machine has sufficient load capacity to lift a CEA or fuel assembly, and (3) the core internals ano pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL - SPENT FVEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly and CEA over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.
This assumption is consistent with the activity release assumed in the accident analyses. The Spent Fuel Cask Handling Crane, which has a critical load capacity of 125/15 ton, meets the
' single-failure-proof" criteria of NUREG-0554 and NUREG-0612.
3/4.9.8 -COOLANT CIRCVLATION The requirement that at least one shutdown cooling loop be in operation ensures that (1) sufficient cooling capacity is avrilable to remove decay heat and maintain the water in the reactor pret are vessel 0
below 140 F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.
The requirement to have two shutdown cooling loops OPERABLE when there is less than 23 feet of water above the core ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of Lcay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling, thus in the event of a failure of thb operating shutdown cooling loop, adequate time is provided to initiate emergency procedures to cool the core.
In MODE 6, shutdown cooling flow must provide sufficient heat removal to match core decay heat generation rates and maintain the core exit temperature within the MODE limit.
Thus, as decay heat production is reduced with time, shutdown cooling flow may be proportionally reduced.
Pursuant to NRC Gene. c Letter 88-17, flow reduction is necessary for operations near the mi1-point of the hot leg piping to prevent vortex formation at the shutdown cooling suction nozzle.
Prevention af vortex l
formation reduces the potential for a loss of shutdown cooling due to air binding of the low pressure safety injection (LPSI) pump (s) operating to provide shutdown cooling flow.
In accordance with the recommendations of NRC Bulletin 88-04, " Safety Related Pump Loss," a minimum flow rate CAVLERT CLIFFS - UNIT 2 B 3/4 9-2 Amendment No. JE/ # #, 146 4
RELVIL1HLDELIM11035 1
BASES m_
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requirement of 1500 99m is imposed. 'ihis protects the vendor recommended minimum continuous duty flow rate of 1340 gpm for the LPSI pumps.
The 1500 gpm minimum flow rate is also more than adequate to pred ude a boron dilution event in MODE 6 operation and in no way restricts the ability to l'
increase flow as necessary to remove decay heat.
1/4.9.9 CORIAllMNT pVRELyALyLJ10LE101L1151W q
1ho 0PERABIL11Y of this system ensures that the containment purge valves will be automatically isolated upon detection of high radiation levels within the containmer.t.
1he OPERABill1Y of this system is required to restrict the release of radioactive material from the conta bment atmosphere to the environment.
)].AJL10_And 3/4.9.11_.)lAlifL11YEL-REAC10R VESSIL AND SPENT FULLEQQL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.
The minimum water depth is consistet with the assumptions of the accident analysis.
3/4.9.12 SNHf FUEL POOL VENillATION SYSl B The limitations on the spent fuel pool ventilation syFlem ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the H PA filters and charconi adsorber prior to discharge to the atmosphere.
The OPERABIL11Y or this system and the resulting iodine removal capacity are consistent udh the assumptions of the accident analyses.
3/4.9.14 CONTAINMENLVIRLISDLATION VALVES 1h4 OPERABILITY and closure restrictions on the containment vent isolation valves are sufficient to restrict radioactive material release from a fuci element rupture based uaon the lack of containment pressurization potential while in tie REIUELING MODE.
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CALVERT CL,
- UNIT 2 B 3/4 9-3 Amendment No. El#UJ/9 146 l
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