Text

Amend 1 to License DPR-21,adding New Reporting Requirements, Consolidating Specs,Designating salt-splitting Capacity for Replacement Condensate Demineralizer Resin & Clarifying Functional Test Requirements
	ML20210A286
Person / Time
Site:	Millstone
Issue date:	01/29/1987
From:	Charemagne Grimes; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20210A248	List: ML20210A243; ML20210A286; ML20210A304; ML20210A345; ML20210A362; ML20210A400;
References
	NUDOCS 8702060448
	Download: ML20210A286 (118)
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NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-245 AMENDMENT TO FACILITY OPERATING LICENSE >

Amendment No. 1 License No. DPR-21

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The applications for amendment by Northeast Nuclear Energy Company, (the licensee), dated July 9, 1985; October 7 and 16, 1985; January 10, 1986; January 28, 1986; and February 27, 1986; comply 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 a:nendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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8702060448 870129 PDR ADOCK 05000245 P PDR

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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-21 is hereby amended to read as follows:

(2) Technical Specifications The technical specifications contained in Appendix A revised through Amendment No. 1, are hereby incorporated in the license.

The Northeast Nuclear Energy Company shall operate the facility in accordance with the technical specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION C.o.oArcp I. %

Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - fi

Attachment:

Changes tc the Technical Specifications Date of Issuance: January 29, 1987

L ATTACHMENT TO LICENSE AMENDMENT NO. 1 1 FACILITY OPERATING LICENSE NO. DPR-21 DOCKET NO. 50-245 Replace the following pages of the Appendix A Tech.ical Speciff:ations with the enclosed pages as indicated. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change. As indicated by a single asterisk, some overleaf pages without changes have been provided to maintain document completeness. Double asterisks indicate rearrangement of the text only. .

REMOVE INSERT Table of Contents Table of Contents 1-5 1-5 1-6 1-6 1-7 1-7 2-1 2-1 2-2 2-2 B 2-5 8 2-5 8 2-6 8 2-6 8 2-7 8 2-7 8 2-8 8 2-8 8 2-9 8 2-9 8 2-10 B 2-10 8 2-11 8 2-11 3/4 1-3 3/4 1-3 3/4 1-4 3/4 1-4*

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3-REMOVE INSERT 3/4 12-13 3/4 12-13 3/4 12-14 3/4 12-14*

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6-18 6-18 6-19 6-19 6-20 6-20 6-21 6-21 6-22 6-22 6-23 6-24 6-25 The pages indicated by triple asterisks were inadvertently omitted during l reproduction and distribution of the full-term operating license. New pages have been included in this package as necessary.

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i TECHNICAL SPECIFICATIONS 1

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MILLSTONE NUCLEAR POWER STATION UNIT NO. 1 4

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Docket No. 50-245 l Appendix A to License No. DPR-21

L TABLE OF CONTENTS Page No.

1.0 DEFINITIONS ............................................... 1-1 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1.1 FUEL CLADDING INTEGRITY............. 2.1.2 ................ 2-1 2.2.1 REACTOR COOLANT SYSTEM.............. 2.2.2 ................ 2-7 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.0 GENERAL ..................................... 4.0 ......... 3/4 0-1 3.1 REACTOR PROTECTION SYSTEM ................... 4.1 ... ..... 3/4 1-1 3.2 PROTECTIVE INSTRUMENTATION ......... ........ 4.2 ......... 3/4 2-1 A. Primary Containment Isolation Functions ................. 3/4 2-1 B. Emergency Core Cooling Subsystems Actuation ............. 3/4 2-1 C. Control Rod Block Actuation ............................. 3/4 2-1 D. Air Ejector Off-Gas System............................... 3/4 2-12 E. Reactor Building Ventilation, Steam Tunnel Ventilation Isolation, and Standby Gas Treatment Sy s t em I n i t i a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-12 3.3 REACTIVITY CONTROL .......................... 4.3 ......... 3/4 3-1 A. Reactivity Limitations ..................... A ........ 3/4 3-1 B. Control Rod Wi thdrawal . . . . . . . . . . . . . . . . . . . . . B ........ 3/4 3-2 C. Scram Insertion Times ...................... C ........ 3/4 3-5 D. Control Rod Accumulators ................... D ........ 3/4 3-7 E. Reactivity Anomalies ....................... E ........ 3/4 3-8 F. Shutdown Reauirements ................................... 3/4 3-8 G . The rmal Power - Core Fl ow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-8 3.4- STANDBY LIQUID CONTROL SYSTEM ................ 4.4 ........ 3/4 4-1 A. Normal Operation ....................... A ............ 3/4 4-1 B. Operation with Inoperable Components .................... 3/4 4-3 C. Boron Requirements ..................... C ............ 3/4 4-4 D. Shutdown Requirement .................................... 3/4 4-4 i

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i Amendment No. 1 l

J Surveillance Pace No.

3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 3/4 5-1 A. Core Spray and LPCI Subsystems .............. A ...... 3/4 5-1 B. Containment Cooling Subsystems............... B ...... 3/4 5-3 C. FWCI Subsystem .............................. C ...... 3/4 5-5 D. Automatic Pressure Relief Subsystems ........ D ...... 3/4 5-6 E. Isolation Condenser System .................. E ...... 3/4 5-7 F. Minimum Core and Containment Cooling Sy s t em A v a i l a b i l i ty . . . . . . . . . . . . . . . . . . . . . . . F . . . . . . 3/4 5-8 3.6 PRIMARY SYSTEM BOUNDARY 4.6 3/4 6-1 A. The rmal Limi ta ti ons . . . . . . . . . . . . . . . . . . . . . . . . . A . . . . . . 3/4 6-1 B. Pressurization Temperature .................. B ...... 3/4 6-2 C. Coolant Chemistry ........................... C ...... 3/4 6-5 D. Coolant Leakage.............................. D ...... 3/4 6-11 E. Safety and Relief Valves..................... E ...... 3/4 6-11 F. Struc tu ral I n teg ri ty . . . . . . . . . . . . . . . . . . . . . . . . F . . . . . . 3/4 6-12 G. Jet Pumps ................................... G ...... 3/4 6-13 H. Recirculation Pump Flow Mismatch ............ H ...... 3/4 6-14

1. Snubbers .................................... I ...... 3/4 6-15 J. Condensate Demineralizers ................... J ...... 3/4 6-21 3.7 CONTAINMENT SYSTEMS 4.7 3/4 7-1 A . P rima ry Conta i nment . . . . . . . . . . . . . . . . . . . . . . . . . A . . . . . . 3/4 7-I B. Standby Gas Treatment System ................ B ...... 3/4 7-10 C. Seconda ry Contai nment . . . . . . . . . . . . . . . . . . . . . . . C . . . . . . 3/4 7-13 D. Primary Containment Isolation Valves . . . . . . . . D . . . . . . 3/4 7-14 3.8 RADI0 ACTIVE MATERIALS 4.8 3/4 8-1 A. Radioactive Liquid Effluent ~ Instrumentation . A ...... 3/4 8-1 B. Radioactive Gasecus Effluent Instrumentation. B ...... 3/4 8-6 C. Radioactive Liquid Effluents . . . . . . . . . . . . . . . . C . . . . . . 3/4 8-12

0. Radioactive Gasious Effluents. . . . . . . . . . . . . . . . D . . . . . . 3/4 0-14 3.9 AUXILIARY ELiCTRICAL SYSTEM 4.9 3/4 9-1 3.10 REFUELING 4.10 3/4 10-1 A . Re fuel i ng I nterl oc ks . . . . . . . . . . . . . . . . . . . . . . . . A . . . . . . 3/4 10-1 B. Core Monitoring ............................. B ...... 3/4 10-2 C. Fuel Storage Pool Water Level . . . . . . . . . . . . . . . C . . . . . . 3/4 10-3 D . C ra n e O p e ra b i l i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . D . . . . . . 3/4 10-3 E. Crane Interlocks and Switches. . . . . . . . . . . . . . . . E . . . . . . 3/4 10-3 ii Amendment No. 1 l l

L Surveillance Page No.

3.11 REACTOR FUEL ASSEMBLY 4.11 3/4 11-1 A. Averaoe Planar Linear Heat Generation Rate (APLHGR)................... A ...... 3/4 11-1 B. Linear Heat Generation Rate (LHGR) . . . . . . . . . . . B . . . . . . 3/4 11-5 C. Minimum Critical Power Ratio (MCPR) . . . . . . . . . . C . . . . . . 3/4 11-6 3)12 FIRE PROTECTION SYSTEMS 3/4 12-1 A. Fire Suppression Water System . . . . . . . . . . . . . . . A . . . . . . 3/4 12-1 B. Spray and/or Spri nkler Systems . . . . . . . . . . . . . . B . . . . . . 3/4 12-5 C. Carbon Dioxide and Halon 1301 Systems ....... C ...... 3/4 12-7 D. Fire Hose Stations .......................... D ...... 3/4 12-9 E. Fire Protection Instrumentation ............. E ...... 3/4 12-13 F. Penetra ti on Fi re Barriers . . . . . . . . . . . . . . . . . . . F . . . . . . 3/4 12-17 3.13 INSERVICE INSPECTION 4.13 3/4 13-1 3.14 PLANT SYSTEMS 4.14 3/4 14-1 1

5.0 DESIGN FEATURES 5-1 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 Responsibility .................................. 6-1 6.2 Organization .................................... 6-1 6.3 Facili ty Staf f Quali fications . . . . . . . . . . . . . . . . . . . 6-4 6.4 Training ........................................ 6-6 6.5 Review and Audit ................................ 6-6 i 6.6 Reportable Event Action .......................... 6-15 6.7 Safety Limi t Vi ol ati on . . . . . . . . . . . . . . . . . . . . . . . . . . 6-15 6.8 Procedures ...................................... 6-16 6.9 Reporting Requirements .......................... 6-17 6.10 Record Retention ................................ 6-19 6.11 Radiation Protection Program .................... 6-20 6.12 High Radiation Area ............................. 6-21 6.13 Systems Integrity ............................... 6-21 6.14 Iodine Monitoring ............................... 6-22 6.15 Radiological Effluent Monitoring and Offsite Dose Calculation Manual............ 6-22 6.16 Radioactive Waste Treatment ..................... 6-22 i

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1.0 V 1. Hot shutdown means the conditions as above with the reactor coolant temperature greater than 212 F.

2. Cold shutdown means conditions as above with the reactor coolant temperature equal to or less than 212 F and the reactor vessel vented.

ft. Simulated Automatic Actuation Simulated automatic actuation means applying a simulated signal to actuate the circuit in question.

X. Surveillance Surveillance is that process whereby systems and components which are essential to plant nuclear safety during all modes of operation or which are necessary to prevent or mitigate the consequences of incidents are checked, tested, calibrated and/or inspected, as warranted, to verify performance and availability at optimum intervals.

Unless otherwise specified, each Surveillance Requirement shall be performed within the specified time interval with:

1. A maximum allowable extension not to exceed 25% of the test interval, and l

2. A total maximum combined interval time for any 3 consecutive j tests not to exceed 3.25 times the specified test interval.

Performance of a surveillance requirement within the specified time interval shall constitute compliance with Operability requirements for a Limiting Condition for Operation.

Surveillance Requirements shall be applicable during the operational modes associated with Limiting Conditions for Operation. Surveillance need not be performed if the system or component to be tested is not required to be operational as specified by the Limiting Conditions for Operation. However, the required surveillance shall be perforn'ed prior to returning the component to an operational status as required by the Limiting Conditions for Operation.

Y. Reportable Event l A Reportable Event shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

AA. Transition Boiling Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

BB. Emergency Power Sources Emergency power sources means the on-site gas turbine generator and diesel generator.

Amendment No. 1 Millstone Unit 1 1-5

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1.0 CC. Staggered Test Basis )

1. Staggered Test Basis shall consist of:

a) A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and b) The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

DD. Dose Equivalent I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (uCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation, shall be those listed on Table E-7 of Regulatory Guide 1.109, Revision 1.

EE. E-Average Disintegration Energy E shall be the average sum of the beta and gamma energies per disintegration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

FF. Radioactive Waste Treatment Systems Radioactive Waste Treatment Systems are those liquid, gaseous, and solid waste systems which are required to maintain control over radioactive materials in order to meet the LC0's set forth in the Specifications.

GG. Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM)

A RADIOLOGICAL EFFLUENT MONITORING MANUAL shall be a manual containing the site and environmental sampling and analysis programs for measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures to individuals from station operation. An 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall be a manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints. Requirements of the REM 0DCM are provided in Specification 6.15.

HH. Purge - Purging Purge or Purging is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

Amendment No. 1 Millstone Unit 1 1-6

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1.0 II. Venting VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that

replacement air or gas is not provided or required during venting.

-Vent, used in system names, does not imply a VENTING process.

JJ. Member (s) of the Public Member (s) of the Public shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or its vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational occupational or other purposes not associated with the plant.

The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location.

KK. Site Boundary The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.

LL. Source Check A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

MM. Unrestricted Area An UNRESTRICTED AREA shall be any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional and/or recreational purposes.

NN. Refuel Condition The reactor is in the Refuel Condition when all of the following conditions are met:

1) All Control Rods are fully inserted or the Refuel Interlocks are operable with no more than one rod not fully inserted.

2) The reactor mode switch is in the refuel position.

3) The Reactor Coclant Temperature is <200*F.

4) The Reactor Vessel is vented.

NOTE: It is permissible to perform the Refuel Interlock Functional Check, and still be considered in the Refuel Condition.

Amendment No. 1 Millstone Unit 1 1-7

s SAFETY LIMITS l 2.1.1 FUEL CLADDING INTEGRITY Applicability:

Applies to the interrelated variables associated with fuel thermal behavior.

Objective:

To establish limits below which the integrity of the fuel cladding is preserved.

Specification:

A. When the reactor pressure is greater than 800 psia and the core flow is greater than 10% of rated design, a minimum critical power ratio (MCPR) less than 1.07 shall constitute a violation of the fuel cladding integrity safety limit.

LIMITING SAFETY SYSTEM SETTINGS 2.1.2 FUEL CLADDING INTEGRITY Applicability:

Applies to trip settings of-the instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.

Objective: .,

To define the level of the process variables at which automatic protective action is initiated to prevent the safety limits from being exceeded.

Specification:

The limiting safety system settings shall be as specified below:

A. Neutron Flux Scram

1. APRM Flux Scram Trip Setting (Run Mode)

a. When the Mode Switch is in the RUN position, the APRM flux scram trip setting shall be as shown on Figure 2.1.2 and shall be:

S 5 0.58 W + 62 1

i Amendment No.1 Millstone Unit 1 2-1 L

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100 Normal APRM Scram Trip L5 = .58W + 62 /

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FIGURE 2.1.2 MILLSTONE APAM SCRAM AND A00 SLOCK TRIP LIMITING SAFETY SYSTEMS SETTINGS L

fillistone Unit 1 2-2 Amendment No. 1

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L 2.1.2 FUEL CLADDING INTEGRITY BASES This choice of using conservative values of controlling parameters and initiating transients at the licensed maximuu steady state power level produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

Steady-state operation without forced recirculation will not be permitted, except during startup testing. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.

In summary:

1 - The licensed maximum steady-state power level is 2011 MWt.

2 - The abnormal operational transients were analyzed to a thermal power level of 2011 MWt.

3 - Analyses of transients employ adequately conservative values of the controlling reactor parameters.

4 - The analytical procedures now used result in a more logical answer than the alternative method of assuming a higher starting power in conjunction with the expected values for the parameters.

A. Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads

, percent of rated thermal power. Since fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during transients induced by disturbances and with APRM scram settings as specified by Section 2.1.2A, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that, even with a fixed 120% scram trip setting, none of the postulated transients result in violation of the fuel safety limit and there is a substantial margin from fuel damage. Therefore, use of a flow-biased scram provides even additional margin.

Increasing the APRM scram setting would decrease the margin present before the thermal huydraulic safety limit is reached. The APRM scram setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. A reduction in this operating margin would increase the frequency of spurious scrams which have an adverse affect on reactor safety. Thus, the specified APRM setting was selected to provide adequate margin from the thermal-hydraulic safety limit and allow operating margin to minimize the frequency of unnecessary scrams.

Amendment No. 1 Millstone Unit 1 B 2-5

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2.1.2 FUEL CLADDING INTEGRITY BASES Analyses of the limiting transients show that no scram adjustment is required to assure MCPR 2 1.07 when the transient is initiated from MCPR's specified in Section 3.11.C. In order to assure adequate core margin during full load rejections in the event of failure of the select rod insert, it is necessary to reduce the APRM scram trip setting to 90%

of rated power following a full load rejection incident. This is necessary because, in the event of failure of the select rod insert to function, the cold feedwater would slowly increase the reactor power level to the scram trip setpoint. A trip setpoint of 90% of rated has been established to provide substantial margin during such an occurrence. The trip setdown is delayed to prevent scram during the initial portion of the transient. The specified maximum setdown delay of 30 seconds is conservative because the cold feedwater transient does not produce significant increases in reactor power before approximately 60 seconds following the load rejection.

For operation in the REFUEL or STARTUP/ HOT STANDBY modes while the reactor is at low pressure, the APRM reduced flux trip scram setting of 5 15% of rated power provides adequate thermal margin between the maximum power and the safety limit, 25% of rated power. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor.

Cold water from sources available during startup is not much colder than that already in the system. Temperature coefficients are small and control rod patterns are constrained to be uniform by' operating procedures backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.

In an assumed uniform rod withdrawal approach to the scram level, the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The APRM reduced trip scram remains active until the mode switch is placed in the RUN position. This switch occurs when the reactor pressure is greater than 880 psig.

The IRM trip at 5 120/125 of full scale remains as a backup feature.

The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.

During steady-state operation, with one recirculation pump operating, the equalizer line shall be isolated. Analyses of transients from this operating condition are less severe than the same transients from the two pump operation.

Amendment No. 1 Millstone Unit 1 8 2-6

L 2.1.2 FUEL CLADDING INTEGRITY BASES B. APRM Control Rod Block Trips Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thus to protect against a condition of MCPR

< 1. 07. This rod block setpoint, which is automatically varied with recirculation flaw rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The specified flow variable setpoint provides substantial margin from fuel damage, assuming steady-state operation at the setpoint, over the entire recirculation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship.

Therefore, the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

When the maximum fraction of limiting power density exceeds the fraction of rated thermal reactor power, the rod block setting is adjusted in accordance with the formula in Specification 2.1.2.B. If the APRM rod bicck setting should require a change due to an abnormal peaking condition, it will be done by increasing the APRM gain, thus reducing the slope and intercept point of the flow referenced rod block curve by the reciprocal of the APRH gain change.

The APRM rod block setpoint is reduced to 5 12% of rated thermal power with the mode switch in REFUEL or STARTUP/ HOT STANDBY position.

C. Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained.

D. Reactor Low Low Water Level ECCS Initiation Trip Point The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the decay heat associated with the

! loss-of-coolant accident; to limit fuel clad temperature to well below the clad melting temperature; to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%. To accomplish

this function, the capacity of each emergency core cooling system l component was established based on the reactor low low water level. To lower the setpoint of the low water level scram would require an increase in the capacity of each of the ECCS components. Thus, the reactor vessel

! low water level scram was set low enough to permit margin for operation, l yet will not be set lower because of ECCS capacity requirements.

Amendment No. 1 Millstone Unit 1 B 2-7

2.1.2 FUEL CLADDING INTEGRITY BASES The design of the ECCS components to meet the above criteria was dependent on three previously set parameters: the maximum break size, the low water level scram setpoint, and the ECCS initiation setpoint. To lower the setpoint for initiation of the ECCS would prevent the ECCS components from meeting their design criteria. To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.

E. Turbine Stop Valve Scram The turbine stop valve scram, like the load rejection scram, anticipates the pressure, neutron flux, and heat flux increase caused by the rapid closure of the turbine stop valves and failure of the bypass. With a scram setting of 5 10% of valve closure, the resultant increase in surface heat flux is limited such that MCPR remains above 1.07 even during the worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine steam flow is < 45% of rated, as measured by the turbine first stage pressure.

F. Turbine Control Valve Fast Closure The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass; i.e., it prevents MCPR from becoming less than 1.07 for this transient. For the load rejection from 100% power, the heat flux increases to only 106.5% of its rated power value, which results in only a -

small decrease in MCPR. This trip is bypassed below a generator output of 307 MWe because, below this power level, the MCPR is greater than 1.07 throughout the transient without the scram.

In order to accommodate the full load rejection capability, this scram trip must be bypassed because it would be actuated and would scram the reactor during load rejections. This trip is automatically bypassed for a maximum of 280 millisec following initiation of load rejection. After 280 millisec, the trip is bypassed providing the bypass valves have opened.

If the bypass valves have not opened after 280 millisec, the bypass is removed and the trip is returned to the active condition. This bypass does not adversely affect plant safety because the primary system pressure is within limits during the worst transient even if the trip fails. There are many other trip functions which protect the system during such transients.

G. Main Steam Line Isolation Valve Closure Scram The low pressure isolation of the main steam lines at 825 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed, Amendment No. 1 Millstone Unit 1 8 2-8

L 2.1.2 FUEL CLADDING INTEGRITY BASES to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the STARTUP position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients, which occur during normal or inadvertent isolation valve closure. With the scrams set at 10% valve closure, there is no increase in neutron flux during valve closure.

H. Main Steam Line Low Pressure Initiates Main Steam Isolation Valve Closure The low pressure isolation at 825 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are clased to provide for reactor shutdown so that operation at pressures loser than those specified in the thermal-hydraulic safety limit does not occur, although operation at a pressure lower than 825 psig would not necessarily constitute an unsafe condition.

Amendment No. 1 Millstone Unit 1 B 2-9

2.2.1 REACTOR COOLANT SYSTEM 1

BASES The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The pressure safety limit of 1325 psig, as measured by the vessel steam space pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value is derived from the design pressures of the reactor pressure vessel, coolant system piping, and isolation condenser. The respective design pressures are 1250 psig at 575 F, 1175 psig at 564 F, and 1250 psig at 575 F. The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes:

ASME Boiler and Pressure Vessel Code,Section III for the pressure vessel and isolation condenser and USAS Code B31.1 for the reactor coolant system piping. '

The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10%

over design pressure (110% x 1250 = 1375 psig), and the USAS Code permits pressure transients up to 20% over the design pressure (120% x 1175 = 1410 psig). The Safety Limit pressure of 1375 psig is referenced to the lowest elevation of the primary coolant system.

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig.

The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is more than a factor of 1.5 below the yield strength of 43,300 psi at 575 F. At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yield strength.

The relationships of stress levels to yield strength are comparable for the isolation condenser and primary system piping and provide a similar margin of protection at the established safety pressure limit.

l The normal operating pressure of the reactor coolant system is 1035 psig. For the turbine trip or loss of electrical load transients, the turbine trip scram or generator load rejection scram, together with the turbine bypass system, l limits the pressure to less than 1085 psig. The safety / relief valves are set

! at 1095 psig, 1110 psig, and 1125 psig and are sized to keep the reactor

! coolant system pressure below 1375 psig with no credit taken for the turbine l bypass system. Credit is taken for the neutron flux scram, however.

During operation, reactor pressure is continuously displayed in the control room on a 0-1500 psig pressure recorder.

l Amendment No. 1 i Millstone Unit 1 B 2-10 1

A 2.2.2 REACTOR COOLANT SYSTEM BASES In compliance with Section III of the ASME Boiler and Pressure Vessel Code, 1965 Edition, the specified settings of the pressure relieving devices are below 103% of design pressure. As described in the General Electric Topical Report, NEDE-24011-P-A, Generic Reload Fuel Applicatien (Section 5.3), the most severe isolation event with indirect scram has been evaluated. The most severe isolation is the MSIV closure from steady-state operation at 2011 MWt. The evaluation assures that the sizing and settings of the pressure relieving devices are adequate to assure that the peak allowable pressure of 110% of vessel design pressure is not exceeded.

Evaluations indicate that a total of six dual purpose safety / relief valves set at the specified pressures maintain the peak pressure during the transient well within the code allowable and safety limit pressure.

Amendment No. 1 Millstone Unit 1 B 2-11

e .

TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS Minimum Number cf Operable Modes in which Function Inst. Channels Trip Function Trip Level Setting Must Be Operable Action

per Trip (1) REFUEL / STARTUP/ HOT System SHUTDOWN (8,11) STANOBY RUN 1 Mode Switch in SHUTDOWN X X X A 1 Manual Scram X X X A IRM:

3 High Flux 5120/125 of full scale X X (5) A 3 Inoperative A. HI Voltage <80 voit DC X X X (10) A B. IRM Module Unplugged C. Selector Switch not in Operate Position APRM:

2 Flow Blased High Flux See Section 2.1.2A X X X A or B 2 Reduced High Flux See Section 2.1.2A X X X A or B 2 Inoperable A. > 50% LPRM Inputs ** X X X A or B B. Circuit Board Removed C. Selector Switch not in Operate Position 2 High Reactor Pressure 51085 psig X X X A 2 High Drywell Pressure 52 psig X (9) X (7) X (7) A 2 Reactor Low Water Level 21.0 inch *** X X X A 2 Scram Discharge Vol. 526 inches above the center- X (2) X X A High Level line of the lower end cap to SDIV pipe weld Millstone Unit 1 3/4 1-3 Amendment No. 1

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a TABLE 3.1.1 (Continued)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRt> MENTATION REQUIREMENTS Minimus Number cf Operable Inst. Channels Trip Function Modes in which Function Trip Level Setting Must Be Operable Action

per Trip (1)

System REFUEL / STARTUP/ HOT SHUTDOWN (8,11) STAND 8Y RUN 2 Turbine C ser Low ;:23 in. Hg. Vacuum X (3) X (3) X A or C 2 Main Steamline Radiation 57 x Normal Full Power X X X A or C

Background

4 (6) Main Steamline Isolation 510% Valve Closure Valve Closure X (3) X (3) X A or C 2 Turbine n Ive 1 See Section 2.1.2 F X (4) X (4) X (4) A or C 2 Turbine Stop Valve 510% Valve Closure X (4) X (4) X (4) A or C Notts: 1. There shall be two operable or tripped trip systems for each function.

2. Permissible to bypass, with control rod block, for reactor protection system reset in REFUEL and SHUTDOWN positions of the reactor mode switch.

3. Bypassed when reactor pressure is <600 psig.

4. Bypassed when first stage turbine pressure is less than that which corresponds to 45% rated steam flow (generator output approximately 307 MWe).

Millstone Unit 1 3/4 1-4 , ,

TABLE 3.1.1 (Continued)

J Reactor Protection System (Scram) Instrumentation Requirements 4

Notas:

5. IRM's are bypassed when mode switch is placed in RUN. The detector for each operable IRM channel shall be fully inserted until the associated APRM channel is operable and indicating at least 3/125 full scale. l

6. The design permits closure of any one valve without a scram being initiated.

7. May be bypassed when necessary by closing the manual instrument isolation valve for scram of PS-1621 A through D during purging for containment inerting or deinerting.

8. When the reactor is subcritical and the reactor water temperature is less than 212*F, only the following trip functions need to be operable:

a. Mode Switch in SHUTDOWN

b. Manual Scram

c. High Flex IRM

d. Scram Discharge Volume High Level

e. APRM Reduced High Flux

9. Not required to be operable when primary containment integrity is not required.

10. With the mode switch in RUN position an inoperative trip function also requires an associated APRM "downscale alarm."

11. Trip functions are not required to be operable if all control rods are fully inserted, and either electrically or hydraulically disarmed in accordance with Specification 4.1.D.

Action: If the first column cannot be met for one of the trip systems, that trip system shall be tripped. If the first column cannot be met for both trip systems, the appropriate actions listed below shall be taken:

A. Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

B. Reduce power level to IRM range and place mode switch in the STARTUP/ HOT STANDBY position within eight hours.

C. Reduce tarbine load and close main steam line isolation valves within eight hours.

- An APRM will be considered inoperable if there are less than two LPRM inputs per level or there are less than 50% of the normal l

compliment of LPRM's to an APRM.

*** One inch on the water level instrumentation is 127 inches above the top of the active fuel.

! Millstone Unit 1 3/4 1-5 Amendment No. 1 l

1

TABLE 4.1.1 SCRAM INSTRUMENTATION FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS Instrument Channel Group (3) Functional Test Minimum Frequency (4)

High Reactor Pressure A Trip Channel and Alarm (1)

High Drywell Pressure A Trip Channel and Alarm (1)

Low Reactor Water Level A Trip Channel and Alarm (2) (1)

High Water Level in Scram Discharge A Trip Channel and Alarm (1)

Condenser Low Vacuum A Trip Channel and Alarm (8) (1)

Main Steam Line Isolation Valve Closure A Trip Channel and Alarm (1)

Turbine Stop Valves Closure A Trip Alarm (1)

Manual Scram A Trip Channel and Alarm (1)

Turbine Control Valve Fast Closure A Trip Channel and Alarm (6) (8) (1)

Flow Biased High Flux APRM B Trip Output Relays (6) (7) (8) (1)

Reduced High Flux APRM B Trip Output Relays (8) Before each startup (5)

IRM C Trip Channel and Alarm (5) (8) Before each startup (5)

High Steam Line Radiation B Trip Channel and Alarm (2) (8) (1)

Mode Switch in SHUTDOWN A Place Mode Switch in SHUTDOWN Each refueling outage Millstone Unit 1 3/4 1-6 s '.

4 4

l TABLE 3.2.2 INSTRUMENTATION THAT INITIATES AND CONTROLS THE EMERGENCY CORE COOLING SYSTEMS Minimum Number of Operable Inst. Trip Function Trip. Level Setting Remarks Channels Per Trip System (1) 2 Reactor Low Low Water 79 (+4-0) inches above 1- In conjunction with Level top of active fuel low reactor pressure initiates core spray and LPCI.

2- .In conjunction with

, high drywell pressure, 120 sec. time delay, and LP core cooling pump interlock initiates auto blowdown.

3- Initiates FWCI and

, Isolation Condenser.

4- Initiates starting of diesel generator and gas turbine generator.

2 High Drywell Pressure 52 Psig 1 - Initiates core spray, LPCI, FWCI, and SBGTS. -

l Millstone Unit 1 3/4 2-3

TABLE 3.2.2 (Continued)

INSTRUMENTATION THAT INITIATES AND CONTROLS THE EMERGENCY CORE COOLING SYSTEMS Minimum Number of Operable Inst. Trip Function Trip Level Setting Remarks Channels Per Trip System (1) 2- In conjunction with low low water level, 1

120 sec. time delay, ,

and LP core cooling '

pump interlock initiates auto blowdown.

3- Initiates starting of diesel and gas turbine generator.

1 Reactor Low Pressure 300 Psig $P $350 Psig 1- Permissive for Permissive opening core Spray and LPCI admission valves.

2- In conjunction with low low reactor water level initiates core spray and LPCI.

1 High Reactor Pressure 51085 Psig 1- In conjunction with 15 second time delay, initiates Isolation l Condenser.

Millstone Unit 1 3/4 2-4 Amendment No. 1 * -

1

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TABLE 4.2.1 (continued) -

MINIMUM TEST AND CALIBRATION FREQUENCY FOR CORE COOLING INSTRUMENTATION ROD BLOCKS AND ISOLATIONS 4

I 5) The individ6a1 power available on emergency bus relays will be functionally tested at the frequency i specified by (1) above. A full functional test including the actuation of the permissives will be performed every refueling outage.

6) This instrumentation is excepted from the functional test definition. The functional test will consist i of injecting a simulated electrical signal into the measurement channel. Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibrations including the sensors will be performed during each refueling outage. Instrument checks shall be performed at least once per day during those periods when the instruments are required to be operable.

i 1

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Millstone Unit 1 3/4 2-11 '

d LIMITING CONDITION FOR OPERATION 3.2 PROTECTIVE INSTRUMENTATION C. 2. The minimum number of operable instrument channels specified in Table 3.2.3 for the Rod Block Monitor may be reduced by one for maintenance and/or testing for periods not in excess of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 30-day period.

D. Air Ejector Off-Gas System

1. Except as specified in 3.2.D.2 below, bcth air ejector off gas system radiation monitors shall be operable during reactor power operation. The trip settings for the monitors shall be set at a value not to exceed the equivalent of the instantaneous stack release limit specified in Specification 3.8. The time delay setting for closure of the steam jet-air ejector off gas isolation valve shall not exceed 15 minutes.

2. From and after the date that one of the two air ejector off gas system radiation monitors is made or found to be inoperable, reactor' power operation is permissible only during the succeeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , provided the inoperable monitor is tripped, unless such system is sooner made operable.

E. Reactor Building Ventilation Isolation, Steam Tunnel Ventilation Isolation and Standby Gas Treatment System Initiation

1. Except as specified in 3.2.E.2 below, six radiation monitors shall be operable at all times.

2. Oae of the two radiation monitors in the reactor building ventilation duct, one of the two radiation monitors on the refueling floor and one of the two radiation monitors in the steam tunnel ventilation may be inoperable for 24 hrs. If it is not restored to service in this time, the reacter building ventilation system and

, steam tunnel ventilation system shall be isolated and the standby I

gas treatment operated until repairs are complete.

3. The radiation monitors shall be set to trip as follows:

a. Ventilation duct - 11 mr/hr.

b. Refueling floor - 100 mr/hr.

l

c. Steam tunnel ventilation - 12 mr/hr.

i l

1 Amendment No. 1 Millstone Unit 1 3/4 2-12 l

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LIMITING CONDITION FOR OPERATION 3.3 REACTIVITY CONTROL Applicability:

Applies to the operational status of the control rod system.

Objective:

To assure the ability of the control rod system to control reactivity.

Specification:

A. Reactivity Limitations

1. Reactivity Margin - Core Loading The core loading shall be limited to that which can be made subcritical in the most reactive condition during the operating cycle with the strongest operable control rod in its full-out position and all other operable rods felly inserted.

SURVEILLANCE REQUIREMENT 4.3 REACTIVITY CONTROL Applicability:

Applies to the surveillance requirements of the control rod system.

Objective:

To verify the ability of the control rod system to control reactivity.

Specification:

A. Reactivity Limitations

1. Reactivity Margin - Core Loading Sufficient control rods shall be withdrawn following a refueling outage when core alterations were performed to demonstrate with a margin of 0.33% A K that the core can be made subcritical at any time in the subsequent fuel cycle with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.

Millstone Unit 1 3/4 3-1

LIMITING CONDITION FOR OPERATION 3.3 REACTIVITY CONTROL 3.3.A.2. Reactivity Margin - Stuck Control Rods Control rod drives which cannot be moved with control rod drive pressure shall be considered inoperable. The control rod directional control valves for inoperable control rods shall be disarmed ele-trically and the rods shall be in such positions that Specification 3.3.A.1 is met. In no case shall the number of non-fully inserted rods disarmed be greater than eight during power operation. If a partially or fully withdrawn control rod drive cannot be moved with drive or scram pressure the reactor shall be brought to a shutdown condition within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless investigation demonstrates that the cause of the failure is not due to a failed control rod drive mechanism collet housing.

SURVEILLANCE REQUIREMENT 4.3.A.2. Reactivity Margin - Stuck Control Rods Each partially or fully withdrawn operable control rod shall be exercised one notch at least once each week. This test shall be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in the event power operation is continuing with three or more inoperable control rods or in the event power operation is continuing with one fully or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out. The surveillance need not be completed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the number of inoperable rods has been reduced to less than three and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause

. of an immovable control rod.

I i

Amendment No. 1 Millstone Unit 1 3/4 3-2

e ,

LIMITING CONDITION FOR OPERATION p

3.3 REACTIVITY CONTROL

!s B. Control Rod Withdrawal

1. Each control rod shall be coupled to its drive or completely

, inserted and the control rold directional control valves disarmed electrically. .However, for purposes of removal of a control rod drive,=as many as one drive in each quadrant may be uncoupled from its control rod so Yong as the reactor is in the shutdown or refuel ccadition and Specification 3.3.A.1 is met.

2. The control rod drive housing support sy; tem shall be in place durirg power operation and when the reactor coolant system is pressurized above atnespheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.3.A.1 is met.

3. Whenever the reactor is in the STARTUP or RUN mode below 20% rated thermal power, no control rods shall be moved unless the rod worth minimizer is operable or a second independent operator, or engineer, verifies that the operator at the reactor console is following the control rod program. The second operator may be used as a substitute for an inoperable rod worth minimizer during a startup only if the rod worth minimizer fails after withdrawal of at least twelve control rods.

4. Cont.ol rods shall not be withdrawn for startup or refueling unless at least two source range channels have an observed count rate equal to or greater than three counts per second.

5. During operation with limiting control rod patterns, as determined by the reactor engineer, either:

a. Both RBM channels shall be operable; or

b. Control rod withdrawal shall be blocked; or

c. The operating power level shall be limited so that the MCPR will remain equal to or greater than 1.07 assuming a single error that results in complete withdrawal of any single operable control rod.

Amendment No. 1 Millstone Unit 1 3/4 3-3

SURVEILLANCE REQUIREMENT 4.3 REACTIVITY CONTROL B. Control Rod Withdrawal

1. The coupling integrity shall be verified for each withdrawn control rod as follows:

a. When the rod is fully withdrawn the first time subsequent to each refueling outage or after maintenance, observe that the drive does not go to the overtravel position; and

b. When the rod is withdrawn the first time subsequent to each refueling outage or after maintenance, observe discernible response of the nuclear instrumentation; however, for initial rods when response is not discernible, subsequent exercising of these rods after the reactor is critical shall be performed to verify instrumentation response.

2. The control rod drive housing support system shall be inspected after reassembly and the results of the inspection shall be recorded.

3.a. To consider the rod worth minimizer operable, the following steps must be performed:

(i) The control rod withdrawal sequence for the rod worth minimizer computer shall be verified as correct.

(ii) The rod worth minimizer diagnostic test shall be successfully completed.

(iii) Proper annunciation of the select error of at least one out-of-sequence control rod in each fully inserted group shall be verified.

(iv) The rod block function of the rod worth minimizer shall be verified by attempting to withdraw an out-of-sequence control rod beyond the block point.

b. If the rod worth minimizer is inoperable while the reactor is in the STARTUP or RUN mode below 10% rated thermal power, and a second independent operator or engineer is being used, he shall verify that all rod positions are correct prior to commencing withdrawal of each rod group.

, 4. Prior to control rod withdrawal for startup or during refueling, l verify that at least two source range channels have an observed l count rate of at least three counts per second.

l l 5. When a limiting control rod pattern exists, an instrument functional test of the RBM shall be performed prior to withdrawal of the designated rod (s) and daily thereafter.

Amendment No. 1 Millstone Unit 1 3/4 3-4

LIMITING CONDITION FOR OPERATION 3.3. REACTIVITY CONTROL C. Scram Insertion Times

1. The average scram insertion time, based on the deenergization of the scram pilot valve solenoids as time zero, of all operable control rods in the reactor power operation condition shall be no greater than:

% Inserted From Average Scram Fully Withdrawn Insertion Times (Sec.)

5 0.375 20 0.900 50 2.000 90 3.500

2. The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two by two array shall be no greater than:

% Inserted From Average Scram Fully Withdrawn Insertion Times (sec.)

5 0.398 20 0.954 50 2.120 90 3.800

3. a. The maximum scram insertion time for 90% insertion of any operable control rod shall not exceed 7.00 seconds.

b. The scram discharge volume drain and vent valves will close in less than 30 seconds after receipt of a signal for control rods to scram.

Amendment No. 1 Millstone Unit 1 3/4 3-5

SURVEILLANCE REQdIREMENT _

4.3 REACTIVITY CONTROL C. Scram Insertion Times

1. During each operating cycle, each operable control rod shall be subjected to scram time tests from the fully withdrawn position. If testing is not accomplished during reactor power operation, the measured scram insertion times shall be extrapolated to the reactor power operation condition utilizing previously determined correlations.

2. The scram discharge volume drain and vent valves shall be verified open at least once per month.

3. The following conditions of operability of the scram discharge volume drain and vent valves shall be verified at least once per operating cycle in accordance with Section 3.13, Inservice Inspection:

a. Closing time after signal for control rods to scram and

b. Verification of opening when scram signal is resct and when the scram discharge volume trip is bypassed.

Amendment No. 1 Millstone Unit 1 3/4 3-6

1

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LIMITING CONDITION FOR OPERATION 3.3 REACTIVITY CONTROL D. Control Rod Accumulators At all reactor operating pressures, a rod accumulator may be inoperable provided that no other control rod in the nine-rod square array around this rod has a:

1. Inoperable accumulator.

2. Directional control valve electrically disarmed while in a non-fully inserted position.

3. Scram insertion greater than maximum permissible insertion time.

If a control rod with an inoperable accumulator is inserted " full-in" and its directional control valves are electrically disarmed, it shall not be considered to have an inoperable accumulator.

SURVEILLANCE REQUIREMENT 4.3 REACTIVITY CONTROL D. Control Rod Accumulators Once a shift, check the status in the control room of the pressure and level alarms for each accumulator.

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Millstone Unit 1 3/4 3-7 l

LIMITING CONDITION FOR OPERATION 3.3 REACTIVITY CONTROL E. Reactivity Anomalies The reactivity equivalent of the difference between the actual critical rod configuration and the expected configuration during power operation shall not exceed 1% A K. If this limit is exceeded, the reactor will be shutdown until the cause has been determined and corrective actions have been taken if such actions are appropriate.

F. Shutdown Requirements If Specification 3.3 A through D above are not met, a normal orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Thermal Power Core Flow .

G.

Allowable combinations of thermal power and total core flow shall be restricted to Curve 1 shown in Figure 3.3.1.

SURVEILLANCE REQUIREMENT <

4.3 REACTIVITY CONTROL E. Reactivity Anomalies During startups following refueling outages, the critical rod configura-tions will be compared to the expected configurations at selected

{ operating conditions. These comparisons will be used as base data for I reactivity monitoring during subsequent power operation throughout the fuel cycle. At specific power operating conditions, the critical rod configuration will be compared to the configuration expected based upon appropriately corrected past data. This comparison will be made at least every equivalent full power month.

Amendment No. 1 Millstone Unit 1 3/4 3-8

ALLO'.JABLE TDMBINATIONS OF TOTAL CORE FLOW AND POWER LEVEL

, FIGURE 3 3 1 (See Note 1 TS Page B 2-4)

, 140 t

100% Power Line i i 120 100'4 Intercept g Pof,nt (100/87)

K l ! I APR'i Rod Block Line

' (108/100)

(0 58 + 50%) ; ,

100 g 7 (100/100)

/

./

? 80 / ,

/

a. '

. i /

IU A / Typical 100% Power /

100% Flow Load Line 60 (Reference Only)

/

/ I' I l Minlanum Pump e Speed Line l

(Reference Only) i ; i

[ Natural Circulation

. . - - c- -

20 '

j

/

10 20 30 40 50 60 80 70 90 100 110 Core Flow (%)

Hillstone Unit 1 3/4 3-9 Ref. N 4 66 Sept. 1981

LIMITING CONDITION FOR OPERATION 3.4 STANDBY LIQUID CONTROL SYSTEM Applicability:

Applies to the operating status of the standby liquid control system.

Objective:

To assure the availability of the standby liquid control system.

Specification:

A. Normal Operation During periods when fuel is in the reactor, the liquid poison system shall be operable, except when the reactor is in the COLD SHUTDOWN or REFUEL CONDITION and Specification 3.3.A is met, and except as noted in Specification 3.4.8 below.

SURVEILLANCE REQUIREMENT 4.4 STANDBY LIQUID CONTROL SYSTEM Applicability:

Applies to the periodic testing requirements for the standby liquid control system.

Objective:

To verify the operability of the standby liquid control system, i Specification:

A. Normal Operation The operability of the standby liquid control system shall be verified by performance of the following tests:

1. At least once per month Demineralized water shall be recycled to the test tank. Pump minimum flow rate of 32 gpm shall be verified against a system head corresponding to a reactor vessel pressure of 1225 psig.

l Amendment No. 1 Millstone Unit 1 3/4 4-1

SURVEILLANCE REQUIREMENT (Continued) 4.4 STANDBY LIQUID CONTROL SYSTEM 4.4.A.2. At least once during each operating cycle

a. Manually initiate one of the two standby liquid control systems and pump demineralized water into the reactor vessel at or near operating pressures. This test checks explosion of the charge associated with the tested system, proper operation of the valves and pump capacity. The replacement charges to be installed will be selected from the same ba.tch as those tested.

Both systems shall be tested and inspected, including each explosion valve in the course of two operating cycles.

b. Manually initiate each system, except the explosion valve and pump solution in the recirculation path back to the storage tank.

c. Test that the setting of the system pressure relief valves is ,

between 1350 and 1450 psig.

Millstone Unit 1 3/4 4-2

. .g sc -

J LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS Applicability:

Applies to the operational status of the emergency cooling subsystems.

Objective:

To assure adequate cooling capability for heat removal in the event of a

. loss of coolant accident or isolation from the normal reactor heat sink.

Specification: -

! A. Core Spray and LPCI Subsystems

1. Except as specified in 3.5.A.2, 3.5.F.6, 3.5.F.7 and 3.5.F.8,

! both core spray subsystems shall be operable whenever irradiated

fuel is in the reactor vessel.

2. From and after the date that one of the core spray subsystems is made

or found to be inc7erable for any reason, reactor operation is permissible only during the succeeding fifteen days unless such subsystem is sooner made operable, provided that during such

fifteen days all active components of the other core spray subsystem and the LPCI subsystem and both emergency power sources required for j operation of such components, if no external source of power were j available, shall be operable.

3. Except as specified in 3.5.A.4; 3.5.B.3, 4, 5; 3.5.F.6; 3.5.F.7 and 3.5.F.8, the LPCI subsystem shall be operable whenever irradiated fuel is in the reactor vessel.

4. From and after the date that one of the LPCI pumps is made or found

to be inoperable, for any reason, reactor operation is permissible

! only during the succeeding 30 days unless such pump is sooner made -

operable, provided that during such thirty days the remaining active components of the LPCI and containment cooling subsystem and all active components of both core spray subsystems and both emergency power sources required for operation of such components, if no external source of power were available, shall be operable.

5. A maximum of one drywell spray loop may be inoperable for 30 days when reactor water temperature is greater than 212*F.

6. If the requirements of 3.5. A cannot be met, an orderly shutdown of i the reactor shall be initiated and the reactor shall be in the C0' '

l SHUTDOWN or REFUEL condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

! Amendment No. 1 l Millstone Unit 1 3/4 5-1

3 SURVEILLANCE REQUIREMENT l

4.5 CORE AND CONTAINMENT COOLING SYSTEMS Applicability:

Applies to periodic testing of the emergency cooling subsystems.

Objective

To verify the operability of the core and containment cooling subsystems.

Specification:

A. Surveillance of the Core Spray and LPCI Subsystems shall be performed as follows:

1. Core Spray Subsystem Testing: ,

t' Item Frequency 4

a. Simulated Automatic Each Refueling Actuation Test Outage

b. Pump and Valve Per Surveillance Operability Requirement 4.13

~

c. Core Spray header A p instrumentation check Once/ day calibrate Once/3 months test Once/3 mo'ths

2. LPCI Subsystem Testing shall be as specified in 4.5.A.1.a, b and

, c except that three LPCI pumps shall deliver at least 15,000 gpm l against a system head corresponding to a reactor vessel pressure i of 214.7 psia.

l

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

i Amendment No. 1 Millstone Unit 1 3/4 5-2

_ _ _ _ _ . ._ _ _.._ _ _ . ,_.-_- _ _ ___._ - _ ~_ ___. _ __ _ .-_--- _ ___-__. _ .

' LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS B. Containment Cooling Subsystems

1. Except as specified in 3.5.8.2, 3.5.B.3, 3.5.F.6, 3.5.F.7 and 3.5.F.8, both containment cooling subsystems shall be operable whenever irradiated fuel is in the reactor vessel.

2. From and after the date that one of the emergency service water (ESW) subsystem pumps is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding thirty days unless pump is sooner made operable, provided that during such thirty days all other active components of the containment cooling system are operable.

3. From and after the date that one active component in each containment cooling subsystem or a LPCI and ESW in one containment cooling subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding 7 days provided the remaining active components in each containment cooling subsystem, both core spray subsystems and both emergency power sources for operation of such components, if no external source of power were available, shall be operable.

4. From and after the date that one LPCI and one ESW pump in each containment cooling subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding four days provided the remaining active components of the containment cooling subsystems, both core spray subsystems and both emergency power sources for operation of such components, if no external source of power were available, shall be operable.

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

l 6. If the requirements of 3.5.B cannot be met, an orderly shutdown shall i

be initiated and the reactor shall be in the COLD SHUTOOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment No. 1 Millstone Unit 1 3/4 5-3 i

SURVEILLANCE REQUIREMENT 4.5 CORE AND CONTAINMENT COOLING SYSTEMS B. Surveillance of the containment cooling subsystems shall be performed as follows:

. 1. Emergency Service Water Subsystem Testing:

Item Frequency I

a. Pump & Valve Per Surveillance Operability Requirement 4.13 1

h i

I Amendment No. 1 Millstone Unit 1 3/4 5-4

LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS C. FhCI Subsystem

1. Except as specified in 3.5.C.3 below, the FWCI subsystem shall be operable whenever the reactor coolant temperature is greater than 330'F and irradiated fuel is in the reactor vessel.

2. There shall be a minimum of 225,000 gallons of water in the condensate storage tank for operation of the FWCI.

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

4. If the requirements of 3.5.C cannot be met, an orderly shutdown shall be initiated and the reactor coolant temperature shall be less than 330 F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENT i 4.5 CORE AND CONTAINMENT COOLING SYSTEMS C. Surveillance of FWCI Subsystems shall be performed as follows:

1. Item Frequency

a. Pump and valve Per Surveillance operability Requirement 4.13

b. Simulated Automatic Every refueling Actuation Test outage c

2. Once a week the quantity of water in the condensate storage tank shall be logged.

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Millstone Unit 1 3/4 5-5

LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS D. Automatic Pressure Relief (APR) Subsystems

1. Except as specified in 3.5.0.2 below, the APR subsystem shall be operable whenever the reactor coolant temperature is greater than 330*F and irradiated fuel is in the reactor vessel.

2. From and after the date that one of the four relief / safety valves of the automatic pressure relief subsystem is made or found to be inoperable when the reactor coolant temperature is above 330 F with irradiated fuel in the reactor vessel, reactor operation is permissible only during the succeeding seven days unless repairs are completed and the subsystem made fully operable and provided that during such time the remaining automatic pressure relief valves, FWCI subsystem, and gas turbine generator are operable.

3. If the requirements of 3.5.D cannot be met, an orderly reactor shutdown shall be initiated and the coolant temperature shall be less than 330 F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENT 4.5 CORE AND CONTAINMENT COOLING SYSTEMS

0. Surveillance of the Automatic Pressure Relief Subsystem shall be performed as follows:

1. During each operating cycle, the following shall be performed:

a. A simulated automatic initiation of the system throughout its operating sequence, but excluding actual valve opening, and

b. With the reactor at low pressure, each relief valve shall be manually opened until valve operability has been verified by torus water level instrumentation, or by an audible discharge detected by an individual located outside the torus in the vicinity of each relief line.

2. When it is determined that one safety / relief valve of the automatic pressure relief subsystem is inoperable, the actuation logic of the remaining APR valves and FWCI subsystem shall be demonstrated to be operable immediately and daily thereafter.

Amendment No. 1 Millstone Unit 1 3/4 5-6 L

LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS E. Isolation Condenser System

1. Whenever the reactor coolant temperature is greater than 330*F and irradiated fuel is in the reactor vessel, the isolation condenser shall be operable except as specified in 3.5.E.2, and the shell side water level shall be greater than 66 inches.

2. From and after the time that the Isolation Condenser is made or found to be inoperable, for any reason, power operation shall be restricted to a maximum of 40% of full power, i.~e., (804 MW g) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until su:h time the Isolation Condenser is ,dturned f

to service provided that all active components of the core spray subsystems and LPCI subsystems are operable.

3. If the requirements of 3.5.E cannot be met, an orderly shutdown shall be initiated and the reactor coolant temperature shall ba less than 330 F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENT 4.5 CORE AND CONTAINMENT COOLING SYSTEMS E. Surveillance of the Isolation Condenser System shall be performed as follows:

1. Isolation Condensor System Testing:

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

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

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

d. Calibrate vent line radiation monitors quarterly.

e. Motor operated valves shall be tested per surveillance requirement 4.13.

i i

i Amendment No. 1 Millstone Unit 1 3/4 5-7 i

L

LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS F. Minimum Core and Containment Cooling System Availability

1. Except as specified in 3.5.F.2, 3.5.F.3, 3.5.F.7 and 3.5.F.8 below, both emergency power sources shall be operable whenever irradiated fuel is in the reactor.

2. From and after the date that the diesel generator is made or found to be inoperable, for any reason, continued reactor operation is permissible only during the succeeding seven days provided that the gas turbine generator, FWCI, Automatic Pressure Relief Subsystem, all compcnents of the low pressure core cooling and the containment cooling subsystems shall be operable.

3. From and after the date that the gas turbine generator is made or '

found to be inoperable, for any reason, continued reactor operation is permissible only during the succeeding four days provided that the diesel generator, all components of the APR subsystem, all components of the low pressure core cooling and the containment cooling subsystems shall be operable.

4. If the requirements of 3.5.F.1 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

5. Any combination of inoperable components in the core and containment cooling systems shall not defeat the capability of the remaining operable components to fulfill the core and containment cooling functions.

6. Except as specified in 3.5.F.7, when irradiated fuel is in the vessel and the reactor is in the COLD SHUTDOWN CONDITION all low pressure l core and containment cooling subsystems may be inoperable provided that no work is being done which has the potential for draining the reactor vessel.

SURVEILLANCE REQUIREMENT

4. 5 CORE AND CONTAINMENT COOLING SYSTEMS F. Surveillance of Core and Containment Cooling System

1. The surveillance requirements for normal operation are in Section 4.9.

Amendment No. 1 Millstone Unit 1 3/4 5-8

LIMITING CONDITION FOR OPERATION 3.5 CORE AND CONTAINMENT COOLING SYSTEMS I

3.5.F.7. When irradiated fuel is in the reactor vessel and the reactor is in the REFUEL CONDITION a single control rod may be withdrawn and the drive mechanism replaced or fuel removal and replacement may be conducted provided that the following conditions are satisfied.

(a) The reactor vessel head is removed.

(b) The cavity is flooded.

(c) The spent fuel pool gates are removed.

(d) Water level is maintained within the limits of specification 3.10.C.

(e) Either (i) both core spray systems, (ii) both low pressure coolant injection systems, or (iii) one core spray system and one low pressure coolant injection system, each combination being supplied by independent electrical power, shall be operable or available for operation with the respective 4160 volt supply breaker (s) racked out.

(f) With the torus drained,(i) the ECCS configuration required in 3.5.F.7(e) shall be aligned with the condensate storage tank and the condensate storage tank suction valve V7-58 l locked open, (ii) the condensate storage tank shall ;

contain at least 414,000 gallons of usable water. l shall contain at least 383,000 gallons of water.

(g) The minimum electrical power source requirements shall be the same as specified in paragraph 3.7.B.4. l (h) No work will be performed in the reactor vessel other than fuel sipping while a control rod drive housing is open. !

(i) Fuel removal and replacement will not be done without a full complement of control rods.

(j) During fuel movement no work being done which has the potential for draining the vessel.

Amendment No. 1 Millstone Unit 1 3/4 5-9

LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM B0UNDARY C. Coolant Chemistry

1. a. When the reactor is in the STARTUP/ HOT STANDBY or RUN mode with the reactor coolant at any temperature, or in the SHUTDOWN mode with the average reactor coolant temperature greater than 212 F, the reactor coolant specific activity shall be limited to less than 0.2 microcuries per gram DOSE EQUIVALENT I-131.

When the reactor is in the STARTUP/ HOT STANDBY or RUN mode, if the reactor coolant specific activity is greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram, operation in that mode may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided that the cumulative operating time under these circumstances does not exceed 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months in any consecutive 12-month period.

When the reactor is in the STARTUP/ HOT STANDBY or RUN mode with the reactor coolant at any temperature, or in the SHUTDOWN mode with the average reactor coolant temperature greater than 212 F, if the reactor coolant specific activity is greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 during any continuous period of time longer than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or is greater than 4.0 microcuries per gram, have the reacter in COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. When the reactor is in the STARTUP/H0T STANDBY or RUN mode with the reactor coolant at any temperature or in the SHUTDOWN mode with the average reactor coolant temperature greater than 212 F, the reactor coolant specific activity shall be limited to less than 100/E microcuries per gram total activity.

When the reactor is in the STARTUP/ HOT STANDBY or RUN mode with the reactor coolant at any temperature or in the SHUTDOWN mode with the average reactor coolant temperature greater than 212 F, if the reactor coolant specific activity is greater than 100/E microcuries per gram have the reactor in COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM BOUNDARY C. Coolant Chemistry

1. a. A reactor coolant sample shall be taken at least every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and analyzed for dose equivalent I-131 and total activity content.

Amendment No. 1 Millstone Unit 1 3/4 6-5

SURVEILLANCE REQUIREMENT (Continued) 4.6 PRIMARY SYSTEM BOUNDARY.

4.6.C.1.b. The 100/E activity for the reactor coolant shall be calculated every six months.

i

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! Millstone Unit 1 3/4 6-6

LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM BOUNDARY D. Coolant Leakage Any time irradiated fuel is in the reactor vessel, reactor coolant leakage into the primary containment from unidentified sources shall not exceed 2.5 gpm. In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm. If these conditions cannot be met, or if leak rate cannot be determined, initiate an orderly shutdown and have the reactor in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

E. Safety and Relief Valves

1. During power operation or whenever the reactor coolant pressure is greater than 90 psig with irradiated fuel in the reactor vessel, the safety valve function of the six relief / safety valves shall be operable, except as specified in 3.6.E.5 below. (The solenoid activated relief function of the relief / safety valves shall be operable as required by Specification 3.5.D.).

SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM B0UNDARY D. Coolant Leakage Reactor coolant system leakage into the primary containment shall be checked and recorded at least once per day.

E. Safety and Relief Valves

1. Three of the relief / safety valves top works shall be bench checked or replaced with a bench checked top works each refueling outage.

All six valves top works shall be checked or replaced every two refueling outages. The set pressure shall be adjusted to correspond with a steam set pressure of:

No. of Valves Set Point (psig) 1 1095 1 1%

1 1110 1 1%

4 1125 i 1%

Millstone Unit 1 3/4 6-11 I

LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM BOUNDARY E.2. If Specification 3.6.E.1 is not met, initiate an orderly shutdown and have the reactor coolant pressure below 90 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3. When the safety / relief valves are required to be operable per Specification 3.6.E.1, the Valve Position Indication shall be operable. Two of the six channels may be out of service provided backup indication for the affected valves is provided by the Valve Discharge Temperature Monitor.

4. If Specification 3.6.E.3 is not met, reactor operation is permissible only fo'r the succeeding 30 days unless the Valve Position Indication System is made operable sooner.

5. During reactor vessel hydrostatic testing with all control rods

inserted, the safety valve function of at least two of the six safety / relief valves shall be operable.

F. Structural Integrity The structural integrity of the primary boundary shall be maintained as specified in Technical Specification 3.13.

SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM B0UNDARY E.2. At least one of the relief / safety valves shall be disassembled and inspected each refueling outage.

3. During each operating cycle with the reactor at low pressure, each safety valve shall be manually opened until operability has been verified by torus water level instrumentation, or by the Valve Position Indication System, or by an audible discharge detected by an individual located outside the torus in the vicinity of each discharge.

4. The Valve Position Indication System shall be functionally tested once every three months and calibrated once per operating cycle.

Due to the inaccessability of the pressure switches (in the drywell),

the functional test shall consist of a simulated signal into the monitoring channel rather than the instrument.

5. The valve discharge temperature monitor shall be calibrated at least once per operating cycle.

F. Structural Integrity Inservice Inspection and Testing of primary system boundary components shall be performed as specified in Surveillance Requirement 4.13.

Amendment No. 1 Millstone Unit 1 3/4 6-12

LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM B0UNDARY G. Jet Pumps

1. Whenever the reactor is in the STARTUP/ HOT STANDBY or RUN modes, all jet pumps shall be intact and all operating jet pumps shall be operable. If it is determined that a jet pump is inoperable, an

, orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. Flow indication from each of the twenty jet pumps shall be verified

, prior to initiation of reactor startup from a cold shutdown condition.

} 3. The indicated core flow is the sum of the flow indication from each

of the twenty jet pumps. If flow indication failure occurs for two or more jet pumps, immediate corrective action shall be taken. If flow indication cannot be obtained for at least nineteen jet pumps, an orderly shutdown shall be initiated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and the reactor shall be in the COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM BOUNDARY G. Jet Pumps

1. Whenever there is a recirculation flow with the reactor in the STARTUP/ HOT STANDBY or RUN modes, jet pump integrity and operability shall be checked daily by verifying that the following two conditions do not occur simultaneously:

a. The recirculation pump flow differs by more than 10% from the established speed-flow characteristics.

b. The indicated total core flow is more than 10% greater than the core flow value derived from established power-core flow relationships.

2. Additionally, when operating with one recirculation pump with the equalizer valves closed, the diffuser to lower plenum differential pressure shall be checked daily, and the differential pressure of any jet pump in the idle loop shall not vary by more than 10%

from established patterns.

3. The baseline data required to evaluate the conditions in i

Specification 4.6.G.1 and 4.6.G.2 will be acquired each operating cycle.

l Amendment No. 1 Millstone Unit 1 3/4 6-13

LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM BOUNDARY H. Recirculation Pump Flow Mismatch

1. Whenever both recirculation pumps are in operation, pump speeds shall be maintained within 10% of each other when power level is greater than 80% and within 15% of each other when power level is less than 80%.

2. If Specification 3.6.H.1 cannot be met, one recirculation pump shall be tripped. Operation with a single recirculation pump is per.nitted for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the recirculation pump is sooner made operable.

If the pump cannot be made operable, the reactor shall be in COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3. The reactor shall not be operated unless the equalizer line is isolated.

4. With the mode switch in the STARTUP/ HOT STANDBY or RUN MODE, operation without forced circulation shall not be permitted.

SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM BOUNDARY H. Recirculation Pump Flow Mismatch

1. Recirculation pump speed shall be checked daily for mismatch.

l l

l Amendment No. 1 Millstone Unit 1 3/4 6-14

LIMITING CONDITION FOR OPERATION

3. 6 PRIMARY SYSTEM B0UNDARY I. SNUBBERS

1. The snubbers listed in Tables 3.6.1.a and 3.6.1.b are required to protect the reactor coolant system or other safety related systems or components.

2. During all modes of operation except COLD SHUTDOWN and REFUEL CONDITION, all snubbers shall be OPERABLE.

3. If a snubber is determined to be inoperable, continued reactor operation is permissible only during the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following such determination unless the snubber is sooner replaced, made OPERABLE, or an engineering evaluation determined the supported system /

component to be OPERABLE with the inoperable snubber.

4. If the requirements of 3.6.I.2 and 3.6.I.3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN or REFUEL CONDITION within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

5. If a snubber is determined to be inoperable while the reactor is in the COLD SHUTDOWN or REFUEL CONDITION, the snubber shall be made OPERABLE, or replaced, prior to reactor startup.

6. Snubbers may be added to or deleted from safety related systems without prior License Amendment to Tables 3.6.1.a and 3.6.1.b provided that safety evaluations, documentation, and reporting are provided in accordance with 10 CFR 50.59 and that a proposed revision to Tables 3.6.1.a and 3.6.1.b is included with the next license amendment request.

SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM B0UNDARY I. SNUBBiRS Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program.

1. Visual Inspection All snubbers shall be visually inspected in accordance with the following schedule.

en m n N.1 Millstone Unit 1 3/4 6-15

' o {

SURVEILLANCE REQUIREMENT (Continued) 4.6.I.1. # of Snubbers Found Inoperable During Inspection or During Next Required Inspection Interval Inspection Interval 0 18 months i 25%

1 12 months i 25%

2 6 months i 25%

3,4 124 days i 25%

5,6,7 62 days i 25%

28 31 days i 25%

The required inspection interval shall not be lengthened more than one step at a time.

Snubbers may be categorized in two groups, mechanical and hydraulic.

Each group may be divided into two subgroups; those accessible and those inaccessible during reactor operation. Each group and subgroup may be inspected independently in accordance with the above schedule.

2. Visual Inspection Acceptance Criteria Visual inspections shall be conducted in the following manner:

(a) in the case where the attachments to tra foundation or supporting structure are found to be insecure, the cause of the rejection shall be clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; or (b) if there are visual indications of damage or impaired OPERABILITY, the affected snubber shall be tested in the as-found condition and, if determined OPERABLE per specifications 4.6.I.4(a) or 4.6.I.4(b), as applicable, may be excluded from the number of snubbers counted as having failed the visual inspection.

Snubbers which appeared inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval provided that the above procedure is adhered to.

3. Snubber Tests At least once per eighteen (18) months, during shutdown, a representative sample (10% of the total of each type of snubbers, mechanical and hydraulic, in use in the plant) shall be tested either in place or in a bench test. For each snubber that does not meet the test acceptance criteria of Specification 4.6.I.4(a) or 4.6.I.4(b), as applicable, an additional 10% of that type of snubber shall be tested. Tables 3.6.1.a and 3.6.1.b may be used jointly or separately as the basis for the sampling plan.

Millstone Unit 1 3/4 6-16

l LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM BOUNDARY J. Condensate Demineralizers

1. Regeneration of and/or replacement of a condensate demineralizing l resin charge shall occur before the unused capacity of the resin reaches a minimum value of 30 pounds as chloride ions. .

2. If the charge is regenerated, anion resin in the condensate demineralizing system shall have a minimum salt-splitting capacity of 0.75 milliequivalents per milliliter in the wet, chloride form.

Anion resins which do not have a capacity of 0.75 milliequivalents per milliliter will be replaced with new resin as will the cation re, sin which occupies the same bed.

3. If the charge is replaced, the new anion resin shall have a minimum salt-splitting capacity of 1.2 milliequivalents per milliliter in the wet, chloride form.

4. At least one condensate demineralizer influent conductivity l instrument shall be operable.

5. Whenever a demineralizer is on-line, the conductivity of either its l effluent or the condensate-booster pump discharge shall be continuously monitored.

6. Flow rate and/or integrating flow instrumentation shall be operable l and recorded for each demineralizer.

K. Mechanical Condenser Vacuum Pump

1. The mechanical condenser vacuum pump shall be capable of being isolated and secured on a signal of high radioactivity whenever the main steam line isolation valves are open.

SURVEILLANCE REQUIREMENT 4.6 PRIMARY SYSTEM BOUNDARY s J. Condensate Demineralizers

1. The percent of the remaining ion exchange capacity of the anion resins shall be calculated and logged: ,

a. Weekly when the influent conductivity is between 0.055 and 0.3 umho/cm;

b. Daily when the influent conductivity is equal to or greater than 0.3 umho/cm.

Amendment No. 1 Millstone Unit 1 3/4 6-21

SURVEILLANCE REQUIREMENT (Continued) 4.6 PRIMARY SYSTEM BOUNDARY 4.6.J.2. All REGENERATED condensate demineralizer charges shall have the anion resin analyzed quarterly for salt-splitting capacity.

3. If resin is replaced instead of regenerated, new samples of anion resin shall be analyzed for salt-splitting capacity as follows:

a. At least once per year or at each replacement, whichever is longer, if resin is replaced with material of the same type.

b. Prior to use in the condensate demineralizers if the type of anion resin changed.

K. Mechanical Condenser Vacuum Pump. ,

At least once during each operating cycle, verify automatic securing and isolation of the mechanical condenser vacuum pump.

1 Amendment No. 1 Millstone Unit 1 3/4 6-22 s

l LIMITING CONDITION FOR OPERATION 3.7 CONTAINMENT SYSTEMS Applicability Applies to the operating status of the primary and secondary containment systems.

Objective To assure the integrity of the primary and secondary containment systems.

Specification A. Primary Containment

1. Suppression Chamber Water Level and Temperature The volume a.:d temperature of the water in the suppression chamber shall be maintained within the following limits whenever primary containment is required:

a. Maximum water volume 100,400 ft3 (corresponding to a downcomer submergence of 3.33 ft. at 1.0 psid)

b. Minimum water volume 98,000 ft3 (corresponding to a downcomer submergence of 3.0 ft. at 1.0 psid)

c. Maximum water temperature:

(1) During normal power operation - 90 F.

(2) During testing which adds heat to the suppression pool, the water temperature shall not exceed 10 F above the normal power operation limit specified in (1) above. In connection with such testing, the pool temperature must be reduced to below the normal power operation limit specified in (1) above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(2) The reactor shall be scrammed from any operating condition if the pool temperature reaches 110*F. Power operation shall not be resumed until the pool temperature is reduced below the normal power operation limit specified in (1) above.

(4) During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig at normal cooldown rates if the pool temperature reaches 120 F.

d. At least one of the two existing narrow range torus water level monitoring systems shall be operable whenever primary containment is required, except as specified in 3.7.A.1.e.

Millstone Unit 1 3/4 7-1

LIMITING CONDITION FOR OPERATION (Continued) 3.7 CONTAINMENT SYSTEllS 3.7.A.1.e. If the torus water level monitoring system is disabled and cannot be restored in six (6) hours, an orderly shutdown shall be initiated ar.d the reactor shall be in COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the level monitoring system is made operable.

SURVEILLANCE REQUIREMENT 4.7 CONTAINMENT SYSTEMS Applicability Applies to the primary and secondary containment integrity.

Objective To verify the integrity of the primary and secondary containment.

Specification A. Primary Containment

1. The suppression chamber water level and bulk temperature shall be checked once per shift. The interior painted surfaces above the water line of the pressure suppression chamber shall be inspected at each refueling outage.

a. Whenever there is indication of relief valve operation which adds heat to the suppression pool, the bulk pool temperature shall be continually monitored and also observed and logged every 5 minutes until the hea'. addition is terminated.

b. Whenever there is indication of relief valve operation with the local temperature of the suppression pool reaching 200 F or more an external visual examination of the suppression chamber shall I be conducted before resuming power operation.

c. Torus water level instrumentation shall be calibrated once per 6 months if both systems are operable.

d. Torus water level instrumentation shall be calibrated once per month if only one system is operable.

l Amendment No. 1 Millstone Unit 1 3/4 7-2

LIMITING CONDITION FOR OPERATION 3.7 CONTAINMENT SYSTEMS 3.7.A.2. Drywell to Suppression Chamber Differential Pressure

a. Differential pressure between the drywell and suppression chamber shall be maintained equal to or greater than 1.0 psid, except as specified below.

(1) The differential pressure shall be established within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of entering the RUN mode, and may be reduced to less than 1.0 psid 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a scheduled shutdown.

(2) The differential pressure may be reduced to less than 1.0 psid for a maximum of four (4) hours during required operability testing of the torus / reactor building and drywell/ torus vacuum breakers, and during venting and purging of the containment.

(3) Differential pressure may be less than 1.0 psid for a period not to exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for purposes of conducting a drywell entry.

(4) If the provisions of (1) and (2) above cannot be met, the differential pressure shall be restored within the subsequent six (6) hour period or the provisions of 3.7.A.7 shall apply,

b. At least one (1) drywell to torus differential pressure monitoring system shall be operable whenever primary containment is required, except as specified in 3.7.A.2.c.

c. If the drywell to torus differential pressure monitoring system is disabled and cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the differential pressure monitoring system is made operable.

SURVEILLANCE REQUIREMENT 4.7.A.2. Drywell to Suppression Chamber Differential Pressure

a. The differential pressure between the drywell and suppression chamber shall be recorded once per shift.

b. Drywell to torus differential pressure instrumentation shall be calibrated once per month if only one system is operable.

c. The drywell to torus differential pressure instrumentation shall be calibrated once per six months if both systems are operable.

Amendment No. 1 Millstone Unit 1 3/4 7-3

LIMITING CONDITION FOR OPERATION 3.7 CONTAINMENT SYSTEMS 3.7.A.3. Primary containment integrity, as defined in Section 1.0, shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212*F and fuel is in the reactor vessel, except while performing low power physics test at atmosphere pressure during or after refueling at power levels not to exceed 5 Mw(t).

SURVEILLANCE REQUIREMENT 4.7.A.3. The primary containment integrity shall be demonstrated as follows:

a. Integrated Primary Containment Leak Test (IPCLT)

The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972 and BN-TOP-1

1. Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40 1 10 month intervals during shutdown at P (43 psig) during each ten year service period. Thefhirdtestofeachsetshallbeconducted during the shutdown for the ten year plant inservice inspection.

2. If any periodic Type A test fails to meet 0.75 L , the test .

scheduleforsubsequentTypeAtestsshallbere0iewedand approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L a Type A test shall be performed at least every 18 months 6n,til two consecutive Type A tests meet 0.75 L,, at which time the above schedule may be resumed.

3. The accuracy of each Type A test shall be verified by a supplemental test which:

a. Confirms the accuracy of the test by verifying that ,

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

b. Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test,

c. Requires the quantity of gas injected into containment or bled from containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at P,.

Millstone Unit 1 3/4 7-4 ;

i

LIMITING CONDITION FOR OPERATION 3.7 CONTAINMENT SYSTEMS 3.7.A.4. Pressure suppression chamber - reactor building vacuum breakers:

a. Except as specified in 3.7.A.4.b below, two pressure suppression chamber-reactor building vacuum breakers shall be operable at all times when primary containment integrity is required. The set-point of the differential pressure instrumentation which actuates the pressure suppression chamber-reactor building vacuum breakcrs shall be from 0.4 to 0.5 psid.

b. From and after the date that one of the pressure suppression chamber-reactor building vacuum breakers is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless such vacuum breaker is sooner made operable, provided that the repair procedure does not violate primary containment integrity.

SURVEILLANCE REQUIREMENT 4.7.A.4. Pressure suppression chamber - reactor building vacuum breakers:

a. The pressure suppression chamber-reactor building vacuum breakers and associated instrumentation, including setpoint, shall be checked for proper operation every three months.

l Millstone Unit 1 3/4 7-7

LIMITING CONDITION FOR OPERATION 3.7 CONTAINMENT SYSTEMS 3.7.A.5. Pressure suppression chamber - drywell vacuum breakers:

a. When primary containment is required, all suppression chamber-drywell vacuum breakers shall be operable except during testing and as stated in Specification 3.7.A.5.b and c below.

Suppression chamber-drywell vacuum breakers shall be operable if:

(1) The valve is demonstrated to open fully with the applied force at all valve positions not exceeding that equivalent to 0.5 psi acting on the face of the valve disc.

(2) The valve can be closed by gravity, when released after being manually opened, to within not greater than 0.075 inch or less, as measured at the bottom of the valve disc.

(3) The position alarm system will annunciate in the control room if any valve opening exceeds the equivalent of 0.075 inch as measured at the bottom of the disc.

b. Up to two (2) of the ten (10) suppression chamber-drywell vacuum breakers may be determined to be inoperable provided that they are secured, or known to be in the closed position.

c. If Specification 3.7.A.5.a or b cannot be met, the situation shall be corrected within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the reactor shall be placed in the COLD SHUTOOWN or REFUEL CONDITION within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

6. Oxygen concentration:

a. After completion of the startup test program and demonstration of plant electrical output, the primary containment atmosphere shall be reduced to less than 4% oxygen with nitrogen gas whenever the reactor coolant pressure is greater than 90 psig and during reactor power operation except as specified in 3.7.A.6.b or 3.7.A.6.c.

b. Within the 24-hour period subsequent to placing the reactor in the RUN mode following a shutdown, the containment atmosphere oxygen concentration shall be reduced to less than 4% by volume and maintained in this condition. Deinerting may commence 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a shutdown.

c. Oxygen concentration may be greater than 4% by volume for a period not to exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for purposes of conducting drywell entries relating to testing, surveillances, or maintenance on equipment important to safety.

7. If the specifications of 3.7.A cannot be met, initiate an orderly shutdown and have the reactor in the COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment No. 1 Millstone Unit 1 3/4 7-8

LIMITING CONDITION FOR OPERATION 3.7 CONTAINMENT SYSTEMS C. Secondary Containment

1. Secondary containment integrity, as defined in Section 1, shall be maintained during all modes of plant operation except when all of the following conditions are met.

a. The reactor is in the COLD SHUTDOWN condition and Specification 3.3.A is met.

b. The Fuel Cask or Irradiated Fuel is not being moved within the Reactor Building.

SURVEILLANCE REQUIREMENT 4.7 CONTAINMENT SYSTEMS l

C. Secondary Containment l

1. Secondary containment surveillance shall be performed as indicated below:

a. A secondary containment capability test shall be conducted after isolating the reactor building and placing either standby gas treatment system filter train in operation. Such tests shall demonstrate the capability of the secondary containment to maintain a 1/4 inch of water vacuum with a filter train flow rate of not more than 1100 scfm. Secondary containment capability shall be demonstrated at three or more points within the containment prior to fuel movement and may be demonstrated up to 10 days prior to fuel movement. Secondary containment capability need not be demonstrated more than once per operating cycle unless damage or modifications to the secondary containment have violated the integrity of the pressure retaining boundary of that structure.

Amendment No. 1 Millstone Unit 1 3/4 7-13

LIMITING CONDITION FOR OPERATION 3.7 CONTAINMENT SYSTEMS D. Primary Containment Isolation valves

1. During reactor power operating conditions, all isolation valves listed in Table 3.7.1 and all instrument line flow check valves shall be operable except as specified in 3.7.D.2.

SURVEILLANCE REQUIREMENT 4.7 CONTAINMENT SYSTEMS D. Primary Containment Isolation Valves

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

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

b. At least once oer operating cycle, the instrument line flow check valves shall be tested for proper operation.

c. At least once per quarter:

1) All normally open power-operated isolation valves (except for the main steam line power-operated isolation valves) shall be fully closed and reopened.

2) With the reactor power less than 75% of rated, trip main steam isolation valves (one at a time) and verify closure time.

d. At least once per month, the main steam line power-operated isolation valves shall be exercised by partial closure and subsequent reopening.

i

! Amendment No. 1 Millstone Unit 1 3/4 7-14

! LIMITING CONDITION FOR OPERATION I

3.7 CONTAINMENT SYSTEMS 3.7.D.2. In the event any isolation valve specified in Table 3.7.1 becomes inoperable, reactor power operation may continue provided at least one valve in each line having an inoperable valve is in the mode corresponding to the isolated condition.

3. If Specification 3.7.0 cannot be met, initiate an orderly shutdown and have reactor in the COLD SHUTDOWN or REFUEL CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENT 4.7.D.2. Whenever an isolation valve listed in Table 3.7.1 is inoperable, the position of at least one other valve in each line having an inoperable valve shall be recorded daily.

Amendment No. 1 Millstone Unit 1 3/4 7-15 l

TABLE 3.7.1 PRIMARY CONTAINMENT ISOLATION Issittion Valve (Valve Number) Number of Power Grsup Identification Operated Valves Maximum Action on Operating Initiating Inboard Outboard Time (Sec) Position Signal 1 Main Steam Line Isolation (MS-1A, 2A, 18, 28, IC, 2C, 4 4 35T55 0 GC ID, 2D) 1 Main Steam Line Drain (MS-5) 1 35 C SC 1 Main Steam Line Drain (MS-6) 1 35 C SC 1 Recirculation Loop Sample Line (SM-1, 2) 1 1 5 C SC 1 Isolation Condenser Vent to Main Steam Line (IC-6, 7) 2 5 0 GC 2 Drywell Floor Drain (55-3, 4) 2 20 0 GC 2 Drywell Equipment Drain (55-13, 14) 2 20 0 GC 2 Drywell Vent (AC-7) 1 10 C SC 2 Drywell Vent Relief (AC-9) 1 15 C SC 2 Drywell and Suppression Chamber Vent from Reactor 1 10 C SC Building (AC-8) 2 Drywell Vent to Standby Gas Treatment System (AC-10) 1 10 C SC 2 Suppression Chamber Vent (AC-11) 1 10 C SC 2 Suppression Chamber Vent Relief (AC-12) 1 15 C SC 2 Suppression Chamber Supply (AC-6) 1 10 C SC 2 Drywell Supply (AC-5) 1 10 C SC 2 Drywell and Suppression Chamber Supply (AC-4) 1 10 C SC 3 Cleanup Demineralizer System (CU-2) 1 18 0 GC 1 3 Cleanup Demineralizer System (CU-3, 28) 2 IB 0 GC 3 Shutdown Cooling System (50-1) 1 48 C SC 3 Shutdown Cooling System (SD-2A, 2B, 4A, 4B) 4 48 C SC 3 Shutdown Cooling System (5D-5) 1 48 C SC 3 Reactor Head Cooling Line (HS-4) 1 45 C SC 4 Isolation Condenser Steam Supply (IC-1) 1 24 0 GC 4 Isolation Condenser Steam Supply (IC-2) 1 24 0 GC 4 Isolation Condenser Condensate Return (IC-3) 1 19 C SC 4 Isolation Condenser Condensate Return (IC-4) 1 19 0 GC l Feedwater Check Valves (FW-9A, 10A, 98, 10B) 2 2 NA 0 Process Control Rod Hydraulic Return Check Valves (301-95, 98) 1 1 NA 0 Process Reactor Head Cooling Check Valves (HS-5) 1 NA C Process Standby Liquid Control Check Valves (SL-7, 8) 1 1 NA C Process 3 Cleanup Demineralizer System (CU-5) 1 18 C SC ,

Millstone Unit 1 3/4 7-16 Amendment No. 1 *

TABLE 3.8-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum # Alarm Instrument Operable Setpoints Applicability Action

1. Main Condenser Augmented Offgas Treatment System Explosive Gas Monitor (For System Designed to Withstand Effects of a Hydrogen Explosion)

(a) Hydrogen Monitor 1 Yes ** A

2. Condenser Air Ejector Noble Gas Activity Monitor (a) SJAE Off-Gas Monitor 2 Yes B

3. MP1 Main Stack

. (a) Noble Gas Activity 1 Yes C Monitor (b) Iodine Sampler 1 No

D (c) Particulate Sampler 1 No D (d) Stack Flow Rate 1 No E Monitor (e) Sampler Flow Rate 1 Yes

E Monitor At all times which means that channels shall be OPERABLE and in service on a continuous, uninterrupted basis, except that outages are permitted, within the time frame of the specified action statement, for the purpose of maintenance and performance of required tests, checks and calibrations.

- During augmented off gas treatment system (recombiner) operation.

Amendment No. 1 Millstone Unit 1 3/4 8-7

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _J

'~

. l 1

TABLE 3.8-2 (Continued)

ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by a Minimum Channels OPERABLE requirement, operation of the main condenser augmented offgas treatment system may continue provided that best efforts are made to repair the instrument and that gas samples are collected once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed for hydrogen within the ensuing 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Action B With one monitor inoperable, releases via this pathway may continue provided the inoperable monitor is placed in the tripped position. With both monitors inoperable, releases may continue for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the augmented gas system is not bypassed and the main stack monitor is operable, otherwise, be in at least HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that grab samples are taken once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Action D With the number of samplers OPERABLE less than required by the Minimum number OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that samples are continuously collected with auxiliary sampling equipment for periods of seven (7) days and analyzed for principal gamma emitters with half lives greater than 8 days within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after the end of the sampling period.

Action E

! With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and l

that the flow rate is estimated once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l Millstone Unit 1 3/4 8-8 l

LIMITING CONDITION FOR OPERATION 3.8 RADI0 ACTIVE MATERIALS 3.8.C.2. Liquid Effluents - Dose

1. The dose or dose commitment to any REAL MEMBER OF THE PUBLIC from radioactive materials in liquid effluents from Unit 1 released from the site (See Figure 3.8-1) shall be limited:

a. During any calendar quarter to 5 1.5 mrem to the total body and to 5 5 mrem to any organ; and,

b. During any calendar year to 5 3 mrem to the total body and to 5 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluent exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and the calendar year so that that cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 3 mrem to the total body and 10 mrem to any organ.

SURVEILLANCE REQUIREMENT 4.8 RADI0 ACTIVE MATERIALS 4.8.C.2. Liquid Effluent - Dose

1. Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with Section II of the REM 0DCM once per 31 days.

2. Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in the REM 0DCM.

I Amendment No. 1 Millstone Unit 1 3/4 8-13

- ~ . - . . . .

LIMITING CONDITION FOR OPERATION 3.8 RADI0 ACTIVE MATERIALS 4

0. Radioactive Gaseous Effluents

1. Gaseous Effluents - Dose Rate

1. The instantaneous dose rate offsite (see Figure 3.8-1) due to

. radioactive materials released in gaseous effluents from the l

site shall be limited to the following values:

! a. The dose rate limit for noble gases shall be 6 500 mrem /yr to the total body and 5 3000 mrem /yr to the skin; and,

b. The dose rate limit for Iodine-131, Iodine-133, Tritium, and for all radioactive materials in particulate form with

, half lives greater than 8 days shall be 5 1500 mrem /yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate (s) exceeding the above limits, decrease the release rate j to comply with the limit (s) given in Specification 3.8.D.1.1 within 15 minutes.

- SURVEILLANCE REQUIREMENT l

4.8 RADI0 ACTIVE MATERIALS D. Radioactive Gaseous Effluents

1. Gaseous Effluents - Dose Rate

1. The instantaneous release rate corresponding to the above dose rate shall be determined in accordance with the methodology Section II of the REM 0DCM.

I 2. The instantaneous release rate shall be monitored in

, accordance with the requirements of Table 3.8-2.

3. Sampling and analysis shall be performed in accordance with i Section I of the REM 0DCM to assure that the limits of l Specification 3.8.D.1.1 are met. '

e Millstone Unit 1 3/4 8-14

.i

.,__,m.____. , . . . . . . . , ,_ .._ .. _ _._._

LIMITING CONDITION FOR OPERATION 3.9 AUXILIARY ELECTRICAL SYSTEM Applicability:

Applies to the auxiliary electrical power system.

Objective:

To assure an adequate supply of electrical power during plant operation.

Specification:

A. The reactor shall not be made critical unless all of the following conditions are satisfied:

1. One 345 kv line, associated switchgear, and auxiliary startup transformer capable of automatically supplying auxiliary power.

2. Both emergency power sources are operable.

3. An additional source of power consisting of one of the following:

a. The 27.6 kv line, associated switch gear, shutdown transformer to supply power to the emergency 4160 volt buses,

b. One 345 kv line fully operational and capable of carrying auxiliary power to the emergency buses.

4. 4160 volt buses five and six are energized and the associated 480 volt buses are energized.

5. All station and switchyard 24 and 125 volt batteries and associated battery chargers are operable.

SURVEILLANCE REQUIREMENT 4.9 AUXILIARY ELECTRICAL SYSTEM Applicability:

Applies to the periodic testing requirements of the auxiliary electrical system.

Objective:

Verify the operability of the auxiliary electrical system.

l l

Millstone Unit 1 3/4 9-1

- I

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SURVEILLANCE REQUIREMENT (Continued) l 4.9 AUXILIARY ELECTRICAL SYSTEM Specification:

A. Emergency Power Sources

1. Diesel Generator

a. The diesel generator shall be started and loaded once a month to demonstrate operational readiness. The test shall continue until the diesel engine and the generator are at equilibrium temperature at full load output. During this test, the diesel starting air compressor will be checked for operation and its ability to recharge air receivers.

o. During each refueling outage, the conditions under which the diesel generator is required will be simulated and a test conducted to demonstrate that it will start and be ready to accept load within 13 seconds.

c. During the monthly generator test, the diesel fuel oil transfer pumps shall be operated.

2. Gas Turbine Generator

a. The gas turbine generator shall be fast started and the output breaker closed within 48 seconds once a month to demonstrate l operational readiness. The test shall continue until the gas ,

turbine and generator are at equilibrium temperature at full load output. Use of this unit to supply power to the system electrical network shall constitute an acceptable demonstration of operability.

, b. During each refueling outage, the conditions under which the l gas turbine generator is required will be simulated and a test l conducted to verify that it will start and be able to accept .

emergency loads within 48 seconds.

I l

l I

Amendment No. 1 Millstone Unit 1 3/4 9-2

LIMITING CONDITION FOR OPERATION 3.9 AUXILIARY ELECTRICAL SYSTEM B. When the mode switch is in RUN, the availability of power shall be as specified in 3.9.A, except as specified below:

1. From and after the date that power is available from only one 345 kv l line, reactor operation is permissible only during the succeeding seven days unless an additional 345 kv line is sooner placed in service.

2. From and after the date that incoming power is not available from any 345 kv line, reactor operation shall be permitted provided both emergency power sources are operable. The NRC shall be notified, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , of the precautions to be taken during this situation and the plans for restoration of incoming power. The minimum fuel supply for the gas turbine during this situation shall be maintained above 20,000 gallons.

3. From and after the date that power cannot be made available from the RSST, the plant shall be isolated from the grid within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after which time reactor operation is permissible according to specification 3.9.B.2. During the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period, both emergency power sources shall be operable. The minimum fuel supply for the gas turbine during this situation shall be maintained above 20,000 gallons. If during the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period it is determined that the plant cannot be operated isolated from the grid, be in at least HOT STANDBY within the next six (6) hours and in COLD SHUTDOWN within the following thirty (30) hours.

4. From and after the date that either emergency power source or its associated bus is made or found to be inoperable for any reason, I reactor operation is permissible according to Specification 3.5.F/ g 4.5.F unless such emergency power source and its bus are sooner made

! operable, provided that during such time two offsite lines (345 or 27.6 kv) are operable.

l l S. From and after the date that one of the two 125 volt or 24 volt l l battery systems is made or found to be inoperable for any reason I reactor operation is permissible only during the succeeding seven days unless such battery system is sooner made operable.

l SURVEILLANCE REQUIREMENT 4.9 AUXILIARY ELECTRICAL SYSTEM i

l B. Batteries l

l 1. Station Batteries l

t I

1 Amendment No. 1 Millstone Unit 1 3/4 9-3

SURVEILLANCE REQUIREMENT (^,ntinued) 4.9 AUXILIARY ELECTRICAL SYSTEM 4.9.B.1.a. Every week the specific gravity and voltage of the pilot cell and temperature of adjacent cells and overall battery voltage shall be measured.

b. Every three months the measurements shall be made of voltage of each cell to nearest 0.01 volt, specific gravity of each cell, and temperature of every fifth cell.

c The following tests will be performed in accordance with IEEE Standard 450-1975 "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."

(1) At least once every refuel outage, a battery service test will be performed in accordance with section 5.6 of IEEE Standard 450-1975 to verify that the battery capacity is adequate to supply and maintain in operable status all of the actual emergency loads for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

(2) At least once every 60 months, during shutdown, a performance discharge test will be performed in accordance with Section 5.4 of IEEE Standard 450-1975 to verify that the battery capacity is at least 80 percent of the manufacturer's rating. Once per 60-month interval, this performance discharge test may be performed in lieu of the battery service test.

2. Switchyard Batteries

a. Every week the specific gravity and voltage of the pilot cell and temperature of adjacent cells and overall battery voltage shall be measured.

b. Every three months the measurements shall be made of voltage of each cell to nearest 0.01 volt, specific gravity of each cell, and temperature of every fifth cell.

I i

Millstone Unit 1 3/4 9-4 -

SURVEILLANCE REQUIREMENT (Continued) 4.12 FIRE PROTECTION SYSTEMS 4.12.A.1.g.2. At least once per 92 days, by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-65, is within the acceptable limits specified in Table 1 of ASTM-D975-74 with respect to viscosity, water content, and sediment.

3. At least once per 18 months by:

. (a) Subjecting the diesel to an inspection in accordance

. with procedures prepared in conjunction with its manufacturer's recommendations for the class of service, and (b) Verifying the diesel starts from ambient conditions on the auto-start signal and operates for 220 minutes while loaded with the fire pump.

h. The fire pump diesel starting 12-volt batteries and charger shall be demonstrated OPERABLE:

1. At least once per 7 days, by verifying that:

(a) The electrolyte level of each battery cell is above the plates, and (b) The individual overall battery voltages are 212 volts.

2. At least once per 92 days, by verifying that the specific gravity is appropriate for continued service of the batteries.

3. At least once per 18 months, by verifying that:

(a) The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and (b) The terminal connections are clean, tight, free of corrosion, and coated with anti-corrosion material.

Millstone Unit 1 3/4 12-3

LIMITING CONDITION FOR OPERATION 3.12 FIRE PROTECTION SYSTEMS 3.12.A.2. With one pump and/or one water supply inoperable, restore the inoperable equipment to OPERABLE status within 7 days or, in lieu of any other report required by Specification 6.6.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the plans and procedures to be used to provide for the loss of redundancy in this system.

3. With two pumps inoperable, establish a continuous fire watch with backup fire suppression equipment for the turbine building within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; restore the system to operable status within 14 days or, in lieu of any other report required by Specification 6.6.1 prepare and submit a special report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the pumps to operable status.

4. With the fire suppression water system otherwise inoperable:

a. Establish a backup fire suppression water system within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

b. Submit a Special Report in accordance with Specification 6.9.2;

1. By telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

2. Confirmed by telegraph, mailgram, or facsimile transmission no later than the first working day following the event, 4

3. In writing within 14 days following the event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status, and

! 4. If 4.a above cannot be fulfilled, place the reactor in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN or REFUEL CONDITION within the following twenty-four (24) hours.

I l

Amendment No. 1 Millstone Unit 1 3/4 12-4

O LIMITING CONDITION FOR OPERATION 3.12 FIRE PROTECTION SYSTEMS B. Spray and/or Sprinkler Systems

1. The following spray and/or sprinkler systems located in the following areas shall be OPERABLE at all times when equipment in the area is required to be OPERABLE:

a. Diesel Generator Room - manually operated

b. Diesel Generator Day Tank Room i

c. Hydrogen Seal Oil Unit

d. Gas Turbine Building - manually operated

e. Condenser Bay

f. Turbine Lubrication - Oil Room Deluge

g. Boiler Room and Machine Shop - wet pipe

h. Bearing Lift Pump and Seal Oil Detraining Tank -

wet pipe

i. Reactor Building 14'6" Elevation - wet pipe

2. With one or more of the above required spray and/or sprinkler systems inoperable, establish a continuous fire watch with backup fire suppression equipment for the unprotected area (s) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; restore the system to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.6.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of inoperability, and the plans and schedule for restoring the system to OPERABLE status.

SURVEILLANCE REQUIREMENT 4.12 FIRE PROTECTION SYSTEMS B. Spray and/or Sprinkler Systems

1. Each of the spray / sprinkler systems in 3.12.8 shall be demonstrated OPERABLE:

a. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

b. At least once per 18 months:

Amendment No. 1 Millstone Unit 1 3/4 12-5

SURVEILLANCE REQUIREMENT (Continued) 4.12 FIRE PROTECTION SYSTEMS 4.12.B.1.b.(1) By performing a system functional test which includes simulated automatic actuation of the system, and:

(a) Verifying that the automatic valves in the flow path actuate to their correct positions on a simulated test signal, and (b) Cycling each valve in the flow path that is not testable during plant operation (except valve 1-F-71) through at least one complete cycle of full travel, and (2) By inspection of the spray headers to verify their integrity, and (3) By inspection of each nozzle to verify no blockage.

c. At least once per 3 years, by performing an air or water flow test through each open head spray / sprinkler header and verifying each open head spray / sprinkler nozzle is unobstructed.

l l

l Millstone Unit 1 3/4 12-6

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LIMITING CONDITION FOR OPERATION 3.12 FIRE PROTECTION SYSTEMS C. Carbon Dioxide and Halon 1301 Systems

1. The following high pressure CO, systems shall be OPERABLE with the storage tanks at least 90% of Tull charge weight whenever equipment in the high pressure CO2 protected areas is required to be OPERABLE.

a. Gas Turbine Enclosure

2. With one or more of the above required high pressure CO, systems

. inoperable, establish a continuous fire watch with backDp fire suppression equipment for the unprotected area (s) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; restore the system to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.6.1, prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

3. The Halon 1301 System for the fire pump house shall be operable, with the bottles connected and available for service and the bottle at 95% of full charge weight.

4. From and after the time the fire pump house Halon System is determined to be inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment for the unprotected equipment and/or area; restore the system to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.6.1, prepare l and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

SURVEILLANCE REQUIREMENT 4.12 FIRE PROTECTION SYSTEMS C. Carbon Dioxide and Halon 1301 Systems

1. Each of the high pressure C02 systems in 3.12.C.1 shall be demonstrated OPERABLE:

a. At least once per 6 months, by verifying CO2 storage tank weight,

b. At least once per 18 months, by:

Amendment No. 1 Millstone Unit 1 3/4 12-7

SURVEILLANCE REQUIREMENT (Continued) 4.12 FIRE PROTECTION SYSTEMS 4.12.C.1.b.(1) Verifying the system, including associated ventilation dampers, actuates automatically upon receipt of a simulated test signal and manually through operator action, and (2) Performance of a visual inspection of the discharge nozzles to assure no blockage.

2. The Halon 1301 System referenced in 3.12.C.3 shall be demonstrated OPERABLE:

a. At least once per 6 months the weight and pressure of the refillable container shall be checked. If the container shows a loss in net weight of more than 5% or the pressure drops to 325 psig it shall be refilled or replaced.

b. At least once per 18 months, by:

(1) Verifying the system, including associated ventilation dampers, actuates automatically upon receipt of a simulated test signal and manually through operator action, and (2) Performance of a visual inspection of the discharge nozzles to assure no blockage.

Millstone Unit 1 3/4 12-8

J O

LIMITING CONDITION FOR OPERATION l

3.12 FIRE PROTECTION SYSTEMS E. Fire Detection Instrumentation

1. The minimum required fire detection instrumentation fcr each fire detection zone shown in Table 3.12.2 shall be OPERABLE whenever equipment in that fire detection zone is required to be OPERABLE.

2. With less than the minimum required number of the fire detection instrument (s) shown in Table 3.12.2 OPERABLE:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , establish a watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, and

b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.6.1, prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrument (s) to OPERABLE status.

SURVEILLANCE REQUIREMENT 4.12 FIRE PROTECTION SYSTEMS E. Fire Detection Instrumentation

1. The fire detection instruments listed in Table 3.12.2 shall be demonstrated OPERABLE at least once per 6 months by performance of an INSTRUMENT FUNCTIONAL TEST with the exception that the functional test may consist of injecting a r.imulated electrical signal into the measurement channel rather than the instrument. Due to the inaccessability of the fire detectors located in the condenser bay, a sample consisting of 1/3 of the detectors per channel will be tested during every refuel outage. The sample test cycle will be completed every third refueling outage.

2. The non-supervised circuits between the above required detection instruments and the control room shall be demonstrated OPERABLE at least once per 31 days, per approved procedures.

Amendment No. 1 Millstone Unit 1 3/4 12-13

c, .

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TABLE 3.12.2 FIRE DETECTION INSTRUMENTS NOTE: No two (2) adjacent detectors inoperable.

HEAT SM0KE MINIMUM MINIMUM TOTAL INSTRUMENTS TOTAL INSTRUMENTS INSTRUMENT LOCATION AVAILABLE REQUIRED AVAILABLE REQUIRED

1. Cable Vault - -

15 12

2. Hg Seal Oil Unit 1 1 - -

3. Condenser Bay 36 25 - -

4. Diesel Generator Room 6 4 - -

5. D/G Fuel Oil Day Tank Room 1 1 - -

6. Gas Turbine Enclosure 3 2 - -

7. RX Bldg (1-FDS-1)

a. R-2A RX Bldg 14'6" N.W. Corner Including CRD Bank 7 5

b. R-28 RX Bldg 14'6" N.E. Corner 4 3 r,

c. R-2C RX Bldg 14'6" S.E. Corner 3 2 l

i

d. R-2D RX Bldg 14'6" 5.W. Corner including CRD Bank 3 2

e. R-3 TIP Room - -

1 1

f. R-5 Shutdown Cooling - -

1 1 Pump Room l 8. RX Bldg (42' elev) 1-FDS-2 i

a. R-17 Clean-up Pump Room - -

3 2

b. R-18 Shut Down Heat Exchanger Room 1 1

c. R-19 RX Bldg 42' N.W. Corner to SE Corner 9 7

9. RX Bldg (Elev. 65' and 82') 1-FDS-3) l l a. R-12 RX Bldg Elev. 65'

! except gated area along North Wall 9 7 I

Millstone Unit 1 3/4 12-14

e s

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I LIMITING CONDITION FOR OPERATION 3.12 FIRE PROTECTION SYSTEMS F. Penetration Fire Barriers

1. All penetration fire barriers (including cable penetration fire C' barriers, fire doors, and fire dampers) which protect safety-related X areas shall be functional whenever safety-related equipment in the area is required to be operable. ,

2. With one or more of the above required penetration fire barriers '~

non-functional, within I hour establish a temporary fire barter of equal effectiveness or establish a continuous fire watch on at least one side of the affected penetration. Restore the fire barrier (s) to functional status within 30 days or, in lieu of any other report required by Specification 6.6.1, prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within the next 30 days outlining the action taken, the cause of the fire barrier (s) being non-functional and the plans and schedule for restorir.n the fire barrier (s) to functional status.

4 SURVEILLANCE REQUIREMENT J

4.12 FIRE PROTECTION SYSTEMS F. Penetration Fire Barriers >

1. Penetration fire barriers shall be verified to be functional:

a. At least once per 18 months, by a visual inspection.

b. Prior to returning a penetratior. fire barrier to functional status following repairs or maintenance, by performance of a visual inspection of the affected penetration fire barrier (s).

Amendment No. 1 ,

Millstone Unit 1 3/4 12-17

y. 4

._ .. s s

[)'.4 STANDBY LIQUID CONTROL SYSTEM BASES s Ai Normal Operation D .

yA s. .

u s

The design' objective of the ligtM control system is to provide the .

s capability of bringing the reactor from full power to a cold, xenon-free shutdown ass'udng that none cof Jhe withdrawn control rods can be inserted.

To meet this' objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of 660 ppm of boron in

., the reactor: core in less than 125 minutes. The 660 ppm concentration in J the reactor core would bring the reactor from full power to a minimum of 2.6% delta K subcritical' condition.considering the hot to cold reactivity '

swing, xenon poisoning,% analyticals biases and uncertainties, etc. '.

1_ An additionalJ5% of boron solution is provided for possible imperfect mixing of thC chemical solution in the reactor coolant. A minimum quantity of'2720 net gallons of colution having a 13.4% sodium pentaborate concentration is required to meet this shutdown requirement. Actual ,

system volume for this quantity is 2960 gallons. (240 gallons are contained below the pump suction and, therefore, cannot be inserted.)

( Ths; time requirement (125 minutes) for insertion of the boron solution waP selected to override the rate of reactivity insertion due to cooldown N

of the reactor following the xenon poison peak 7 For the minimum requ}{ed ~

punging rate' of '32 gallons per minute, the maximum storage volume of tre y L~ boron'soluticn is established as 4190 gallons. p N Boron co'ncentration, solution temperature Nithin the tank and connecting piping including check of tank heater and pipe heat tracing system),p'and volume are checked on a frequency to assure a high reliability of s operation of the system should it ever be required. Experience with'pumo operability indicates that monthly testing is adequate to detect if:

i failures have occurred.

'N'2 o

( ,s Components of the system are checked periodically, as described above, and h a functional test is made of the entire system on a frequency of less A than once during each operating cycle unnecessary. A test of one installed explosive charge is made at least once during each operating cycle to assure that the charges are satisfactory. The replacement charge will be selected from the same batch as the tested charge. A continual check of the firing circuit continuity is provided by pilot lights in the control q, y ,; room.

'O The relief valves in the standby liquid control system protect the system piping and positive displacement pumps, which are nominally designed for

. 1500 psi, from overpressure. The pressure relief valves discharge back

, to the standby liquid control solution tank. '

\

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e Millstoi' Unit 1 B 3/4 4-1 L

l)~'

3.4 STANDBY LIQUID CONTROL SYSTEM BASES B. Operation with Inoperable Components Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made.

C. Boron Requirements The solution saturation temperature of 13.4% sodium pentaborate, by weight, is 59"F. The solution shall be kept at least 10*F above the saturation temperature within the tank and suction piping to guard against boron precipitation. The 10*F margin is included in Figure 3.4.2. Temperature and liquid level alarms for the system are annunciated in the control room.

Pump operability is checked on a frequency to assure a high reliability of operation of the system should it ever be required.

Once the solution has been made up, boron concentration will not vary unless more boron or more water is added or removed. Level indication and alarm indicate whether the solution volume has changed which might indicate a possible solution concentration change. Considering these factors, the test interval has been established.

Millstone Unit 1 B 3/4 4-2

" 3.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES A. Core Spray and LPCI This specification assures that adequate emergency cooling capability is available.

Based on the loss of coolant analysis included in Section VI FSAR, either of the two core spray subsystems provides sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident and to limit fuel clad temperature (around 2000 F) to well below the clad melting 1 temperature to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%.

In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additionally conservative in that no credit is taken for spray coolant entering the reactor before the coolant temperature has fallen to 330 F (90 psig).

The LPCI subsystem is designed to provide emergency cooling to the core by flooding in the event of a loss of coolant accident. This system is completely independent of the core spray subsystem; however, it does function in combination with the core spray system to prevent excessive fuel clad temperature. The LPCI subsystem, in combination with the core spray subsystem, provides adequate cooling for break areas of approximately 0.2 square feet up to and including 5.8 square feet, the latter being the double-ended recirculation line break, without assistance from the high pressure emergency core cooling subsystems.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in Reference (1). Using the results developed in this reference, the repair period is found to be less than 1/2 the test interval. This assumes that the core spray and LPCI subsystems constitute a 1 out of 3 system, however, the combined effect of the two systems to limit excessive clad temperature must also be considered. The test interval specified in Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the basic risk considering single failures should be less than 45 days and this specification is within this period. For multiple failures, a shorter interval is specified.

Although it is recognized that the information in Reference (1) provides a quantitative method to estimate allowable repair times, the lack of operating data to support the analytical approach prevents complete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due r'egard to judgment.

(1) APED 5736, " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards," April 1969, I. M. Jacobs and P. W.

Marriott.

Amendment No. 1 Millstone Unit 1 B 3/4 5-1 ;

3.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES Should one core spray subsystem become inoperable, the remaining core spray and the entire LPCI system are available should the need for core cooling arise.

Should the loss of one LPCI pump cccur, a nearly full complement of core and containment cooling equipment is available. Three LPCI pumps in conjunction with one core spray subsystem will perform the core cooling function. Because of the availability of the majority of the core cooling equipment, which will be demonstrated to be operable, a 30-day repair period is justified. If more than one LPCI pump is inoperable, the repair time is set considering the containment cooling function of the LPCI pumps.

B. Containment Cooling Subsystems The two containment subsystems are provided to remove heat energy from _

the containment in the event of a loss-of-coolant accident. Each single containmentcoolingsubsystemincludestwosergicewaterpumps, associated valves, one heat exchanger (40 x 10 BTU /hr), two LPCI pumps and necessary instrumentation, control and power equipment. With two heat exchangers (i.e., both loops) operable, it is possible to degrade system performance to one LPCI and one service water pump operating per loop and still not exceed significantly the equipment design temperatures and not rely completely on containment pressure for net positive suction head (NPSH). An interlock to prevent containment spray actuation is included in the design of engineered safety features to prevent inadvertent pressure reduction below that required for NPSH. The heat removal' capacity of a single cooling loop is adequate to prevent the torus water temperature from exceeding the equipment temperature capability which is specified to be 203 F. It also provides sufficient subcooling so that adequate NPSH could be assured without reliance on containment pressure except for short intervals during the postulated accident. In the event that only one heat removal loop is operable, station operation will be permitted for four days unless necessary repairs are made to make the other loop operable. A four-day period is selected to permit reasonable time for operator action and maintennce operations.

C. Feedwater Coolant Injection Subsystem The feedwater coolant injection subsystem is provided to adequately tool the core for all pipe breaks smaller than those for which the LPCI or core spray subsystems can protect the core. The FWCI meets this requirement without the use of off-site electrical power. For the pipe breaks for which the FWCI is intended to function, the core never uncovers and is continuously cooled, thus no clad damage occurs. The repair times for tha limiting conditions of operation were set considering the use of the FWCI as part of the emergency core cooling system and isolation cooling system.

Millstone Unit 1 8 3/4 5-2

3.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES The FWCI utilizes portions of the normally operating feedwater system; e.g., condensate, condensate booster and feedwater pumps. Therefore, the reliability of the pumps, valves and motors is constantly being demonstrated. Thus the system has an inherently higher degree of reliability than normally non-operating systems. Since an operating string of pump and valves is programmed for FWCI operation, it is not expected that the normally operating portions of the FWCI would be out of operation during normal operation.

D. Automatic Pressure Relief (APR) Systems The relief valves of the automatic pressure relief subsystem are a backup to the FWCI subsystem. They enable the core spray or LPCI to provide protection against the small pipe break in the event of FWCI failure, by depressurizing the reactor vessel rapidly enough to actuate the core sprays or LPCI. The core spray and/or LPCI provide sufficient flow of coolant to limit fuel clad temperature to well below clad melt to assure that core geometry remains intact.

APR testing at low reactor pressure is required during each operating cycle. It has been demonstrated that the blowdown of the APR to the torus causes a wave action that is detectable on the torus water level instrumentation. The discharge of a relief line is audible to an individual located outside the torus in the vicinity of the line, as experienced at other BWR's.

E. Isolation Condenser System The turbine main condenser is normally available. The isolation condenser is provided for care decay heat removal following reactor isolation and scram. The isolation condenser has a heat removal capacity sufficient to l handle the decay heat production at 300 seconds following a scram. Water i will be lost from the reactor vessel through the relief valves in the l 300 seconds following isolation and scram. This represents a minor loss relative to the vessel inventory.

The system may be manually initiated at any time. The system is automatically initiated on high reactor pressure in excess of 1085 psig sustained for 15 seconds. The time delay is provided to prevent unnecessary actuation of the system during turbine trips. Automatic initiation is provided to minimize the coolant loss following isolation from the main condenser. Make-up water to the shell side of the isolation

, condenser can be provided by the condensate transfer pumps from the l condensate storage tank. The condensate transfer pumps are operable from l on-site power. The fire protection system is also available as a supply of make-up water. An alternate method of cooling the core, upon isolation from the main condenser, is by using the relief valves and FWCI subsystem in a feed and bleed manner. The minimum shell side water volume in the isolation condenser is 15,500 gallons.

l Millstone Unit 1 B 3/4 5-3

3.5 CORE AND CONTAINMENT COOLING SYSTEMS BASES The function of the Isolation Condenser during a small break accident is to assist the automatic pressure relief system in depressurizing the reactor as a backup to the FWCI system. The two effects of isolation condenser depressurization are: (1) the minimization of reactor inventory loss which normally occurs during APR blowdown; this reduces the time of core uncovery prior to reflooding; and (2) earlier onset of low pressure core spray cooling.

Analysis performed by General Electric in March 1976, in support of extended operation of Millstone while the isolation condenser was being retubed indicated that from 40% rated power, over 30 minutes is available to initiate operator action to mitigate the consequences of a loss of all feedwater. This is based upon manual depressurization with APR and coolant supplied by all LPCI and core spray subsystems. The FWCI was assumed lost as part of the non-mechanistic assumption of loss of feedwater.

The successful mitigation of this postulated event was no uncovering of the fuel. Operators are instructed regarding special procedures to be utilized during this mode of plant operation.

F. Ernergency Cooling Availability The purpose of Specification F is to assure a minimum of core cooling equipment is available at all times. If, for example, one core spray system were out of service and the emergency power source which powered the opposite core spray system were out of service, only two LPCI pumps would be available. Likewise, if two LPCI pumps were out of service and two emergency service water pumps on the opposite side were also out of service, no containment cooling would be available. It is during refueling outages that major maintenance is performed and during such time that low pressure core cooling systems may be out of service depending on the activities being performed. Specification F allows removal of one CRD mechanism or fuel removal and replacement while the torus is in a drained condition without compromising core cooling capability. The specification establishes the minimum operable low pressure core cooling systems, water inventories, electrical power l supplies and other additional requirements that must exist to allow such activities as CRD mechanism maintenance or fuel removal and replacement l to be performed in parallel with other major activities. The available core cooling capability for a potential draining of the reactor vessel while this work is performed is based on an estimated drain rate and the maintained minimum water level in the refueling cavity to be supplied to the reactor vessel. In addition, the available low pressure core cooling systems are lined up to the condensate storage tank which supplements the reactor cavity water with an additional 450,000 gallons of water. Thus, with the torus drained, a volume of approximtely 800,000 gallons of water will be maintained available to be supplied to the reactor vessel.

Amendment No. 1 Millstone Unit 1 8 3/4 5-4

3.6 PRIMARY SYSTEM B0UNDARY BASES I. Snubbers All snubbers are required CPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained. Snubbers excluded from this inspection program

, are those installed on non-safety related systems and then only if their failure, or failure of the system on which they are installed, would have no adverse effect on any safety related system.

, The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections, performed before the original required time interval has elapsed (nominal time less 25%), may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

When the cause of the rejection of a snubber is clearly established and ,

remedied for that snubber and for any other snubbers that may be generically susceptible, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration.

When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety related component or system has been adversely affected by the inoperability of the snubber.

The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.

To provide assurance of snubber reliability, a representative sample of the installed snubbers will be tested during plant shutdowns at eighteen (18) month intervals. Observed failures of these sample snubbers shall require testing of additional units.

Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above surveillance programs.

Millstone Unit 1 8 3/4 6-7

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3.6 PRIMARY SYSTEM BOUNDARY BASES J. Co,ndensate Demineralizers The criteria of the resin monitoring program and the resin replacement program have been established to protect the reactor from high chloride level should a seawater leak occur in the main condenser. Should a seawater leak occur when a resin has 30 pounds of capacity remaining, this criteria will allow a sufficient buffer for an orderly plant shutdown. Therefore, the resin must be replaced, or regenerated, before the calculated unused capacity of the resin reaches 30 punds of chloride ion. Should a demineralizer be regenerated, the anion minimum salt-splitting capacity of 0.75 milliequivalents per milliliter will ensure that the resin shall be replaced prior to reaching a point where a regeneration provides a recovery of less than 60 percent of its original salt-splitting capacity.

The resin depletion can be calculated using the measured salt-splitting '

capacity, the flow through the bed, and the average influent conductivity.

Based on this result, a depletion can be calculated which will assure a 6 30 pound chloride ion exchange reserve. Regeneration prior to this level of depletion will assure a sufficient ion exchange reserve for removal of chloride from the condensate system.

These factors form the basis for the frequency of sampling, analyzing, calculation and logging surveillance requirements. A yearly or once per replacement sampling frequency will be sufficient to verify supplier resin specifications. A quarterly sampling frequency will be sufficient to detect the slow, long-term degradation of the regenerated resin. As conductivity increases, the calculation and logging will be increased to a weekly basis and ultimately on a daily basis when and if influent conductivity reaches 0.3 umho/cm or greater.

K. Mechanical Condenser Vacuum Pump The purpose of selecting the mechanical condenser vacuum pump line is to limit the release of activity from the main condenser in the unlikely

, event of a control rod drop accident. During the postulated accident, fission products would be transported from the reactor to the main steam lines to the main condenser. The fission product radioactivity would be sensed by the main steamline radioactivity monitors and isolation would be initiated.

1 l

Amendment No. 1 Millstone Unit 1 B 3/4 6-8

. l 3.7 CONTAINMENT SYSTEMS !

BASES containment is normally slightly pressurized during periods of reactor operation assuring no air in-leakage through the primary containment. However, at least once a week, the oxygen concentration will be determined as added assurance.

B. Standby Gas Treatment Systems The standby gas treatment system is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment

, isolation conditions. Both standby gas treatment system fans are designed to automatically start upon containment isolation and to maintain the reactor building pressure to the design negative pressure so that all leakage should be in-leakage. Each of the two fans has 100 percent capacity.

High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers. The charcoal should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 95 percent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelinns for the accidents analyzed. Operation of the fans signify different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

Only one of the two standby gas treatment systems is needed to clean up the reactor building atmosphere upon containment isolation. If one system is found to be inoperable, there is no immediate threat to the containment j system performance, and reactor operation or refueling operation may i continue while repairs are being made. During refueling, two off-site power sources (345 KV or 27.6 KV) and one emergency power source would provide an adequate and reliable source of power and allow completion of annual diesel or gas turbine preventative maintenance.

C. Secondary Containment The secondary containment is designed to minimize any ground level release of radioactive materials which might result from a serious accident. The reactor building provides secondary containment during reactor operation, when the drywell is sealed and in service; the reactor building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the secondary containment is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required.

Amendment Nc. 1 Millstone Unit 1 8 3/4 7-5

3.7 CONTAINMENT SYSTEMS l BASES D. Primary Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment. Closure of one of the valves in each lire would be sufficient to maintain the integrity of the pressure suppression system.

Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.

Millstone Unit 1 B 3/4 7-6

l 3.9 AUXILIARY ELECTRICAL SYSTEM BASES A. The objective of the auxiliary electric power specification is to assure that adequate power will be available to operate the emergency safeguards equipment. Adequate power can be provided by any one of the following power sources: one 345 kv line, the 27.6 kv system, the gas turbine-generator and the diesel generator.

This specification assures that at least two offsite and two onsite power sources will be available before the reactor is started up. In addition to assuring power source operability, all of the associated switch gear and vital equipment must be operable as specified to assure that the emergency cooling equipment can be operated, if required, from the power sources.

B. Normally, three 345 kv lines will be available to provide emergency power to the plant when the reactor is operating. However, adequate power is available with only one 345 kv line in service. Therefore, reactor operation is permitted for up to seven days with only one 345 kv line in service to accommodate necessary maintenance, etc.

In the event that all 345 kv lines are out of service, continued reactor operation is permitted provided both onsite emergency power sources are operable with an adequate fuel supply. Two operational power sources provide an adequate assurance of emergency power availability under these circumstances.

In the event that power cannot be made available from the RSST, continued reactor operation is permitted for the succeeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided both onsite emergency power sources are available with an adequate supply of fuel. Seventy-two hours of reactor operation provides adequate time for repairs of a reasonable nature or to isolate from the grid. When isolated from the grid, the plant is not susceptible to offsite induced transients.

If neither repair nor isolation is possible, an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is allowed for an orderly shutdown. Two operable emergency power sources provide adequate assurance of emergency power availability under these circumstances.

Normally both the gas turbine generator and diesel generator are required to be operable to assure adequate emergency power with no offsite power sources. However, due to the redundancy and reliability of offsite power, one of the two emergency onsite power sources may be out of service for limited periods of time providing two offsite power sources are available during these periods.

C. Either of the two station batteries has enough capability to energize the vital buses and power the other emergency equipment. Due to the high reliability of battery systems, one of the two batteries may be out of service for up to 7 days. This minimizes the probability of unwarranted shutdowns by providing adequate time for reasonable repairs.

Millstone Unit 1 B 1/4 9-1 Amendment No. 1

3.9 AUXILIARY ELECTRICAL SYSTEM BASES D. The diesel fuel supply of 20,000 gallons will supply the diesel generator with about iive days of full load operation. The gas turbine generator fuel supply of 35,000 gallons is sufficient to operate the unit for at least two and one-half days considering the fuel consumption vs. load and load vs. time requirements during the postulated accident. Reference Amendment 18 to the Provisional Operating License. Additional fuel can be supplied to the site within twelve hours.

Millstone Unit 1 B 3/4 9-la Amendment No. 1 l

4.9 AUXILIARY ELECTRICAL SYSTEM BASES A. The monthly test of the diesel generator and gas turbine generator is conducted to check for equipment failures and deterioration. Testing is conducted up to equilibrium operating conditions to demonstrate proper operation at these conditions. Operation of the gas turbine generator for peaking purposes may substitute for the monthly performance check. The units will be manually started, synchronized to the bus, and load picked up. Generator experience at other generating stations indicates that the testing frequency is adequate to assure a high reliability of operation should the system be required. In addition, during the test when the generator is synchronized to the bus it is also synchronized to the off-site power source and thus not completely independent of this source. To maintain the maximum amount of independence, a thirty day testing interval is also desirable.

Both the diesel generator and the gas turbine generator have air compressors and air receber tanks for starting. It is expected that the air compressors will run only infrequently. During the monthly check of the units, the receivers will be drawn down below the point at which the compressor automatically starts to check operation and the ability of the compressors to recharge the receivers. Pressure indicators are provided on each of the receivers.

Following the tests or peaking operation, of the unit and at least weekly, the fuel volume remaining will be checked. At the end of the monthly load test of the diesel generator, the fuel oil transfer pump will be operated to refill the day tank and to check the operation of this pump. Peaking operation shall be controlled so that major maintenance operations on the gas turbine will not be scheduled during an operating cycle.

The test of the diesel and gas turbine generators during the refueling outage will be more comprehensive in that it will functionally test the system; i.e., it will check starting and closure of breakers and sequencing of loads. The units will be started by simulation of a loss of coolant accident. In addition, a loss of normal power condition will be imposed to stimulate a loss of off-site power. The timing sequence will be checked to assure proper loading in the time required. Periodic tests between refueling outages check the capability of the units to run at full load. Periodic testing of the various components plus a functional test at a refueling interval are sufficient to maintain adequate reliability.

B. Although the station and switchyard batteries will gradually deteriorate with time, the surveillance specified is that which will provide an indication of all degradation long before the battery would have insufficient capacity to meet the design load which could be placed upon

) it. Battery cell replacements will be made in accordance with Section 6 of IEEE Standard 450-1972, " Battery Replacement Criteria."

C. Logging the diesel and gas turbine generator fuel supply weekly and after each operation, assures that the minimum fuel supply requirements will be maintained.

Millstone Unit 1 B 3/4 9-2

6.0 /tDMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Station Superintendent shall be responsible for overall operation of the Millstone Station Site while the Unit Superintendent shall be responsible for operation of the unit. The Station Superintendent and Unit Superintendent shall each delegate, in writing, the succession to these responsibilities during their absence.

6.2 ORGANIZATION Offsite 6.2.1 The Offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.

Facility Staff 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Operator shall be in the control room when fuel is in the reactor.

c. At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips,

d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.(Table 6.2-1)

e. ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operatcr Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f. A site Fire Brigade of at least 5 members shall be maintained onsite at all times.(Table 6.2-1) The Fire Brigade shall not include 2 members of the minimum shift crew necessary for safe shutdown of the unit or any personnel required for other essential functions during a fire emergency.

g. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions. These procedures should follow the general guidance of the NRC Policy Statement on working hours (Generic Letter No. 82-12).

Millstone Unit 1 6-1 l

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N111 stone Unit 1 6-3 Amendment No. 1

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Protection Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, Revision 1, and (2) the Shift Technical Advisor who shall have a Bachelor Degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

Millstone Unit 1 6-4

ADMINISTRATIVE CONTROLS Meeting Frequency 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman.

Quorum 6.5.1.5 A quorum of the PORC shall consist of the Chairman, or Vice Chairman,

, or Station Superintendent and four members including elternates.

Responsibilities 6.5.1.6 The PORC shall be responsible for:

a. Review of 1) all procedures, except common site procedures, required by Specification 6.8 and changes thereto, 2) any other proposed procedures, or changes thereto, as determined by the Unit Superintendent to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

c. Review of all proposed changes to Sections 1.0 - 5.0 of these Technical Specifications.

d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technical Specifications and preparation and forwarding of a report covering evaluation j and recommendations to prevent recurrence to the Vice President '

Nuclear Operations and to the Chairman of the Nuclear Review Board.

f. Review of all REPORTABLE EVENTS.

g. Review of facility operations to detect potential safety hazards.

h. Performance of special review and investigations and reports thereon as requested by the Cnairman of the Nuclear Review Board.

i. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

Amendment No. 1 Millstone Unit 1 6-7

ADMINISTRATIVE CONTROLS Authori ty 6.5.1.7 The PORC shall:

a. Recommend to the Unit Superintendent written approval or disapproval of items considered under 6.5.1.6(a) through (d) above.

b. Provide immediate written notification to the Station Superintendent, Vice President Nuclear Operations and the Chairman of the Nuclear Review Board of disagreement between the PORC and the Unit Superintendent; however, the Unit Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

Records

6. 5.1. 8 The PORC shall maintain written minutes of each meeting and copies shall be provided to the Station Superintendent, Vice President Nuclear Operations and Chairman of the Nuclear Review Board.

6.5.2 Site Operations Review Committee (SORC)

Function 6.5.2.1 The 50RC shall function to advise the Station Superintendent on all matters related to nuclear safety of the entire Millstone Station Site.

Composition i

6.5.2.2 The 50RC shall be composed of the:

Chairman: Station Superintendent Member: Unit 1 Superintendent Member: Unit 2 Superintendent Member: Unit 3 Superintendent Member: Station Services Superintendent Member: Designated Member of Unit 1 PORC Member: Designated Member of Unit 2 PORC Member: Designated Member of Unit 3 PORC or Unit No. 3 Operations Supervisor Alternates 6.5.2.3 Alternate members shall be appointed in writing by the SORC Chairman ,

to serve on a temporary basis; however, no more than two alternates shall participate in 50RC activities at one time.

l Millstone Unit 1 6-8

ADMINISTRATIVE CONTROLS Consultants 6.5.3.3 Consultants shall be utilized as determined by the NRB Chairman to provide expert advice to the NRB.

Meeting Frequency 6.5.3.4 The NRB shall meet at least once per six months and as convened i by the NRB Chairman.

Quorum 6.5.3.5 The minimum quorum of the NRB necessary for the performance of the NRB review and audit functions of these technical specifications shall consist of the Chairman, or his designated alternate, and at least four members. No more than a minority of the quorum shall have line responsibility for operation of the facility.

Review 6.5.3.6 The NRB shall review:

a. The safety evaluation for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the l provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.

b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question, as defined in Section 50.53, 10 CFR.

c. Proposed tests or experiments which involve an unreviewed safety question, as defined in Section 50.59, 10 CFR.

d. Proposed changes in Sections 1.0 - 5.0 of these technical specifications or licenses.

e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

I f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

l g. All REPORTABLE EVENTS requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission. .

l

h. Indications of a significant unanticipated deficiency, affecting nuclear safety, in some aspect of design or operation of safety related structures, systems or components.

i Amendment No. 1 l Millstone Unit 1 6-11 l -- _. - _ - - . . - __

ADMINISTRATIVE CONTROLS

i. Report and meetings minutes of the PORC.

Audits 6.5.3.7 Audits of Unit activities shall be performed under the cognizance of the NRB. These audits shall encompass:

a. The conformance of Unit 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 unit staff at least once per 12 months.

c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems, or method of operation that affect nuclear safety, at least once par six months.

d. Any other area of unit operation considered appropriate by the NRB or the Senior Vice President Nuclear Engineering and Operations.

Authority 6.5.3.8 The NRB shall report to and advise the Senior Vice President Nuclear Engineering and Operations on those areas of responsibility specified in Sections 6.5.3.6 and 6.5.3.7. Meeting minutes may be used for this purpose.

Records 6.5.3.9 Records of NRB activities shall be prepared, approved, and distributed as indicated below:

a. Minutes of each NRB meeting shall be prepared, approved and forwarded to the Senior Vice President Nuclear Engineering and Operations within 14 days following each meeting.

, b. Reports of reviews encompassed by Section 6.5.3.6 above, shall

. be prepared, approved and forwarded to the Senior Vice President Nuclear Engineering and Operations within 14 days following completion of the review.

! c. Audit reports encompassed by Section 6.5.3.7 above, shall be I forwarded to the Senior Vice President Nuclear Engineering and l Operations and to the management positions responsible for the

! areas audited within 30 days after completion of the audit.

I Millstone Unit 1 6-12 l

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' ADMINISTRATIVE CONTROLS Records 6.5.4.9 Records of SNRB activities shall be prepared, approved, and forwarded to the Senior Vice President Nuclear Engineering and Operations within 14 days following each meeting.

a. Minutes of each SNRB meeting shall be prepared, approved, and forwarded to the Senior Vice President Nuclear Engineering and Operations within 14 days following each meeting.

b. Reports of reviews encompassed by Section 6.5.4.6 above shall 3 be prepared, approved, and forwarded to the Senior Vice President Nuclear Engineering and Operations within 14 days following completion of the review.

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c. Audit reports encompassed by Section 6.5.4.7 above, shall be forwarded to the Senior Vice President Nuclear Engineering and Operations and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted l pursuant to the requirement of 10 CFR 50.73.

b. Each REPORTABLE EVENT shall be reviewed by the PORC and the results of this review shall be submitted to the NRB and the Vice President of Nuclear Operations.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The unit shall be placed in at least HOT STANDBY within two hours.

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President Nuclear Operations and the NRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PORC. 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.

Amendment No. 1 Millstone Unit 1 6-15

ADMINISTRATIVE CONTROLS

d. The Safety Limit Violation Report shall be submitted to the Commission, the NRB, and the Vice President Nuclear Operations within 14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and i.adintained Covering the activities referenced below.

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

b. Refueling operations.

c. Surveillance activities of safety related equipment.

d. Security Plan implementation

e. Emergency Plan Implementation I

f. Fire Protection Program Implementation

g. Quality Control for effluent monitoring using the guidance in Regulatory Guide 1.21 Rev. 1, June 1974.

h. Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM) implementation, except for Section I.E.,

Radiological Environmental Monitoring.

6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PORC/50RC, as applicable, and approved by the Unit Superintendent / Station Superintendent prior to implementation and reviewed periodically as set forth in each document.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a. The intent of original procedure is not altered.

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.

c. The change is documented, reviewed by the PORC/50RC, as applicable, and approved by the Unit Superintendent / Station Superintendent within 14 days of implementation.

6.8.4 Written procedures shall be established, implemented and maintcined covering Section I.E, Radiological Environmental Monitoring, of the REMODCH.

Millstone Unit 1 6-16

ADMINISTRATIVE CONTROLS 6.8.5 All procedures and procedure changes required for the Radiological Environmental Monitoring Program of 6.8.4 above shall be reviewed by an individual (other than the author) from the Radiological Assessment Branch or the Production Operation Service Laboratory (POSL) and approved by appropriate supervisiun.

Temporary changes may be made provided the intent of the original procedure is not altered and the change is documented and reviewed by an individual (other than the author) from the Radiological Assessment Branch or the POSL, within 14 days of implementation.

6.9 REPORTING REQUIREMENTS Routine Reports 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 Regional Administrator, Region I, U.S. Nuclear Regulatory Commission 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 that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance 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 of characteristics obtained during the test program and a comparison of these values with design predictions and specificatinns. 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.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completien of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e.,

initial criticality, completion of startup test program, and resumption or commencement of commercial power operation),

supplementary reports shall be submitted at least every three months until all three events have been completed.

Millstone Unit 1 6-17

ADMINISTRATIVE CONTROLS Annual Reports 1#

6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

6.9.1.5 A tabulation, on an annual basis, of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their aspciated man-rem exposure according to work and job functions, - e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (described maintenance), waste processing and refueling.

The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

Monthly Operating Report 6.9.1.6 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington D.C. 20555, with a copy to the Regional Administrator, Region I, U.S. Nuclear Regulatory Commission, no later than the 15th of each month following the calendar month covered by the report.

Annual Radiological Environmental Operating Report 6.9.1.7 Routine Annual Radiological Environmental Operating reports l covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The Annual Radiological Environmental Operating Report shall include that information delineated in the REMODCM.

Semiannual Radioactive Effluent Release Report

6. 9.1. 8 Routine radioactive effluent release reports covering the operstion of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.

A supplemental report containing dose assessments for the previous year shall be submitted annually within 90 days after January 1.

1/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

2/ This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.

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Millstone Unit 1 6-18

ADMINISTRATIVE CONTROLS The report shall include that information delineated in the REMODCM.

Any changes to the REM 00CM shall be submitted in the Semiannual Radioactive Effluent Release Report.

Special Reports 6.9.2 Special reports shall be submitted to the Regional Administrator, Region I, U.S. Nuclear Regulatory Commission, within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirement of the applicable reference specification:

a. In-service Inspection Results, Specification 4.6.F.

b. Primary Containment Leak Rate Test Results, Specification 4.7. A.3.

c. Materials Radiation Surveillance Specimen Examination and Results, Specification 4.6.B.3.

d. Fire detection instrumentation, Specification (3.12.E.2)

e. Fire suppression systems, Specifications (3.12.A.2, 3.12.B.2, and 3.12.C.2).

f. Radiological Effluent Reports required by Specifications in 3.8.C.2, 3.8.D.2, 3.8.D.3, and 3.8.D.4.

6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

a. Records and logs of facility operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related

. to nuclear safety.

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c. All REPORTABLE EVENTS.

dJ Records of surveillance activities, inspections, and calibrations required by these technical specifications.

e. Records of reactor tests and experiments.

f. Records of changes made to operating procedures.

g. Records of radioactive shipments.

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Millstone Unit 1 6-19 l i

l

ADMINISTRATIVE CONTROLS

h. Records of sealed source leak tests and results.

i. Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be-retained for the duration of the facility operating license:

a. Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories.

c. Records of facility radiation and contamination surveys.

d. Records of radiation exposure for all individuals entering radiation control areas.

e. Records of gaseous and liauid radioactive material released to the environs.

f. Records of transients or operational cycles for those facility components designed for a limited number of transients or cycles.

g. Records of training and qualification for current members of the plant staff.

h. Records of inservice inspections performed pursuant to these Technical Specifications.

i. Records of quality assurance activities required by the QA Manual.

J. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR Part 50.59.

k. Records of meetings of the PORC, the NRB, the 50RC and the SNRB.

1. Reccrds for environmental qualification.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

Amendment No. 1 Millstone Unit 1 6-20 I

' ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit *. Any individual or grcup of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.

This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit. The surveillance frequency shall be established by the Health Physics Supervisor.

6.12.2 The requirements of 6.12.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Health Physics Supervisor.

6.13 SYSTEMS INTEGRITY The licensee shall implement a program to reduce leakage from systems outside containment that would, or could, contain highly radioactive fluids during a serious transient, or accident, to as low as practical levels. This program -

shall include the following:

1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt for the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.

Amendment No. 1 Millstone Unit 1 6-21

ADMINISTRATIVE CONTROLS

2. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

6.14 I0 DINE MONITORING The licensee shall implement a program which will ensure the capability to .

accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1. Training of personnel,

2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equipment.

6.15 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REM 0DCM)

Section I, Radiological Effluent Monitoring Manual, shall outline the sampling and analysis programs to determine the concentration of radioactive materials released offsite as well as dose commitments to individuals in those exposure pathways and for those radionuclides released as a result of station operation.

It shall also specify operating guidelines for radioactive waste treatment systems and report content.

Changes in Section I shall be submitted to the Commission for approval prior to implementation.

Section II, the Offsite Dose Calculation Manual (ODCM), shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these technical specifications.

Changes to Section II need not be submitted to the Commission for approval prior to implementation, but shall be included in the next Semi-Annual Radioactive Effluent Release Report.

6.16 RADI0 ACTIVE WASTE TREATMENT Procedures for liquid and gaseous radioactive effluent discharges from the Unit shall be prepared, approved, maintained, and adhered to for all operations involving offsite releases of radioactive effluents. These procedures shall specify the use of appropriate

waste treatment utilizing the guidance provided in the REM 0DCM.

The solid radioactive waste treatment system shall be operated in accordance with the Process Control Program to process wet radioactive wastes to meet shipping and burial ground requirements.

Amendment No. 1 Millstone Unit 1 6-22

9 94%

8 UNITED STATES

!I ) e'j NUCLEAR REGULATORY COMMISSION E 9 y WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAP. REACTOR REGULATI0fi SUPPORTING AMENDMENT N0. 1 TO FACILITY OPERATING LICENSE NO. DPR-21 NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION, UNIT N0. 1 DOCKET NO. 50-245

1.0 INTRODUCTION

tiRC Generic Letter 83-43, dated December 19, 1983, discussed revisions to notification and reporting requirements in 10 CFR Part 50.72 and Part 50.73 and requested licensees to revise Technical Specifications to be consistent with the new requirements. By .Proposed Tech Spec Revs Encl|letter dated July 9, 1985]], Northest Nuclear Energy Company submitted a request for a proposed amencment to Appendix A of Operating License No. DPR-21, Millstone Nuclear Power Station, Unit ho.1 to accomplish these revisions.

A Notice of Consideration of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing related to the requested action was published in the Federal Register on August 14, 1985 (50 FR 32791). No comments or requests for hearing were received.

2.0 EVALUATION The proposed revisions include changing the definition of " reportable occurrence" to that of " reportable event," deleting unnecessary and conflicting references to reporting requirements in the limiting conditions for operation and surveillance requirements section, and revising the administrative controls section to reference 10 CFR Parts 50.72 and 50.73 and to delete the previous reporting requirements, now unnecessary or conflicting.

The proposed revisions are administrative in nature since they only revise the reporting requirements for reportable events. The revisions do not involve physical changes in plant safety related systems, components, or structures. The revisions will not increase the likelihood of a malfunction of safety related equipment, will not increase the consequences of an accident previously analyzed, nor create the possibility of a malfunction different from those previously evaluated in the Final Safety Analysis Report. The proposed changes contained several administrative errors that have been corrected with the agreement of the licensee.

3.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change to a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) nc environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such

activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.

Dated: January 29, 1987 Principal Contributor: Robert J. Sumers

ML20210A286

Text

Attachment:

Background

1.0 INTRODUCTION

3.0 ENVIRONMENTAL CONSIDERATION

4.0 CONCLUSION

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