ML20151H671

Amends 82 & 65 to Licenses DPR-53 & DPR-69,respectively, Increasing Steam Generator Min Pressurization Temp, Correcting Typographical Error & Limiting Overtime for Personnel Involved in safety-related Activities

ML20151H671
Person / Time
Site:	Calvert Cliffs
Issue date:	04/21/1983
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20151H674	List: ML20151H671 ML20151H681 ML20151H684
References
NUDOCS 8305040789
Download: ML20151H671 (34)

Text

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UNITED STATES Docket File

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NUCLEAR REGULATORY COMMISSION ORB #3 Rdg k

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• s s. *,o Docket No.50-317 and 318 Docketing and Service Section Office of the Secretary of the Commission r

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SUBJECT:

BALTIMORE GAS & ELECTRIC COMPANY,CCalvert Cliffs Units Nos.1&2 l

Two signed originals of the Federal Register Notice identified below are enclosed for your transmittal to the Office of the Federal Register for publication. Additional conformed copies ( 12 ) of the Notice are enclosed for your use.

O Noticrf of Receipt of Application for Construction Permit (s) and Operating License (s).

O Notice of Receipt of Partial Application for Construction Permit (s) and Facility License (s): Time for Submission of Views on Antitrust Matters.

O Notice of Availability of Applicant's Environmental Report.

O Notice of Proposed issuance of Amendment to Facility Operating License.

O Notice of Receipt of Application for Facility License (s); Nohce of Availability of Applicant's Environmental Report; and Notice of Consideration of issuance of Facility License (s) and Notice of Opportunity for Hearing.

O Notice of Availability of NRC Draft / Final Environmental Statement.

O Notice of Limited Work Authorization.

O Notice of Availability of Safety Evaluation Report.

O Notice of issuance of Construction Permit (s).

O Notice of issuance of Facility Operating License (s) or Amendment (s).

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,t g, og and 65

@ Other:.

bferenceddoen= ants _havebeen provided PDR.

OkdeN0cfe$rNeg0rhegulation c s

Enclosure:

As Stated omee-.- _0RB# :D.._._

summe-.- _ Pt r utzer.:cd.

DATE--.-,4

[83 NRC FOM 102 7 - 79 8305040789 830421 PDR ADOCK 05000317 P

PDR

P,altimore Gas and, Electric Company cc:

James A. Biddison, Jr.

Ms. Mary Harrison, President General Counsel Calvert County Board of County Commissioners Baltimore Gas and Electric Company Prince Frederick, MD 20768 P. O. Box 1475 Baltimore, MD 21203 U. S. Environmental Protection Agency Region III Office Grorge F. Trowbridge, Esquire Attn:

Regional Radiation Representative Shaw, Pittman, Potts and Trowbridge Curtis Building (Sixth Floor) 1800 M Street, N. W.

Sixth and Walnut Streets Washington, D. C.

20036 Philadelphia, PA 19106 Mr. R. C.' L. Olson, Principal Engineer Mr. Ralph E. Architzel Nuclear Licensing Analysis Unit Resident Reactor Inspector Baltimore Gas and Electric Company NRC Inspection and Enforcement Room 922 - G&E Building P. O. Bos 437 P. O. Box 1475 Lusby, MD 20657 Baltimore, MD 21203 Mr. Charles B. Brinkman Mr. Leon B. Russell Manager - Washington Nuclear Operations Plant Superintend.ent Combustion Engineering, Inc.

'Calvert Cliffs Nuclear Power Plant 7910 Woodmont Avenue Maryland Routes 2 & 4 Bethesda, MD 20814 Lusby, MD 20657 Mr. J. A. Tiernan, Manager Bechtel Power Corporation Nuclear Power Department Attn: Mr. J. C. Ventura Calvert Cliffs Nuclear Power Plant Calvert Cliffs Project Engineer Maryland Routes 2 & 4 15740 Shady Grove Road Lusby, MD 20657 Gaithersburg, MD 20760 Mr. W. J. Lippold, Supervisor Combustion Engineering, Inc.

Nuclear Fuel Management Attn: Mr. R. R. Mills, Manager Baltimore Gas and Electric Company

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Engineering Services Calvert Cliffs Nuclear Power Plant P. O. Box 500 P. O. Box 1475 Windsor, CT 06095 Baltimore, Maryland 21203 Mr. R. E. Denton, General Supervisor Training & lechnical Services Calvert Cliffs Nuclear Power Plant Maryland Routes 2 & 4 Lusby, MD 20657 Mr. R. M. Douglass, Manager Quality Assurance Department Baltimore Gas & Electric Company Administrator, Power Plant Siting Program Fort Smallwood Road Complex Energy and Coastal Zone Administrat. ion P. 0. Box 1475 Department of Natural Resources Baltimore, MD 21203 Tawes State Office Building Mr. S. M. Davis, General Supervisor Operations Quality Assurance Calvert Cliffs Nuclear Pcwer Plant Regional Administrator,

Maryland Routes 2 & 4 Nuclear Regulatory Commission, Region I Lusby, MD 20657 Office of Executive Director for Operations 631 Park Avenue King of Prussia, Pennsylvania 19406

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UjYlTED STATES

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. NUCLEAR REGULATORY COMMIESION 4

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BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT.NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 82 License No. DPR-53 1.

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

A.

The application for amendment by Baltimore Gas & Electric Company (the licensee) dated February 24, 1983, 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 (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission' regulations; D.

The issuance of this amesdment will not be inimical to the

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l 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 t

Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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2.

Accordingly, the license is arended by changes to the Tcchnical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License

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No. DPR-53 is hereby amended to read as follows:

(2) Technical Specifications

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The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 82, 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.

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FOR THE NUCLEAR REGULATORY COMMISSION Rober

. Clark, Chief

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Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: April 21,1983

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ATTACHMENT TO LICE l1SE AMEf4DMEllT fl0. 82 FACILITY OPERATIf;G LICENSE NO. DPR-53 DOCKET NO. 50-317

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Replace the following pages of the Appendix A Technical Specifications-with the enclosed pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change. Corresponding overleaf pages are also provided to maintain document completeness.

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Pages 3/4 3-41 3/4 3-42 3/4 7-13 3/4 7-22 B 3/4 4-1 B 3/4 7-3 6-11 6-13 e

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TABLE 3.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION 9

E MINIMUM l

CHANNELS

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INSTRUMENT OPERABLE 1.

Power Range Nuclear Flux 2

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e 2.

Containment Pressure 2

3.

Wide Range Logarithmic Neutron Flux Monitor 2

4.

Reactor Coolant Outlet Temperature 2

5.

Reactor Coolant Total Flow 2

6.

Pressurizer Pres'sure 2

7.

Pressurizer Level 2

S 8.

Steam Generator ' Pressure 2/ steam generator 9.

Steam Generator Lt. vel 2/ steam generator 10.

Feedwater Flow 2

F 11.

Auxiliary Feedwater Flow Rate 1/ steam generator i

12.

RCS Subcooled Margin Monitor 1

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13.

PORV/ Safety Valve Acoustic Flow Monitoring 1/ valve O'

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14.

PORY Solenoid Power Indication

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TABLE 4.3-10

_OST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS P

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A CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION nCq 1.

Power Range Nuclear Flux M

Q v,

2.

Containment Pressure M

R E

l Q

3.

Wide Range Logarithmic Neutron Flux Monitor M

N.A.

a 4.

Reactor Coolant Outlet Temperature M

R 5.

Reactor Coolant Total Flow M

R g

6.

Pressurizer Pressure M

R R

Y 7.

Pressurizer Level M

A 8.

Steam Generator Pressure M

R 9.

Steam Generator Level M

R 3.

10.

Feedwater Flow M

R I

11.

Auxiliary Feedwater Flow Rate M

R k

12.

RCS Subcocled Margin Monitor M

R E

M 13.

PORV/ Safety Valve Acoustic Monitor N.A.

R 5

14.

PORV Solenoid Power Indication N.A.

N.A.

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J_nNT SYSTEMS 3/4.7.

STEAM GENERATOF RESSURE/ TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be > 80 F when the pressure of either coolant l

in the steam generator is > 200 psig.

APPLICABILITY: At all times.,

ACTION:

With the requirements of the above specification not satisfied:

Reduce the steam generator pressure of the applicable side to a.

< 200 psig within 30 minutes, and b.

Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator. Determina that the steam generator remains acceptable for continued operation prior to increasing its i

temperatures above 200*F.

. SURVEILLANCE REQUIREMENTS 4.7.2.1 The pressure in each side of the steam generators shall be determined to be < 200 psig at least once per hour when the temperature of either the primary or secondary coolant < 80 F.

l CALVERT CLIFFS-UNIT 1 3/4 7-13 Amendment No. 82

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM L

LIMITING. CONDITION FOR-0PERATION 3.7.3.1 At least two component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one component cooling water loop OPERABLE, restore at least two loops 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 at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE. REQUIREMENTS i

4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed or otherwise secured in position, is in its correct position.

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b.

At least once per 18 m'onths during shutdown, by verifying that l

each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection Actuation test signal.

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l CALVERT CLIFFS-UNIT 1 3/4 7-14

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PLANT SYSTEMS 3/a.7.7-ECCS PUMP ROOM EXHAUST AIR FILTRATION SYSTEM LIMITING CONDITION FOR 0PERATION 3.7.7.1 The ECCS pump room exhaust ventilation system shall be OPERABLE i

with one HEPA filter and charcoal adsorber train and two exhaust fans.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a.

With one ECCS pump room exhaust fan inoperable, restore the inoperable fan to OPERABLE status within 7 days or be in'at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the ECCS exhaust filter train inoperable, restore the filter train to OPERABLE' status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.7.1 The ECCS pump room exhaust ventilation system shall be demon-strated OPERABLE:

a.

At least once per 31 days by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that each exhaust fan operates for at least 15 minutes.

b.

At least once per 18 ' months or (1) after any structural main-

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tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti-lation zone communicating with the system by:

CALVERT CLIFFS - UNIT 1 3/4 7-21 y

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FLANT SYSTEMS i

SURVEILLANCE REOUIREMENTS (Continued) 1.

Verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the filter train at a flow rate of 3000 cfm 1,10%.

2.

Verifying that the HEPA filter banks remove > 99% of the DOP when they*are tested in-place in accordance with ANSI N510-1975 while operating the filter train at a flow rate of 3000 cfm + 10%.

3.

Verifying within 31 days after removal that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal effi-ciency of > 90% for radioactive methyl iodide when the sample' is tested in accordance with ANSI N510-1975 (130 C, 95% R.H.).

The carbon samples not obtained from test canisters shall be prepared by either:

a)

Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining sampics at least two inches in diameter and with a j

length equal to the thickness of the bed, or

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Emptying a longitudinal sample from an adscrber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a

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length equal to the thickness of the bed.

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Verifying a system flow rate of 3000 cfm + 10% during system operation when tested in accordance with ANSI i

N510-1975.

c.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by either:

1.

Verifying within 31 days after removal that a laboratory analysis of a carbon sample obtained from a test canister demonstrates a removal efficiency of > 90% for radioactive methyl iodide when the sample is tested in accordance l

with ANSI N510-1975 (130 C, 95% R. H.); or l

CALVERT CLIFFS-UNIT 1 3/4 7-22 Amendment No. 82 l

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RE'CTOR"COOLA"~ SYSTE" A

'l BASES l'3/4.4.i COOLANT LOOPS AliD COOLAlC CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.195 during all normal operations and anticipated transients.

A single reactor coolant loop with its steam generator filled above the icw level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure considerations require plant shutdown if component.r,epairs and/or corrective actions cannot be reade within the allowable out-of-service time.

In MODES 4 and 5, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two shutdown cooling loops to be OPERABLE.

The operat' ion of one Reactor Co'olant Pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump during MODES 4 and 5 with one or more RCS cold legs < 275 F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients, and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restrict-ing starting of the RCPs to wheg the gecondary water temperature of each steam generator is less than 46 F (34 F when measured by a surface contact instrument) above the coolant temperature in the reactor vessel.

3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurizedaboveitsSafetyLigitof2750 psia.

Each safety valve is designed to relieve approximately 3 x 10 lbs per hour of saturated steam at the valve l

setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating shutdown cooling. loop, connected to the RCS, provides overpressure relief capability and will prevent RCS over-pressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.

The combined relief capacity of these valves is sufficient to CALVERT CLIFFS - UNIT 1 AmendmentNo.34,E3,pp,82 B 3/4 4-1

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' REACTOR COOLANT SYSTEM BASES f

limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating a.t RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint- (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of thz ASME Boiler and Pressure Vessel Code, j

3/4.4.3 RELIEF VALVES i

The power operated relief valves (PORVs) operate to relieve.RCS' pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for. both the relief valves and the block valve: is capable of being supplied from an 4

emergency power source to ensure the ability to seal this possible RCS leak-age path.

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer with the levgl as programmed ensurei that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The operating band for pressurizer level bounds the programmed level and ensures that RCS pressure remains within the bounds of an analyzed condition during the excessive charging event as well as during the limiting depressurization event, Excess Load. The operating band also protects the pressurizer code safety valves and power operated relief valve against water relief. The power operated relief valves function to relieve RCS pressure during all design transients. Operation of the power operated relief f

valve in conjunction with a reactor trip on a Pressurizer--Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

The requirement that 150 kw of pressurizer heaters and their associated i

controls be capable of being supplied electrical power from an emergency bus 4

provides assurance that these heaters can be energized during a loss of off-site power condition to maintain natural circulation at HOT STANDBY.

i 3/4.4.5 STEAll GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will i

be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to CALVERT CLIFFS - UNIT 1 B 3/4 4-2 Amendment No. 34, E3,AO,82'

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3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with steam discharge to atmosphere with concurrent and total loss of offsite power. The contained water volume limit includes an allow-ance for water not usable because of tank discharge line location or other physical characteristics.

,3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the result-ant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the accident analyses.

3/4.7.1.5 MAIN STEAri LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 SECONDARY WATER CHEMISTRY The secondary water chemistry program is designed to provide maximum protec-tion to both the steam generator and secondary system internals.

The most damag-ing chemical reactants enter the system via condenser cooling water ingress.

Accumulation of these impurities in the steam generators may lead to loss of metallurgical integrity and/or eventual component failure. The limits presented in _ Table 3.7-3 are those prescribed by the NSSS supplier as " limited-operation" chemistry parameters and are consistent with the most recent industry standards.

By routine monitoring of these parameters, plant personnel are able to rapidly detect and limit the duration of ingress of chemically detrimental species and thereby maintain steam generator tube integrity.

3L4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allow-able fracture toughness stress limits.

The limitations of 80 F and 200 psig are based on steam generator secondary side limitations and are sufficient to prevent brittle fracture.

CALVERT CLIFFS - UNIT 1 B 3/4 7-3 Amendment No. 35, pp.

82

_ ANT SYSTEMS 3ASES 3/4.7.3 CCT ONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses'.

3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service water system ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.5 SALT WATER SYSTEM The OPERABILITY of the salt water system ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumenta-tion cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its l

equivalent. This limitation is consistent with the requirements of General Design Criteria 10 of Appendix "A",10 CFR 50.

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j 3/4.7.7 ECCS~ PUMP ROOM EXHAUST AIR FILTRATION SYSTEM I

The OPERABILITY of the ECCS pump room exhaust air filtration system ensures that radioactive materials leaking from the ECCS equipment within the pump room following a LOCA are filtered prior to reaching the 1

CALVERT CLIFFS-UNIT 1 B 3/4 7-4 l

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.A;":,:: :: :l C ONTP.: _.

'6.7 IATETv LII'IT VIOLAT10" 6.7.1 Tne following actions shall be taken in the event a Safety Limit is, violated:

The facility shall be placed in at least HOT STANDBY within one a.

hour.

b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Manager - Nuclear Power Department and the OSSRC shall be notified within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,s.

c.

A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the POSRC. This report shall describe (1). applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d..The Safety Limit Violation Report shall be submitted to the Commission, the OSSRC and.the Manager - Nuclear Power Department withii.14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referencad below:

The applicable procedures recommended in Appendix "A" of a.

Regulatory Guide 1.33, Ravision 2, February 1978.

b.

Refueling operations.

c.

Surveillance and test activities of safety related equipnent.

d.

Security Plan implementation.

e.

Emergency Plan implementation.

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f.

Fire Protection Program implemer.tation.

The amount of overtime worked by plant staff members performing g.

safety related functions must be limited in accordance with the NRC Policy Statement on working hcurs (Generic Letter No. 82-12).

l 6.8.2 Each procedure and adninistrative policy of 6.8.1 above and changes l

thereto shall be reviewed by the POSRC and approved by the Plant Superin-l tendent prior to implementation and reviewed periodically as set forth in administrative procedures.

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CALVERT CLIFFS - UNIT 1 6-13 Amendment No.26, 63, 75,82 l

ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made pro-vided:

The intent of the originial procedure 'is not altered.

a.

b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on th,e unit affected.

c.

The change is documented, reviewed by the POSRC and approved by the Plant Superintendent within 14 days of implementation.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10 Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel tha~t has 'a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perfor-mance of the plant.

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6.9.1.2 The startuo report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating cnnditions or characteristics obtained during the test program and a comparison of these values with design credictions and l

specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

CALVERT CLIFFS - UNIT 1 6-14 AmendmentNo.[/,11,82

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g!INISTR;,TI"ECONTROLS AUDITS 6.5.2.8.1 Audits of facility activities shall be performed under.the cognizance l of the OSSRC. These audits shall encompass:

s The conformance of facility operation to drovisions.tontained within a.

the Technical Specifications and applicable license conditions at least once per 12 months.

The performance, training and qualificatkon of the entire facility b.

staff at least oitce.per 12 months.

c.

The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.

d.

The performance of activities required by the Quality Assurance k

Program to meet the criteria of Appendix "B",10 CFR 50, at least once per 24 months.

I e.

Deleted f.

The Safeguards Contingency Plan and implementing procedures at le'st a

once per 12 months in accordance with 10 CFR 73.40(d).

g.

Any other area of facility operation considered ar-opriate by the OSSRC or the Vice President-Supply.

h.

The Facility Fire Protection Program and implementing procedures at least once per 24' months.

i.

An independent fire protection and loss prc.vention program inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire s

protection firm.

~j.

An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.

6.5.2.8.2 P.eview of facility activities shall be performed under the cognizance of the OSSRC.

These reviews shall encompass:

a.

The Facility Emergency Plan and implementing procedures at least once per 12 months in accordance with'10 CFR Part 50.54(t).

AUTHORITY 6.5.2.9 The OSSRC shall report to and advise the Vice Pr.esident-Supply on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.

CALVERT CLIFFS - UNIT 1 6-11 Amendment No. 26,82

/

r'.'

ADMINISTRATIVE CONTROLS RECORDS 6.5.2.10 Records of OSSRC activities shall be prepared, approved and distributed as indicated below:

a.

Minutes of each OSSRC meeting shall be prepared, approved and forwarded to the Vice President-Supply within 14 days following each meeting.

b.

Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Vice President-

' Supply within 14 days following completion of the review.

c.

Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice President-Supply and to the management

. i positions responsible for the areas audited within 30 days after completion of the audit.

6.6 REPORTABLE OCCURRENCE ACTION

'6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

a.

The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

- s b.

Each REPORTABLE.0CCURRENCE requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission shall be reviewed by the POSRC and submitted to the OSSRC and the Manager - Nuclear Power Department.

l I

i I

CALVERT CLIFFS - UNIT 1 6-12 Amendment No. AB,82 L

UNITED STATEE j.

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NUCLEA :'EGULATORY COMT/lSSION j

J.5 H i t. G T O f., C 0. - - " ~

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BALTIM0RE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 License No. DPR-69 1.

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

A.

The application for amendment by Baltimore Gas & Electric '

Company (the licensee) dated February 24, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and r.egulations 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' 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. DPR-69 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 65, are hereby incorporated in the license.

The licensee shall operate the facility in acco,,rdance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 7

Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: April 21,1983

ATTACH!iENT TO LICEt;SE AfiENDMEf;T NO. 65 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET l10. 50-318 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Corresponding over. leaf pages are also provided to maintain document completeness.

Pages 3/4 ~ 7-13 3/4 7 '

B 3/4 4-1 B 3/4 7-3 6-11 6-13

PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be > 90*F when the pressure of either coolant j

in the steam generator is > 200 psig.

APPLICABILITY: At all times.

ACTION:

With the requirements of the above specification not satisfied:

a.

Reduce the steam generator pressure of the applicable side to

. j:,200 psig within 30 minutes, and b.

Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.

Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F.

SURVEILLANCE REQUIREMENTS 4.7.2.1 The pressure in each side of the steam generators shall be determined to be < 200 psig at least once per hour when the temperature of either the primary or secondary coolant < 90*F.

CALVERT CLIFFS-UNIT 2 3/4 7-13 Amendment No. 65

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION I

3.7.3.1 At least two component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one component cooling water loop OPERABLE, restore at least -

two loops 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 at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN w'ithin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE:

At least once per 31 days by verifying that each valve (manual, a.

power operated or automatic) servicing safety related equipment that is not locked, sealed or otherwise secured in position, is in its correct position.

b.

At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection Actuation test signal.

CALVERT CLIFFS-UNIT 2 3/4 7-14

PLANT SYSTEMS 3/4.7.7 ECCS PUMP'R00M EXHAUST AIR FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7.1 ' The ECCS pump room exhaust ventilation system shall be OPERABLE with one HEPA filter and charcoal adsorber train and two exhaust fans.

APPLICABILITY: MODES 1, 2,.3 and 4.

ACTION:

a.

With one ECCS pump room exhaust fan inoperable, restore the t

inoperable fan to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the ECCS exhaust filt'er train inoperable, restore the filter train to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.7.1 The ECCS pump room exh'aust ventilation system shall be demon-strated OPERABLE:

a.

At least once per 31 days by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that each exhaust fan operates for at least 15 minutes.

b.

At least once per 18 months or (1) after any structural main-tenance on the hEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti-lation zone communicating with the system by:

4 CALVERT CLIFFS - UNIT 2 3/4 7-21

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e.-,

e

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e

PLANT SYSTEMS SUFVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the filter train at a flow rate of 3000 cfm + 10%.

2.

Verifying th'at the HEPA filter banks remove > 99% of the D0P when they are tested in-place in accordaiice with ANSI N510-1975 while operating the filter train at a flow rate of 3000 cfm i 10%.

l 3.

Verifying within 31 days after removal that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one '

of the charcoal adsorbers demonstrates a removal effi-ciency of > 90% for radisactive methyl iodide when the sample is tested in accordance with ANSI N510-1975 (130*C, 95% R.H.).

The carbon samples not obtained from test canisters shall be prepared by either:

a)

Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b)

Emptying a' longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4.

Verifying a system flow rate of 3000 cfm i 10% during system operation when tested in accordance with ANSI N510-1975.

l After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by either:

c.

1.

Verifying within 31 days after removal that a laboratory analysis of a carbon sample obtained from a test canister demonstrates a removal efficiency of > 90% for radioactive methyl iodide when the sample is tested in accordance with ANSI N510-1975 (130 C, 95% R. H.); or CALVERT CLIFFS-UNIT 2 3/4 7-22 Amendment No. 65

EEA:T;E C Z ANT SYSTEF

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m LA T LOOPS AND CODLAC CIRCULATION I

ine plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.195 during all normal operations and anticipated transients.

A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat rencval capability for core cooling while in MODES 2 and 3; however, single failure considerations require pl, ant shutdown if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

In MODES 4 and 5, a single reactor coolant loop or shutdown cooling loop provides. sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires two shutdown cooling loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and produce i

gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump during H0 DES 4 and 5 with one or more RCS cold legs 1 275 F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restrict-ing starting of the RCPs to wheg the gecondary water temperature of each steam generator is less than 46 F (34 F when measured by a surface contact instrument) above the coolant temperature in the reactor vessel.

3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

Each safety valve is designed to relieve approximately 3 x 105 lbs per hour of saturated steam at the valve l

setpoint.

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.

The combined relief capacity of these valves is sufficient to CALVERT CLIFFS - UNIT 2 B 3/4 4-1 Amendment No. 78, H, M,65

REACTOR COOLANT SYSTEM BASES limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.

Denonstration of the safety valves' lift settings will occur only during shutdown and will be. performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

3/4.4.3 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leak-age path.

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer with the level as programned ensures'that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation.

The operating band for pressurizer level bounds the programmed level and ensures that RCS pressure rem: ins within the bounds of an analyzed condition during the excessive charging event as well as during the limiting depressurization event, Excess Load. The operating band also protects the pressurizer code safety valves and power operated relief valve against water relief.

The pow 6r~ operated relief valves function to relieve RCS pressure during all design transients.

Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer--Pressure-High signal, minimizes ti,e undesirable opening of the spring-loaded pressurizer code safety valves.

The requirement that 150 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a-loss of off-site power condition to maintain natural circulation.at HOT STANDBY.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to CALVERT CLIFFS - UNIT 2 B 3/4 4-2 Amendmcnt No. 76, 36, pp,65

PL AN~ SYSTEMS BASES 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with steam discharge to atmosph'ere with concurrent and total loss of offsite power. The contained water volume limit includes an allow-ance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure.that the result-ant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose-also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the accident analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION OALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Ccolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 SECONDARY WATER CHEMISTRY The secondary water chemistry program is designed to provide maximum protec-tion to both the steam generator and secondary system internals.

The most damag-ing chemical reactants enter the system via condenser cooling water ingress.

Accumulation of these impurities in the steam generators may lead to loss of metallurgical integrity and/or eventual component failure.

The limits presented in Table 3.7-3 are those prescribed by the NSSS supplier as " limited-operation" chemistry parameters and are consistent with the most recent industry standards.

~

By routine monitoring of these parameters, plant personnel are able to rapidly detect and limit the duration of ingress of chemically detrimental species and thereby maintain steam generator tube integrity.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION-The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allow-able fracture toughness stress limits.

The limitations of 90 F and 200 psig are based on steam generator secondary side limitations and are sufficient to prevent brittle fracture,

.i CALVERT CLIFFS - UNIT 2 B 3/4 7-3 Amendment No. 76, //'65 S

A PLANT SYSTEMS BASES 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumpt. ions used in the accident analyses.

3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service wa:er system ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptio~ns used in the accident analyses.

3/4.7.5 SALT WATER SYSTEM The OPERABILITY of the salt water system ensures that sufficient cooling capacity is available fnr continued operation of equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allow-able temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

The OPERABILITY of this system in con-junction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.

This limitation is consistent with the requirements of General Design Criteria 10 of Appendix "A",

10 CFR 50.

3/4.7.7 ECCS PUMP ROOM EXHAUST AIR FILTRATION SYSTEM The OPERABILITY of the ECCS pump room exhaust air filtration system ensures that radioactive naterials leaking from the ECCS equipment within the pump room following a LOCA are filtered prior to reaching the CALVERT CLIFFS-UNIT 2 8 3/4 7-4

i j A37.:..;5TRATIVE CC,TRCL5 AUDITS 6.5.2.8.i Audits of facility activities shall be performed under the cognizance of the OSSRC. These audits shall encompass:

a.

The conformance of facility operation to -provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b.

The performance, training and qualifications of the entire facility staff at least onc,e per 12 months.

s c.

The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.

d.

The performance of activities required by the Quality Assurance Program to meet the criteria of Appendix "B",10 CFR 50, at least once per

.24 months.

e.

Deleted f.

The Safeguards Contingency Plan and implementing procedures at least

._ once per 12 months in accordance with 10 CFR 73.40(d).

g.

Any other area of facility operation considered appropriate by the OSSRC or the Vice President-Supply.

h.

The Facility Fire Protection Program and implementing procedures at least once per 24 mon,ths.

i. An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.

j. An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.

6.5.2.8.2 Review of facility activities shall be performed under the cognizance of the OSSRC.

These reviews shall encompass:

a.

The Facility Emergency Plan and implementing procedures at least once per 12 months in accordance with 10 CFP, 50.54(t).

AUTHORITY 6.5.2.9 The OSSRC shall report to and advise the Vice President-Supply on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.

CALVERT CLIFFS - UtilT 2 6-11 kaendment.*40.77,65

ADMINISTRATIVE CONTROLS RECORDS 6.5.2.10 Records of OSSRC activities shall be prepared, ' approved and distributed as indicated below:

a.

Minutes of each OSSRC meeting shall be prepared, approved and forwarded to the Vice President-Supply within 14 days following each meeting.

b.

Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Vice President-Supply within 14 days following completion of the review.

Audit reports encompassed by Section 6.5.2.8 above, shall be c.

forwarded to the Vice President-Supply and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

a.

The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

b.

Each REPORTABLE-0CCURRENCE requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission shall be reviewed by the POSRC and submitted to the OSSRC and the Manager - Nuclear Power Department.

l CALVERT CLIFFS - UNIT 2 6-12 Amendment No.,2f,65

)

m_

~

AD"H:ISTRATI'.'E CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is.

violated:

a.

The facility shall be placed in at least HOI STANDBY within one hour.

b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Manager - Nuclear Power Department and the OSSRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, c.

A Safety Limit Violation Repart shall be prepared. The report shall be reviewed by the POSRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structu'res, and (3) corrective action taken to prevent recu rrence.

d.

The Safety Limit Violation Report shall be submitted to the Commission, the OSSRC and the Manager - Nuclear Power Department within 14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shell be established, implemented and maintained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix "A: of Regulatory Guide 1.33, Revision 2, February 1978.

b.

Refueling operations.

c.

Surveillance and test activities of safety related equipment.

d.

Security Plan inplementation.

e.

Emergency Plan implementation.

f.

Fire Protection Program implementation, g.

The amount of overtime worked by plant staff members performing safety related functions must be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).

6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the POSRC and approved by the Plant Superin-tendent prior to implementation and reviewed periodically as set forth in administrative procedures.

f CALVERT CLIFFS - UNIT 2 6-13 Amendment No. 77, 26, 5'6,65

J ADMINISTRANVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made pro-vided:

The intent of the originial procedure is not altered.

a.

b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on ;the unit affected.

The change is documented, reviewed by the POSRC and approved by c.

the Plant Superintendent within 14 days of implementation.

l 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES

6. 9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

STARTUP REPORT 6.9.1.1 A sunnary report of plant startup and power escalation testin shall be submitted following (1) receipt of an operating license, (2) g amendment to the license involving a planned increase in power level, (3) installation of fuel 'that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perfor-mance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and s peci fica tions. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

AmendmentNo./f($8,65 CALVERT CLIFFS - UNIT 2 6-14

-_