ML17058B775

Proposed Tech Specs,Reflecting Editorial Changes, Administrative Corrections & Retyping of TS

ML17058B775
Person / Time
Site:	Nine Mile Point
Issue date:	05/14/1993
From:	NIAGARA MOHAWK POWER CORP.
To:
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ML17058B774	List: ML17058B773 ML17058B775 ML18040A082
References
NUDOCS 9305280003
Download: ML17058B775 (762)

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RADIOLOGICALTECHNICAL SPECIFICATION APPENDIX A TO FACILITY OPERATING LICENSE NO. DPR-63 FOR THE NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT NUCLEAR STATION UNIT 1 DOCKET NO. 50-220 DECEMBER 26, 1974 9305280003 9305i4 PDR ADDCR 05000220 P PDR

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FOREWORD These revised specifications supersede in their entirety the previous technical specifications and are issued as Appendix A to Full-Term Operating License DPR-63 issued to Niagara Mohawk Power Corporation by the Atomic Energy Commission. The Environmental Technical Specifications are issued as Appendix 8 to License DPR-63.

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NINE MILE POINT NUCLEAR STATION UNIT 1 - TECHNICAL SPECIFICATIONS CONTENTS SECTION DESCRIPTION PAGE 1.0 Definitions 2.0 Safety Limits and Limiting Safety System Setting Safe Limits Limitin Safe S stem Se in 2.1.1 Fuel Cladding Integrity 2.1.2 Fuel Cladding Integrity 2.2.1 Reactor Coolant System 2.2.2 Reactor Coolant System 23 3.0 Limiting Condition for Operation and Surveillance Requirements 27 3.1.0 Fuel Cladding 28 Limitin Condi ion for 0 era i n Surveillance Re uir m n 3.1.1 Control Rod System 4.1.1 Control Rod System 29 3.1.2 Liquid Poison System 4.1.2 Liquid Poison System 3.1.3 Emergency Cooling System 4.1.3 Emergency Cooling System 50 3.1.4 Core Spray System 4.1 4 Core Spray System 3.1.5 Solenoid-Actuated Pressure Relief Valve 4.1.5 Solenoid-Actuated Pressure Relief Valve 60 3.1.6 Control Rod Drive Coolant Injection 4.1.6 Control Rod Drive Coolant Injection 62 3.1.7 Fuel Rods 4.1.7 Fuel Rods 65 3.1.8 High Pressure Coolant Injection 4.1.8 High Pressure Coolant Injection 76 AMENDMENT NO.

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SECTION DESCRIPTION PAGE 3.2.0 Reactor Coolant System 80 Limitin Condi ion for 0 eration Surveillance Re uiremen s 3.2.1 Reactor Vessel Heatup and Cooldown Rates 81 3.2.2 Minimum Reactor Vessel Temperature for 4.2.2 Minimum Reactor Vessel Temperature for 83 Pressurization Pressurization 3.2.3 Coolant Chemistry 4.2.3 Coolant Chemistry 96 3.2.4 Coolant Activity 4.2.4 Coolant Activity 99 3.2.5 Leakage Rate 4.2.5 Leakage Rate 101 3.2.6 Inservice Inspection and Testing 4.2.6 Inservice Inspection and Testing 105 3.2.7 Isolation Valves 4.2.7 Isolation Valves 108 3.2.8 Safety Valves 4.2.8 Safety Valves 118 3.2.9 Solenoid-Actuated Pressure Relief Valves 4.2.9 Solenoid-Actuated Pressure Relief Valves 120 3.3.0 Primary Containment 123 Limitin Condition for 0 eration Surveillance Re uirements 3.3.1 Oxygen Concentration 4.3.1 Oxygen Concentration 124 3.3.2 Pressure and Suppression Chamber Water 4.3.2 Pressure and Suppression Chamber Water 127 Temperature and Level Temperature and Level 3.3.3 Leakage Rate 4.3.3 Leakage Rate 131 AMENDMENT NO.

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SECTION DESCRIPTION PAGE 3.3.4 Isolation Valves 4.3.4 Isolation Valves. 143 3.3.5 Access Control 4.3.5 Access Control 151 3.3.6 Vacuum Relief 4.3.6 Vacuum Relief 153 3.3.7 Containment Spray 4.3.7 Containment Spray 159 3A.O Secondary Containment Limitin Condition for 0 eration Limitin Condition for 0 eration 3 4.1 Leakage Rate 4.4.1 Leakage Rate 165 3 4.2 Isolation Valves 4 4.2 Isolation Valves 168 3 4.3 Access Control 4 4.3 Access Control 170 3.4.4 Emergency Ventilation 4.4.4 Emergency Ventilation 173 3.4.5 Control Room Ventilation 4A.S Control Room Ventilation 178 3.5.0 Shutdown and Refueling 182 Limi in Condi i n for 0 era ion Limi in Condi for era i n 3.5.1 Source Range Monitoring 4.5.1 Source Range Monitoring 183 3.5.2 Refueling Platform Interlock 4.5.2 Refueling Platform Interlock 186 3.6.0 General Reactor Plant 191 Limitin Condi ion for 0 eration Limitin Condition for 0 eration 3.6.1 . Station Process Effluents 4.6.1 Station Process Effluents 192 3.6.2 Protective Instrumentation 4.6.2 Protective Instrumentation 194 3.6.3 Emergency Power Sources 4.6.3 Emergency Power Sources 255 3.6.4 Shock Suppressors (Snubbers) 4.6.4 Shock Suppressors (Snubbers) 259 AMENDMENT NO.

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SECTION DESCRIPTION PAGE 3.6.5 Radioactive Material Sources 4.6.5 Radioactive Material Sources 265 3.6.6 Through 3.6.10 (Deleted) 4.6.6 Through 4.6.10 (Deleted) 3.6.11 Accident Monitoring Instrumentation 4.6.11 Accident Monitoring Instrumentation 268 3.6.12 Reactor Protection System and Reactor 4.6.12 Reactor Protection System and Reactor 274 Trip System Power Supply Monitoring Trip System Power Supply Monitoring 3.6.13 Remote Shutdown Panels 4.6.13 Remote Shutdown Panels 277 3.6.14 Radioactive Effluent Instrumentation 4.6.14 Radioactive Effluent Instrumentation 282 3.6.15 Radioactive Effluents 4.6.15 Radioactive Effluents 295 3.6;16 Radioactive Effluent Treatment Systems 4.6.16 Radioactive Effluent Treatment Systems 314 3.6.17 Explosive Gas Mixture 4.6.17 Explosive Gas Mixture 317 3.6.18 Mark I Containment 4.6.18 Mark I Containment 319 3.6.19 Liquid Waste Holdup Tanks 4.6.19 Liquid Waste Holdup Tanks 321 3.6.20 Radiological Environmental Monitoring 4.6.20 Radiological Environmental Monitoring 323 Program Program 3.6.21 Interlaboratory Comparison Program 4.6.21 Interlaboratory Comparison Program 334 3.6.22 Land Use Census 4.6.22 Land Use Census 336 3.7.1 Special Test Exceptions - Shutdown 4.7.1 Special Test Exceptions - Shutdown 339 Margin Demonstration Margin Demonstration AMENDMENT NO. IV

SECTION DESCRIPTION PAGE 5.0 Design Features 342 5.1 Site 342 5.2 Reactor 342 5.3 Reactor Vessel 342 5.4 Containment 345 5.5 Storage of Unirradiated and Spent Fuels 346 5.6 Seismic Design 346 6.0 Administrative Controls 347 6.'I Responsibility 347 6.2 Organization 347 6.3 Facility Staff Qualifications 351 6.4 Training 351 6.5 Review and Audit 351 6.6 Reportable Event Action 360 6.7 Safety Limit Violation 360 6.8 Procedures 360 6.9 Reporting Requirements 361 AMENDMENT NO.

SECTION DESCRIPTION PAGE 6.10 Record Retention 370 6.11 Radiation Protection Program 371 6.12 High Radiation Area 371 6.13 Fire Protection Inspection 373 6.14 Systems Integrity 373 6.15 Iodine Monitoring 373 AMENDMENT NO. Vl

1.0 DEFINITIONS 1.1 Reac or 0 era in Conditions The various reactor operating conditions are defined below. Individual technical specifications amplify these definitions when appropriate.

Shu down Condition - C Id (1) The reactor mode switch is in the shutdown position or refuel position. "

(2) No core alterations leading to an addition of reactivity are being performed.

(3) Reactor coolant temperature is less than or equal to 212 F.

b. Shu down Condi ion - Ho (1) The reactor mode switch is in the shutdown position.

(2) No core alterations leading to an addition of reactivity are being performed.

(3) Reactor coolant temperature is greater than 212'F.

C. Refuelin Condi i n (1) The reactor mode switch is in the refuel position.

(2) The reactor coolant temperature is less than 212'F.

(3) Fuel may be loaded or unloaded.

(4) No more than one operable control rod may be withdrawn.

d. Power 0 era in ondi ion (1) Reactor mode switch is in startup or run position.

(2) Reactor is critical or criticality is possible due to control rod withdrawal.

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e. Ma or Main enance Condi ion (1) No fuel is in the reactor.

The reactor mode switch may be placed in the startup position to perform the shutdown margin demonstration. See Special Test Exception 3.7.1.

The reactor mode switch may be placed in the refuel position to perform reactor coolant system pressure testing, control rod scram time testing and scram recovery operations.

AMENDMENT NO.

Pea kin Factor The ratio of the fuel rod heat flux to the heat flux of an average rod in an identical geometry bundle operating at the average core power.

g. Total Peakin Factor The Total Peaking Factor (TPF) is the highest product of radial, axial, and local peaking factors simultaneously operative at any segment of fuel rod.

h. Critical Power That assembly power which causes some point in the assembly to experience transition boiling.

i. Critical Power Ra io CPR The ratio of critical power to the bundle power at the reactor condition of interest.

j. Minimum Cri i al Power Ra io MCPR The minimum in-core critical power ratio.

AMENDMENT NO.

1.2 ~OerabIe A system, subsystem, train, component or device shall be operable when it is capable of performing its specified function(s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, except as noted in 3.0, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

A verification of operability is an administrative check, by examination of appropriate plant records (logs, surveillance test records) to determine that a system, subsystem, train, component or device is not inoperable. Such verification does not preclude the demonstration (testing) of a given system, subsystem, train, component or device to determine operability.

1.3 ~Oeratin Operating means that a system or component is performing its required functions in its required manner.

1.4 Pro ec ive lns rumen a ion Lo ic Defini ions

a. Ins rum n Channel An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.

A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

1.5 Sensor Check A sensor check is a qualitative determination of acceptable operability by observation of sensor behavior during operation.

This determination shall include, where possible, comparison of the sensor with other independent sensors measuring the same variable.

AMENDMENT NO.

1.6 Ins rument Channel Test Instrument channel test means injection of a simulated signal into the channel to verify its proper response including, where applicable, alarm and/or trip initiating action.

1.7 Ins rumen Channel Calibra ion Instrument channel calibration means adjustment of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip.

1.8 Ma'or Refueiin Ou a e For the purpose of designating frequency of testing and surveillance, a major refueling outage shall mean a.regularly scheduled refueling outage; however, where such outages occur within 8 months of the end of the previous refueling outage, the test or surveillance need not be performed until the next regularly scheduled outage.

An operating cycle is that portion of Station operation between reactor startups following each major refueling outage.

0 The test intervals specified are only valid during periods of power operation and do not apply in the event of extended Station shutdown.

1.11 Primar Con ainmen In e ri Primary containment integrity means that the drywell and absorption chamber are closed and all of the following conditions are satisfied:

a. All non-automatic primary containment isolation valves which are not required to be open for plant operation are closed.

b. At least one door in the airlock is closed and sealed.

AMENDMENT NO.

c. All automatic containment isolation valves are operable or are secured in the closed position.

d. All blind flanges and manways are closed.

1.12 Reactor Buildin Inte ri Reactor Building Integrity means that the reactor building is closed and the following conditions are met:

a. At least one door at each access opening is closed.

b. The standby gas treatment system is operable.

c. AII Reactor Building ventilation system automatic isolation valves are operable or are secured in the closed position.

A core alteration is the addition, removal, relocation, or other manual movement of fuel or controls in the reactor core.

Control rod movement with the control rod drive hydraulic system is not considered to be a core alteration.

1.14 ~Ra ed Fl x Rated flux is the neutron flux that corresponds to a steady-state power level of 1850 thermal megawatts. The use of the term 100 percent also refers to the 1850 thermal megawatt power level.

1.'1 5 ~urveillan Surveillance means that process whereby systems and components which are essential to p'lant 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.

AMENDMENT NO.

1.16 (Deleted) 1.17 (Deleted) 1.18 as ous Radwas e Trea men S stem A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents by collecting main condenser offgas and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

1.19 Member s of he Public Member(s) of the public shall include persons who are not occupationally associated with the Nine Mile Point Nuclear Station.

This category does not include employees of Niagara Mohawk Power Corporation, the New York State Power Authority, its contractors or vendors who are occupationally associated with Nine Mile Point Unit 1. 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 Nine Mile Point Unit 1.

1.20 Milk Sam lin Location A milk sampling location is that location where 10 or more head of milk animals are available for the collection of milk samples.

1.21 ffsi e Dos Calcula ion Man al OD M The Offsite Dose Calculational Manual shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the environmental radiological monitoring program.

AMENDMENT NO.

1.22 Proces Con rol Pro ram PCP The process control program shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of radioactive waste, based on demonstrated processing of actual or simulated wet or liquid wastes, will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71, and Federal and State regulations and other requirements governing the transport and disposal of radioactive waste.

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. The purge is completed when the oxygen concentration exceeds 19.5 percent.

The site boundary shall be that line around the Nine Mile Point Nuclear Station beyond which the land is neither owned, leased, nor otherwise controlled by Niagara Mohawk Power Corporation or the New York Power Authority.

Solidification shall be the conversion of wet or liquid waste into a form that meets shipping and burial ground requirements.

A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

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• The unrestricted area shall be any area at or beyond the site boundary access that is not controlled by Niagara Mohawk Power Corporation or the New York Power Authority for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. That area outside the restricted area (10 CFR 20.3(a)(14)) but within the site boundary will be controlled by the owner as required.

AMENDMENT NO.

1.28 Ven ilation Exhaus Treatmen S stem A ventilation exhaust treatment system is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be ventilation exhaust treatment system components.

1.29 ~Ven in 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.

1.30 R ac or Coolant Leaka e (1) Leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered and conducted to a sump or collecting tank, or (2) Leakage into the primary containment atmosphere from sources that are both specifically located and known not to be from a through-wall crack in the piping within the reactor coolant pressure boundary.

b. Unid n ifie Leaka All other leakage of reactor coolant into the primary containment area.

1.31 Core 0 ra in Limi s Re or The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1f. Plant operation within these operating limits is addressed in individual specifications.

AMENDMENT NO.

SAFETY LIMIT LIMITINGSAFETY SYSTEM SETTING 2.1.1 FUEL CLADDIN INTEGRITY 2.1.2 FUEL CLADDING INTEGRITY A licabili Applies to the interrelated variables associated with Applies to trip settings on automatic protective fuel thermal behavior. devices related to variables on which the fuel loading safety limits have been placed.

~Ob'ec ive: ~Ob'ective:

To establish limits on the important thermal-hydraulic To provide automatic corrective action to prevent variables to assure the integrity of the fuel cladding. exceeding the fuel cladding safety limits.

a. When the reactor pressure is greater than 800 Fuel cladding limiting safety system settings shall be psia and the core flow is greater than 10%, the as follows:

existence of a Minimum Critical Power Ratio (MCPR) less than the Safety Limit Critical Power a. The flow biased APRM scram trip settings shall Ratio (SLCPR) (Reference 12) shall constitute be less than or equal to that shown in Figure violation of the fuel cladding integrity safety 2.1.1.

limit.

b. The IRM scram trip setting shall not exceed 12%

b. When the reactor pressure is less than or equal of rated neutron flux.

to 800 psia or core flow is less than 10% of rated. the core power shall not exceed 25% of c. The reactor high pressure scram trip setting shall rated thermal power. be a1080 psig.

AMENDMENT NO.

SAFETY LIMIT LIIVllTINGSAFETY SYSTEM SETTING C. The neutron flux shall not exceed its scram d. The reactor water low level scram trip setting setting for longer than 1.5 seconds as indicated shall be no lower than -12 inches (53 inches by the process computer. When the process indicator scale) relative to the minimum normal computer is out of service, a safety limit violation water level (302'9").

shall be assumed if the neutron flux exceeds the scram setting and control rod scram does not e. The reactor water low-low level setting for core occur. spray initiation shall be no less than -5 feet (5 inches indicator scale) relative to the minimum To ensure that the Safety Limit established in normal water level (Elevation 302.'9").

Specifications 2.1.1a and 2.1.1b is not exceeded, each required scram shall be initiated f. The flow biased APRM rod block trip settings by its expected scram signal. The Safety Limit shall be less than or equal to that shown in shall be assumed to be exceeded when scram is Figure 2.1.1.

accomplished by a means other than the expected scram signal.

d. Whenever the reactor is in the shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be more than 6 feet, 3 inches (-10 inches indicator scale) below minimum normal water level (Elevation 302'9")

except as specified in "e" below.

e. For the purpose of performing major maintenance (not to exceed 12 weeks in duration) on the reactor vessel; the reactor water level may be lowered 9'elow the minimum normal water level (Elevation 302'9"). Whenever the reactor water level is to be lowered below the low-low-low level setpoint redundant instrumentation will be provided to monitor the reactor water level.

AMENDMENT NO.

SAFETY LIIVIIT LIMITINGSAFETY SYSTEM SETTING Written procedures will be developed and g. The reactor low pressure setting for main-steam-followed whenever the reactor water level is line isolation valve closure shall be a850 psig lowered below the low-low level set point (5 feet when the reactor mode switch is in the run below minimum normal water level). The position.

procedures will define the valves that will be used to lower the vessel water level. All other h. The main-steam-line isolation valve closure scram valves that have the potential of lowering the setting shall be <10 percent of valve closure vessel water level will be identified by valve (stem position) from full open.

number in the procedures and these valves will be red tagged to preclude their operation during i. The generator load rejection scram shall be the major maintenance with the water level initiated by the signal for turbine control valve below the low-low level set point. fast closure due to a loss of oil pressure to the acceleration relay any time the turbine first stage In addition to the Facility Staff requirements ~

steam pressure is above a value corresponding to given in Specification 6.2.2.b, there shall be - 833 Mwt, i.e., 45 percent of 1850 Mwt.

another control room operator present in the control room with no other duties than to The turbine stop valve closure scram shall be monitor the reactor vessel water level. initiated at ~10 percent of valve closure setting (Stem position) from full open whenever the turbine first stage steam pressure is above a value corresponding to 833 Mwt, i.e., 45 percent of 1850 Mwt.

AMENDMENT NO.

FIGURE 2.1.3 FLOW BIASED SCRAM AND APRM ROD BLOCK 160 NOTES:

1. Rated Power ls lddg MW 0 140 2. Design Flow Is dl.d X 10 Ib/hr.

3. Caloul ~ led Tot ~ I Peaking Faotor ( MTPF

-I- 4. Core Pressure Is ) 800 psla SCRAM

~ 120 0 OD BLOC

+ 100

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80 For Caloulated Total Peaking Faotore ) MTPF X 60 S'a~

MTPF PKFL X S o

Wherei NOTE:

In oases where for ~ shprt period the tot ~ I Sa The new Soram and Rod Stook Lines peaking faotor (PKFL) exoeeds the maximum z 40 total peaking faotor lMTPF), rather than MTPF ~ Maximum Total Peaking Faotor MTPF values lor eaoh fuel design 0 adlusllng the APRM setpolnls, the APRM gain may be ad)usted so that the APRM readings are are obl ~ Ined from the Core Operating Llmlts Report greater than or equal to oore power X PKFL/MTPF 20 provided that the ad)usted ARRM reading does PKFL w Caloulated Total Peaking Faotor not exoeed 100% ol rated thermal power and ~ 8 wSoram and Rod Blook Lines notloe ol ad)ustment ls posted on the reaotor shown above oontrol panel 0

0 20 40 60 80 100 120 140 RECIRCULATION FLOW, PERCENT OF DESIGN AMENDMENT NO.

BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient.

Because fuel damage is not directly observable, a step-back approach is used to establish a safety limit such that the Minimum Critical Power Ratio (MCPR) is no less than the Safety Limit Critical Power Ratio (SLCPR) (Reference 12). The SLCPR represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding safety limit is defined with margin to the conditions which would produce onset of transition boiling, (MCPR of 1.0) These

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conditions represent a significant departure from the condition intended by design for planned operation.

Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure. However, the existence of critical power, or boiling transition, is not a directly observable parameter in an operating reactor.

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Therefore, at reactor pressure >800 psia and core flow 10% of rated the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the Critical Power Ratio (CPR) which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the Minimum Critical Power Ratio (MCPR). It is assumed that the plant operation is controlled to the nominal protective set points via the instrumented variables, by the nominal expected flow control line. The SLCPR has sufficient conservatism to assure that in the event of an abnormal operational transient initiated from a normal operating condition more than 99.9% of the fuel rods in the core are expected to avoid boiling transition. The margin between MCPR of 1.0 (onset of transition boiling) and the SLCPR is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the core operating state including uncertainty in the boiling transition correlation as described in References 1 and 12.

Because the boiling transition correlation is based on a large quantity of full scale data, there is a very high confidence that operation of a fuel assembly at the condition of the SLCPR would not produce boiling transition. Thus, although it is not required to establish the safety limit, additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity.

AMENDMENT NO. 13

BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT However, if boiling transition were to occur, clad perforation would not be expected. Cladding temperatures would increase to approximately 1100'F which is below the perforation temperature of the cladding material. This has bean verified by tests in the General Electric Test Reactor (GETR) where similar fuel operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation.

If reactor pressure should ever exceed 1400 psia during normal power operating (the limit of applicability of the boiling transition correlation), it would be assumed that the fuel cladding integrity safety limit has been violated.

In addition to the boiling transition limit SLCPR, operation is constrained to a maximum LHGR as specified in the Core Operating Limits Report. At 100% power, this limit is reached at a given Maximum Total Peaking Factor (MTPF). The value of MTPF for each fuel design is contained in the Core Operating Limits Report. During steady-state operation where the Calculated Total Peaking Factor (PKFL) is above the MTPF, the equation on Figure 2.1 1 will be used to adjust the flow biased scram and APRM rod block setpoints.

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At pressure equal to or below 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.56 psi. At low power and all core flows, this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and all flows will always be greater than 4.56 psi.

Analyses show that with a bundle flow of 28x10 Ib/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Therefore, due to the 4.56 psi driving head, the bundle flow will be greater than 28x10 Ib/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at 28x10 Ib/hr is approximately 3.35 MWt. With the design peaking factor, this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative.

AMENDMENT NO. 14

BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT During transient operation the heat flux (thermal power-to-water) would lag behind the neutron flux due to the inherent heat transfer time constant of the fuel which is 8 to 9 seconds. Also, the limiting safety system scram settings are at values which will not allow the reactor to be ogerated above the safety limit during normal operation or during other plant operating situations which have been analyzed in detail.i Sl In addition, control rod scrams are such that for normal operating transients the neutron flux transient is terminated before a significant increase in surface heat flux occurs. Scram times of each control rod are checked periodically to assume adequate insertion times. Exceeding a neutron flux scram setting and a failure of the control rods to reduce flux to less than the scram setting within 1.5 seconds does not necessarily imply that fuel is damaged; however, for this specification a safety limit violation will be assumed any time a neutron flux scram setting is exceeded for longer than 1.5 seconds.

If the scram occurs such that the neutron flux dwell time above the limiting safety system setting is less than 1.7 seconds, the safety limit will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected. These analyses show that even if the bypass system fails to operate, the design limit of the SLCPR is not exceeded. Thus, use of a 1.5-second limit provides additional margin.

The process computer has a sequence annunciation program which will indicate the sequence in'which scrams occur such as neutron flux, pressure, etc. This program also indicates when the scram set point is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 2.1.1.c will be relied on to determine if a safety limit has been violated.

AMENDMENT NO. 15

BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT During periods when the reactor is shut down, consideration must also be given to water level requirements, due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds of the core height.

The lowest point at which the reactor water level can normally be monitored is approximately 7 feet 11 inches below minimum normal water level or 4 feet 8 inches above the top of the active fuel. This is the location of the reactor vessel tap for the low-low-low water level instrumentation. The actual low-low-low water level trip point is 6 feet 3 inches (-10 inches indicator scale) below minimum normal water level (Elevation 302'-9"). The 20 inch difference resulted from an evaluation of the recommendations contained in General Electric Service Information Letter 299 "High Drywell Temperature Effect on Reactor Vessel Water Level Instrumentation." The low-low-low water level trip point was raised 20 inches to conservatively account for possible differences in actual to indicated water level due to potentially high drywell temperatures. The safety limit has been established here to provide a point which can be monitored and also can provide adequate margin. However, for performing major maintenance as specified in Specification 2.1.1.e, redundant instrumentation will be provided for monitoring reactor water level below the low-low-low water level set point. (For example, by installing temporary instrument lines and refererice points to redundant level transmitters so that the reactor water level may be monitored over the required range). In addition, written procedures, which identify all the valves which have the potential of lowering the water level inadvertently, are established to prevent their operation during the major maintenance which requires the water level to be below the low-low level set point.

The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a safety limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a safety limit provided scram signals are operable is supported by the extensive plant safety analysis.

AMENDMENT NO. 16

BASES FOR 2.1.2 FUEL CLADDING - LS The abnormal operational transients applicable to operation of the plant have been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 1850 MWt. The analyses were based upon plant operation in accordance with the operating map given in Reference 11. In addition, 1850 MWt is the licensed maximum power level, and represents the maximum steady-state power which shall not knowingly be exceeded.

Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic performance. Results obtained from a General Electric boiling water reactor have been compared with predictions made by the model. The comparisons and results are summarized in Reference 2.

The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about 25% greater than the nominal maximum value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent to approximately 80% of the total scram worth of the control rods. The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications. The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion. By the time the rods are 60% inserted, approximately four dollars of negative reactivity have been inserted which strongly turns the transient, and accomplishes the desired effect. The times for 50% and 90% insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

a. The Average Power Range Monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated thermal power. Because 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 abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses 'emonstrate that with a 120% scram trip setting, none of the abnormal operational transients analyzed violate the fuel safety limit and there is a substantial margin from fuel damage.

AMENDMENT NO. 17

BASES FOR 2.1.2 FUEL CLADDING - LS However, in response to expressed beliefsl I that variation of APRM flux scram with recirculation flow is a prudent measure to assure safe plant operation during the design confirmation phase of plant operation, the scram setting will be varied with recirculation flow.

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity safety limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency. of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity safety limit yet allows operating margin that reduces the possibility of unnecessary scrams.

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of MTPF and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Figure 2.1.1 when the maximum total peaking factor is greater than the limiting total peaking factor.

b. Normal operation of the automatic recirculation pump control will be in excess of 30% rated flow; therefore, little operation below 30% flow is anticipated. For operation in the startup mode while the reactor is at low pressure, the IRM scram setting is 12% of rated neutron flux. Although the operator will set the IRM scram trip at 12% of rated neutron flux or less, the actual scram setting can be as much as 2.5% of rated neutron flux greater. This includes the margins discussed above. This provides adequate margin between the setpoint and the safety limit at 25% of rated power. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. There are a few possible sources of rapid reactivity input to the system in the low power flow condition. 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. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated per minute, and the IRM system would be more than adequate to assure a scram before the power could exceed the safety limit.

AMENDMENT NO. 18

BASES FOR 2.1.2 FUEL CLADDING - LS Procedural controls will assure that the IRM scram is maintained up to 20% flow. This is accomplished by keeping the reactor mode switch in the startup position until 20% flow is exceeded and the APRM's are on scale. Then the reactor mode switch may be switched to the run mode, thereby switching scram protection from the IRM to the APRM system.

In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale. This condition exists at quarter rod density. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited to 1%

of rated power, thus maintaining a limit above the SLCPR. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

c. As demonstrated in Appendix E-I'nd the Technical Supplement to Petition to Increase Power Level, the reactor high pressure scram is a backup to the neutron flux scram, turbine stop valve closure scram, generator load rejection scram, and main steam isolation valve closure scram, for various reactor isolation incidents. However, rapid isolation at lower power levels generally results in high pressure scram preceding other scrams because the transients are slower and those trips associated with the turbine generator are bypassed.

The operator will set the trip setting at 1080 psig or lower. However, the actual set point can be as much as 15.8 psi above the 1080 psig indicated set point due to the deviations discussed above.

AMENDMENT NO. 19

BASES FOR 2.1.2 FUEL CLADDING - LS

d. A reactor water low level scram trip setting -12 inches (53 inches indicator scale) relative to the minimum normal water level (Elevation 302'9") will assure that power production will be terminated with adequate coolant remaining in the core. The analysis of the feedwater pump loss in the Technical Supplement to Petition to Increase Power Level, dated April 1970, has demonstrated that approximately 4 feet of water remains above the core following the low level scram.

The operator will set the low level trip setting no lower than -12 inches relative to the lowest normal operating level.

However, the actual set point can be as much as 2.6 inches lower due to the deviations discussed above.

e. A.reactor water low-low level signal -5 feet (5 inches indicator scale) relative to the minimum normal water level (Elevation 302'9") will assure that core cooling will continue even if level is dropping. Core spray cooling will adequately cool the core, as discussed in LCO 3.1.4.

The operator will set the low-low level core spray initiation point at no less than -5 feet (5 inches indicator scale) relative to the minimum normal water level (Elevation 302'9"). However, the actual set point can be as much as 2.6 inches lower due to the deviations discussed above.

f. 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 the condition of a MCPR less than the SLCPR. This rod block trip setting, which is automatically varied with recirculation flow rate,-prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, 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 110%

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. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum total peaking factor exceeds the design peaking factor, thus, preserving the APRM rod block safety margin.

AMENDMENT NO. 20

BASES FOR 2.1.2 FUEL CLADDING - LS g-h. The low pressure isolation of the main steam lines at 850 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 closed, 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 850 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 isolation on reactor low pressure 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 and peak pressure in the vessel dome is limited to 1141 psig.

(8, 9, 10)

The operator will set the pressure trip at greater than or equal to 850 psig and the isolation valve stem position scram setting at less than or equal to 10% of valve stem position from full open. However, the actual pressure set point can be as much as 15.8 psi lower than the indicated 850 psig and the valve position set point can be as much as 2.5% of stem position greater.

These allowable deviations are due to instrument error, operator setting error and drift with time.

In addition to the above mentioned LS, other reactor protection system devices (LCO 3.6.2) serve as a secondary backup to the LS chosen. These are as follows:

High fission product activity released from the core is sensed in the main steam lines by the high radiation main steam line monitors. These monitors provide a backup scram signal and also close the main steam line isolation valves.

The scram dump volume high level scram trip assures that scram capability will not be impaired because of insufficient scram dump volume to accommodate the water discharged from the control rod drive hydraulic system as a result of a reactor scram (Section X-C.2.10) ".

The generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to the worst case transient of a load rejection and subsequent failure of the bypass.

In fact, analysis shows that heat flux does not increase from its initial value at all because of the fast action of the load rejection scram; thus, no significant change in MCPR occurs.

The turbine stop valve closure scram is provided for the same reasons as discussed in i above. With a scram setting of

~10% valve closure, the resultant transients are nearly the same as for those described in i above; and, thus, adequate margin exists.

%FSAR AMENDMENT NO. 21

REFERENCES FOR BASES 2.1.1 AND 2.1.2 FUEL CLADDING (1) General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO-10958 and NEDE-10958.

(2) Linford, R. B., "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NEDO-10801, February 1973.

(3) FSAR, Volume II, Appendix E.

(4) FSAR, Second Supplement.

(5) FSAR, Volume II, Appendix E.

(6) FSAR, Second Supplement.

(7) Letters, Peter A. Morris, Director of Reactor Licensing, USAEC, to John E. Logan, Vice-President, Jersey Central Power and Light Company, dated November 22, 1967 and January 9, 1968.

(8) Technical Supplement to Petition to Increase Power Level, dated April 1970.

(9) Letter, T. J. Brosnan, Niagara Mohawk Power Corporation, to Peter A. Morris, Division of Reactor Licensing, USAEC, dated February 28, 1972.

(10) Letter, Philip D. Raymond, Niagara Mohawk Power Corporation, to A. Giambusso, USAEC, dated October 15, 1973.

(11) Nine Mile Point Nuclear Power Station Unit 1 Load Line Limit Analysis, NEDO 24012, May, 1977.

(12) Licensing Topical Report "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A, latest approved revision.

NEDO-(13) Nine Mile Point Nuclear Power Station Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal (Cycle 6),

24185, April 1979.

(14) General Electric SIL 299 "High Drywell Temperature Effect on Reactor Vessel Water Level Instrumentation."

(15) Letter (and attachments) from C. Thomas (NRC) to J. Charnley (GE) dated May 28, 1985, "Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-B, Amendment 10."

AMENDMENT NO. 22

SAFET.Y LIMIT LIMITINGSAFETY SYSTEM SETTING 2.2.1 REA TOR COOLANT SYSTEM 2.2.2 REACTOR C OLANT SY TEM bib% a. The settings on the safety valves of the pressure vessel shall be as shown below. The allowable Applies to the limit on reactor coolant system ~

initial set point error on each setting will be a1 pressure. percent.

~e~cive: Set Point Number of

~Psi Saf Valv To define those values of process variables which shall assure the integrity of the reactor coolant 1218 4 system to prevent an uncontrolled release of 1227 3 radioactivity. 1236 3 1245 3 1254 16 The reactor vessel or reactor coolant system pressure shall not exceed 1375 psig at any time with fuel in b. The reactor high-pressure scram trip setting shall the vessel. be ~1080 psig.

c. The flow biased APRM scram trip settings shall be as shown in Figure 2.1.1.

AMENDMENT NO. 23

BASES FOR 2.2.1 REACTOR COOLANT SYSTEM SAFETY LIMIT The pressure safety limit of 1375 psig was derived from the design pressures and applicable codes for the reactor pressure vessel and the reactor coolant system piping. (ASME Boiler and Pressure Vessel Code Section I applies to the reactor pressure vessel and ASA Piping Code, Section B31.1 applies to the coolant system piping.) The ASME Code permits pressure transients up to 10 percent over design pressure (110% x 1250 = 1375 psig) and the ASA Code permits pressure transients up to 15 percent over the design pressure (115% x 1200 = 1380 psig).

Data presented in Volume IV, Section I-B" includes the design analyses which were performed to demonstrate that the reactor pressure vessel would meet the applicable code requirements. As a part of these analyses, both design and non-design events (Tables 7 and 8) were postulated to evaluate their strain effect to the vessel. Among the non-design events, a postulated over-pressure of 3750 psig was expected to result in vessel destruction. Comparable data concerning the piping system is not available, however, ASA Code (B31.1) indicates a margin of safety factor, code allowable (10,800 psi at 600'F) vs. yield strength (75,000 psi), of 6.8 for the process piping system while the margin of safety factor (15,000 psi vs. 60,000 psi) for the high pressure feedwater system is 4. Additional data in Supplement 2, Table IV-1" indicates a calculated feedwater valve burst pressure of 13,000 psi based upon a yield strength of 36,000 psi.

Based upon the available data and for safety valve sizing calculations, 1375 psig was selected as a safety limit for the reactor coolant system. The maximum pressure of the critical hydro test of the unfueled system was selected as 1800 psig, while the normal system operating pressure will be 1030 psig.

"FSAR AMENDMENT NO. 24

BASES FOR 2.2.2 REACTOR COOLANT SYSTEM - LS

a. The range of set points for a safety valve actuation is selected in accordance with code requirements. A safet~ valve capability study presented in the Technical Supplement to Petition to Increase Power Level using the stated LS values has demonstrated the maximum pressures occurring at the bottom of the reactor vessel and the bottom of the recirculation piping are 1303 psig and 1315 psig, respectively, some 72 psig below the 1375 psig safety limit. This analysis has assumed the highly improbable event of reactor isolation occurring without scram, in spite of separate and redundant scram signals such that the power output reached 167 percent of rated (1850 Mwt).

In addition to the safety valves, the solenoid-actuated relief valves are used to prevent safety valve lift during rapid reactor isolation at power coupled with failure of the bypass system. Any five of these valves opening at 1090 psig to 1100 psig will keep the maximum vessel pressure below the lowest safety valve setting, as demonstrated in Appendix E-I.3.11 (p. E-35)". (The Technical Supplement to Petition to Increase Power Level, and letter from T. J. Brosnan, Niagara Mohawk Power Corporation, to Peter A.

Morris, Division of Reactor Licensing, USAEC, dated February 28, 1972). Subsequently, six valves were provided due to the blowdown requirements, following a small line break. The capacity of a solenoid-actuated relief valve is about the same as a safety valve. Therefore, even without scram any combination of 16 safety valves and solenoid-actuated valves will limit the pressure below the safety limit following the worst isolation situation.

b. The reactor high pressure scram setting is relied upon to terminate rapid pressure transients if other scrams, which would normally occur first, fail to function. As demonstrated in Appendix E-I of the FSAR and the Technical Supplement to Petition to Increase Power Level, Page II-12, the reactor high pressure scram is a backup to the neutron flux scram, generator load rejection scram, and main steam isolation-valve closure scram for various reactor isolation incidents. However, rapid isolation at lower power levels generally results in high pressure scram preceding other scrams because the transients are slower and those trips associated with the turbine-generator are bypassed.

The operator will set the trip setting at 1080 psig or lower. However, the actual set point can be as much as 15.8 psi above the 1080 psig indicated set point due to the deviations discussed above.

%FSAR AMENDMENT NO. 25

BASES FOR 2.2.2 REACTOR COOLANT SYSTEM - LS

c. As shown in Appendix E-I.3.8 and 3.11", rapid Station transients due to isolation valve or turbine trip valve closures result in coincident high-flux and high-pressure transients. Therefore, the APRM trip, although primarily intended for core protection, also serves as backup protection for pressure transients.

Although the operator will set the scram setting at less than or equal to that shown in Figure 2.1.1, the actual neutron flux setting can be as much as 2.7 percent of rated neutron flux above the line. This includes the errors discussed above. The flow bias could vary as much as one percent of rated recirculation flow above or below the indicated point.

In addition to the above-mentioned LS, other reactor protection system devices (LCO 3.6.2) serve as secondary backup to the LS chosen. These are as follows:

The primary containment high-pressure scram serves as backup to high reactor pressure scram in the event of lifting of the safety valves. As discussed in Vol. I, Vill, 2.0.c (p. Vill-9)" a pressure in excess of 3.5 psig due to steam leakage of blowdown to the drywell will trip a scram well before the core is uncovered.

A low condenser vacuum situation will result in loss of the main reactor heat sink, causing an increase in reactor pressure.

The scram feature provided, therefore, anticipates the reactor high-pressure scram. A loss of main condenser vacuum is analyzed in Appendix E-I.3.17".

The scram dump volume high-level scram trip assures that scram capability. will not be impaired because of insufficient scram dump volume to accommodate the water discharge from the control-rod-drive hydraulic system as a result of a reactor scram (Section X-C.2.10) ".

In the event of main-steam-line isolation valve closure, reactor pressure will increase. A reactor scram is, therefore, provided on main-steam-line isolation valve position and anticipates the high reactor pressure scram trip.

NFSAR AMENDMENT NO. 26

3.0 LIMITINGCONDITIONS FOR OPERATION 3.0.1 PERABILITY REQUIREMENT When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered operable for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is operable; and (2) all of its redundant system(s), subsystem(s), train(s), component(s) and device(s) are operable, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in a condition stated in the individual specification.

In the event a Limiting Condition for Operation and/or associated surveillance requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in a condition consistent with the individual specification unless corrective measures are completed that permit operation under the permissible surveillance requirements for the specified time interval as measured from initial discovery or until the reactor is placed in an operational condition in which the specification is not applicable.

4.0 URVEILLANCE REQUIREMENT 4.0.1 RVEILLANCE INTERVALS Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

BASES Specification 4.0.1 establishes the limit for which the specified time interval for. Surveillance Requirements may be extended.

It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It also provides flexibilityto accommodate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with a 24 month surveillance interval. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outages. The limitation of Specification 4.0.1 is based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

AMENDMENT NO. 27

3.1.0 FUEL CLADDING A) GENERAL APPLICABILITY Applies to the power level regulation, control rod system, liquid poison system, emergency cooling system, and core spray system. LCOs for the minimum allowable circuits corresponding to the LS3 settings are included in the Reactor Protection System LCO (3.6.2).

B) GENERAL OBJECTIVE LIMITINGCONDITIONS FOR OPERATION - To define the lowest functional capability or performance level of the systems and associated components which will assure the integrity of the fuel cladding as a barrier against the release of radioactivity.

SURVEILLANCE REQUIREMENTS - To define the tests or inspections required to assure the functional capability or performance level of the required systems or components.

AMENDMENT NO. 28

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.1 CONTROL ROD SYSTEM 4.1.1 CONTR L ROD SYSTEM Applies to the operational status of the control rod Applies to the periodic testing requirements for the system. control rod system.

~Ob'ec ive: ~ob ec ive:

To assure the capability of the control rod system to To specify the tests or inspections required to assure control core reactivity. the capability of the control rod system to control core reactivity.

e ifica ion The control rod system surveillance shall be performed as indicated below.

a. Reactivity Limitations a. Reactivity Limitations (1) Reactivity margin - core loading (1) Reactivity margin - core loading The core loading shall be limited to that Sufficient control rods shall be withdrawn which can be made subcritical in the most following a refueling outage when core reactive condition during the operating cycle. alterations were performed to demonstrate with the strongest control rod in its full-out with a margin of 0.25 percent hk that the position and all other operable rods fully core can be made subcritical at any time in inserted. the subsequent fuel cycle with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.

AMENDMENT NO. 29

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) Reactivity margin - stuck control rods (2) Reactivity margin - stuck control rods Control rods which cannot be moved with Each partially or fully withdrawn control rod control rod drive pressure shall be shall be exercised at least once each week.

considered inoperable. Inoperable control This test shall be performed at least once rods shall be valved out of service, in such per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power operation is positions that Specification 3.1.1a(1) is met. continuing with two or more inoperable In no case shall the number of non-fully control rods or in the event power operation inserted rods valved out of service be is continuing with one fully or partially greater than six during power operation. If withdrawn rod which cannot be moved and this specification is not met, the reactor for which control rod drive mechanism shall be placed in the cold shutdown damage has not been ruled out. The condition. If a partially or fully withdrawn surveillance need not be completed within control rod drive cannot be moved with drive 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable rods or scram pressure the reactor shall be has been reduced to less than two and if it brought to a shutdown condition within 48 has been demonstrated that control rod drive hours unless. investigation demonstrates that mechanism collat housing failure is not the the cause of the failure is not due to a failed- cause of an immovable control rod.

control rod drive mechanism collet housing.

b. Control Rod Withdrawal b. Control Rod Withdrawal (1) The control rod shall be coupled to its drive (1) The coupling integrity shall be verified for or completely inserted and valved out of each withdrawn control rod by either:

service. When removing a control rod drive...

for inspection, this requirement does not (a) Observing the drive does not go to the apply as long as the reactor is in a shutdown overtravel position, or or refueling condition.

(b) A discernible response of the nuclear instrumentation.

AMENDMENT NO. 30

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) The control rod drive housing support (2) The control rod drive housing support system shall be in place during power system shall be inspected after reassembly.

. operation and when the reactor coolant system is pressurized above atmospheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.1.1a(1) is met.

(3)(a) Control rod withdrawal sequences shall be (3)(a) To consider the rod worth minimizer established so that maximum reactivity that operable, the following steps must be could be added by dropout of any increment performed:

of any one control blade would not make the core more than 0.013 hk supercritical. (i) The control rod withdrawal sequence for the rod worth minimizer computer shall be verified as correct.

(ii) The rod worth minimizer computer on-line 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.

AMENDMENT NO.

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS (b) Whenever the reactor is in the startup (b) If the rod worth minimizer is inoperable or run mode below 20/ rated thermal while the reactor is in the startup or run power, no control rods shall be moved mode below 20% rated thermal power unless the rod worth minimizer is and a second independent operator or operable, except as noted in 4.1.1.b engineer is being used he shall verify (3)(a)(iv), or a second independent that all rod positions are correct prior to operator or engineer verifies that the commencing withdrawal of each rod operator at the reactor console is group.

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.

If the rod worth minimizer fails prior to the complete withdrawal of the first twelve rods, then the withdrawn rods shall be inserted in the reverse order in which they were withdrawn. A second independent operator or engineer shall verify that the operator at the reactor controls is following the control rod program in reverse order.

(4) Control rods shall not be withdrawn for approach to criticality unless at least three source range channels have an observed count rate equal to or greater than three counts per second.

AMENDMENT NO. 32

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT C. Scram Insertion Times c. Scram Insertion Times (1) The average scram insertion time of all (1) After each major refueling outage and operable control rods, in the power operation prior to power operation with reactor condition, shall be no greater than: pressure above 800 psig, all operable control rods shall be scram time tested

% Inserted Average Scram from the fully withdrawn position.

From Fully Insertion W~ihdrawn 5 0.375 20'0 0.90 2.00 90 5.00 (2) Except as noted in 3.1.1.c{3), the maximum {2) Following each reactor scram from rated insertion scram time, in the power operation pressure, the mean 90% insertion time condition, shall be no greater than: shall be determined for eight selected rods. If the mean 90% insertion time of

% Inserted Maximum Scram the selected control rod drives does not From Fully Insertion fall within the range of 2 4 to 3.1

~Wi hdrawn seconds or the measured scram time of any one drive for 90% insertion does not 5 0.398 fall within the range of 1.9 to 3.6 20 0.954 seconds, an evaluation shall be made to 50 2.12 provide reasonable assurance that proper 90 5.30 control rod drive performance is maintained.

AMENDMENT NO. 33

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (3) Control rods with longer scram insertion (3) Following any outage not initiated by a time will be permitted provided that no other reactor scram, eight rods shall be scram control rod in a nine-rod square array around tested with reactor pressure above 800 psig.

this rod has a: The same criteria of 4.1.1.c(2) shall apply.

(a) Scram insertion time greater than the maximum allowed, (b) Malfunctioned accumulator, (c) Valved out of service in a non-fully inserted position.

d. Control Rod Accumulators d. Control Rod Accumulators At all reactor operating pressures, a rod Once a shift check the status of the accumulator accumulator may be out of service provided that pressure and level alarms in the control room.

no other control rod in a nine-rod square array around this rod has a:

(1) Malfunctioned accumulator, (2) Valved out of service in a non-fully inserted position, (3) Scram insertion greater than maximum permissible insertion time.

AMENDMENT NO. 34

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT If a control rod with a malfunctioned accumulator is inserted "full-in" and valved out of service, it shall not be considered to have a malfunctioned accumulator.

e. Scram Discharge Volume e. Scram Discharge Volume (SDV)

With one scram discharge volume vent valve Scram Discharge Volume Vent and Drain Valves and/or one scram discharge volume drain valve shall be demonstrated OPERABLE during Power inoperable and open, restore the inoperable Operations by:

valve(s) to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1. At least once per month verifying each valve With any scram discharge volume vent valve(s) to be open and/or any scram discharge volume drain valve(s) otherwise inoperable, restore at least one vent 2. At least once per quarter cycling each valve and one drain valve to OPERABLE status within 8 through at least one complete cycle of full hours. travel; and The Scram Discharge Volume Drain and Vent valves shall be demonstrated OPERABLE at least once per Operating Cycle by verifying that:

1. Valves close within 10 seconds after receipt of a signal for control rods to scram;

2. Valves open when the scram signal is reset;

3. Level instrumentation response proves that no blockage in the system exists.

These valves may be closed intermittently for testing under administrative controls.

AMENDMENT NO. 35

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

f. If specification 3.1.1.a through e, above, are not met, the reactor shall be placed in the hot shutdown condition within ten hours except as noted in 3.1.1.a(2).

g. Reactivity Anomalies g. Reactivity Anomalies The difference between an observed and The observed control rod inventory shall be predicted control rod inventory shall not exceed compared with a normalized computed prediction the equivalent of one percent in reactivity. If this of the control rod inventory during startup, limit is exceeded, the reactor shall be brought to following refueling or major core alteration.

the cold,. shutdown condition by normal orderly shutdown procedure. Operation shall not be These comparisons will be used as base data for permitted until the cause has been evaluated and reactivity monitoring during subsequent power the appropriate corrective action has been operation throughout the fuel cycle. At specific completed. power operating conditions, the actual control rod configuration will be compared with the expected configuration based upon appropriately corrected past data. This comparison will be made every equivalent full power month.

AMENDMENT NO. 36

BASES FOR 3.1.1 AND 4.1.1 CONTROL ROD SYSTEM Reactivity Limitations (1) Reactivity margin - core loading The core reactivity limitation is a restriction to be applied to the design of new fuel which may be loaded in the core or into a particular refueling pattern. Satisfaction of the limitation can only be demonstrated at the time of loading or reloading and must be such that it will apply to the entire subsequent fuel cycle. It is sufficient that the core in its maximum reactivity condition be subcritical with the control rod of highest worth fully withdrawn and all other rods fully inserted. In order to implement this requirement, it will be required that the amount of shutdown margin will be at least R + 0.25 percent hk in the cold, xenon-free condition. In this generalized expression, the value of R is the difference between the calculated value of core reactivity anytime later in the cycle where it may be greater than at the beginning. R must be a positive quantity or zero. A core which contains temporary control curtains or other burnable neutron absorbers may have a reactivity characteristic which increases the core lifetime, goes through a maximum, and then decreases thereafter.

A 0.25 percent hk in the expression R + 0.25 percent hk is provided as a finite, demonstrable, subcriticality margin.

For the first fuel cycle, core reactivity is calculated never to be greater than the beginning-of-life value;-hence, R = 0.

The new value of R must be determined for each fuel cycle.

(2) Reactivity margin - stuck control rods The specified limits provide sufficient scram capability to accommodate failure to scram of any one operable rod. This failure is in addition to any inoperable rods that exist in the core, provided that those inoperable rods met the core reactivity Specification 3.1.1 a(1).

Control rods which cannot be moved with control rod drive pressure are indicative of an abnormal operating condition on the affected rods and are, therefore, considered to be inoperable. Inoperable rods are valved out of service to fix their position in the core and assure predjctabie behavior. If the rod is fully inserted and then valved out of service, it is in a safe position of maximum contribution to shutdown reactivity. If it is valved out of service in a non-fully inserted position, that position is required to be consistent with the shutdown reactivity limitation stated in Specification 3.1.1a(1), which assures the core can be shut down at all times with control rods.

AMENDMENT NO. 37

BASES FOR 3.1.1 AND 4-1.1 CONTROL ROD SYSTEIVI The allowable inoperable rod patterns will be determined using information obtained in the startup test program supplemented by calculations. During initial startup, the reactivity condition of the as-built core will be determined.

Also, sub-critical patterns of widely separated withdrawn control rods will be observed in the control rod sequences being used. The observations, together with calculated strengths of the strongest control rods in these patterns will comprise a set of allowable separations of malfunctioning rods. During the fuel cycle, similar observations any cold shutdown can be used to update and/or increase the allowable patterns.

made'uring The number of rods permitted to be valved out of service could be many more than the six allowed by the specification, particularly late in the operating cycle; however, the occurrence of more than six could be indicative of a generic problem and the reactor will be shut down. Placing the reactor in the shutdown condition inserts the control rods and accomplishes the objective of the specifications on control rod operability. This operation is normally expected to be accomplished within ten hours. The weekly control rod exercise test serves as a periodic check against deterioration of the control rod system. Experience with this control rod drive system has indicated that weekly tests are adequate, and that rods which move by drive pressure will scram when required as the pre sure applied is much higher.

Also if damage within, the control rod drive mechanism and in particular, cracks in drive internal housings, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out. Circumferential cracks resulting from stress assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods. Limiting the period of operation with a potentially severed collet housing and requiring increased surveillance after detecting one stuck rod will assure that the reactor will not be operated with a large number of rods with failed collat housings.

b. Control Rod Withdrawal (1) Control rod dropout accidents as discussed in Appendix E" can lead to significant core damage. If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides an indirect verification that the rod is coupled to its drive. Details of the control rod drive coupling are given in Section IV.B.6.1 ".

"FSAR AMENDMENT NO. 38

BASES FOR 3.1.1 AND 4.1.1 CONTROL ROD SySTEM (2) The rod housing support is provided to prevent control rod ejection. accidents. Its design is discussed in Section Vll-E". Procedural control shall assure that the housing supports are in place for all control rods.

(3) Control rod withdrawal and insertion sequences are established to assure that the maximum in-sequence individual control rod or control rod segments which are withdrawn could not be worth enough to cause the core to be more than 0.013 b,k supercritical if they were to drop out of the core in the manner defined for the Rod Drop Accident.( )

These sequences are developed prior to initial operation of the unit following any refueling outage and the requirement that an operator follow the sequences is backed up by the operation of the RWM. This 0.013 b,k limit, together with the integral rod velocity limiters and the action of the control rod drive system, limits potential reactivity insertion such that the results of a control rod drop accident will not exceed a maximum fuel energy content of 280 cal/gm. The peak fuel enthalpy content of 280 cal/gm is below the energy content at which rapid fuel dispersal and primary system damage have been found to occur based on experimental data as is discussed in reference 1.

Recent improvements in analytical capability have allowed more refined analysis of the control rod drop accident.

These techniques have been described In a topical report, two supplements and letters to the AEC. ( ) ) . By using the analytical models described in these reports coupled with conservative or worst-case input parameters, it has been determined that for power levels less than 20% of rated power, the specified limit on in-sequence control rod or control rod segment worths will limit the peak fuel enthalpy content to less than 280 cal/gm. Above 20% power, even multiple operator errors cannot result in a peak fuel enthalpy content of 280 cal/gm should a postulated control rod drop accident occur.

The following conservative or worst-case bounding assumptions have been made in the analysis used to determine the specified 0.013 hk limit on in-sequence control rod or control rod segment worths. The allowable boundary conditions used in the analysis are quantified in references (4) and (5). Each core reload will be analyzed to show conformance to the limiting parameters.

NFSAR AMENDMENT NO. 39

BASES FOR 3.1.1 AND 4.1.1 CONTROL ROD SYSTEM

a. A startup inter-assembly local power peaking factor of 1.30 or less.l )

b. An end of cycle delayed neutron fraction of 0.005.

c. A beginning of life Doppler reactivity feedback.

d. The Technical Specification rod scram insertion rate.

e. The maximum possible rod drop velocity (3.11 ft/sec).

f. The design accident and scram reactivity shape function.

g. The moderator temperature at which criticality occurs.

It is recognized. that these bounds are conservative with respect to expected operating conditions. If any one of the above conditions is not satisfied, a more detailed calculation will be done to show compliance with the 280 cal/gm design limit.

In most cases the worth of in-sequence rods or rod segments will be substantially less than 0.013 bk. Further, the addition of 0.013 hk worth of reactivity as a result of a rod drop in conjunction with the actual values of the other important accident analysis parameters described above would most likely result in a peak fuel enthalpy substantially less than the 280 cal/gm design limit. However, the 0.013 4k limit is applied in order to allow room for future reload changes and ease of verification without repetitive Technical Specification changes.

Should a control rod drop accident result in a peak fuel energy content of 280 cal/gm, less than 660 (7 x 7) fuel rods are conservatively estimated to perforate. This would result in offsite doses greater than previously reported in the FSAR, but still well below the guideline values of 10 CFR 100. For 8 x 8 fuel, less than 850 rods are conservatively estimated to perforate, which has nearly. the same consequences as for the 7 x 7 fuel case because of the operating rod power rod differences.

AMENDMENT NO. 40

BASES FOR 3.1.1 AND 4.1.1 CONTROL ROD SySTEM The RWM provides automatic supervision to assure that out-of-sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. It serves as an independent backup of the normal withdrawal procedure followed by the, operator. In the event that the RWM is out of service when.=

required, a second independent operator or engineer can manually fulfill the operator-follower control rod pattern conformance function of the RWM. In this case, procedural control is exercised by verifying all control rod positions after the withdrawal of each group, prior to proceeding to the next group. Allowing substitution of a second independent operator or engineer in case of RWM inoperability recognizes the capability to adequately monitor proper rod sequencing in an alternate manner without unduly restricting plant operations. Above 20% power, there is no requirement that the RWM be operable since the control rod drop accident with out-of-sequence rods will result in a peak fuel'energy content of less than 280 cal/gm. To assure high RWM availability, the RWM is required to be operating during a startup for the withdrawal of a significant number of control rods for any startup.

(4) The source range monitor (SRM) system performs no automatic safety function. It does provide the operator with a visual indication of neutron level which is needed for knowledgeable and efficient reactor startup at low neutron levels.

The results of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 cps assures that any transient at or above the initial value of 10 of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to critical using homogeneous patterns of scattered control rods. A minimum of three operable SRMs is required as an added conservation.

C. Scram Insertion Times The revised scram insertion times have been established as the limiting condition for operation since the postulated rod drop analysis and associated maximum in-sequence control rod worth are based on the revised scram insertion times. The specified times are based on design requirements for control rod scram at reactor pressures above 950 psig. For reactor pressures above 800 psig and below 950 psig the measured scram times may be longer. The analysis discussed in the next paragraph is still valid since the use of the revised scram insertion times would result in greater margins to safety valves lifting.

AMENDMENT NO. 41

BASES FOR 3.1.1 AND 4.1.1 CONTROL ROD SYSTEM The insertion times previously selected were based on the large number of actual scrams of prototype control rod drive mechanisms as discussed in Section IV-B.6.3". Rapid control rod insertion following a demand to scram will terminate Station transients before any possibility of damage to the core is approached. The primary consideration in setting scram time is to permit rapid termination of steam generation following an isolation transient (i.e., main-steam-line closure or turbine trip without bypass) such that operation of solenoid-actuated relief valves will prevent the safety valves from lifting.

Analyses presented in Appendix E-1", the Second Supplement and the Technical Supplement to Petition to Increase Power Level were based on times which are slower than the proposed revised times.

The scram times generated at each refueling outage when compared to previous scram times demonstrate that, the control rod drive scram function has not deteriorated.

d. Control Rod Accumulators The basis for this specification was not described in the FSAR and, therefore, is presented in its entirety. Requiring no more than one rnalfunctioned accumulator in any nine-rod square array is based on a series of XY PDQ-4 quarter core problems of a cold, clean core. The worst one in a nine-rod withdrawal sequence resulted in a k ff (1.0-other repeating rod sequence

)

with more rods withdrawn resulted in keff 1.0. At reactor pressures in excess of 800 psig, even those control rods with malfunctioned accumulators will be able to meet required scram insertion times due to the action of reactor pressure. In addition, they may be normally inserted using the control-rod-drive hydraulic system. Procedural control will assure that control rods with malfunctioned accumulators will be spaced in a one-in-nine array rather than grouped together.

Scram Discharge Volume The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram, isolate the reactor coolant system from the containment when required, and to comply with the requirements of the NRC Confirmatory letter of June 24, 1983. The fill/drain test was determined to be an acceptable alternative to a reactor scram test at approximately 50% ROD DENSITY. Performance of a water fill/drain test during cold shutdown will verify that the Scram Discharge Volume is OPERABLE and instrument lines are not plugged. The volume comparison test of water drained equal.yyater used to fill will demonstrate that there is no blockage in the system. By comparing the response of the individual instrument lines during the drain test, partial or complete blockage in one line can be detected.

The SDV Instrumentation/valve response surveillance test will be satisfied anytime a scram occurs (less than or equal to 50%

rod density) or by the fill/drain test not to exceed an operating cycle.

'FSAR AMENDMENT NO. 42

BASES FOR 3.1.1 AND 4.1.1 CONTROL ROD SYSTEM Reactivity Anomalies During each fuel cycle excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary controls is burned. The magnitude of this excess reactivity is indicated by the integrated worth of control rods inserted into the core, referred to as the control rod inventory in the core. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of actual rod inventory at any base equilibrium core state to predicted rod inventory at that state. Equilibrium xenon, samarium and power distribution are considered in establishing the steady-state base condition to minimize any source of error. During an initial period, (on the order of 1000 MWD/T core average exposure following core reloading or modification) rod inventory predictions can be normalized to actual rod patterns to eliminate calculational uncertainties. Experience with other operating BWR's indicates that the control rod inventory should be predictable to the equivalent of one percent in reactivity. Deviations beyond this magnitude would not be expected and would require thorough evaluation. One percent reactivity limit is considered safe since an insertion of this reactivity into the core would not lead to transients exceeding design conditions of the reactor system.

(1) Paone, C. J., Stirn, R.C., and Wooley, J.A., "Rod Drop Accident Analysis for Large Boiling Water Reactors," NEDO-10527, March 1972.

- NEDO-(2) Stirn, R. C., Paone, C. J., and Young, R. M., "Rod Drop Accident Analysis for Large BWRs," Supplement 1 10527, July 1972.

(3) Stirn, R. C., Paone, C. J., and Haun, J. M., "Rod Drop Accident Analysis for Large Boiling Water Reactors Addendum No. 2 Exposed Cores," Supplement 2- NED0-10527, January 1973.

(4) Report entitled "Technical Basis for Changes to Allowable Rod Worth Specified in Technical Specification 3.3.B.3,"

transmitted by letter from L. O. Mayer (NSP) to J. F. O'eary (USAEC) dated October 4, 1973.

(5) Letter, R. R. Schneider, Niagara Mohawk Power Corporation to A. Giambusso, USAEC, dated November 15, 1973.

,r ~

(6) To include the power spike effect caused by gaps between fuel pellets.

AMENDMENT NO. 43

0 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT 3.1.2 LI UID POISON SY TEM 4.1.2 LIQUID POISON SYSTEM Applies to the operating status of the liquid poison Applies to the periodic testing requirements for the system. liquid poison system.

~Ob'ec ive: ~ob ec ive:

To assure the capability of the liquid poison system to To specify the tests required to assure the capability of function as an independent reactivity control the liquid poison system for controlling core reactivity.

mechanism.

a. During periods when fuel is in the reactor and the The liquid poison system surveillance shall be performed reactor is not shutdown by the control rods, the as indicated below:

liquid poison system shall be operable except as specified in 3.1.2.b. a. v rail S m Tes:

b. If a redundant component becomes inoperable, (1) Iea on e urin each era in le-Specification 3.1.2.a shall be considered fulfilled, provided that the component is returned to an Manually initiate the system from the control operable condition within 7 days and the additional room. Demineralized water shall be pumped to surveillance required is performed. the reactor vessel to verify minimum flow rates and demonstrate that valves and nozzles aie not clogged.

AMENDMENT NO.

0 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. The liquid poison tank shall contain a minimum of Remove the squibs from the valves and verify that 1185 gallons of boron bearing solution. The no deterioration has occurred by actual field firing of solution shall have a sufficient concentration of the removed squibs. In addition, field fire one squib sodium pentaborate enriched with Boron-10 isotope from the batch of replacements.

to satisfy the equivalency equation.

C 13% wt x 628300 M

x~x 86 GPM 6

19.8% Atom Disassemble and inspect the squib-operated valves to verify that valve deterioration has not occurred.

(2) A Ieas once ermonh-Where: C = Sodium Pentaborate Solution Concentration (Wt %) Demineralized water shall be recycled to the test tank. Pump discharge pressure and minimum flow M = Mass of Water in Reactor Vessel and rate shall be verified.

Recirculation piping at Hot Rated Conditions (501500 Ib) b. Bor n olui nCh k:

Q = Liquid Poison Pump Flow Rate (30 GPM (1) A Ie nc rm nh-nominal)

Boron concentration shall be determined.

E = Boron-10 Enrichment (Atom %)

(2) A leas once r

d. The liquid poison solution temperature shall not be less than the temperature presented in Figure Solution volume shall be checked. In addition, the 3.1.2.b. sodium pentaborate concentration shall be determined and conformance with the requirements

e. If Specifications "a" through "d" are not met, of the equivalency equation shall be checked any initiate normal orderly shutdown within one hour. time water or boron are added or if the solution temperature drops below the limits specified by Figure 3.1.2.b.

AMENDMENT NO. 45

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (3) A leas once er da The solution temperature shall be checked.

(4) A leas once er era in I Verify enrichment by analysis.

b When a component becomes inoperable its redundant component shall be verified to be operable immediately and daily thereafter.

AMENDMENT NO. 46

Figure 3.1;2b MINIMUH ALLOWABLE SOLUTION TEMPERATURE 160 140 rx 120 100 Pn 80 60 40 10 20 30 40 50

. AMENDMENT NO.

47

0 BASES FOR 3.1.2 AND 4.1.2 LIQUID POISON SYSTEM The liquid poison system (Section Vll-C)" acting alone does not prevent fuel clad damage for any conceivable type of Station transient.

This system provides a backup to permit reactor shutdown in the event of a massive failure of the control rods to insert.

The liquid poison system is designed to provide the capability to bring the reactor from full design rating (1850 thermal megawatts) to a cold, xenon free shutdown condition assuming none of the control rods can be inserted. A concentration of 120 ppm of boron-10 (the boron isotope with a high neutron cross section) in the reactor coolant will bring the reactor from full design rating (1850 thermal megawatts) to greater than 3 percent b,k subcritical (0.97 keff) considering the combined effects of the control rods, coolant voids, temperature change, fuel doppler, xenon, and samarium.

In order to provide good mixing, the injection time has to be greater than 17 minutes. The rate of boron-10 injection must also be sufficient to achieve hot shutdown during ATWS events.

The liquid poison storage tank minimum volume assures that the above requirements for boron solution insertion are met with one 30 gpm liquid poison pump. The quantity of Boron-10 isotope required to be stored in solution includes an additional 25 percent margin beyond the amount needed to shutdown the reactor to allow for any unexpected non-uniform mixing. The relationship between sodium pentaborate concentration and sodium pentaborate Boron-10 enrichment must satisfy the equivalency equation:

C x628300x~x 6 ~ 1 13% wt M 86 GPM 19.8% Atom Where: C Sodium Pentaborate Solution Concentration (Wt %)

M Mass of Water in Reactor Vessel and Recirculation piping at Hot Rated Conditions (501500 Ib)

Q Liquid Poison Pump Flow Rate (30 GPM nominal)

E Boron-10 Enrichment (Atom %)

The tank volume requirements include consideration for 197 gallons of solution which is contained below the point where the pump takes suction from the tank and therefore cannot be inserted into the reactor.

The solution saturation temperature varies with the concentration of sodium pentaborate. Figure 3.1.2.b includes a 5'F margin above the saturation temperature to guard against precipitation. Temperature and liquid level alarms for the system are annunciated in the Control Room.

NFSAR (1) GE Topical Report NEDE-31096-P-A, "Anticipated Transients Without Scram. Response to ATWS Rule 10 CFR 50.62."

(2) GE Report NEDC-30921, "Assessment of ATWS Compliance Alternatives."

AMENDMENT NO.

BASES FOR 3.1.2 AND 4.1.2 LIQUID POISON SYSTEM Nearly all maintenance can be completed within a few days. Infrequently, however, major maintenance might be required. Replacement of principal system components could necessitate outages of more than 7 days. In spite of the best efforts of the operator to return equipment to service, some maintenance could require up to 6 months.

The system test specified demonstrates component response such as pump starting upon manual system initiation and is similar to the operating requirement under accident conditions. The only difference is that demineralized water rather than the boron solution will be pumped to the reactor vessel. The test interval between operating cycles results in a system failure probability of 1.1 x 10 (Fifth Supplement, p. 115)" and is consistent with practical considerations.

Pump operability will be demonstrated on a more frequent basis. A continuity check of the firing circuit on the explosive valves is provided by pilot lights in the control room. Tank level and temperature alarms are provided to alert the operator of off-normal conditions.

The functional test and other surveillance on components, along with the monitoring instrumentation, gives a high reliability for liquid poison system operability.

"FSAR AMENDMENT NO. 49

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.3 EMERGENCY COOLING SY TEM 4.1.3 EMERGENCY COOLIN YSTEM UIILJIJis Applies to the operating status of the emergency Applies to periodic testing requirements for the cooling system. emergency cooling system.

~Ob ei~iv: ~Ob c ive:

To assure the capability of the emergency cooling To assure the capability of the emergency cooling system to cool the reactor coolant in the event the system for cooling of the reactor coolant, normal reactor heat sink is not available.

The emergency cooling system surveillance shall be

a. During power operating conditions and whenever performed as indicated below:

the reactor coolant temperature is greater than 212'F except for hydrostatic testing with the a. A legs ne vr five er reactor not critical, both emergency cooling systems shall be operable except as specified, in 3.1.3.b. The system heat removal capability shall be determined.

b. If one emergency cooling system becomes inoperable, Specification 3.1.3.a shall be considered fulfilled, provided that the inoperable system is returned to an operable condition within 7 days and. The shell side water level and makeup tank water the additional surveillance required in 4.1.3.f is level shall be checked.

performed.

AMENDMENT NO. 50

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

c. Make up water shall be available from the two c. At least once r m n h-gravity feed makeup Water tanks.

The makeup tank level control valve shall be

d. During Power Operating Conditions, each manually opened and closed.

emergency cooling system high point vent to torus shall be operable. d. A least once each hif-

1. With a vent path for one emergency cooling The area temperature shall be checked.

system inoperable, restore the vent path to an operable condition within 30 days. e. Durin each ma or refuelin ou a e-

2. With vent paths for both emergency cooling Automatic actuation and functional system testing systems inoperable, restore one vent path to an shall be performed during each major refueling operable condition with 14 days and both vent outage and whenever major repairs are completed paths within 30 days. on the system.

e. If Specification 3.1.3.a, b, c, or d are not met, a Each emergency cooling vent path shall be demon-normal orderly shutdown shall be initiated within strated operable by cycling each power-operated one hour, and the reactor shall be in the cold valve (05-01R, 05-11, 05-12,05-04R, 05-05 and shutdown conditions within ten hours. 05-07) in the vent path through one complete cycle of full travel and verifying that all manual valves are in the open position.

f. urveillan e wi h-an Ino era le When one of the emergency cooling systems is inoperable, the level control valve and the motor-operated isolation valve in the operable system shall be verified to be operable immediately and daily thereafter.

AMENDMENT NO.

BASES FOR 3.1.3 AND 4.1.3 EMERGENCY COOLING SYSTEM The turbine main condenser is normally available. The emergency cooling system (Section V-E)" is provided as a redundant backup for core decay heat removal following reactor isolation and scram. One emergency condenser system has a heat removal capacity at normal pressure of 19.0 x 10 Btu/hr, which is approximately three percent of maximum reactor steam flow. This capacity is sufficient to handle the decay heat production at 100 seconds following a scram. If only one of the emergency cooling systems is available, 2000 pounds of water will be lost from the reactor vessel through the relief valves in the 100 seconds following isolation and scram. This represents a minor loss relative to the vessel inventory of about 450,000 pounds (Section V-E.3.1)".

The required heat removal capability is based on the data of Table V-1" adjusted to normal operating pressures. The only difference is manual system initiation rather than automatic initiation.

The system may be manually initiated at any time. The system is automatically initiated on high reactor pressure in excess of 1080 psig sustained for 12 seconds. The time delay is provided to prevent unnecessary actuation of the system during anticipated turbine trips (Section XV-B.3.15)". Automatic initiation is provided to minimize the coolant loss following isolation from the main condenser."" To assist in depressurization for small line breaks the system is initiated on low-low reactor water level five feet (5 inches indicator scale>

below the minimum normal water level (Elevation 302'9") sustained for 12 seconds. The timers for initiation of the emergency condensers will be set at 12 seconds delay based on the analysis (Section XV-B.3.15)". For the MSIV closure analysis (Section XV-B.3.5)",

emergency condenser action is ignored.

The initial water volume in each emergency condenser is 21,360 a1500 gallons which keeps the level within a6 inches of the normal water level. About 72,000 gallons are available from the two gravity feed condensate storage tanks. To assure this gallonage, a level check shall be done at least once per day.

This is sufficient to provide about eight hours of continuous system operation. This time is sufficient to restore additional heat sinks or pump makeup water from the two-200,000 gallon condensate storage tanks. The fire protection is also available as a makeup water supply.

"UFSAR

""Technical Supplement to Petition to Increase Power Level AMENDMENT NO. 52

BASES FOR 3.1.3 AND 4.1.3 EMERGENCY COOLING SYSTEM Nearly all maintenance can be completed within a few days. Infrequently, however, major maintenance might be required. Replacement of principal system components could necessitate outages of more than 7 days. In spite of the best efforts of the operator to return equipment to service, some maintenance could require up to 6 months.

The system heat removal capability shall be determined at five-year intervals. This is based primarily on the low corrosion characteristics of the stainless steel tubing. During normal plant operation the water level will be observed at least once daily on emergency condensers and makeup water tanks. High and low water level alarms are also provided on the above pieces of equipment. The test frequency selected for level checks and valve operation is to assure the reliability of the system to operate when required.

The emergency cooling system is provided with high point vents to exhaust noncondensible gases that could inhibit natural circulation cooling. Valve redundancy in the vent path serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path. The function, capabilities and testing requirements of the emergency cooling vent paths are consistent with the requirements of item II.B.1 of NUREG 0737, "Clarification of TMI Action Plan Requirement," November 1980.

AMENDMENT NO. 53

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.4 CORE SPRAY SYSTEM 4.1.4 CORE SPRAY Y TEM b'i Applies to the operating status of the core spray Applies to the periodic testing requirements for the systems. core spray systems.

~Ob ec ice: ~beehive:

To assure the capability of the core spray systems to To verify the operability of the core spray systems.

cool reactor fuel in the event of a loss-of-coolant accident.

a. Whenever irradiated fuel is in the reactor vessel The core spray system surveillance shall be and the reactor coolant temperature is greater performed as indicated below.

than 212'F, each of the two core spray systems shall be operable except as specified in a. At each major refueling outage automatic Specifications b and c below. actuation of each subsystem in each core spray system shall be demonstrated.

b. If a redundant component of a core spray system becomes inoperable, that system shall be b. At least once per quarter pump operability shall considered operable provided that the component be checked.

.is returned to an operable condition within 7 days and the additional surveillance required is c. At least once per quarter the operability of performed. power-operated valves required for proper system operation shall be checked.

c. If a redundant component in each of the core spray systems becomes inoperable, both systems shall be considered operable provided that the component is returned to an operable condition within 7 days and the additional surveillance required is performed.

AMENDMENT NO. 54

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

d. If Specifications a, b and c are not met, a normal d. Core spray header h,P instrumentation orderly shutdown shall be initiated within one hour and the reactor shall be in the cold check Once/day shutdown condition within ten hours. calibrate Once/3 months test Once/3 months

e. During reactor operation, except during core e. urv illan wi h Ino ra I m n n spray system surveillance testing, core spray isolation valves 40-02 and 40-12 shall be in the When a component becomes inoperable its open position and the associated valve motor redundant component or system shall be. verified starter circuit breakers for these valves shall be to be operable immediately and daily thereafter.

locked in the off position. In addition, redundant valve position indication shall be available in the control room.

f. Whenever irradiated fuel is in the reactor vessel f. With a core spray subsystem suction from the and the reactor coolant temperature is less than CST, CST level shall be checked once per day.

or equal to 212'F, two core spray subsystems shall be operable except as specified in g and h below.

g. If one of the above required subsystems g. At least once per month when the reactor becomes inoperable, restore at least two coolant temperature is greater than 212'F, verify subsystems to an operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> that the piping system between valves 40-03, or suspend all operations that have a potential ... 13 and 40-01, 09, 10, 11 is filled with water.

for draining the reactor vessel.

AMENDMENT NO. 55

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

h. If both of the above required subsystems become inoperable, suspend core alterations and all operations that have a potential for draining the reactor vessel. Restore at least one subsystem to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or establish secondary containment integrity within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With the downcomers in the suppression chamber having less than three and one half foot submergence, two core spray subsystems and the associated raw water pumps shall be operable with the core spray suction from the condensate storage tanks (CST), and the CST inventory shall not be less than 300,000 gallons.

AMENDMENT NO. 56

BASES FOR 3.1.4 AND 4.1 4 CORE SPRAY SYSTEM The core spray system consists of two automatically actuated, independent systems capable of cooling reactor fuel for a range of loss-of-coolant accidents. Each of the two independent systems consists of 2 subsystems having one pump set of a core spray pump and core spray topping pump. Both systems (at least one subsystem in each system) are required to operate to limit peak clad temperatures below 2200'F (10 CFR 50 Appendix K model) for the worst case line break (recirculation line break at the point where the emergency condenser return line connects to the recirculation loop). When a component/subsystem is in a LCO state, additional surveillance requirements are imposed for the redundant component/subsystem. Consequently, application of the single failure criteria to the redundant component/

subsystem is not a design requirement during the LCO period.

Allowable outage time is specified to account for redundant components that become inoperable.

Both core spray systems contain redundant supply pump sets and blocking valves. Operation of one pump set and blocking valve is sufficient to establish required delivery rate and flow path. Therefore, even with the loss of one of the redundant components, the system is still capable of performing its intended function. If a redundant component is found to have failed, corrective maintenance will begin promptly. Nearly all maintenance can be completed within a few days. Infrequently, however, major maintenance might be required.

Replacement of principal system components could necessitate outages in excess of those specified. In spite of the best efforts of the operator to return equipment to service, some maintenance could require up to 6 months.

In determining the operability of a core spray system the required performance capability of its various components shall be considered. For example:

Periodic tests will demonstrate that adequate core cooling is provided to satisfy the core spray flow requirements used in the 10 CFR 50 Appendix K analysis.

2. The pump shall be capable of automatic initiation from a low-low water level signal in the reactor vessel or a high containment pressure signal. The blocking valves shall be capable of automatically opening from either a low-low water signal or high containment pressure signal simultaneous with low reactor pressure permissive signal (Section VII)".

"FSAR AMENDMENT NO. 57

BASES FOR 3.1.4 AND 4.1 4 CORE SPRAY SYSTEM Instrumentation has been installed to monitor the integrity of the core spray piping within the reactor pressure vessel.

The testing specified for each major refueling outage will demonstrate component response upon automatic system initiation. For example, pump set starting (low-low level or high drywell pressure) and valve opening (low-low level or high drywell pressure and low reactor pressure) must function, under simulated conditions, in the same manner as the systems are required to operate under actual conditions.

The only differences will be that demineralized water rather than suppression chamber water will be pumped to the reactor vessel and the reactor will be at atmospheric pressure. The core spray systems are designed such that demineralized water is available to the suction of one set of pumps in each system (Section Vll-Figure Vll-1)".

The system test interval between operating cycles results in a system failure probability of 1.1 x 10 (Fifth Supplement, page 115) and is consistent with practical considerations. The more frequent component testing results in a more reliable system.

At quarterly intervals, startup of core spray pumps will demonstrate pump starting and operability. No flow will take place to the reactor vessel due to the lack of a low-pressure permissive signal required for opening of the blocking valves. A flow restricting device has been provided in the test loop which will create a low pressure loss for testing of the system. In addition, the. normally closed power operated blocking valves will be manually opened and re-closed to demonstrate operability.

The intent of Specification 3.1.4i is to allow core spray operability at the time that the suppression chamber is dewatered which will allow normal refueling activities to be performed. With a core spray pump taking suction from the CST, sufficient time is available to manually initiate one of the two raw water pumps that provide an alternate core spray supply using lake water. Both raw water pumps shall be operable in the event the suppression chamber was dewatered.

%FSAR AMENDMENT NO. 58

tp BASES FOR 3.1.4 AND 4.1 4 CORE SPRAY SYSTEM Based on the limited time involved in performance of the concurrent refueling maintenance tasks, procedural controls to minimize the potential and duration of leakage and available coolant makeup (CST) provides adequate protection against drainage of the vessel while the suppression chamber is drained.

Specification 3.1.4e establishes provisions to eliminate a potential single failure mode of core spray isolation valves 40-'02 and 40-12.

These provisions are necessary to ensure that the core spray system safety function is single failure proof. During system testing, when the isolation valve(s) are required to be in the closed condition, automatic opening signals to the valve(s) are operable if the core spray

. system safety function is required.

ln the cold shutdown and refuel conditions, the potential for a LOCA due to a line break is much less than during operation. In addition, the potential consequences of the LOCA on the fuel and containment is less due to the lower reactor coolant temperature and pressures.

Therefore, one subsystem of a core spray system is sufficient to provide adequate cooling for the fuel during the cold shutdown or refueling conditions. Therefore, requiring two core spray subsystems to be operable in the cold shutdown and refuel conditions provides sufficient redundancy.

AMENDMENT NO. 59

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.5 S LENOID-ACTUATED PRESSURE RELIEF VALVES 4.1.5 SOLENOID-A T ATED PRE RE RELIEF VALVE A T MATIC DEPRE SURIZATION SYSTEM A TOMATI DEPRE S RIZATION SY TEM Applies to the operational status of the solenoid- Applies to the periodic testing requirements for the actuated relief valves. solenoid-actuated pressure relief valves.

~be~iv: ~Ob'ective:

To assure the capability of the solenoid-actuated To assure the operability of the solenoid-actuated pressure relief valves to provide a means of. pressure relief valves to perform their intended depressurizing the reactor in the event of a small line functions.

break to allow full flow of the core spray system.

a. During power operating condition. whenever the The solenoid-actuated pressure relief valve reactor coolant pressure is greater than 110 psig surveillance shall be performed as indicated below.

and the reactor coolant temperature is greater than saturation temperature, all six solenoid- a. At least once during each operating cycle with actuated pressure relief valves shall be operable. the reactor at pressure, each valve shall be manually opened until acoustic monitors or

b. If specification 3.1.5a above is not met, the thermocouples downstream of the valve indicate reactor coolant pressure and the reactor coolant. that the valve has opened and steam is flowing temperature shall be reduced to 110 psig or less from the valve.

and saturation temperature or less, respectively, within ten hours. b. At least once during each operating cycle, automatic initiation shall be demonstrated.

AMENDMENT NO. 60

BASES FOR 3.1.5 AND 4.1.5 SOLENOID-ACTUATED PRESSURE RELIEF VALVES Pres r Blowd own In the event of a small line break, substantial coolant loss could occur from the reactor vessel while it was still at relatively high pressures.

A pressure blowdown system is provided which in conjunction with the core spray system will prevent significant fuel damage for all sized line breaks (Appendix E-11.2.0) ".

Operation of three solenoid-actuated pressure relief valves is sufficient to depressurize the primary system to 110 psig which will permit full flow of the core spray system within required time limits (Appendix E-11.2)". Requiring all six of the relief valves to be operable, therefore, provides twice the minimum number required. Prior to or following refueling at low reactor pressure, each valve will be manually opened to verify valve operability. The malfunction analysis (Section II.XV, "Technical Supplement to Petition to Increase Power Level," dated April 1970) demonstrates that no serious consequences result if one valve fails to close since the resulting blowdown is well within design limits.

In the event of a small line break, considerable time is available for the operator to permit core spray operation by manually depressurizing the vessel using the solenoid-actuated valves. However, to ensure that the depressurization will be accomplished, automatic features are provided. The relief valves shall be capable of automatic initiation from simultaneous low-low-low water level (6 feet, 3 inches below minimum normal water level at Elevation 302'-9", -10 inches indicator scale) and high containment pressure (3.5 psig). The system response to small breaks requiring depressurization is discussed in Section VII-A.3.3" and the time available to take operator action is summarized in Table Vll-1". Additional information is included in the answers to Questions III-1 and III-5 of the First Supplement.

Steam from the reactor vessel is discharged to the suppression chamber during valve testing. Conducting the tests with the reactor at nominal operating pressure is appropriate because 1) adequate redundant safety systems are provided to ensure adequate core cooling in the event of a small break loss of feedwater, and multiple relief valve failures, 2) dynamic loads and suppression pool heatups associated with high pressure testing are within allowable limits, and 3) testing at nominal operating pressures enhances plant safety and availability by assuring the relief valves can operate under normal operating conditions.

The test interval of once per operating cycle results in a system failure probability of 7.0 x 10 (Fifth Supplement, p. 115)" and is consistent with practical consideration.

%FSAR AMENDMENT NO. 61

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT 3.1.6 ONTROL R 0 DRIVE PUMP 0 LANT INJECTION 4.1.6 CONTROL ROD DRIVE P MP OLANT IN E TI N bi i Applies to the operational status of the control rod Applies to the periodic testing requirements for the drive pump coolant injection system. control rod drive pump coolant injection system.

~b'ec ive: ~Ob'ec ive:

To assure the capability of the control rod drive pump To assure the capability of the control rod drive pump coolant injection system to: coolant injection system in performing its intended functions.

Provide core cooling in the event of a small line break, and Provide coolant makeup in the event of reactor coolant leakage (see LCO 3.2.5).

The control rod drive pump coolant injection system surveillance shall be performed as indicated below.

a. Whenever irradiated fuel is in the reactor vessel a. A least once r eratin cle-and the reactor coolant temperature is greater ...

than 212'F, the control rod drive pump coolant Automatic starting of each pump shall be injection system shall be operable except as demonstrated.

specified in "b" below.

AMENDMENT NO. 62

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT b.'f a redundant component becomes inoperable, b. A lea nce r u er-the control rod drive pump coolant injection system shall be considered operable provided Pump flow rate shall be determined.

that the component is returned to an operable condition within 7 days and the additional surveillance required is performed.

c. If Specifications "a" or "b" above are not met, c. Surveillan e wi h In arable C m onen the reactor coolant temperature shall be reduced to 212'F or less within ten hours. When a component becomes inoperable, its redundant component shall be verified to be operable immediately and daily thereafter.

AMENDMENT NO. 63

BASES FOR 3.1.6 AND 4.1.6 CONTROL ROD DRIVE PUMP COOLANT INJECTION The high pressure coolant injection capability of the control rod drive pumps is used to provide high pressure makeup for the specified leakage of 25 gpm (see LCO 3.2.5) and to provide core cooling in the case of a small line break. Each pump can supply 50 gpm water makeup to the reactor vessel.

One pump will normally be operating. Electric power for this system is normally available from the reserve transformer. Automatic initiation is provided to start each pump on its respective diesel generator in case offsite power is lost.

The system minimum delivery rate of 50 gpm within 60 seconds of receipt of signal will assure that automatic pressure blowdown is not actuated for the specified leakage rate of 25 gpm.

The 60-second delay in pump starting is acceptable since at least 15 minutes are available before the triple Iow reactor water level signals the automatic pressure blowdown to start. This analysis was based on the following.assumptions; no makeup to the reactor vessel, a 50 gpm (two times allowable) leak rate exists, and the emergency condensers over-perform by 10 percent.

Nearly all maintenance can be completed within a few days. Infrequently, however, major maintenance might be required. Replacement of principal system components could necessitate outages of more than 7 days. In spite of the best efforts of the operator to return equipment to service, some maintenance could require up to 6 months.

The testing specified during an operating cycle will demonstrate component response upon automatic system initiation in the same manner that the system will operate if required. The testing interval results in a calculated failure probability of 1.1 x 10 for a control rod drive pump (Fifth Supplement), and is compatible with practical considerations. Continual monitoring of pump performance is provided since one pump is normally operating and instrumentation and alarms monitor operation of flow and pressure regulation (Section X)"..

"FSAR AMENDMENT NO. 64

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVlENT 3.1.7 FUEL ROD 4.1.7 FIISL RODS eteteev vi:

The Limiting Conditions for Operation associated with The Surveillance Requirements apply to the the fuel rods apply to those parameters which parameters which monitor the fuel rod operating monitor the fuel rod operating conditions. conditions.

~Ob e ive: ~Ob ec ive:

The objective of the Limiting Conditions for Operation The objective of the Surveillance Requirements is to is to assure the performance of the fuel rods. specify the type and frequency of surveillance to be applied to the fuel rods.

a. Av r e Planar Linear Hea Genera ion Ra e APLHGR a. Avera e Planar Linear Hea Gen ra ion Ra APLHGR During power operation, the APLHGR for each type of fuel as a function of average planar The APLHGR for each type of fuel as a function exposure shall not exceed the limiting value of average planar exposure shall be determined provided in the Core Operating Limits Report. If daily during reactor operation at a25 percent at any time during power operation it is rated thermal power.

determined by normal surveillance that the limiting value for APLHGR is being exceeded at any node in the core, action shall be initiated within 15 minutes to restore operation to within" the prescribed limits. If the APLHGR at all nodes in the core is not returned to within the prescribed limits within two (2) hours, reactor power reductions shall be initiated at a rate not less than 10% per hour until APLHGR at all nodes is within the prescribed limits.

AMENDMENT NO. 65

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. Lin ar Hea enera ion Ra e (LHGR) b. Linear He enera i n Ra e (LHGR)

During power operation, the Linear Heat The LHGR as a function of core height shall be Generation Rate (LHGR) of any rod in any fuel checked daily during reactor operation at ~25%

assembly at any axial location shall not exceed rated thermal power.

the limiting value specified in the Core Operating Limits Report.

If at any time during power operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded at any location, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR at all locations is not returned to within the prescribed limits within two (2) hours, reactor power reductions shall be initiated at a rate not less than 10% per hour until LHGR at all locations is within the prescribed limits.

C. Minim m Cri ical Power Ra io MCPR c. Minimum Critical Power Ra io (MCPR)

During power operation, the MCPR for all fuel at MCPR shall be determined daily during reactor rated power and flow shall be within the limit power operation at ) 25% rated thermal power.

provided in the Core Operating Limits Report.

If at any time during power operation it is determined by normal surveillance that the above'imit is no longer met, action shall be initiated within 15 minutes to restore operation to within the prescribed limit. If all the operating MCPRs are not returned to within the prescribed limit within two (2) hours, reactor power reductions shall be initiated at a rate not less than 10% per hour until MCPR is within the prescribed limit.

AMENDMENT NO. 66

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT For core flows other than rated, the MCPR limit shall be the limit identified above times Kf where Kf is provided in the Core Operating Limits Report.

d. Power Flow Rela ion hi Durin 0 era ion d. Power Fl w Rela i n hi This power/flow relationship shall not exceed the Compliance with the power flow relationship in limiting values shown in the Core Operating Section 3.1.7.d shall be determined daily during Limits Report. reactor operation.

lf at any time during power operation it is determined by normal surveillance that the limiting value for the power/flow relationship is being exceeded, action shall be initiated within 1S minutes to restore operation to within the prescribed limits. If the power/flow relationship is not returned to within the prescribed limits within two (2) hours, reactor power reductions shall be initiated at a rate not less than 10% per hour until the power/flow relationship is within the prescribed limits.

e. Pa ial Loo 0 era i n e. Par ial Loo 0 era ion During power operation, partial loop operation is Under partial loop operation, surveillance permitted provided the following conditions are requirements 4.1.7, a, b, c and d above are met. applicable.

AMENDMENT NO. 67

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT When operating with four recirculation loops in operation and the remaining loop unisolated, the reactor may operate at 100% of full licensed power level in accordance with the power/flow limits specified in the Core Operating Limits Report and an APLHGR not to exceed the applicable limiting values provided in the Core Operating Limits Report for the fuel type.

When operating with four recirculation loops in operation and one loop isolated, the reactor may operate at 100 percent of full licensed power in accordance with the power/flow limits specified in the Core Operating Limits Report and an APLHGR not to exceed the applicable limiting values provided in the Core Operating Limits Report for the fuel type, provided the following conditions are met for the isolated loop.

1. Suction valve, discharge valve and discharge bypass valve in the isolated loop shall be in the closed position and the associated motor breakers shall be locked in the open position.

2. Associated pump motor circuit breaker shall.

be opened and the breaker removed.

If these conditions are not met, core power shall be restricted to 90.5 percent of full licensed power.

AMENDMENT NO.

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT When operating with three recirculation loops in operation and the two remaining loops isolated or unisolated, the reactor may operate at 90/o of full licensed power in accordance with the power/flow limits specified in the Core Operating Limits Report and an APLHGR not to exceed the applicable limiting values provided in the Core Operating Limits Report for the fuel type.

During 3 loop operation, the limiting MCPR shall be adjusted as described in the Core Operating Limits Report.

Power operation is not permitted with less than three recirculation loops in operation.

If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR under one and two isolated loop operation is being exceeded at any node in the core, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR at all nodes in the core is not returned to within the prescribed limits for one and two isolated loop operation within two (2) hours, reactor power reduction shall be initiated at a rate not less than.10 percent per hour until APLHGR at all nodes is within the prescribed limits.

AMENDMENT NO. 69

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Recir ula ion Loo s During all operating conditions with irradiated fuel in the reactor vessel, at least two (2) recirculation loop suction valves and their associated discharge valves will be in the full open position except when the reactor vessel is flooded to a level above the main steam nozzles or when the steam separators and dryer are removed.

g. Re or in Re uiremen s If any of the limiting values identified in Specification 3.1.7.a, b, c, d, and e are exceeded, a Reportable Occurrence Report shall be submitted. If the corrective action is taken, as described, a thirty-day written report will meet the requirements of this Specification.

h. 0 erations Be ond he End-of-C cle Coastdown For coastdown operations beyond the End-of-Cycle (i.e., when the core reactivity has decreased such that full power cannot be maintained by further control rod withdrawal),

steady state thermal power shall be limited to forty (40) percent minimum. Increasing core power level via reduced feedwater heating, once operation in the coastdown mode has begun, is not allowed.

AMENDMENT NO. 70

BASES FOR 3.1.7 AND 4.1.7 FUEL RODS Av ra e Pl nar Lin ar H a Generation Rate APLHGR This specification assures that the peak cladding temperature and the peak local cladding oxidation following the postulated design basis loss-of-coolant accident will not exceed the limits specified in 10CFR50, Appendix K.

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod-to-rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than a20 F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10CFR50, Appendix K limit. The limiting value for APLHGR is provided in the Core Operating Limits Report. The APLHGR curves in the Core Operating Limits Report are based on calculations using the models described in References 13, 15 and 16.

The Reference 13 and 15 LOCA analyses are sensitive to minimum critical power ratio (MCPR). In the Reference 15 analysis, an MCPR value of 1.30 was assumed. If future transient analyses should yield a MCPR limit below this value, the Reference 15 LOCA analysis MCPR value would become limiting. The current MCPR limit is provided in the Core Operating Limits Report. For fuel bundles analyzed with the Reference 13 LOCA methodology, assume MCPR values of 1.30 and 1.36 for five recirculation loop and less than five loop operation respectively.

Linear Hea G nera i n Ra e LHGR This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation even if fuel pellet densification is postulated (Reference 12). The LHGR shall be checked daily during reactor operation at z25% power to determine if fuel burnup or control rod movement has caused changes in power distribution.

Minimum Cri ical Power Ratio MCPR At core thermal power levels less than or equal to 25'lo, the reactor will be operating at a minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal-hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin.

With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.

During initial startup testing of the plant, an MCPR evaluation will be made at the 25'/o thermal power level with minimum recirculation pump AMENDMENT NO. 71

BASES FOR 3.1.7 AND 4.1.7 FUEL RODS speed. The MCPR margin will thus be demonstrated such that future MCPR evaluations below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

MCPR limits during operation at other than rated conditions are provided in the Core Operating Limits Report. For the case of automatic flow control, the Kf factor is determined such that any automatic increase in power (due to flow control) will always result in arriving at the nominal required MCPR at 100% power. For manual flow control, the Kf is determined such that an inadvertent increase in core flow (i.e.,

operator error or recirculation pump speed controller failure) would result in arriving at the 99.9% limit MCPR when core flow reaches the maximum possible core flow corresponding to a particular setting of the recirculation pump MG set s'coop tube maximum speed control limiting set screws. These screws are to be calibrated and set to a particular value and whenever the plant is operating in manual flow control, the Kf defined by that setting of the screws is to be used in the determination of required MCPR. This will assure that the reduction in MCPR associated with an inadvertent flow increase always satisfies the 99.9% requirement. Irrespective of the scoop tube setting, the required MCPR is never allowed to be less than the nominal MCPR (i.e., Kf is never less than unity).

PowerFI wR lai nhi The power/flow curve is the locus of critical power as a function of flow from which the occurrence of abnormal operating transients will yield results within defined plant safety limits. Each transient and postulated accident applicable to operation of the plant was analyzed along the power/flow line. The analysis ( ' ") justifies the operating envelope bounded by the power/flow curve as long as other operating limits are satisfied. Operation under the power/flow line is designed to enable the direct ascension to full power within the design basis for the plant.

AMENDMENT NO. 72

BASES FOR 3.1.7 AND 4.1.7 FUEL RODS Par ial Loo 0 era ion The requirements of Specification 3.1.7e for partial loop operation in which the idle loop is isolated, precludes the inadvertent startup of a recirculation pump with a cold leg. However, if these conditions cannot be met, power level is restricted to 90.5 percent power based on current transient analysis (Reference 9). For three loop operation, power level is restricted to 90 percent power based on the Reference 13 and 15 LOCA analyses.

The results of the ECCS calculation are affected by one or more recirculation loops being unisolated and out of service. This is due to the fact that credit is taken for extended nucleate boiling caused by flow coastdown in the unbroken loops. The reduced core flow coastdown following the break results in higher peak clad temperature due to an earlier boiling transition time. The results of the ECCS calculations are also affected by one or more recirculation loops being isolated and out of service. The mass of water in the isolated loops unavailable during blowdown results in an earlier uncovery time for the hot node. This results in an increase in the, peak clad temperature.

For fuel bundles analyzed with the methodology used in Reference 13, MAPLHGR shall be reduced as required in the Core Operating Limits Report for 4 and 3 loop operation. For fuel bundles analyzed with the methodology used in References 15 and 16, MAPLHGR shall be reduced as required in the Core Operating Limits Report for both 4 and 3 loop operation.

Partial loop operation and its effect on lower plenum flow distribution is summarized in Reference 11 Since the lower plenum hydraulic

~

design in a non-jet pump reactor is virtually identical to a jet pump reactor, application of these results is justified. Additionally, non-jet pump plants contain a cylindrical baffle plate which surrounds the guide tubes and distributes the impinging water jet and forces flow in a circumferential direction around the outside of the baffle.

R irculation Loo s Requiring the suction and discharge for at least two (2) recirculation loops to be fully open assures that an adequate flow path exists from the annular region between the pressure vessel wall and the core shroud, to the core region. This provides for communication between those areas, thus assuring that reactor water level instrument readings are indicative of the water level in the core region.

When the reactor vessel is flooded to the level of the main steam nozzle, communication between the core region and annulus exists above the core to ensure that indicative water level monitoring in the core region exists. When the steam separators and dryer are removed, safety limit 2.1.1d and e requires water level to be higher than 9 feet below minimum normal water level (Elevation 302'9"). This level is above the core shroud elevation which would ensure communication between the core region and annulus thus ensuring indicative water level monitoring in the core region. Therefore, maintaining a recirculation loop in the full open position in these two instances is not necessary to ensure indicative water level monitoring.

AMENDMENT NO. 73

BASES FOR 3.1.7 AND 4.1.7 FUEL RODS Re or in Re uirements The LCOs associated with monitoring the fuel rod operating conditions are required to be met at all times, i.e., there is no allowable time in which the plant can knowingly exceed the limiting values of MAPLHGR, LHGR, MCPR, or Power/Flow Ratio. It is a requirement, as stated in Specifications 3.1.7a, b, c, and d that if at any time during power operation it is determined that the limiting values for MAPLHGR, LHGR, MCPR, or Power/Flow Ratio are exceeded, action is then initiated to restore operation to within the prescribed limits. This action is initiated as soon as normal surveillance indicates that an operating limit has been reached. Each event involving operation beyond a specified limit shall be reported as a Reportable Occurrence. If the specified corrective action described in the LCOs was taken, a thirty-day written report is acceptable.

0 era i n Be ond he End-of-C cle Coas down The General Electric generic BWR analysis of coastdown operation (Reference 17) concludes that operation beyond the end-of-cycle (coastdown) is acceptable. Amendment No. 7 to GESTAR (Reference 18) concludes that the analysis conservatively bounds coastdown operation to forty (40) percent power. The margin to all safety limits analyzed increased linearly as the power decreased.

AMENDMENT NO. 74

REFERENCES FOR BASES 3.1.7 AND 4.1.7 FUEL RODS References (1) through (6) intentionally deleted.

(7) "Nine Mile Point Nuclear Power Station Unit 1, Load Line Limit Analysis," NED0-24012.

(8) Licensing Topical Report GE Boiling Water Reactor Generic Reload Fuel Application, NEDE-24011-P-A, August 1978.

(9) Final Safety Analysis Report, Nine Mile Point Nuclear Station, Niagara Mohawk Power Corporation, June 1967.

(10) NRC Safety Evaluation, Amendment No. 24 to DPR-63 contained in letter from G. Lear, NRC, to D. P. Disc dated May 15, 1978.

(11) "Core Flow Distribution in a GE Boiling Water Reactor as Measured in Quad Cities Unit 1," NEDO-10722A.

(12) Nine Mile Point Nuclear Power Station Unit 1, Extended Load Line Limit Analysis, License Amendment Submittal (Cycle 6), NEDO-24185, April 1979.

(13) Loss-of-Coolant Accident Analysis Report for Nine Mile Point Unit 1 Nuclear Power Station, NEDO-24348, August 1981.

(14) GE Boiling Water Reactor Extended Load Line Limit Analysis for Nine Mile Point Unit 1 Cycle 9, NEDC-31126, February 1986.

(15) Nine Mile Point Unit 1, Loss-of-Coolant Accident Analysis, NEDC-31446P, June.1987.

(16) Supplement 1 to Nine Mile Point Generating Station Unit 1 SAFER/CORECOOL/GESTR-LOCA Analysis Report NEDC-31446P-1, Class III, September 1987.

(17) Communication: R. E. Engel (GE) to T. A. Ippolito (NRC) - "End-of-Cycle Coastdown Analyzed with ODYN/TASC," dated September 1, 1981.

(18) Amendment No. 7 to GESTAR, NEDE-24011-P-A-7-US,.dated August 1985.

AMENDMENT NO. 75

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.1.8 HI H PRESSURE OLANT INJECTION 4.1.8 HI H PRES URE OOLANT INJE TI N Applies to the operational status of the high pressure Applies to the periodic testing requirements for the coolant injection system. high pressure coolant injection system.

~Ob ec ive: ~Ob ective:

To assure the capability of the high pressure coolant To verify the operability of the high pressure coolant injection system to cool reactor fuel in the event of a injection system.

loss-of-coolant accident.

a. During the power operating condition" whenever The high pressure coolant injection surveillance shall the reactor coolant pressure is greater than 110 be performed as indicated below:

psig and the reactor coolant temperature is greater than saturation temperature, the high a. A least once er o era in I pressure coolant injection system shall be operable except as specified in Specification "b" Automatic start-up of the high pressure coolant below. injection system shall be demonstrated.

b. If a redundant component of the high pressure b. A least once er uar er-coolant injection system becomes inoperable, the high pressure coolant injection shall be Pump operability shall be determined.

considered operable provided that the component is returned to an operable condition within 15 days and the additional surveillance required is

'erformed.

One Feedwater Pump blocking valve in one HPCI pump train may be closed during reactor startup when core power is equal to or less than 25% of rated thermal power.

AMENDMENT NO. 76

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

c. If Specification "a" and "b" are not met; a c. rv illan wi h In erabl C m n n normal orderly shutdown shall be initiated within one hour and reactor coolant pressure and When a component becomes inoperable, its temperature shall be reduced to less than 1.10 redundant component shall be verified to be psig and saturation temperature within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. operable immediately and daily thereafter.

AMENDMENT NO. 77

BASES FOR 3.1.8 AND 4.1.8 HIGH PRESSURE COOLANT INJECTION The High Pressure Coolant Injection System (HPCI) is provided to ensure adequate core cooling in the unlikely event of small reactor coolant line break. The HPCI System is available for line breaks which exceed the capability of the Control Rod Drive pumps and which are not large enough to allow fast enough depressurization for core spray to be effective.

One set of high pressure coolant injection pumps consists of a condensate pump, a feedwater booster pump and a motor driven feedwater pump. One set of pumps is capable of delivering 3,420 gpm to the reactor vessel at reactor pressure. The performance capability of HPCI alone and in conjunction with other systems to provide adequate core cooling for a spectrum of line breaks is discussed in the Fifth Supplement of the FSAR.

In determining the operability of the HPCI system, the required performance capability of various components shall be considered.

a. The HPCI System shall be capable of meeting at least 3,420 gpm flow at normal reactor operating pressure.

b. The motor driven feedwater pump shall be capable of automatic initiation upon receipt of either an automatic turbine trip signal or reactor low-water-level signal.

c. The Condenser hotwell level shall not be less than 57 inches (75,000 gallons).

d. The Condensate storage tanks inventory shall not be less than 105,000 gallons.

e. The motor-driven feedwater pump will automatically trip if reactor high water level is sustained for ten seconds and the associated pump downstream flow control valve is not closed.

During reactor startup and shutdown, only the condensate and feedwater booster pumps are in operation at reactor pressures below approximately 400 psig. The feedwater pump is in standby. However, if the HPCI initiation signal occurs, the feedwater pump would automatically start. Calculations show that the condensate and feedwater booster pump alone are capable of providing 3,420 gpm at a reactor pressure of approximately 270 psig.

The capability of the condensate, feedwater booster and motor driven feedwater pumps will be demonstrated by their operation as part of the feedwater supply during normal station operation. Stand-by pumps will be placed in service at least quarterly to supply feedwater during station operation. An automatic system initiation test will be performed at least once per operating cycle. This will involve automatic starting of the motor driven feedwater pumps and flow to the reactor vessel.

AMENDMENT NO. 78

BASES FOR 3.1.8 AND 4.1.8 HIGH PRESSURE COOLANT INJECTION During reactor startup with periods of low reactor water feed demand, one feedwater train is operated with a blocking valve closed downstream of the main flow control valve when core power is less than or equal to 25% of rated thermal power. This allows the low flow control valve to control the reactor water flow during the startup period when feedwater flow demand is low. Use of the low flow control valve provides more uniform feedwater flow which reduces thermal cycling at the reactor pressure vessel feedwater nozzles and in the feedwater piping as well as eliminating a severe service condition in the main flow control valves during reactor startup, Under low feedwater flow conditions, the main flow control valves also experience high pressure drops and fluid velocities which shorten the valve's life and can cause plant transients due to control valve failure. Reactor startup with one HPCI train available is acceptable since LOCA makeup requirements are reduced during startup because of lower reactor pressure, less decay heat, and lower reactor power than assumed in LOCA analyses performed to Appendix K 10 CFR 50 requirements. The other feedwater train (other HPCI loop) with its blocking valve open would remain capable of supplying 3,800 gpm of feedwater upon automatic HPCI initiation at all reactor pressure.

AMENDMENT NO. 79

3.2.0 REACTOR COOLANT SYSTEM A) GENERAL APPLICABILITY Applies to the operating conditions of the reactor coolant system and its associated systems and components.

B) GENERAL OB J ECTIVE LIMITINGCONDITIONS FOR OPERATION - To define the lowest functional capability or performance level of the systems which will assure the integrity of the reactor coolant system as a barrier against the uncontrolled release of radioactivity.

SURVEILLANCE REOUIREMENTS - To define the tests or inspections required to assure the functional capability or performance level of the above.

AMENDMENT NO. 80

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT 3.2.1 REA TOR VE SEL HEATUP AND CO LDOWN

~RATE Applies to the reactor vessel heating or cooling rate.

~Ob ec ive:

To assure that thermal stress resulting from reactor heatup and cooldown are within allowable code limits.

During the startup and shutdown operations of the reactor, the reactor vessel temperature shall not be increased more than 100'F in any one hour period nor decreased more than 100'F in any one hour penod.

AMENDMENT NO.

BASES FOR 3.2.1 REACTOR VESSEL HEATUP AND COOLDOWN Design calculations reported in Volume I, Section V-A, 4.0 (page V-6)" have demonstrated that the heatup and cooldown rate of 100 F/hr considered in the fatigue analysis will result in stresses well within code limits. A series of calculations have demonstrated that various extreme heatup and cooldown transients result in thermal strains well within the ASME Code limits stated in Volume I, Section V-C, 3.0 (p.

V-19)'. Cooldown incidents include: failure of the pressure regulator leading to a cooldown of 215'F in 5.5 minutes (Appendix E-l, 3.15 (p. E-45))", inadvertent opening of a single solenoid-actuated pressure relief valve leading to a cooldown of 1050'F/hr sustained for 10 minutes (Vol. I, Section V-B, 1.3 (p. V-11))", and finally, opening all six of the solenoid-actuated relief valves leads to a cooldown of 250'F in 7.5 minutes (Volume IV, Section I-B)". Reactor vessel heatup of 300'F/hr (Volume IV, Section I-B)" also demonstrates stresses well within the code requirements. In view of the reported results, the specified heatup and cooldown rates are believed to be conservative.

"FSAR AMENDMENT NO. 82

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.2 MINIM M REACT R VE EL TEMPERATURE FOR 4.2.2 MINIMUMREA T R VE EL TEMPER T RE F R

~*'i i

Applies to the minimum vessel temperature required Applies to the required vessel temperature for for vessel pressurization. pressurization.

~Ob ec ive: ~Ob ective:

To assure that no substantial pressure is imposed on To assure that the vessel is not subjected to any the reactor vessel unless its temperature is substantial pressure unless its temperature is greater considerably above its Nil Ductility Transition than its Nil Ductility Transition Temperature (NDTT).

Temperature (NDTT).

a. During reactor vessel heatup and cooldown when a. Reactor vessel temperature and pressure shall be the reactor is not critical, the reactor vessel monitored and controlled to assure that the temperature and pressure shall satisfy the pressure and temperature limits are met.

requirements of Figures 3.2.2.a and 3.2.2.b.

b. During reactor vessel heatup and cooldown when b. Vessel material and surveillance samples located the reactor is critical, the reactor vessel within the core region to permit periodic temperature and pressure shall satisfy the- monitoring of exposure and material properties requirements of Figures 3.2.2.c and 3.2.2.d shall be inspected on the following schedule:

except when performing low power physics testing with the vessel head removed at power First capsule (A) - 5.8 EFPY levels not to exceed 5 mw(t). Second capsule (C) - 7.98 EFPY Third capsule (B) - 16 EFPY AMENDMENT NO. 83

11 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. During hydrostatic testing, the reactor vessel In order to generate additional plant-specific temperature and pressure shall satisfy the data, a capsule containing irradiated and requirements of Figure 3.2.2.e if the core is not unirradiated material will be re-inserted at the B critical. capsule location. Re-insertion capsules have already been installed at the A and C locations.

d. The reactor vessel head bolting studs shall not A prime (') is used to indicate a re-insertion be under tension unless the temperature of the capsule. The withdrawal schedule for the re-vessel head flange and the head are equal to or insertion capsules is as follows:

greater than 100'F.

Fourth capsule (A') - 24 EFPY Fifth capsule (C') - 32 EFPY Sixth capsule (B') - 40 EFPY AMENDMENT NO. 84

0-NINE MILE POINT UNIT HEATIJP CORE NOT CRITICAL 1400 1200 2: 1000 800 NON-CRITICAL HEATUP g e00 cc 400 303 V MINlh/LM TEMP ERATLRE tu 200 FOR BOLTLP tOO'F 100 0 50 100 160 200 260 300 360 REACTOR VESSEL BELTLINE OOWNCOMER WATER TEMPERATURE (F)

(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

FIGURE 3.2.2.a MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING HEATUP AND LOW-POWER PHYSICS TESTS (REACTOR NOT CRITICAL) (HEATING RATE 5 100 F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 85

LIMIT FOR NON-CRITICAL OPERATION HEATUP AT UP TO 100 F/HR REA T R VES EL BELTLINE REA T R PRESSURE i DOWNCOMER WATER INT PD ME TEMPERAT RE F 303 100 303 110 303 120 303 130 308 140 319 150 334 160 354 170 379 180 409 190 445 200 488 210 538 220 596 230 664 240 743 250 834 260 940 270 1061 280 1202 290 1363 300 TABLE 3.2.2.a MINIIVIUMTEMPERATURE FOR PRESSURIZATION DURING HEAT-UP (REACTOR NOT CRITICAL) (HEATING RATE w100 F/HR)

FOR UP TO EIGHTEEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

AMENDMENT NO. 86

NINE MILE POINT UNIT 1 COOLDOWN CORE NOT CRlTICAL 1400 0

1200 1000 800 NON-CRITICAL COOLDOWN 600 409 0 400 I

O 360 MINIh/LM 250 TEMPER ATLRE 200 FOR BOLTLP 100 F 100 146 160 0 50 100 150 200 250 300 350 REACTOR VESSEL BELTLINE OOWNCOMER WATER TEMPERATURE {F)

(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

FIGURE 3.2.2.b MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN AND LOW-POWER PHYSICS TESTS (REACTOR NOT CRITICAL) (COOLING RATE 5 100 F/HR)

FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 87

LIMIT FOR NON-CRITICAL OPERATION COOLDOWN AT UP TO 100oF/HR 1

REA T R VE EL BELTLINE REA T R PRESSURE i DOWN MER WATER IN T P DOME TEMPERAT RE F 250 100 268 110 288 120 312 130 340 140 360 150 409 160 451 170 501 180 558 190 624 200 701 210 784 220 868 230 953 240 1049 250 1161 260 1289 270 1437 280 TABLE 3.2.2.b MINIMUMTEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (REACTOR NOT CRITICAL) (COOLING RATE N1000F/HR)

FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF CORE OPERATION (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

AMENDMENT NO. 88

NINE MILE POINT UNIT 1 CORE OPERATfON (HEATUP) 1400 WaTER I.EVEL M S-IN RANCH FOR POWER OPERaTiON iF CORE is 1200 CRIT Cai BE'W o

O 244 F o 1000 Z

800 CORE HEATUP OPERATION 600 462 400 0 Soz MINIMUM TEMPERATURE o~ 200 FOR BOLTS 100 F 100 244 0 50 100 150 200 250 300 350 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (F)

(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

FIGURE 3.2.2.c MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING CORE OPERATION (CORE CRITICAL) (HEATUP AT A HEATING RATE 5 100 F/HR)

FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 89.

LIMIT FOR POWER OPERATION (CORE CRITICAL)

HEATUP AT UP TO 100 F/HR REACT R VESSEL BELTLINE REACTOR PRESSURE i D WN MER WATER IN T P DOME TEMPERATURE F 303 100 303 110 303 120 303 130 303 140 303 150 303 160 303 170 303 180 303 190 303 200 303 210 303 220 303 230 303 240 488 250 538 260 596 270 664 280 743 290 834 300 940 310 1061 320 1202 330 1363 340 TABLE,3.2.2.c MINIMUMTEMPERATURE FOR PRESSURIZATION DURING HEATUP (REACTOR CRITICAL) (HEATING RATE c100 F/HR)

FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF CORE OPERATION (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

AMENDMENT NO. 90

NINE MILE POINT UNIT 1 CORE OPERATION (COOLDOWN) 1400 WATER LEVEl ~ST BE IN RANGE FOR POWER OPERATICN O

5200 IF CORE IS CRITLCAL BELOW 225 F 0 1000 800 CORE COOLING OPERATION 600 g 529 400 360 O 202 MlNlMUM u TEMPERATURE 200 FOR BP.T~

F '00 100 1S6 225 0 60 100 160 200 250 300 360 REACTOR VESSEL BELTLINE OOWNCOMER WATER TEMPERATURE (F)

(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

FIGURE 3.2.2.d MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING CORE OPERATION (CORE CRITICAL) (COOLDOWN AT A COOLDOWN RATE s 100'F/HR) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 91

LIMITFOR POWER OPERATION (CORE CRITICAL)

COOLDOWN AT UP TO 100 F/HR REA T R VE EL BELTLINE REA T R PRESSURE si D WN MER WATER INT PD ME TEMPERAT RE F 202 100 210 110 222 120 235 130 250 140 268 150 288 160 312 170 340 180 360 190 360 200 360 210 360 220 558 230 624 240 701 250 784 260 868 270 953 280 1049 290 1161 300 1289 310 1437 320 TABLE 3.2.2.d MINIMUMTEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (REACTOR CRITICAL) (COOLING RATE x100 FIHR)

FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF CORE OPERATION (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

AMENDMENT NO.

NINE MILE POINT UNIT, 1 NOH-CRITICAL HYDROTEST 1400 1200 CI 0 1000 800 NCN-CRITICAL'PERATION 614 600 400 360 0 MINIMJvl TElvPERATLRE O 200 FOR BOLTS 100 F 100 130 0 50 100 150 200 250 300 350 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (F)

(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

FIGURE 3.2.2.e MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR:

PRESSURIZATION DURING IN-SERVICE HYDROSTATIC TESTING AND LEAK TESTING (REACTOR NOT.CRITICAL) FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF OPERATION AMENDMENT NO. 93

LIMIT FOR IN-SERVICE TEST (CORE NOT CRITICAL, FUEL IN VESSEL)

REACTOR VES EL BELTLINE REACT R PRE SURE si D WN OMER WATER IN T P DOME TEMPERAT RE F 360 100 360 110 360 120 614 130 641 140 671 150 706 160 747 170 794 180 848 190 911 200 983 210 1066 220 1162 230 1273 240 1401 250 TABLE 3.2.2.e IVIINIMUMTEIVIPERATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTING (REACTOR NOT CRITICAL)

FOR UP TO 18 EFFECTIVE FULL POWER YEARS OF CORE OPERATION (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)

AMENDMENT NO. 94

BASES FOR 3.2.2 AND 4.2.2 MINIMUMREACTOR VESSEL TEMPERATURE FOR PRESSURIZATION Figures 3.2.2.a, 3.2.2.b, 3.2.2.c, and 3.2.2.d are plots of pressure versus temperature for heatup and cooldown rates of up to 100 F/hr.

maximum (Specification 3.2.1). Figure 3.2.2.e is a plot of pressure versus temperature for hydrostatic testing. These curves are based on calculations of stress intensity factors according to Appendix G of Section III of the ASME Boiler and Pressure Vessel Code 1980 Edition with Winter 1982 Addenda. In addition, temperature shifts due to fast neutron fluence at eighteen effective full power years of operation were incorporated into the figures. These shifts were calculated using the procedure presented in Regulatory Guide 1.99, Revision 2.

Since the surveillance materials have a slightly lower copper content than the limiting plate material, a chemistry correction factor was applied to the measured data prior to determining the chemistry factor. These curves are applicable to the beltline region at low and elevated temperatures and the vessel flange at intermediate temperatures. Reactor vessel flange/reactor head flange boltup is governed by other criteria as stated in Specification 3.2.2.d. The pressure readings on the figures have been adjusted to reflect the calculated elevation head difference between the pressure sensing instrument locations and the pressure sensitive area of the core beltline region.

The reactor vessel head flange and vessel flange in combination with the double "0" ring type seal are designed to provide a leak-tight seal when bolted together. When the vessel head is placed on the reactor vessel, only that portion of the head flange near the inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the "0" rings of the head and vessel flanges. Both the head and vessel flanges have an NDT temperature of 40'F and they are not subject to any appreciable neutron radiation exposure. Therefore, the minimum vessel. flange and head flange temperature for bolting is established at 40'F + 60'F or 100'F.

Figures 3.2.2.a, 3.2.2.b, 3.2.2.c, 3.2.2.d, and 3.2.2.e have incorporated a temperature shift due to the calculated fast neutron fluence.

The neutron flux at the vessel wall is calculated from core physics data and has been determined using flux monitors installed inside the vessel. The curves are applicable for up to eighteen effective full power years of operation.

Vessel material surveillance samples are located within the core region to permit periodic monitoring of exposure and changes in material properties. The material sample program conforms with ASTM E185-66 except for the material withdrawal schedule which is specified in Specification 4.2.2.b.

AMENDMENT NO. 95

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.3 OOLANT CHEMISTRY 4.2.3 COOLANT CHEMISTRY Applies to the reactor coolant system chemical Applies to the periodic testing requirements of the requirements. reactor coolant chemistry.

~ob ective: ~ob ective:

To assure the chemical purity of the reactor coolant To determine the chemical purity of the reactor water. coolant water.

a. The reactor coolant water shall not exceed the Samples shall be taken and analyzed for conductivity following limits with steaming rates less than and chloride ion content at least 3 times per week 100,000 pounds per hour except as specified in with a maximum time of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> between samples.

3.2.3c: In addition, if the conductivity becomes abnormal (other than short term spikes) as indicated by the Conductivity 2 pmho/cm continuous conductivity monitor, samples shall be Chloride ion 0.1 ppm taken and analyzed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and daily thereafter until conductivity returns to normal levels.

b. The reactor coolant water shall not exceed the following limits with steaming rates greater than When the continuous conductivity monitor is or equal to 100,000 pounds per hour except as... inoperable, a reactor coolant sample shall be taken specified in 3.2.3c: and analyzed for conductivity and chloride ion content at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Conductivity 5 pmho/cm Chloride ion 0.2 ppm AMENDMENT NO. 96

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. The limits specified in 3.2.3a and 3.2.3b may be exceeded for a period of time not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In no case shall (1) the conductivity exceed a maximum limit of 10 pmho/cm, or (2) the chloride ion concentration exceed a maximum limit of 0.5 ppm.

d. If Specifications 3.2.3a, b, and c are not met, normal orderly shutdown shall be initiated within one hour and the reactor shall be in the cold shutdown condition within ten hours.

e. If the continuous conductivity monitor is inoperable for more than 7 days the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

AMENDMENT NO.

BASES FOR 3.2.3 AND 4.2.3 COOLANT CHEMISTRY Materials in the primary system are primarily 304 stainless steel and the Zircaloy fuel cladding. The reactor water chemistry limits are established to prevent damage to these materials. Limits are placed on chloride concentration and conductivity. The most important limit is that placed on chloride concentration to prevent stress corrosion cracking of the stainless steel. When the steaming rate is less than 100,000 pounds per hour, a more restrictive limit of 0.1 ppm has been established. At steaming rates of at least 100,000 pounds pef hour, boiling occurs causing deaeration of the reactor water, thus maintaining oxygen concentration at low levels.

A short term spike is defined as a rise in conductivity such as that which could arise from injection of additional feedwater flow for a duration of approximately 30 minutes in time.

When conductivity is in its proper normal range, pH and chloride and other impurities affecting conductivity'must also be within their normal range. When and if conductivity becomes abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operating values. This would not necessarily be the case. Conductivity could be high due to the presence of a neutral salt, e.g., Na2SO4, which would not have an affect on pH or chloride. In such a case, high conductivity alone is not a cause for shutdown. In some types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives. In the case of BWR's, however, where no additives are used and where neutral pH is maintained, conductivity provides a very good measure of the quality of the reactor water. Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting boundaries of the reactor coolant, are exceeded. Methods available to the operator for correcting the off-standard condition include, operation of the reactor clean-up system, reducing the input of impurities and placing the reactor in the cold shutdown condition. The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the clean-up system to re-establish the purity of the reactor coolant. During start-up periods, which are in the category of less than 100,000 pounds per hour, conductivity may exceed 2 ymho/cm because of the initial evolution of gases and the initial addition of dissolved metals. During this period of time, when the conductivity exceeds 2 ymho (other than short term spikes), samples will be taken to assure that the chloride concentration is less than 0.1 ppm.

The conductivity at the reactor coolant is continuously monitored. The samples of the coolant which are taken every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the monitors. If conductivity is

'ithin its normal range, chlorides and other impurities will also be within their normal ranges. The reactor coolant samples will also be used to determine the chlorides. Therefore, the sampling frequency. is considered adequate to detect long-term changes in the chloride ion content. However, if the conductivity changes significantly, chloride measurements will be made to assure that the chloride limits of Specification 3.2.3 are not exceeded.

AMENDMENT NO. 98

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.4 REA T R CO LANT A TIVITY 4.2A REA TOR OLANT ACTIVITY Applies to the limits on reactor coolant activity at all Applies to the periodic testing requirements of the operating conditions. reactor coolant activity.

~ob'ec ive: ~Ob ective:

To assure that in the event of a reactor coolant To assure that limits on coolant activity are not system line break outside the drywell permissible exceeded.

doses are not'xceeded.

a. The reactor coolant system radioactivity a. Samples shall be taken at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> concentration in water shall not exceed 25 and analyzed for gross gamma activity.

microcuries of total iodine per gram of water.

b. If Specification 3.2A a, above, cannot be met b. Isotopic analyses of samples shall be made at after a routine surveillance check, the reactor least once per month.

shall be placed in the cold shutdown condition within ten hours.

AMENDMENT NO. 99

BASES FOR 3.2.4 AND 4.2.4 REACTOR COOLANT ACTIVITY The primary coolant radioactivity concentration limit of 25 pCi total iodine per gram of water was calculated based on a steamline break accident which is isolated in 10.5 seconds. For this accident analysis, all the iodine in the mass of coolant released in this time period is assumed to be released to the atmosphere at the top of the turbine building (30 meters). By limiting the thyroid dose at the site boundary to a-maximum of 30 Rem, the iodine concentration in the primary coolant is back-calculated assuming fumigation meteorology, Pasquill Type F at 1m/sec. The iodine concentration in the primary coolant resulting from this analysis is 25 pCi/gm.

A radioactivity concentration limit of 25 yCi/g total iodine could only be reached if the gaseous effluents were near the limit based on the assumed effluent isotopic content (Table A-12 of the FSAR) and the fact that the primary coolant cleanup systems were inoperative. When the cleanup system is operating, it is expected that the primary coolant radioactivity would be about 12 yCi/g total iodine. The concentrations expected during operations with a gaseous effluent of about 0.1 yCi/sec would be about 1.5 yCi/g total iodine.

The reactor water sample will be used to assure that the limit of Specification 3.2A is not exceeded. The total radioactive iodine activity would not be expected to change rapidly over a period of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. In addition, the trend of the stack offgas release rate, which is continuously monitored, is a good indicator of the trend of the iodine activity in the reactor coolant.

Since the concentration of radioactivity in the reactor coolant is not continuously measured, coolant sampling would be ineffective as a means to rapidly detect gross fuel element failures. However, as discussed in the bases for Specification 3.6.2, some capability to detect gross fuel element failures is inherent in the radiation monitors in the offgas system and on the main steam lines.

AMENDMENT NO. 100

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.5 REA T R COOLANT Y TEM LEAKAGE 4.2.5 REA T R OOLANT TEM LEAKA E bi i Applies to the limits on reactor coolant system Applies to the monitoring of reactor coolant leakage rate and leakage detection systems.

system'eakage.

~ob ective: ~ob'ec ive:

To assure that the makeup capability provided by the To determine the reactor coolant system leakage rate control rod drive pump is not exceeded. and assure that the leakage limits are not exceeded.

a. Any time irradiated fuel is in the reactor vessel a. A check of the reactor coolant leakage shall be and the reactor temperature is above 212'F, made every four hours.

reactor coolant leakage into the primary containment shall be limited to:

1. Five gallons per minute unidentified leakage.

2. A two gallon per minute increase in unidenti-fied leakage within any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less.

~ J v

3. Twenty-five gallons per minute total leakage (identified plus unidentified) averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

AMENDMENT NO.

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. Any time irradiated fuel is in the reactor vessel b. The following surveillance shall be performed on and reactor coolant temperature is above 212'F, each leakage detection system:

at least one of the leakage measurement channels associated with each sump (one for the (1) An instrument calibration once each drywell floor drain and one for the equipment refueling outage.

drain) shall be operable.

(2) An instrument functional test once every If the conditions a or b cannot be met, the reactor three months.

will be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

AMENDMENT NO. 102

BASES FOR 3.2.5 AND 4.2.5 REACTOR COOLANT SYSTEM LEAKAGE RATE Allowable leakage-rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite a-c power. The normally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were also considered in establishing the limits. The behavior of cracks in piping systems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary Coolant System Rupture Study (the Pipe Rupture Study). Work utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth. This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly. However, the establishment of allowable unidentified leakage greater than that given in 3.2.5 on the basis of the data presently available would be premature because of uncertainties associated with the data. For leakage of the order of 5 gpm as specified in 3.2.5, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation. Leakage of the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the plant should be shut down to allow further investigation and corrective action.

Inspection and corrective action is initiated when unidentified leakage increases at a rate in excess of 2 gpm, within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period or less. This minimizes the possibility of excessive propagation of intergranular stress corrosion cracking.

A total leakage of 25 gpm is well within the capacity of the control rod drive system makeup capability (page lll-7 of the First Supplement)". As discussed in 3.1.6 above, for leakages within this makeup capability, the core will remain covered and automatic pressure blowdown will not be actuated.

The primary means of determining the reactor coolant leakage rate is by monitoring the rate of rise in the levels of the drywell floor and equipment drain lines. Checks will be made every four hours to verify that no alarms have been actuated due to high leakage. For sump inflows of one gpm, changes on the order of 0.2 gpm can be detected within 40 minutes. At inflows between one and five gpm, changes on the order of 0.5 gpm can be detected in eight minutes.

"FSAR AMENDMENT NO. 103

BASES FOR 3.2.5 AND 4.2.5 REACTOR COOLANT SYSTEM LEAKAGE RATE Leakage is detected by having all unidentified leakage routed to the drywell floor drain tank and identified leakage routed directly to the drywell equipment drain tanks. Identified leakage includes such items as recirculation pump seal leakage and recirculation pump suction and discharge valve packing leakoff.

Another method will monitor the time required to fill the tanks between two accurately determined levels. When the level in the tank reaches the low-level switch setting, a timer will start and operate for a preset time interval. If the timer resets before the high-level switch setting is reached indicating a leakage rate within allowable limits, no action will result, and the system resets for the next filling and timing cycle. If the leakage is high enough to cause the level to reach the high level switch setting before the timer resets automatically, an alarm is actuated indicating leak rate above the predetermined limit (First and Fifth Supplements)".

Additional information is available to the operator which can be used for the shift leakage check if the drywell sumps level alarms are out of service. The integrated flow pumped from the sumps to the waste disposal system can be checked.

Qualitative information is also available to the operator in the form of indication of drywell atmospheric conditions. Continuous leakage from the primary coolant system would cause an increase in drywell temperature. Any leakage in excess of 15 gpm of steam would cause a continuing increase in drywall pressure with resulting scram (First Supplement) ".

Either the rate of rise leak detection system, the timer leak detection system or the integrated flow can be utilized to satisfy Specification 3.2.5.b.

%FSAR AMENDMENT NO. 104

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.6 IN ERVICE INSPECTION AND TESTIN ~

4.2.6 INSERVICE IN PE TION AND TESTIN A licabili Applies to components which are part of the reactor Applies to periodic inspection and testing of coolant pressure boundary and their supports and components which are part of the reactor coolant other safety-related pressure vessels, piping, pumps, pressure boundary and their supports and other and valves. safety-related pressure vessels, piping, pumps, and valves.

~Ob ective: ~Ob ective:

To assure the integrity of the reactor coolant pressure To verify the integrity of the reactor coolant pressure boundary and the operational readiness of safety- boundary and the operational readiness of safety-related pressure vessels, piping, pumps, and valves. related pressure vessels, piping, pumps, and valves.

a. Inservice Ins ec ion a. Inservi In i n

1. To be considered operable, Quality Group A, 1. Inservice inspection of Quality Group A, B B and C components shall satisfy the and C components shall be performed in requirements contained in Section XI of the accordance with the requirements for ASME ASME Boiler and Pressure Vessel Code and Code Class 1, 2 and 3 components, respec-applicable Addenda for continued service of. tively, contained in Section XI of the ASME ASME Code Class 1, 2 and 3 components, Boiler and Pressure Vessel Code and respectively, except where relief has been applicable Addenda as required by 10CFR50, granted by the Commission pursuant to Section 50.55a(g), except where relief has 10CFR50, Section 50.55a(g)(6)(i). been granted by the Commission pursuant to 10CFR Part 50, Section 50.55a (g)(6)(i).

AMENDMENT NO. 105

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

2. The Inservice Inspection Program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule, methods, personnel and sample expansion included in this generic letter.

1. To be considered operable, Quality Group A, 1. Inservice testing of Quality Group A, B and B and C pumps and valves, shall satisfy the C pumps and valves shall be performed in requirements contained in Section XI of the accordance with the requirements for ASME ASME Boiler and Pressure Vessel Code and Code Class 1, 2 and 3 components applicable Addenda for continued service of contained in Section XI of the ASME Boiler ASME Code Class 1, 2 and 3 components, and Pressure Vessel Code and applicable respectively, except where relief has been Addenda as required by 10CFR50, Section granted by the Commission pursuant to 50.55a(f), except where relief has been 10CFR50, Section 50.55a(f)(6)(i). granted by the Commission pursuant to 10CFRPart 50, Section 50.55a(f)(6)(i).

c. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the require-.

ments of any Technical Specification.

AMENDMENT NO. 106

BASES FOR 3.2.6 AND 4.2.6 INSERVICE INSPECTION AND TESTING The inservice inspection and testing programs for the Nine Mile Point Unit 1 plant conform to the requirements of 10CFR50, Section 50.55a(f) and (g). Where practical, the inspection of components, pumps and valves classified into NRC Quality Groups A, B and C conforms to the requirements of ASME Code Class 1, 2 and 3 components, pumps, and valves, respectively, contained in Section XI of the ASME Boiler and Pressure Vessel Code. If a Code required inspection is impractical for the Nine Mile Point Unit 1 facility, a request for relief from that requirement is submitted to the Commission in accordance with 10CFR50, Section 50.55a(f)(6)(i) and Section 50.55a(g)(6)(i) ~

Request for relief from the requirements of Section XI of the ASME Code and applicable Addenda will be submitted to the Commission prior to the beginning of each 10-year inspection interval if they are known to be required at the time. Requests for relief which are identified during the course of inspection will be submitted quarterly throughout the inspection interval.

The inservice inspection program for piping conforms to the staff positions on schedules, methods, personnel and sample expansion contained in Generic Letter 88-01. It is performed in order to detect and survey intergranular stress corrosion cracking of BWR austenitic stainless steel piping that is four inches or larger in nominal diameter and contains reactor coolant at a temperature above 200'F during power operation. Inspections shall be performed by individuals qualified 'to: (A) the ASME Boiler and Pressure Vessel Code,Section XI, and (B) Ultrasonic Testing Operator Training for the Detection of Intergranular Stress Corrosion Cracking developed by the EPRI Non-Destructive Examination Center. As an alternate, Niagara Mohawk may use other qualification programs approved by the NRC.

~Rf<~rn~e (1) Letter from the Nuclear Regulatory Commission (D. B. Vassallo) to Niagara Mohawk Power Corporation (G. K. Rhode), dated September 19, 1983.

(2) Letter from Niagara Mohawk Power Corporation (D. P. Disc) to the Nuclear Regulatory Commission (T. A. Ippolito), dated August 7, 1981.

(3) Generic Letter 88-01 endorses NUREG 0313 Revision 2, ".Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," dated January 1988.

AMENDMENT NO. 107

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.7 REACTOR COOLANT YSTEM ISOLATION VALVE 4.2.7 REACT R C LANT Y TEM I LATION VALVE Applies to the operating status of the system of Applies to the periodic testing requirement for the isolation valves on lines connected to the reactor reactor coolant system isolation valves.

coolant system.

~ob'ec ive: OIIOec ive:

To assure the capability of the reactor coolant system To assure the capability of the reactor coolant system isolation valves to minimize reactor coolant loss in the isolation valves to minimize reactor coolant loss in the event of a rupture of a line connected to the nuclear event of a rupture of a line connected to the nuclear steam supply system. steam supply system.

a. During power operating conditions whenever the The reactor coolant system isolation valves reactor head is on, all reactor coolant system surveillance shall be performed as indicated below isolation valves on lines connected to the reactor (see Table 3.2.7).

coolant system shall be operable except as specified in "b" below. a. A Ieas on e er r in cl the operable automatically initiated power-operated isolation

b. In the event any isolation valve becomes valves shall be tested for automatic initiation and inoperable the system shall be considered closure times.

operable provided at least one valve in each line haying an inoperable valve is in the mode b. A lea on er ar er all normally open corresponding to the isolated condition. power-operated isolation valves (except the feedwater and main-steam-line power-operated isolation valves) shall be fully closed and reopened.

AMENDMENT NO. 108

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. If Specifications 3.2.7a and b above are not met, c. A lea nce er uar er the feedwater and initiate normal orderly shutdown within one hour main-steam line power-operated isolation valves and have reactor in the cold shutdown condition shall be exercised by partial closure and within ten hours. subsequent reopening.

d. A lea n er lan c ld hu wn the feedwater and main steam line power-operated isolation valves shall be. fully closed and reopened, unless this test has been performed within the previous 92 days.

AMENDMENT NO. 109

LIMITINGCONDITIONS FOR OPERATION Table 3.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Looation Relative Maximum Initiating Signal (All No. of Valves to Primary Normal Oper. Time Action on Valves have Remote Une or System IEach Une) Containment Position Motive Powers ISeo) InitiatIng Signal Manual Backup)

Reactor water level low-low or low reactor pressure, (with mode switch in run) or main Main Steamll>> Inside Open AC Motor 10 Close steam line high ITwo Lines) Outside Open Pn/DC Solenoid 10 Close radiation, or main steam line high flow, or low-low-low condenser vacuum, or high temperature in the steam tunnel Feedwaterlt >> Outside Open AC Motor 80 Remote Manual ITwo Lines) Outside Open Self Act. Ck.

Emer enc Coolin Steam Leevin Reectorl Outside Open AC Motor 38 Close High emergency DC Motor 38 Close cooling system flow (Two Lines) Outside Open {

Condensate Return to Resctorl >> Inside Closed Self Act. Ck.

ITwo Lines) Outside Closed Pn/DC Solenoid BO Close High emergency cooling system flow eo Open Reactor water level, low-low or high reactor pressure AMENDMENT NO. 110

LIMITING CONDITIONS FOR OPERATION Table 3.2.7 (cont'd)

REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Nlax)mum Initiating Signal (All No. of Valves to Primary Normal Oper. Time Act)on on Valves Have Remote (Each Une) Containment Position Motive Powers (Sec) Inltlatlng Signal Manual Backup)

Une or System Reactor Cleanu Reactor Inside Open AC Motor 18 Close Reactor water level Water Leevin Outside Open DC Motor 18 Close low-low or high area (One Line) temperature or liquid Water Return to Reactor( ) Inside Open AC Motor 18 Close poison initiation (One Line) Outside Open Self Act. Ck.

Shutdown Coolin Water Leavin Reector>> ) Inside Closed AC Motor 40 Close (One Line) Outside Closed DC Motor 40 Close Reactor water level low-low, or high area temperature Water Return to Reactor Inside Closed AC Motor 40 Close (One Line) Outside Closed Self Act. Ck.

AMENDMENT NO.

LIMITING CONDITIONS FOR OPERATION Table 3.2.7 (cont'd)

REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Maximum Inltlatlng Signal (All No. of Valves To Primary Normal Oper. Time Action on Valves Have Remote Une or System (Each Une) Containment Position Motive Power~ (Seo) InMat(ng Signal Manual Backup)

~Li uid Potsont i. Inside Closed Self Act. Ck.

(One Line) Outside Closed Self Act. Ck.

Control Rod Drive H draulic( I Inside Open Self Act. Ck.

(One Line) Outside Open Self Act. Ck.

Scram Dischsr e Volume(1) Outside Open Pn/AC Solenoid 10 Close

~Sstem Vent" Automatic or manual (One Line) reactor scram Scram Discher e Volume( ) Outside Open Pn/AC Solenoid 10 Close

~Satem Drain" (One Line)

~Core S re Reactor water level Core S re In'ection Inside Closed AC Motor 22.5 Open low-low or high (Two Lines) Outside Open AC Motor 22.5 Open drywell pressure coincident with reactor vessel pressure less than 365 psig Core S ra Hi h Point Vent ") Inside Closed AC Motor 27 Close Reactor water level (Two Lines) Outside Closed Pn/DC Solenoid 27 Close low-low or high drywell pressure Core S ra Condensate Su I Outside Open Self Act. Ck.

(Keep Fill)

(Two Lines)

Core S re S stem Valves Outside Closed Self Act. Ck.

(Two Lines)

Core S ra Pum Discher e Outside Closed AC Motor 27 Close Reactor water level (Two Test Lines to Suppression Chamber) Iow-low or high drywell pressure AMENDMENT NO. 112

LIMITINGCONDITIONS FOR OPERATION Table 3.2.7 (cont'd)

REACTOR COOLANT SYSTEM ISOLATION VALVES Location Relative Maximum Inftletfng Signal (All No. of Valves To Primary Normal Oper. Time Aotlonon Valves Have Remote Une or System (Each Une) Containment Position Motive Power~ {Sec) Initfetfng Sfgnal Manual Backup)

Post Accident Reactor Sam lin { )( Outside Open Self Act. Flow

{One Line) Fuse Outside Closed Pn/DC Solenoid 30 Close Reactor water level low-low or main steam tine high radiation or low-low-fow condenser vacuum or reactor low pressure, (with mode Reactor Recircufation S stem Sam fin {I) inside Closed AC Motor 20 Close switch in run) or high lOne Line) Outside Closed DC Motor 20 Close temperature in the steam tunnel or main steam line high flow AMENDMENT NO. 113

Notes:

" Pn - Pneumatically Operated

"" Section 3.1.1e for LCO Requirements (1) These valves do not have to be vented during the Type A test. However, Type C leakage from these valves is added to the Type A test results, if not vented.

(2) These valves have flow through them during and following an accident (a water seal) and receive a water leak rate test in accordance with the IST Program.

(3) The inside core spray injection isolation valves are water sealed during and after an accident. These valves are leak rate tested with water in accordance with the IST Program. The outside core spray injection isolation valves are open with their breakers locked in the off position. Therefore, the outside core spray injection isolation valves do not have to be tested under the IST or Appendix J Leakage Program.

(4) These valves are provided with a water seal. Valves shall be tested during each refuel outage not to exceed two years consistent with Appendix J water seal testing requirements. Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm.

(5) These valves are tested in accordance with Section 4.2.7.1a.

(6) The self actuating flow fuse is tested in accordance with Section 4.3.4c.

AMENDMENT NO. 114

BASES FOR 3.2.7 AND 4.2.7 REACTOR COOLANT SYSTEM ISOLATION VALVES Double isolation valves are provided in lines which connect to the reactor coolant system to assure isolation and minimize reactor coolant loss in the event of a line rupture. The specified valve requirements assure that isolation is already accomplished with one valve shut or provide redundancy in an open line with two operative valves. Except where check valves are used as one or both of a set of double isolation valves, the isolation valves shall be capable of automatic initiation and the closure times presented in Table 3.2.7. These closure main-times were selected to minimize coolant losses in the event of the specific line rupturing. Using the longest closure time on the steam-line valves following a main-steam-line break (Section XV C.1.0), the core is still covered by the time the valves close. Following (1) a specific system line break, the cleanup and shutdown cooling closing times will upon initiation from a low-low level signal limit coolant loss such that the core is not uncovered. Feedwater flow would quickly restore coolant levels to prevent clad damage. Closure times are discussed in Section VI-D.1.0 The valve operability test intervals are based on periods not likely to significantly affect operations, and are consistent with testing of other systems. Results obtained during closure testing are not expected to differ appreciably from closure times under accident conditions as in most cases, flow helps to seal the valve.

The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 (Fifth Supplement,

p. 115) that a line will not isolate. More frequent testing for valve operability results in a more reliable system.

(1) UFSAR (2) FSAR AMENDMENT NO. 115

3.2.7.1 b'i'e:

LIMITING CONDITION FOR OPERATION PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Applies to the operating status of isolation valves for systems connected to the primary coolant system.

~Ob'e ive:

To increase the reliability of primary coolant system 4.2.7.1

~VALVE SURVEILLANCE REQUIREIVlENT PRIMARY OOLANT Y TEM PRE RE I OLATI N Applies to the periodic testing of primary coolant system pressure isolation valves.

~b'ec lve:

To increase the reliability of primary coolant system pressure isolation valves thereby reducing the pressure isolation valves thereby reducing the potential of an intersystem loss of coolant accident. potential of an intersystem loss of coolant accident.

a. The integrity of all pressure isolation valves listed a. Periodic leakage testingI on each valve listed in in Table 3.2.7.1 shall be demonstrated. Valve Table 3.2.7.1 shall be accomplished prior to leakage shall not exceed the amounts indicated. exceeding 2% power while in the power operating condition every time the plant is placed

b. If Specification a cannot be met, an orderly in a cold shutdown condition for refueling, each shutdown shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the time the plant is placed in a cold shutdown reactor shall be in the cold shutdown condition condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. accomplished in the preceding 9 months, and prior to returning the valve to service after maintenance, repair or replacement work is performed.

ii

• i .i k b indirectly (as from the performance of pressure indicators) d if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

AMENDMENT NO.

TABLE 3.2.7.1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES

~Sstem Valve No. INaximum(a) Allow le L aka

1. Core Spray System 40-03 a5.0 gpm 40-13 s5.0 gpm

2. Condensate Supply to Core Spray 40-20 a5.0 gpm (Keep Fill System) 40-21 x5.0 gpm 40-22 x5.0 gpm 40-23 s5.0 gpm

~F~nLe:

(a) 1. Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm.

2. Test differential pressure shall not be less than 150 psid.

3. The observed leakage at test differential pressure shall be adjusted to the functional maximum pressure differential.

AMENDMENT NO.

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.8 PRE URE RELIEF SYSTEMS-SAFETY VALVES 4.2.8 PRESSURE RELIEF Y TEM - AFETY VALVE v>>"

Applies to the operational status of the safety valves. Applies to the periodic testing requirements for the safety valves.

~Ob ective: ~Ob'ec ive:

To assure the capability of the safety valves to limit To assure the capability of the safety valves to limit reactor overpressure below the safety limit in the reactor overpressure to below the safety limit.

event of rapid reactor isolation and failure of all pressure relieving devices.

a. During power operating conditions and whenever At least once during each operating cycle at least the reactor coolant pressure is greater than 110 eight of the sixteen safety valves shall be removed, psig and temperature greater than saturation tested for set point and partial lift, and then returned temperature all sixteen of the safety valves shall to operation or replaced.

be operable.

b. If specification 3.2.8a is not met, the reactor coolant pressure and temperature shall be reduced to 110 psig or less and saturation temperature or less, respectively, within ten hours.

AMENDMENT NO.

BASES FOR 3.2.8 AND 4.2.8 PRESSURE RELIEF SYSTEM-SAFETY VALVES The required,number of operable safety valves is based on the analysis presented in Appendix E-I.3.7" which assumed reactor isolation with no scram. Operation of all 16 safety valves will limit reactor pressure below the safety limit of 1375 psig. Partial redundancy is provided by the solenoid-actuated pressure relief valves as the relieving capacity of each of these valves is approximately the same as a safety valve, as discussed in 2.2.2 above.

The safety valve testing and intervals between tests are based on manufacturer's recommendations and past experience with spring actuated safety valves.

%FSAR AMENDMENT NO. 119

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.9 PRE URE RELIEF SYSTEMS - OLENOID- 4.2.9 ~ PRE RE RELIEF YSTEMS - OLENOID-ACT ATED PRESS RE RELIEF VALVES ACT ATED PRE RE RELIEF VALVE OVERPRE URIZATION OVERPRE RIZATION eeeeee:

Applies to the operational status of the solenoid- Applies to the periodic testing requirements for the actuated pressure relief valves. solenoid-actuated pressure relief valves.

~Ob ec ice: ~ective:

To assure the capability of the solenoid-actuated To assure the operability of the solenoid-actuated pressure relief valves to limit reactor overpressure pressure relief valves to limit reactor overpressure in

'below the lowest safety valve setpoint in the event of the event of rapid reactor isolation.

rapid reactor isolation.

a. During the power operating condition and The solenoid-actuated pressure relief valve whenever the reactor coolant pressure is greater surveillance shall be performed as indicated below.

than 110 psig and temperature greater than saturation, five of the six solenoid-actuated a. The setpoints of the six relief valves shall be as pressure relief valves shall be operable. follows:

b. If Specification 3.2.9a is not met, the reactor No. of coolant pressure and temperature shall be Valves ~Se cin reduced to 110 psig or less and saturation temperature or less, respectively, within ten 2 (1090 psig hours. 2 (1095 psig 2 <1100 psig AMENDMENT NO. 120

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. At least once during. each operating cycle with the reactor at pressure, each valve shall be manually opened until acoustic monitors or thermocouples downstream of the valve indicate that the valve has opened and steam is flowing from the valve.

c. At least once during each operating cycle, relief valve setpoints shall be verified.

AMENDMENT NO.

BASES FOR 3.2.9 AND 4.2.9 PRESSURE RELIEF SYSTEM - SOLENOID ACTUATED PRESSURE RELIEF VALVES As discussed in 2.2.2 and.3.2.8 above, the solenoid-actuated pressure relief valves are used to avoid actuation of the safety valves. The set points of the six relief valves are staggered. Two valves are set at 1090 psig, two are set at 1095 psig, and two are set at 1100 psig.

The operator will endeavor to place the set-point at these figures. However, a set-point error for each valve can be as much as 112 psig.

Six valves are provided for the automatic depressurization function, as described in 3.1.5. However, only five valves are required to prevent actuation of the safety valves, as discussed in the Technical Supplement to Petition to Increase Power Level,Section II.XV, letter, T. J. Srosnan to Peter A. Morris dated February 28, 1972, and letter, Philip D. Raymond to A. Giambusso, dated October 15, 1973.

The basis for the surveillance requirement is given in 4.1.5.

AMENDMENT NO. 122

3.3.0 PRIMARY CONTAINMENT APPLICABILITY Applies to the operating status of the primary containment systems.

8 OBJECTIVE To assure the integrity of the primary containment systems.

Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 215'F and fuel is in the reactor vessel except while performing low power physics tests at atmospheric pressure during or after refueling at power-levels not to exceed 5 Mwt.

AMENDMENT NO. 123

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.1 OXYGEN CONCENTRATION 4.3.1 XY EN CONCENTRATI N Applies to the limit on oxygen concentration within Applies to the periodic testing requirement for the the primary containment system. primary containment system oxygen concentration.

Qbbective: ~jeci~ive:

To assure that in the event of a loss-of-coolant To assure that the oxygen concentration within the accident any hydrogen generation will not result in a primary containment system is within required limits.

combustible mixture within the primary containment system.

a. The primary containment atmosphere shall be A legs once a week oxygen concentration shall be reduced to less than four percent by volume determined.

oxygen concentration with nitrogen gas whenever the reactor coolant pressure is greater than 110 psig and the reactor is in the power operating condition, except as specified in "b" below.

AMENDMENT NO. 124

i LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVlENT

b. Within the 24-hour period subsequent to the reactor being placed in the run mode for the power operating condition, the containment atmosphere oxygen concentration shall be reduced to less than four percent by volume, and maintained in this condition. Deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a major refueling outage or other scheduled shutdown.

c. If Specifications "a" or "b" above are not met, the reactor coolant pressure shall be reduced to 110 psig or less within ten hours.

AMENDMENT NO. 125

BASES FOR 3.3.1 AND 4.3.1 OXYGEN CONCENTRATION The four percent by volume oxygen concentration eliminates the possibility of hydrogen combustion following a loss-of-coolant accident (Section VII-G.2.0 and Appendix E-11.5.2)". The only way that significant quantities of hydrogen could be generated would be if all core spray systems failed to sufficiently cool the core. As discussed in Section VII-A.2.0 and illustrated in Figure Vll-2", the core spray system is capable of design flow of 3400 gpm at a reactor pressure of 113 psig. In addition to hydrogen generated by metal-water reaction, significant quantities can be generated by radiolysis. (Technical Specification to Petition for Conversion from Provisional Operating License to Full Term Operating License).

At reactor pressures of 110 psig or less, the reactor will have been shutdown for more than an hour and the decay heat will be at sufficiently low values so that fuel rods will be completely wetted by core spray. The fuel clad temperatures would not exceed the core spray water saturation temperature of about 344'F.

The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-of-coolant accident upon which the specified oxygen concentration limit is based. Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety. Thus to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods when the primary system is at or near rated operating temperature and pressure. The 24-hour period to provide inerting is judged to be reasonable to perform the leak inspection and establish the required oxygen concentration.

The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the containment but air could not leak in to increase the oxygen concentration. Once the containment is filled with nitrogen to the required concentration, no monitoring of oxygen concentration is necessary. However, at least once a week, the oxygen concentration will be determined as added assurance that Specification 3.3.1 is being met.

%FSAR AMENDMENT NO. 126

I LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT.

SYSTEM PRE URE AND 4.3.2 PRES RE SUPPRE SI N Y TEM PRE RE AND 3.3.2 PRE RE SUPPRESSION PPRESSION CHAMBER WATER TEMPERATURE UPPRESSION CHAMBER WATER TEMPERAT RE AND LEVEL AND LEVEL

~Alii~i~li Applies to the interrelated parameters of pressure Applies to the periodic testing of the pressure suppression system pressure and suppression suppression system pressure and suppression chamber water temperature and level. chamber water temperature and level.

Qggg five: ~Ob ec ive:

To assure that the peak suppression chamber To assure that the pressure suppression system pressure does not exceed design values in the event pressure and suppression chamber water temperature of a loss-of-coolant accident. and level are within required limits.

a. The downcomers in the suppression chamber a. At least once per day the suppression chamber shall have a minimum submergence of three and water level and temperature and pressure one half feet and a maximum submergence of suppression system pressure shall be checked.

four and one quarter feet whenever the reactor coolant system temperature is above 215'F.

b. During normal power operation, suppression b. A visual inspection of the suppression chamber chamber water temperature shall be less than or interior, including water line regions, shall be equal to 85'F. made at each major refueling outage.

AMENDMENT NO. 127

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. If Specifications a and b above are not met c. Whenever heat from relief valve operation is within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be shut down being added to the. suppression pool, the pool using normal shutdown procedures. temperature shall be continually monitored and also observed and logged every 5 minutes until the heat addition is terminated.

d. During testing of relief valves which add heat to d. Whenever operation of a relief valve is indicated the torus pool, bulk pool temperature shall not and the bulk suppression pool temperature exceed 10'F above normal power operation limit reaches 160'F or above while the reactor specified in b above. In connection with such primary coolant system pressure is greater than testing, the pool temperature must be reduced 200 psig, an external visual examination of the within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to below the normal power suppression chamber shall be made before operation limit specified in b above. resuming normal power operation.

e. The reactor shall be scrammed from any e. Whenever there is indication of relief valve operating condition when the suppression pool operation with the local temperature of the bulk temperature reaches 110'F. Operation shall suppression pool reaching 200'F or more, an not be resumed until the pool temperature is external visual examination of the suppression reduced to below the normal power operation chamber shall be conducted before resuming limit specified in b above. power operation.

f. During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig at normal coo!down rates if the pool bulk temperature reaches 120'F.

AMENDMENT NO. 128

BASES FOR 3.3.2 AND 4.3.2 PRESSURE SUPPRESSION SySTEM PRESSURE AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL The combination of three and one-half foot downcomer submergence, 85'F suppression chamber water temperature at lake water temperature defined by specification 3.3.7/4.3.7 will maintain post-accident system temperature and pressure within FSAR design limits (FSAR Section Vl, XV, XVI).

The three and one-half foot minimum and the four and one-quarter foot maximum submergence are a result of Suppression Chamber Heat-up Analysis and the Mark I Containment Program respectively. The minimum submergence provides sufficient water to meet the Suppression Chamber Heat-up Analysis post LOCA and the maximum submergence limits the torus levels to be consistent with the Mark I Plant Unique Analysis.

~ I The 215'F limit for the reactor is specified, since below this temperature the containment can tolerate a blowdown without exceeding the 35 psig design pressure of the suppression chamber without condensation.

Actually, for reactor temperatures up to 312'F the containment can tolerate a blowdown without exceeding the 35 psig design pressure of the suppression chamber, without condensation.

Some experimental data suggests that excessive steam condensing loads might be encountered if the bulk temperature of the suppression pool exceeds 160'F during any period of relief valve operation with sonic conditions at the discharge exit. This can result in local pool temperatures in the'vicinity of the quencher of 200'F. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the, event of a relief valve inadvertently opens or sticks open. As a minimum, this action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that,no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

AMENDMENT NO. 129

BASES FOR 3.3.2 AND 4.3.2 PRESSURE SUPPRESSION SYSTEM PRESSURE AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL Continuous monitoring of suppression chamber water level and temperature and pressure suppression system pressure is provided in the control room. Alarms for these parameters are also provided in the control room.

To determine the status of the pressure suppression system, inspections of the suppression chamber interior surfaces at each major refueling outage with water at its normal elevation will be made. This will assure that gross defects are not developing.

AMENDMENT NO. 130

e LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.3 LEAKAGE RATE 4.3.3 LEAKAGE RATE vs@:

Applies to the allowable leakage rate of the primary Applies to the primary containment system leakage containment system. rate.

~Ob'ec ive: OIIDective:

To assure the capability of the containment in limiting To verify that the leakage from the primary radiation exposure to the public from exceeding containment system is maintained within specified values specified in 10 CFR 100 in the event of a loss-- values.

of-coolant accident accompanied by significant fuel cladding failure and hydrogen generation from a metal-water reaction.

a. Inte ra ed Primar Con ainmen Leaka e Rate-To assure that periodic surveillances of reactor T e ATes containment penetrations and isolation valves are performed so that proper maintenance and repairs are (1) Integrated leak rate tests shall be performed made during the service life of the containment, and at the test pressure (Pt) of 22 psig.

systems and components penetrating primary Containment pressure shall not be permitted containment. to decrease more than one (1) psi below Pt.

Whenever the reactor coolant system temperature is above 215'F the primary containment leakage rate shall be within the limits of 4.3.3.b.

AMENDMENT NO.

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT (2) Type B and C tests should be completed prior to each Type A test. Type B and C leakages (penalties) not accounted for in the Type A test shall be incorporated as minimum pathway additions to the Upper Confidence Limit (UCL) to determine the overall as left integrated leakage rate.

(3) If the leakage rate exceeds the acceptance criterion, corrective action shall be required.

If, during the performance of a Type A test, excessive leakage occurs through locally testable penetrations or isolation valves to the extent that it would interfere with the satisfactory completion of the test, these leakage paths may be isolated and the Type A re-test continued until completion.

The Type A test shall be considered a failed test. A local leakage test shall be performed at Pt before and after the repair of each isolated leakage path. The sum of the post repaired local leakage rates and the UCL shall be less than 75 percent of the maximum allowable leakage rate, L (22).

Local leakage rates shall not be subtracted from the Type A test results to determine the acceptability of a test. The as found and as left leakage data values of excessive leakage areas beyond acceptance criteria shall be provided to the NRC.

AMENDMENT NO. 132

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT (4) Closure of the containment isolation valves for the purpose of the test shall be accomplished by the means provided for normal operation of the valves.

(5) A Type A test shall last a minimum of eight (8) hours with leakage rates calculated based on "Total Time" method. If a twenty-four (24) hour test is performed the "Mass Point" method will be used to calculate leakage rates. A verification test shall be performed following each Type A test. The verification test provides a method for assuring that systematic error or bias is given adequate consideration. During the verification test, containment pressure may not decrease more than one (1) psi below Pt.

b. Acce ance Cri eria - T AT The maximum allowable leakage rate Lt (22) shall not exceed 1.19 weight percent of the contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the test pressure of 22 psig (Pt).

(2) The maximum allowable operational leakage, rate L o (22) which shall be met prior to power operation following a Type A test (either as measured or following repairs and retest) shall not exceed 0.75 L (22) (0.892 weight percent per day).

AMENDMENT NO. 133

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT (3) When adding the leakage rate measured

'during a Type C test to the results of a Type A test, the leakage rate shall be determined using minimum pathway analysis.

c. Fre cruin (1) Three Type A tests shall be conducted during each ten year service interval at approximately equal intervals. The third test will be conducted when the plant is shutdown for the 10 year inservice inspections.

(2) Retesting (a), If a Type A test fails to meet the acceptance criteria of 4.3.3.b.(1), a Corrective Action Plan that focuses attention on the cause of the problem shall be developed and implemented.

A Type A test that meets the require-ments of 4.3.3.a.(3) and 4.3.3.b.(2) is required prior to plant start-up. A report of the Corrective Action following the failed Type A shall be submitted to the NRC for review and approval with the Containment Leak Test Report.

AMENDMENT NO. 134

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (b) If any periodic Type A test fails to meet the acceptance criteria of 4.3.3.b.(1), the test schedule for subsequent Type A tests will be reviewed and approved by the NRC.

(c) If two consecutive periodic Type A tests (not including an immediate retest under (a)) fail to meet the acceptance criteria of 4.3.3.a. (3),

4.3.3.b.(1) or 4.3.3.b.(2), not-withstanding the periodic retest schedule of 4.3.3.c.(1), a Type A test must be performed at each refueling outage or every 18 months, which-ever occurs first, unless alternative leak test requirements are accepted by the NRC by means of specific

, exemption from Appendix J per 10CFR50.12. This testing shall be performed until two consecutive periodic Type A tests (not including an immediate retest under (a)) meet the acceptance criteria of 4.3.3.a. (3),

4.3.3.b.(1) and 4.3.3.b.(2), then the .

retest schedule specified in 4.3.3.c.(1) should be resumed.

AMENDMENT NO. 135

LIMITINGCONDITION FOR OPERATION - SURVEILLANCE REQUIREINENT

d. Local Leak R e-T e B and T e T (1) Primary containment testable penetrations and isolation valves required to be Type B or Type C tested by regulatory requirements, shall be tested at a pressure of 35.0 psig (Pa) each major refueling outage, not to exceed two years, except as provided in (a) and (b) below.

(a) Bolted double gasketed seals which shall be tested whenever the seal is closed after being opened and at least at each refueling outage not to exceed a two year interval.

(b) Type B tests for primary containment penetrations employing a continuous leakage monitoring system shall be conducted at intervals not to exceed three years.

AMENDMENT NO. 136

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) When system pressure (Psys) on the opposite side of the isolation valve under test cannot be reduced to atmospheric pressure, then the test pressure shall not be less than Pa + Psys.

(3) Personnel airlocks shall be leak tested in accordance with the foliowing:

(a) The airlocks shall be tested at a test pressure of 35 psig following a refueling outage or maintenance outage requiring drywell access prior to primary containment integrity being required.

(b) Airlocks opened during periods when primary containment integrity is required shall be tested within three days after being opened. For airlock doors opened more frequently than once every three days, the airlocks shall be tested at least once every three days.

(c) The airlocks shall be tested every six .

months at a test pressure of 35 psig.

(d) Leakage rate for airlocks shall not exceed 0.05La at 35 psig.

AMENDMENT NO. 137

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (4) Primary containment penetrations and isolation valves that are not defined as Type B or Type C test components (e.g., seal welded cold instrument lines, CRD lines, drywell to wetwell connections, etc.) shall not be individually tested. The penetrations will be considered as integral parts of the Type A test.

e. Ac e tanceCri erie-T eB ndT eCTe s The combined leakage rate for penetrations and valves subject to Type B and C tests determined by maximum pathway analysis shall be less than 0.60 La. If this value is exceeded, repairs and retests shall be performed to correct the condition.

f. n inu Leak Ra e Moni orin (1) When the primary containment is inerted, the containment shall be monitored for gross leakage by a weekly review of the inerting system makeup requirements.

(2) This monitoring system may be taken out of service for the purpose of maintenance or testing but shall be returned to service as these activities are completed.

AMENDMENT NO. 138

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

g. ~ln ec ion The accessible interior surfaces 'of the primary containment shall be visually inspected each operating cycle for evidence of deterioration.

AMENDMENT NO. 139

BASES FOR 3.3.3 AND 4.3.3 LEAKAGE RATE The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident. The peak drywell pressure would be 35 psig which would rapidly reduce to 22 psig within 100 seconds following the pipe break. The total time the drywell pressure would be above 22 psig is calculated to be about 10 seconds. Following the pipe break, the suppression chamber pressure rises to 22 psig within 10 seconds, equalizes with drywell pressure and thereafter rapidly decays with the drywell pressure decay.

The design pressures of the drywell and suppression chamber are 62 psig and 35 psig, respectively. As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after about 10 seconds. Based on the calculated primary containment pressure response discussed above and the suppression chamber design pressure; primary containment preoperational test pressures were chosen. Also, based on the primary containment pressure response and the fact that the drywall and a suppression chamber function as a unit, the primary containment will be tested as a unit rather than testing the individual components separately.

design basis loss-of-coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1.9%/day at

'5Thepsig. The analysis showed that with this leak rate and standby gas treatment system filter efficiency of 90 percent for halogens, 95 a

percent for particulates, and assuming the fission product release fractions stated in TID-14844, the maximum total whole body passing cloud dose is about 6.0 rem and the maximum total thyroid dose is about 150 rem at the site boundary considering fumigation conditions over an exposure duration of two hours. The resultant doses would occur for the duration of the accident at the low population distance of 4 miles are lower than those stated due to the variability of meteorological conditions that would be expected to occur over a 30-day period. Thus, the doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident.

These doses are also based on the assumption of no holdup in the secondary containment resulting in a direct release of fission products from the primary containment through the filters and stack to the environs. Therefore, the specified primary containment leak rate and filter efficiency (Specification 4.4.4) are conservative and provide margin between expected offsite doses and 10CFR100 guideline limits.

The maximum allowable leakage rate (La) is 1.5%/day at a pressure of 35 psig (Pa). This value for the test condition was derived from the maximum allowable accident leak rate of about 1.9%/day when corrected for the effects of containment environment under accident and test conditions. In the accident case, the containment atmosphere initially would be composed of steam and hot air depleted of oxygen whereas under test conditions the test medium would be air or. nitrogen at ambient conditions. Considering the differences in mixture composition and temperatures, the appropriate correction factor applied was 0.8 and determined from the guide on containment testing.>> )

AMENDMENT NO. 140

BASES FOR 3.3.3 AND 4.3.3 LEAKAGE RATE Although the dose calculations suggest that the allowable test leak rate could be allowed to increase to about 3.0%/day before the guideline thyroid dose limit given in 10CFR100 would be exceeded, establishing the limit at 1.5'/c/day provides an adequate margin of safety to assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime. Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak rate. The operational limit is derived by multiplying the allowable test leak rate by 0.75 thereby providing a 25% margin to allow for leakage deterioration which may occur during the period between leak rate tests.

A reduced pressure test program is used for the integrated test. The test pressures are based on loss-of-coolant accident conditions. The peak primary containment pressure following a loss-of-coolant accident would be 35 psig. This would rapidly reduce to 22 psig. The total time drywell pressure would be above 22 psig would be about 10 seconds. Preoperational integrated leak tests were performed at test pressures at 35 psig and 22 psig. Subsequent integrated tests are performed at a test pressure of 22 psig.

Closure of the containment isolation valves for the purpose of the test is accomplished by the means provided for normal operation of the valves. The reactor is vented to the containment atmosphere during testing.

The acceptance criteria states that the maximum allowable leakage rate (4) shall not exceed 1.19 weight percent of the contained air in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 22 psig (Pt). This corresponds to the maximum allowable leakage rate (La) of 1.5 weight percent at 35 psig (Pa). The maximum allowable test leak rate L, (at 22 psig) shall not exceed the 1.5'k/day times the square root of the ratio of the pressures Pt (at 22 psig) and Pa (at 35 psig), respectively since the ratio of measured leakages for Nine Mile Point Unit 1 is 0.735. The allowable operational leakage rate, L,o (at 22 psig) shall not exceed 75 percent of Lt (at 22 psig) and shall be met prior to resumption of power operation following a test.

The primary containment leak rate test frequency is based on maintaining adequate assurance that the leak rate remains within the specification. The leak rate test frequency is based on 10 CFR 50 Appendix J.

The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends. Whenever a double-gasketed penetration (primary containment head equipment hatches and the suppression chamber access hatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of 35 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure. It is expected that the majority of the leakage from valves, penetrations and seals would be into the reactor building. However, it is possible that leakage into other parts of the facility could occur. Such leakage paths that may affect significantly the consequences of accidents are to be minimized. If the leakage rates of the double-gasketed seal penetrations, testable penetration isolation valves, containment air purge inlets and outlets and the vacuum relief valves are at the maximum specified, they will total 90 percent of the allowed leak rate. Hence, 10 percent margin is left for leakage through walls and untested components.

Leakage from airlocks is measured under accident pressures in accordance with 10 CFR 50 Appendix J.

AMENDMENT NO. 141

BASES FOR 3.3.3 AND 4.3.3 LEAKAGE RATE Monitoring the nitrogen make-up requirements of the inerting system provides a method of observing leak rate trends. This instrumentation equipment must be periodically removed from service for test and maintenance, but this out-of-service time will be kept to a practical minimum.

The test program follows the guidelines stated in the Bechtel Topical Report. " This program provides adequate assurance that the test results realistically estimates the degree of containment leakage following a loss-of-coolant accident. The containment leakage rate is calculated using the Absolute Methodology. Containment leakage results are presented in the test report as calculated using the Total Time and Mass Point techniques. The results of local leak rate tests, including, "as-found" and "as-left" leakages, are also included in the containment leak test report.

The specific treatment of selective valve'rrangements including the acceptability of the interpretations of 10 CFR 50 Appendix J requirements are given in References 5, 6, and 7. They serve as the bases for alternative test configurations (e.g., reverse accident, multi-valve, water leakage flow tests) as well as relaxations from previous leakage limits or constraints.

References:

FSAR, Volume II, Appendix E (2) UFSAR, Section Vl B.2.1 (3) TID-20583, Leakage Characteristics of Steel Containment Vessels and the Analysis of Leakage Determinations (4) BN-TOP-1 "Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants,"

Revision 1, Bechtel Corporation, November 1, 1972 (5) NRC Safety Evaluation Report dated May 6, 1988, "Regarding Proposed Technical Specifications and Exemption Requests Related to Appendix J."

~ r ~

(6) Niagara Mohawk Letter dated July 28, 1988, "Clarifications, Justifications 5 Conformance with 10 CFR 50 Appendix J SER."

(7) NRC Letter dated November 9, 1988, "Review of the July 28, 1988 Letter on Appendix J Containment Leakage Rate Testing at Nine Mile Point Unit 1."

(8) ANSI/ANS - 56.8 - 1987, "Containment System Leakage Testing Requirements."

AMENDMENT NO. 142

I LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.4 PRIMARY CONTAINMENT ISOLATION VALVES 4.3.4 PRIMARY ONTAINMENT I OLATI N VALVE URIC Applies to the operating status of the system of Applies to the periodic testing requirements of the isolation valves on lines open to the free space of the primary containment isolation valve system.

primary containment.

~Ob ec iv: ~ob'e ive:

To assure that potential leakage paths from the To assure the operability of the primary containment primary containment in the event of a loss-of-coolant isolation valves to limit potential leakage paths from accident are minimized. the containment in the event of a loss-of-coolant accident.

a. Whenever the reactor coolant system tempera- The primary containment isolation valves surveillance ture is greater than 215'F, all containment shall be performed as indicated (see Table 3.3.4).

isolation valves on lines open to the free space of the primary containment shall be operable except a. At least once per operating cycle the operable as specified in 3.3.4b below. isolation valves that are power operated and automatically initiated shall be tested for

b. In the event any isolation valve becomes automatic initiation and closure times.

inoperable the system shall be considered operable provided that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at least b. At least once per quarter all normally open power one valve in each line having an inoperable valve operated isolation valves shall be fully closed and is in the mode corresponding to the isolated reopened.

condition.

AMENDMENT NO. 143

0 LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. If Specifications 3.3.4 a and b are not met, the c. At least once per operating cycle, each reactor coolant system temperature shall be instrument-line flow check valve will be tested reduced to a value less than 215'F within ten for operability.

hours.

AMENDMENT NO. 144

LIMITING CONDITION FOR OPERATION Table 3.3e4 PRIMARY CONTAINMENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Location Relative Maximum (nit(sting Signal (All No. of Valves To Primary Normal Oper. Time Aotlon on Valves Have Remote Une or System (Each Une) Containment Position Motive Power~ (Seo) (nit(at(ng Signal Manual Backup)

D well Vent & Pur e N> Connection Outside Closed Pn/DC Solenoid 16 Close Reactor water level (One Line) Outside Closed AC Motor 30 Close low-low or high drywell pressure or Air Connection Outside Closed Pn/DC Solenoid 16 Close high radiation at stack (One Line) Outside Closed AC Motor 30 Close monitoring Su ression Chamber Vent & Pur e N2 Connection 1 ~ Outside Closed Pn/DC Solenoid 16 Close Reactor water level (One Line) 1 Outside Closed AC Motor 30 Close low-low or high drywell pressure or Air Connection Outside Closed Pn/DC Solenoid 16 Close high radiation at stack (One Line) Outside Closed AC Motor 30 Close monitoring Drrweii N~~Makea Reactor water level (One Line) Outside Closed Pn/DC Solenoid 60 Close low-low or drywell high pressure ba reaaiea Chamber bib~Maker Reactor water level (One Line) Outside Closed Pn/DC Solenoid BO Close low-low or drywell high pressure D well E ui ment Drain Line( Inside Open AC Motor 60 Close (One Line) Outside Open Pn/DC Solenoid 60 Close Reactor water level low-low or drywell D ell Floor Drain Line( ) Inside Open AC Motor BO Close high pressure (One Line) Outside Open Pn/DC Solenoid 60 Close Vacuum Relief ~ r ~

Atmosphere to Pressure Suppression System Outside Closed Pn/DC Solenoid Open Negative pressure (Three Lines) relative to atmosphere Outside Closed Self Act. Ck.

Reactor Cleanu S stem Relief Valve

~Diaaher e (One Line to Suppression Chamber) Outside Closed Self Act. Ck.

AMENDMENT NO. 145

LIMITING CONDITIONS FOR OPERATION Table 3.3e4 (oont'd)

PRIMARY CONTAINMENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Location Relative Maximum In)tlat(ng Signal (All-No. of Valves To Primary Normal Oper. Time Act(enon Valves Have Remote Une or System (Each Une) Containment Position Motive Power (Seo) (nit(at(ng Signal Manual Backup)

Outside Open Pn/DC Solenoid 60 Close (Two Lines)

Su ression Chamber Su I Outside Open Pn/DC Solenoid 60 Close Reactor Water level (One Line) low-Iow or high drywell pressure D ell Return Outside Open Pn/DC Solenoid 60 Close (One Line) '0 Su ression Chamber Return Outside Open Pn/DC Solenoid Close (One Line)

He0~~412 Sam iin Outside Open Pn/DC Solenoid 60 Close Reactor water level (Three Lines) low-low or high drywell pressure Su ression Chamber Su I Outside Open Pn/DC Solenoid 60 Close (One Line)

DrrDelI Return it i Outside Open Self Act. Ck.

(One Line)

Su ression Chamber Return(I) Outside Open Self Act. Ck.

(One Line)

AMENDMENT NO. 146

LIMITINGCONDITION FOR OPERATION Table 3.3e4 (cont'd)

PRIMARY CONTAINMENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Location Relative Maximum Initiating Signal (All No. of Valves To Primary Normal Oper. Time Action on Valves Have Remote (Each Une) Containment Poslon Motive Powers (Seo) Initiating Signal Manual Backup)

Une or System

~Cote S to Suctiooi Outside Open AC Motor 90 Remote Manual

~Pum (Four Lines From Suppression Chamber)

Outside Closed AC Motor 27 Close Reactor water level low-low or high (Two Test Lines to Suppression Chamber) drywell pressure Condensate Su I Outside Open Self Act. Ck.

(Keep Fill)

(Two Lines)

Outside Closed Pn/DC Solenoid 27 Close Reactor water level Core S ra Hi h Point Vent Inside Closed AC Motor 27 Close low-low or high (Two Lines) drywall pressure Containment S re 0 ell & Su ression Chamber Outside Open Pn/DC Solenoid 60 Remote Manual Common Su I (Four Lines)

Outside Closed Self Act. Ck.

(Four Lines)

Su ression Chamber Branch 20 ~ Outside Closed Self Act. Ck.

(One Branch for Each System) ression Chamber( Outside Open AC Motor 70 Remote Manual Pum Suction From Su (Four Lines)

I Outside Closed AC Motor 6O Remote Manual Containment S re Test Line to Torusi (One Line)

Outside Closed AC Motor Remote Manual Emer enc Coolin Vent to Torus (One Line)

AMENDMENT NO. 147

LIMITING CONDITIONS FOR OPERATION Table 3.3.4 (oont'd)

PRIMARY CONTAINMENT ISOLATION VALVES LINES ENTERING FREE SPACE OF THE CONTAINMENT Location Relative Maximum Initiating Signal (All No+ of Valves To Primary Normal Oper. Time Act)on on Valves Have Remote Une or System (Each Une) Containment Position Motive Power'Seo) Initiating Signal Manual Backup)

Containment Atmos here Monitorin Su I Outside Open Pn/DC Solenoid 60 Close Reactor water level Line low-low or high (One Line) drywall pressure Containment Poet LOCA Vent Outside Closed Pn/DC Solenoid 60 Close Reactor water level (Two Lines) low-low or high drywell pressure N2 Pur e - TIP Indexers (t1 Outside Closed Self Act. Ck.

(One Line)

Traversin Incore Probe Outside Closed AC Motor 60 Close Reactor water level (Four Lines) low-low or high drywell pressure Breathin Air Connection Inside Closed (One Line) Outside Closed Local Manual Service Water Connection(t) inside Closed (One Line) Outside Closed LINES WITH A CLOSED LOOP INSIDE CONTAINMENTVESSElS Recirculation Pum Coolin Water( I Supply Line Outside Open Self Act. Ck.

Return Line Outside Open DC Motor 60 Remote Manual D ell Cooler Water( I Supply Line Outside Open Self Act. Ck.

Return Line Outside Open DC Motor 60 Remote Manual AMENDMENT NO. 148

Notes:

Pn - Pneumatically Operated One valve in each separate line and one valve in each common line.

~

(1) These valves do not have to be vented during the Type A test. However, Type C leakage from these valves is added to the Type A test results, if not vented.

(2) These valves are provided with a water seal capability. No Appendix J or IST leakage rate testing is required.

(3) These valves are water leak rate tested and acceptance criteria are established in accordance with the IST Program.

(4) These valves are provided with a water seal. Valves shall be tested during each refuel outage not to exceed two years consistent with Appendix J water seal testing requirements. Leakage rates shall be limited to 0.5 gpm per nominal inch of valve diameter up to a maximum of 5 gpm.

(5) These valves do not meet the requirements of 10CFR50 Appendix J, Section II-H. No testing required.

AMENDMENT NO.

BASES FOR 3.3.4 AND 4.3.4 PRIIVIARY CONTAINMENT ISOLATION VALVES Double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Except where check valves are used as one or both of a set of double isolation valves, the isolation closure times are presented in Table 3.3 4. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident. Details of the isolation valves are discussed in Section Vl-D. For allowable leakage rate specification, see Section 3.3.3/4.3.3.

For the design basis loss-of-coolant accident fuel rod perforation would not occur until the fuel temperature reached 1700'F which occurs in approximately 100 seconds. A required closing time of 60 seconds for all primary containment isolation valves will be adequate to prevent fission product release through lines connecting to the primary containment.

For reactor coolant system temperatures less than 215'F, the containment could not become pressurized due to a loss-of-coolant accident.

The 215'F limit is based on preventing pressurization of the reactor building and rupture of the blowout panels.

The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 that a line will not isolate (Fifth Supplement, p. 115).l ) More frequent testing for valve operability results In a more fellable system.

In addition to routine surveillance as outlined in Section VI-D.1.0(") each instrument-line flow check valve will be tested for operability. All instruments on a given line will be isolated at each instrument. The line will be purged by isolating the flow check valve, opening the bypass valves, and opening the drain valve to the equipment drain tank. When purging is sufficient to clear the line of non-condensibles and crud the flow-check valve will be cut into service and the bypass valve closed. The main valve will again be opened and the flow-check valve allowed to close. The flow-check valve will be reset by closing the drain valve and opening the bypass valve depressurizing part of the system. Instruments will be cut into service after closing the bypass valve. Repressurizing of the individual instruments assures that flow-check valves have reset to the open position.

An in-depth review of the NMP-1 design and operation relative to Appendix J requirements has evaluated the various system/valving configurations. " The results of the evaluation and subsequent clarifications are reflected in this specification and its bases.

(1) UFSAR (2) Nine IVlile Point Nuclear Generation Station Unit 1 Safer/Corecool/GESTR-LOCA Loss of Coolant Accident Analysis, NEDC-31446P, Supplement 3, September, 1990.

(3) FSAR (4) NRC Safety Evaluation Report, dated May 6, 1988, "Regarding Proposed Technical Specifications and Exemption Requests Related to Appendix J."

(5) Niagara Mohawk Letter dated July 28, 1988, "Clarifications, Justifications 5 Conformance with 10CFR50 Appendix J SER."

AMENDMENT NO. 150

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT CCCCO Applies to the access control to the primary Applies to the surveillance on primary containment containment. access control.

~Ob e ive: ~ob' ive:

To assure the integrity of the primary containment To assure the operability of the primary containment system. access control interlocks.

Whenever the reactor coolant system temperature is A mechanical interlock will be maintained to prevent above 215'F the following shall be in effect. simultaneous opening of two doors.

a. Only one door in each of the two double-door drywell access locks will be opened at one time.

b. The equipment hatch and drywell head and other flanged openings will be secured.

c. If following a routine surveillance check "a" or "b" is not met, initiate normal orderly shutdown within one hour and have reactor in the cold shutdown condition within ten hours.

AMENDMENT NO.

BASES 3.3.5 AND 4.3.5 ACCESS CONTROL Access to the containment during operation is expected to be infrequent. However, each door of the two double-doored access locks is designed to withstand 62 psig drywell pressure. It is, therefore, possible to open one door at a time and still maintain containment integrity. Access door design is discussed in Section Vl-A 2.2 of the FSAR.

The equipment hatch and drywell head and other flanged openings are provided with double "0" rings and must be secure in order to maintain the integrity of the primary containment system. Maintaining the pressure suppression system integrity when above the stated pressure and temperature will ensure that a reactor coolant system rupture will not result in an overpressurization of the reactor building.

AMENDMENT NO. 152

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.6 VACUUM RELIEF 4.3.6 VACUUM RELIEF bills:

Applies to the operational status of the primary Applies to the periodic testing of the vacuum relief containment vacuum relief system. system.

~ob'ec ive: ~Ob'ec ive:

To assure the capability of the vacuum relief system To assure the operability of the containment vacuum in the event of a loss-of-coolant accident to: relief system to perform its intended functions.

a. Equalize pressures between the drywell and suppression chamber, and

b. Maintain containment pressure above the vacuum design values of the drywell and suppression chamber.

a. When primary containment is required, all a. Periodic 0 erabilit Tests suppression chamber - drywell vacuum breakers shall be operable except during testing and as Once each month and following any release of stated above. Suppression chamber - drywell energy to the suppression chamber, each vacuum breakers shall be considered operable if: suppression chamber - drywell vacuum breaker shall be exercised. Operability of valves, position (1) The valve is demonstrated to open fully with switches, and position indicators and alarms shall the applied force at all valve positions not be verified monthly and following any exceeding that equivalent to 0.5 psi acting maintenance on the valves and associated on the suppression chamber face of the equipment.

valve disk.

AMENDMENT NO. 153

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) The valve can be closed by gravity, when b. Refuelin Outa e Tes released after being opened by remote or manual means, to within not greater than (1) All suppression chamber - drywell vacuum the equivalent of 0.06 inch at the bottom of breakers shall be tested to determine the the disk. force required to open each valve from fully closed to fully open.

(3) The position alarm system will annunciate in the control room if the valve opening (2) All suppression chamber - drywell vacuum exceeds the equivalent of 0.06 inch at the breaker position indication and alarm bottom of the disk. systems shall be calibrated and functionally tested.

b. Any drywell-suppression chamber vacuum breaker may be non-fully closed as indicated by (3) Once each operating cycle, each vacuum the position indication and alarm systems breaker valve shall be visually inspected to provided that drywell to suppression chamber ensure proper maintenance and operation.

differential pressure decay rate is demonstrated to be not greater than 25% of the differential (4) A drywell to suppression chamber. leak rate pressure decay rate for all vacuum breakers open test shall demonstrate that with an initial the equivalent of 0.06 inch at the bottom of the differential pressure of not less than 1.0 psi, disk. the differential pressure decay rate shall not exceed the equivalent of the leakage rate through a 1-inch orifice.

AMENDMENT NO. 154

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

c. When it is determined that one or more vacuum c. Pre sur S re i n Cham r - Reac or B il in

'reaker valves are not fully closed as indicated Vac um Breakers by the position indication system at a time when such closure is required, the apparently (1) The pressure suppression chamber-reactor malfunctioning vacuum breaker valve shall be building vacuum breaker systems and exercised and pressure tested as specified in associated instrumentation, including set 3.3.6 b immediately and every 15 days point, shall be checked for proper operation thereafter until appropriate repairs have been every three months.

completed.

(2) During each refueling outage, each vacuum

d. One drywell-suppression chamber vacuum breaker shall be tested to determine that the breaker may be secured in the closed position. force required to open the vacuum breaker does not exceed the force specified in

e. If Specifications 3.3.6 a, b, c, or d cannot be Specification 3.3.6.f(1) and each vacuum met, the situation shall be corrected within 24 breaker shall be inspected and verified to hours or the reactor shall be placed in a cold meet design requirement.

shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

f. Pressure Su res i n Chamber - Reactor Buildin Vacuum Breakers (1) The three pressure suppression chamber reactor building vacuum breaker systems shall be operable at all times when the primary containment integrity is required.

The set point of the differential pressure instrumentation which actuates the pressure suppression chamber-reactor building air-operated vacuum breakers shall be (0.5 psld.

AMENDMENT NO. 155

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT The self-actuating vacuum breakers shall open fully when subjected to a force equivalent to or less than 0.5 psid acting on the valve disk.

(2) From and after the date that one of the pressure suppression chamber-reactor building vacuum breaker systems is made or found inoperable for any reason, the vacuum

. breaker shall be locked closed and reactor operation is permissible only during the succeeding seven (7) days unless such vacuum breaker system is sooner made operable, provided that the procedure does not violate containment integrity.

AMENDMENT NO. 156

BASES FOR 3.3.6 AND 4.3.6 VACUUM RELIEF Four vacuum relief valves are provided between the drywell and suppression chamber (Section VI-A.1.5 and 2.6)". Each valve is capable of opening on a differential pressure of 0.25 a 0.10 psi. The operation of any one valve will prevent damage to the drywell under the accident conditions expected following the loss-of-coolant accident due to a recirculation line break. As discussed in Section Vl-F", one valve operation will limit maximum pressure differential between the two chambers to approximately 3 psi, well below the maximum allowable pressure differential of 8.94 psi.

At-a coolant temperature of 215'F, the steam generated during a loss-of-coolant accident would not be sufficient to purge the drywell or suppression chamber.

Three sets of vacuum relief valves are provided between the primary containment and atmosphere (Section VI-A.1.5 and 2.6)". Each valve is capable of opening on a differential pressure of 0.25 s 0.10 psi. As discussed in Section VI-A.2.6", operation of all three relief valve sets will prevent containment pressure from dropping below the vacuum ratings of the drywall and the suppression chamber. The selection of these valves is based on the conservative assumption that the ventilation valves on the suppression chamber were left open during a postulated loss-of-coolant accident, permitting the pressure suppression system to blow down to atmospheric pressure. Closure of the ventilation valves followed by startup of the containment spray and core spray pumps leads to a rapid condensation of the steam in the drywell and a consequent drop in pressure below atmospheric., Normally, the ventilation valves are locked closed and there is little likelihood of this series of events occurring. Subsequent calculations showed that with only two valve sets operating, the worst vacuum in the suppression chamber is -3.0 psig. At this pressure a safety factor of about 1.70 still exists to incipient buckling.

Nearly all maintenance can be completed within a few days. Infrequently, however, major maintenance might be required. Replacement of principal system components could necessitate outages of more than 15 days. In spite of the best efforts of the operator to return equipment to service, some maintenance could require up to 6 months.

Using an analysis which is the same as used in the Fifth Supplement (page 115)" results in a failure probability of 1.8 x 10 for the drywell to suppression chamber valves and a failure probability of 9.5 x 10 for the valves between the containment and the atmosphere.

FSAR AMENDMENT NO. 157

SASES FOR 3.3.6 AND 4.3.6 VACUUM RELIEF Each drywelt-suppression chamber vacuum breaker is equipped with two independent switches to indicate the opening of the valve disk.

Redundant control room alarms are provided to permit detection of any drywell-suppression chamber vacuum breaker opening in excess of the described allowable limits. The containment design has been examined to establish the allowable bypass area between the drywall and suppression chamber as 0.053 square feet.

The limit on each individual valve will be set such that with all valves at their limit, the maximum value of cumulative leakage will not exceed the maximum allowable. The value will be at approximately 0.06 inch of disk travel off its seat and will be alarmed in the control room.

The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppression chamber and between the suppression chamber and reactor building so that the structural integrity of the containment is maintained.

The vacuum relief system from the pressure suppression chamber to the reactor building consists of three vacuum relief breakers (3 parallel sets of 2 valves in series). Operation of either system will maintain the pressure differential less than 1 psig; the external pressure is 2.

pslg.

The leak rate testing program is based on AEC guidelines for development of leak rate testing and surveillance schedules for reactor containment vessels.

Surveillance of the suppression chamber-reactor building vacuum breakers consists of operability checks and leakage tests (conducted as part of the containment leak-tightness tests). These vacuum breakers are normally in the closed position and open only during tests or an accident condition. Therefore, a testing frequency of three months for operability is considered justified for this equipment. Inspections and calibrations are performed during the refueling outages, this frequency is based on equipment quality, experience, and engineering judgment.

During each refueling outage, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by approximately 1 psi with respect to the suppression pool pressure and then held constant. The subsequent suppression. chamber transient will be monitored with a sufficiently sensitive pressure instrument. If the drywell pressure cannot be increased by 1 psi over the suppression chamber pressure, it would indicate existence of a significant leakage path which will be identified and eliminated before further drywall vacuum breaker testing.

AMENDMENT NO. 158

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.7 ~ CONTAINMENT SPRAY Y TEM 4.3.7 NTAINMENT PRAY YSTEM Applies to the operating status of the containment Applies to the testing of the containment spray spray system. system.

~Ob ective: ~Ob ec iv:

To assure the capability of the containment spray To verify the operability of the containment spray system to limit containment pressure and temperature system.

in the event of a loss-of-coolant accident.

The containment spray system surveillance shall be

a. During all reactor operating conditions whenever performed as indicated below:

reactor coolant temperature is greater than 215'F and fuel is in the'reactor vessel; each of a. Containment Spray Pumps the two containment spray systems and the associated raw water cooling systems shall be (1) A I a n r er in I, automatic operable except as specified in 3.3.7.b. startup of the containment spray pump shall be demonstrated.

b. If a redundant component of a containment spray system becomes inoperable, Specification (2) A I a n e er uar er, pump operability 3.3.7.a shall be considered fulfilled, provided shall be checked.

that the component is returned to an operable condition within 15 days and that the b. Nozzles required is performed.

additional'urveillance At least once per operating cycle, an air test shall be performed on the spray headers and nozzles.

AMENDMENT NO. 159

-0 0

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. If a redundant component in each of the c. Raw Water Cooling Pumps containment spray systems or their associated raw water systems. become inoperable, both At least once per quarter manual startup and systems shall be considered operable provided operability of the raw water cooling pumps shall that the component is returned to an operable be demonstrated.

condition within 7 days and that the additional surveillance required is performed. d. Surveillance with Inoperable Components

d. If a containment spray system or its associated When a component or system becomes raw water system becomes inoperable and all inoperable its redundant component or system the components are operable in the other shall be verified to be operable immediately and systems, the reactor may remain in operation for daily thereafter.

a period not to exceed 7 days.

B. If Specifications "a" or "b" are not met,

" shutdown shall begin within one hour and the reactor coolant shall be below 215'F within ten hours.

If both containment spray systems become inoperable the reactor shall be in the cold shutdown condition within ten hours and no work shall be performed on the reactor which could result in lowering the reactor water level to more than six feet, three inches (-10 inches indicator scale) below minimum normal water level (Elevation 302'9").

AMENDMENT NO. 160

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

f. The containment spray system shall be f. Lake Water Temperature considered operable by verifying that lake water temperature does not exceed 81'F. Record at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when latest recorded water

g. If specification "f" cannot be met commence temperature is greater than or equal to 75'F and shutdown within one hour and be in hot at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the latest shutdown within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and cold shutdown recorded wate'r temperature is greater than or within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. equal to 79'F.

AMENDMENT NO.

0 h

BASES FOR 3.3.7 AND 4.3.7 CONTAINMENT SPRAY SYSTEM For reactor coolant temperatures less than 215'F not enough steam is generated during a loss.-of-coolant accident to pressurize the

~

containment. For reactor coolant temperatures up to 312'F, the resultant loss-of-coolant accident pressure would not exceed the design pressure of 35 psig.

Operation of only one containment spray pump is sufficient to provide the required containment spray cooling flow. The specified flow of 3600 gpm at 87:7 psid primary, 89 psid secondary (approximately 95 percent to the drywell and the balance to the suppression chamber) is sufficient to remove post accident core energy released (FSAR Section Vll). Requiring both pumps systems operable (400 percent redundancy) will assure the availability of the containment spray system.(")

Allowable outages are specified to account for components that become inoperable in both systems and for more than one component in a system.

The containment spray raw water cooling system is considered operable when the flow rate is not less than 3000 gpm and the pressure on the raw water side of the containment spray heat exchangers is 10 psig greater than that on the torus water side (not less than 141 psig).

The higher pressure on the raw water side will assure that any leakage is into the containment spray system.

Electrical power for all system components is normally available from the reserve transformer. Upon loss of this service the pumping requirement will be supplied from the diesel generator. At least one diesel generator shall always be available to provide backup'lectrical power for one containment spray system.

Automatic initiation of the containment spray system assures that the containment will not be overpressurized. This automatic feature would only be required if all core spray systems malfunctioned and significant metal-water reaction occurred. For the normal operation condition of 85'F suppression chamber water, containment spray actuation would not be necessary for about 15 minutes.

(" With two of the containment spray intertie valves open, operation of two containment spray pumps is required to assure the proper flow distribution to the containment spray headers to reduce containment pressure during the first fifteen minutes of the LOCA. Requiring two containment spray pumps to operate reduces the 400 percent redundance of the containment spray system, but there are still six combinations (two out of four pumps) that will assure two pump operation.

AMENDMENT NO. 162

0 BASES FOR 3.3.7 AND 4.3.7 CONTAINMENT SPRAY SYSTEM In conjunction with containment spray pump operation during each operating cycle, the raw water pumps and associated cooling system performance will be observed. The containment spray system shall be capable of automatic initiation from simultaneous low-low reactor water level and high containment pressure. The associated raw water cooling system shall be capable of manual actuation. Operation of the containment spray system involves spraying water into the atmosphere of the containment. Therefore, periodic system tests are not practical. Instead separate testing of automatic containment spray pump startup will be performed during each operating cycle. During pump operation, water will be recycled to the suppression chamber. Also, air tests to verify that the drywall and torus spray nozzles and associated piping are free from obstructions will be performed each operating cycle. Design features are discussed in Volume I, Section Vll-B.2.0 (page Vll-19)". The valves in the containment spray system are normally open and are not required to operate when the system is called upon to operate.

The test interval between operating cycle results in a system failure probability of 1.1 x 10 (Fifth Supplement, page 115)" and is consistent with practical considerations. Pump operability will be demonstrated on a more frequent basis and will provide a more reliable system.

"FSAR AMENDMENT NO. 163

3.4.0 REA TOR BUILDING APPLICABILITY Applies to the operating status of the reactor building.

OB J ECTIVE To assure the integrity of the reactor building.

SPECIFICATION Reactor building. integrity must be in effect in the refueling and power operating conditions and also whenever irradiated fuel or the irradiated fuel cask is being handled in the reactor building.

AMENDMENT NO. 164

0 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.4.1 LEAKAGE RATE 4.4.1 LEAKAGE RATE Applies to the leakage rate of the secondary Applies to the periodic testing requirements of the containment. secondary containment leakage rate.

~Ob ec ive: Qbbec~iv:

To specify the requirements necessary to limit To assure the capability of the secondary exfiltration of fission products released to the containment to maintain leakage within allowable secondary containment as a result of an accident. limits.

Whenever the reactor is in the refueling or power Once durin each o er in Ie - isolate the reactor operating condition, the reactor building leakage rate building and start emergency ventilation system fan as determined by Specification 4.4.1 shall not exceed to demonstrate negative pressure in the building 2000 cfm. If this cannot be met after a routine relative to external static pressure. The fan flow rate surveillance check, then the actions listed below shall shall be varied so that the building internal differential be taken: pressure is at least as negative as that on Figure 3A.1 for the wind speed at which the test is

a. Suspend immediately irradiated fuel handling, conducted. The fan flow rate represents the reactor fuel pool and reactor cavity activities, and building leakage referenced to zero mph with building irradiated fuel cask handling operations in the internal pressure at least 0.25 inch of water less than reactor building. atmospheric pressure. The test shall be done at wind speeds less than 20 miles per hour.

b. Restore the reactor building leakage rates to within specified limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or initiate normal orderly shutdown and be in a cold shutdown condition within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

AMENDMENT NO. 165

0 FIGURE 3.4.1 REACTOR OUILOIHG PRESSURE

-0.25 0

le 0.30 lai

-0.35 8 10 12 14 16 10 20 MIND SPEEU (HPll)

AMENDMENT NO.

0 it

BASES FOR 3.4.1 AND 4.4.1 LEAKAGE RATE In the answers to Questions II-3 and IV-5 of the Second Supplement and also in the Fifth Supplement", the relationships among wind speed direction, pressure distribution outside the building, building internal pressure, and reactor building leakage are discussed. The curve of pressure in Figure 3 4.1 represents the wind direction which results in the least building leakage. It is assumed that when the test is performed, the wind direction is that which gives the least leakage.

If the wind direction was not from the direction which gave the least reactor building leakage, building internal pressure would not be as negative as Figure 3.4.1 indicates. Therefore, to reduce pressure, the fan flow rate would have to be increased. This erroneously indicates that reactor building leakage is greater than if wind direction were accounted for. If wind direction were accounted for, another pressure curve could be used which was less negative. This would mean that less fan flow (or measured leakage) would be required to establish building pressure. However, for simplicity it is assumed that the test is conducted during conditions leading to the least leakage while the accident is assumed to occur during conditions leading to the greatest reactor building leakage.

As discussed in the Second Supplement and Fifth Supplement, the pressure for Figure 3 4.1 is independent of the reactor building leakage rate referenced to zero mph wind speed at a negative differential pressure of 0.25 inch of water. Regardless of the leakage rate at these design conditions, the. pressure versus wind speed relationship remains unchanged for any given wind direction.

By requiring the reactor building pressure to remain within the limits presented in Figure 3 4.1 and a reactor building leakage rate of less than 2000 cfm, exfiltration would be prevented. This would assure that the leakage from the primary containment is directed through the filter system and discharged from the 350-foot stack.

NFSAR AMENDMENT NO. 167

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.2 . REACTOR BUILDIN INTEGRITY - I OLATION 4A.2 REACT R B ILDIN INTEGRITY - I LATI N

~VALVE V~ALVE Applies to the operational status of the reactor Applies to the periodic testing requirements of the building isolation valves. reactor building isolation valves.

~Ob ec ive: ~Ob'ec ive:

To assure that fission products released to the To assure the operability of the reactor building secondary containment are discharged to the isolation valves.

environment in a controlled manner using the emergency ventilation system.

a. The normal Ventilation System isolation valves At least once per operating cycle, automatic initiation shall be operable whenever the reactor is in the of valves shall be checked.

refueling or power operating conditions, and whenever irradiated fuel or the irradiated fuel cask is being handled in the reactor building.

b. If Specification 3.4.2a is not met, the reactor shall be in the cold shutdown condition within ...

ten hours and handling of irradiated fuel cask shall cease.

AMENDMENT NO. 168

BASES FOR 3.4.2 AND 4.4.2 REACTOR BUILDING INTEGRITY ISOLATION VALVES Isolation of the reactor building occurs automatically upon high radiation of the normal building exhaust ducts or from high radiation at the refueling platform (See 3.6.2). Isolation will assure that any fission products entering the reactor building will be routed to the emergency ventilation system prior to discharge to the environment (Section VII-H.3.0 of the FSAR).

AMENDMENT NO. 169

I LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3A.3 ACCE S CONTROL ..3 Applies to the access control to the reactor building. Applies to the periodic checking of the condition of portions of the reactor building.

~Ob'ec ice: ~Ob t;~ive:

To specify the requirements necessary to assure the To assure that pump compartments are properly integrity of the secondary containment system. closed at all times and to assure the integrity of the secondary containment system by verifying that reactor building access doors are closed, as required by Specifications 3.4.3.a.1 and 3A.3.a.2.

a. Whenever the reactor is in the power operating condition, or when irradiated fuel is being handled in the reactor building, or during core alterations, or during irradiated fuel cask handling a. The core and containment spray operations in the reactor building, the following shall be checked once per week pump'ompartments conditions will be met: and after each entry.

1. Only one door in each of the double-doored access ways shall be opened at one time.

2. Only one door or closeup of the railroad bay shall be opened at one time.

3.'he core spray and containment spray pump compartments'oors shall be closed at all times except during passage in order to consider the core spray system and the containment spray system operable.

AMENDMENT NO. 170

LllVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. If these conditions cannot be met, then the b. Verify at least once per 31 days that:

actions listed below shall be taken:

1. At least one door in each access to the

1. If in the power operating condition, restore secondary containment is closed.

reactor building integrity within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least the hot shutdown condition 2.. At least one door or closeup of the railroad within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in the cold bay is closed.

shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. Suspend any. of the following activities:

a. core alterations,

b. Handling of irradiated fuel in the reactor building, c, irradiated fuel cask handling operations in the reactor building.

AMENDMENT NO.

BASES FOR 3.4.3 AND 4.4.3 ACCESS CONTROL The reactor building serves as a secondary containment during normal Station operations and as a primary containment during refueling and other periods when the pressure suppression system is open. Maintaining the building integrity and an operative emergency ventilation.

system for the conditions listed will ensure that any fission products inadvertently released to the reactor building will be routed through the emergency ventilation system to the stack. The worst such incident is due to dropping a fuel assembly on the core during refueling. The consequences of this are discussed in Section XV.C.3 of the FSAR.

As discussed in Section Vl-F" all access openings of the reactor building have as a minimum two doors in series. Appropriate local alarms and control room indicators are provided to always insure that reactor building integrity is maintained. Surveillance of the reactor building access doors provides additional assurance that reactor building integrity is maintained.

Maintaining closed doors on the pump compartments ensures that suction to the core and containment spray pumps is not lost in case of a gross leak from the suppression chamber.

"FSAR AMENDMENT NO. 172

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT 3.4.4 EMERGENCY VENTILATIONSY TEM 4A.4 EMER ENCY VENTILATI N Y TEM bi i Applies to the operating status of the emergency Applies to the testing of the emergency ventilation ventilation system. system.

~ob ective: OIIDBC~IV To assure the capability of the emergency ventilation To assure the operability of the emergency ventilation system to minimize the release of radioactivity to the system.

environment in the event of an incident within the primary containment or reactor building.

a. Except as specified in Specification 3.4.4e Emergency ventilation system surveillance shall be below, both circuits of the emergency ventilation performed as indicated below:

system and the diesel generators required for operation of such circuits shall be operable at all a. At least once per operating cycle, not to exceed times when secondary containment integrity is 24 months, the following conditions shall be required. demonstrated:

b. The results of the in-place cold DOP and halo- (1) Pressure drop across the combined HEPA genated hydrocarbon tests at design flows on filters and charcoal adsorber banks is less HEPA filters and charcoal adsorber banks shall than 6 inches of water at the system rated show a99% DOP removal and a99% halogen- flow rate (a10%).

ated hydrocarbon removal when tested in accordance with ANSI N.510-1980. (2) Operability of inlet heater at rated power when tested in accordance with ANSI N.510-1980.

AMENDMENT NO. 173

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. The results of laboratory carbon sample analysis .b. The tests and sample analysis of Specification shall show ~90% radioactive methyl iodide 3 4.4b, c and d shall be performed at least once removal when tested in accordance with ANSI per operating cycle or once every 24 months, or N.510-1980 at 80'C and 95% R.H. after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, whichever occurs first or following significant painting, fire

d. Fans shall be shown to operate within a10% or chemical release in any ventilation zone design flow. communicating with the system.

\

e. From and after the date that one circuit of the c. Cold DOP testing shall be performed after each emergency ventilation system is made or found complete or partial replacement of the HEPA to be inoperable for any reason, reactor operation filter bank or after any structural maintenance on and fuel handling is permissible only during the the system housing.

succeeding seven days unless such circuit is sooner made operable, provided that during such d. Halogenated hydrocarbon testing shall be seven days all active components of the other performed after each complete or partial emergency ventilation circuit shall be operable. replacement of the charcoal adsorber bank or after any structural maintenance on the system

f. If these conditions cannot be met, within 36 housing..

hours, the reactor shall be placed in a condition for which the emergency ventilation system is e. Each circuit shall be operated with the inlet not required. heater on at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.

f. Test sealing of gaskets for housing doors downstream of the HEPA filters and charcoal adsorbers shall be performed at and in conformance with each test performed for compliance with Specification 4.4.4b and Specification 3.4.4b.

AMENDMENT NO. 174

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

g. At least once per operating cycle, not to exceed 24 months, automatic initiation of each branch of the emergency ventilation system shall be demonstrated.

h. At least once per operating cycle, not to exceed 24 months, manual operability of the bypass valve for filter cooling shall be demonstrated.

i. When one circuit of the emergency ventilation system becomes inoperable all active compon.

ants in the other emergency ventilation circuit shall be verified to be operable within two hours and daily thereafter.

AMENDMENT NO. 175

BASES FOR 3.4 4 AND 4.4.4 EMERGENCY VENTILATIONSYSTEIVI The emergency ventilation system is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions. Both emergency ventilation system fans are designed to automatically start upon high radiation in the reactor building ventilation duct or at the refueling platform arid to maintain the reactor building pressure to the design negative pressure so as to minimize in-leakage. Should one system fail to start, the redundant system is designed to start automatically. 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 adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates.'he laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90 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 10CFR100 guidelines for the accidents analyzed. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

Only one of the two emergency ventilation systems is needed to cleanup the reactor building atmosphere upon containment isolation. If one system is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling operation may continue while repairs are being made. If neither circuit is operable, the plant is brought to a condition where the emergency ventilation system is not required.

Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreigrr matter. Heater capability and pressure drop should be determined at least once per operating cycle to show system performance capability.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated.

The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified in Table 5-1 of ANSI 509-1980.

AMENDMENT NO. 176

BASES FOR 3.4.4 AND 4.4.4 EMERGENCY VENTILATIONSYSTEM The replacement charcoal for the adsorber tray removed for the test should meet the same adsorbent quality. Any HEPA filters found defective shall be replaced with filters qualified pursuant to ANSI 509-1980.

All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity. Operation of the inlet heater will prevent moisture buildup in the filters and adsorber system.

With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repairs and test repeated.

If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use. The determination of significant shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination.

Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability. If one emergency ventilation system is inoperable, the other system must be verified to be operable daily. This substantiates the availability of the operable system and thus reactor operation or refueling operation may continue during this period.

AMENDMENT NO. 177

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.4.5 ONTROL ROOM AIR TREATMENT SYSTEM 4.4.5 ONTROL ROOM AIR TREATMENT Y TEM SUSLJSJI:

Applies to the operating status of the control room air Applies to the testing of the control room air treatment system. treatment system.

Qggg~ive: OIIO glrive:

To assure the capability of the control room air To assure the operability of the control room air treatment system to minimize the amount of radio- treatment system.

activity or other gases entering the control room in the event of an incident.

a. Except as specified in Specification 3.4.5e a. At least once per operating cycle, or once every below, the control room air treatment system 24 months, whichever occurs first, the pressure and the diesel generators required for operation drop across the combined HEPA filters and of this system shall be operable at all times when charcoal adsorber banks shall be demonstrated to reactor building integrity is required. be less than 6 inches of water at system design flow rate (a 10%).

b. The results of the in-place cold DOP and halo-genated hydrocarbon test design flows on HEPA b. The tests and sample analysis of Specification filters and charcoal adsorber banks shall show 3 4.5b, c and d shall be performed at least once

>99% DOP removal and )99% halogenated per operating cycle or once every 24 months, or hydrocarbon removal when tested in accordance after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, whichever with ANSI N.510-1980. occurs first or following significant painting, fire or chemical release in any ventilation zone communicating with the system.

AMENDMENT NO. 178

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

c. The results of laboratory carbon sample analysis c. Cold DOP testing shall be performed after each shall show ~90% radioactive methyl iodine complete or partial replacement of the HEPA removal when tested in accordance with ANSI filter bank or after any structural maintenance on N.510-1980 at 80'C and 95% R.H. the system housing.

d. 'Fans shall be shown to operate within a10% d. Halogenated hydrocarbon testing shall be design flow. performed after each complete or partial replacement of the charcoal absorber bank or

e. From and after the date that the control room air after any structural maintenance on the system treatment system is made or found to be housing.

inoperable for any reason, reactor operation or refueling operations is permissible only during the e. The system shall be operated at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> succeeding seven days unless the system is every month.

sooner made operable.

f. At least once per operating cycle, not to exceed

f. If these conditions cannot be met, reactor 24 months, automatic initiation of the control shutdown shall be initiated and.the reactor shall room air treatment system shall be demon-be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for reactor strated.

operations and refueling operations shall be terminated within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. g. At least once per operating cycle, not to exceed 24 months, the control room air treatment system shall be shown to maintain a positive pressure within the control room of greater than one sixteenth of an inch (water) relative to areas adjacent to the control room.

AMENDMENT NO. 179

BASES FOR 3.4.5 AND 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM The control room air treatment system is designed to filter the control room atmosphere for intake air. A roughing filter is used for recirculation flow during normal control room air treatment operation. The control room air treatment system is designed to automatically start upon receipt of a high radiation signal from one of the two radiation monitors located on the ventilation intake and to maintain the control room pressure to the design positive pressure (one-sixteenth inch water) so that all leakage should be out leakage.

High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorber.

The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room. The in-place test results 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 90 percent for expected accident conditions. If the efficiencies of the HEPA filter and charcoal adsorbers are as specified, adequate radiation protection will be provided such that resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10CFR Part 50. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

If the system is found to be operable, there is no immediate threat to the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made. If the makeup system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or refueling operations are terminated.

Pressure drop across the combined HEPA filters and charcoal adsorbers of less than six inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once per operating cycle to show system performance capability. In addition, air intake radiation monitors will be calibrated and functionally tested each operating cycle,.not to exceed 24 months, to verify system performance.

The frequency of tests and sample analysis are necessary to show the HEPA filters and charcoal adsorbers can perform as evaluated. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples, Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 5-1 of ANSI 509-1980. The replacement. charcoal for the adsorber tray removed for the test should meet the same adsorbent quality. Any HEPA filters found defective shall be replaced with filters qualified pursuant to ANSI 509-1980.

AMENDMENT NO. 180

BASES FOR 3.4.5 AND 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM Operation of the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month will demonstrate operability of the filters and adsorber system and remove excessive moisture built up on the adsorber.

If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign materials, the same tests and sample analysis shall be performed as required for operational use. The determination of significant shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination.

AMENDMENT NO. 181

3.5.0 SHUTDOWN AND REFUELING A) GENERAL APPLICABILITY Applies to the neutron instrumentation systems required during shutdown and refueling operations.

B) GENERAL OBJECTIVE LIMITINGCONDITIONS FOR OPERATION - To.define the lowest functional capability or performance level of equipment required during shutdown and refueling operations.

SURVEILLANCE REQUIREMENTS - To define the test or inspections required to assure the functional capability or. performance level of the above items.

AMENDMENT NO. 182

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.1 SOURCE RANGE MONITORS 4.5.1 OURCE RAN E MONIT R Applies to the operating status of the source range Applies to the periodic testing of the source range monitors. monitors.

Qtjec~ive: ~Ob ec ive:

To assure the capability of the source range monitors To assure the operability of the source range to provide neutron flux indication required for reactor monitors to monitor low-level neutron flux.

shutdown and startup and refueling operations.

Whenever the reactor is in the shutdown, refueling or The source range monitoring system surveillance will power operating conditions (unless the IRM's or be performed as indicated below.

APRM's are on scale) or whenever core alterations are being made at least three SRM channels will be Durin each o eratin c le - check in-core to out-of-operable except as noted in Specification 3.5.3. To core signal ratio and minimum count rate.

be considered operable, the following conditions must be satisfied:

a. Inserted to normal operating level and available for monitoring the core. May be withdrawn as long as a minimum count rate of 100 cps is maintained.

AMENDMENT NO. 183

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. A 3/1 in-core to out-of-core signal ratio and a minimum count rate of 3 cps at a keff equivalent to the initial clean core with all rods and poison control curtains inserted.

c. If following a routine surveillance check "a" or "b" is not met, the reactor shall be in the cold shutdown condition within ten hours.

AMENDMENT NO. 184

BASES FOR 3.5.1 AND 4.5.1 SOURCE RANGE MONITORS The SRM's are provided to monitor the core during periods of Station shutdown and to guide the operator during refueling operations and Station startup. Requiring three operative SRM's will ensure adequate coverage for all possible critical configurations produced by fuel loading or dispersed withdrawals of control rods during Station startup. Allowing withdrawal of the SRM while maintaining a high count rate will extend the operating range of the SRM's. Evaluation of the SRM operation is presented in Section VIII-C.1.2.1 of the FSAR.

AMENDMENT NO. 185

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.2 REFUELING PLATFORM INTERLOCK 4.5.2 REF ELING PLATFORM INTERLO K bile:

Applies to the refueling platform on interlocks. Applies to the. periodic testing requirements for the refueling platform interlocks.

~Ob ec ice:

To assure that a loaded refueling platform hoist is To assure the operability of the refueling platform never over the core when one or more control rods interlock.

are withdrawn.

During the refueling condition with the mode switch The refueling platform interlocks shall be tested prior in the "refuel" position the following interlocks must to any fuel handling with the head off the reactor be operative: vessel, at weekly intervals thereafter until no longer required and following any repair work associated

a. Control rod withdrawal block with a fuel with the interlocks.

assembly on the hoist over the reactor core.

b. With a control rod withdrawn from the core the refuel platform, if loaded with a fuel assembly, is blocked from travelling over the core.

c. If the interlocks for either "a" or "b" or both are not operable, double procedural control will be used to ensure that "a" and "b" are met.

AMENDMENT NO. 186

e BASES FOR 3.5.2 AND 4.5.2 REFUELING PLATFORM INTERLOCK The control rod withdrawal block and refueling platform travel blocks are provided to back up normal procedural controls to prevent

~

inadvertent large reactivity additions to the core. These interlocks are provided even though no more than one control rod can be removed from the core at a time during refueling with the mode switch in the "refuel" position. Even in the fresh fully loaded core if a new assembly is dropped into a vacant position adjacent to the withdrawn rod, no excursion would result. This is discussed in detail in Appendix E-ll.3.0 of the FSAR.

There are normally two Station personnel directly involved in refueling the reactor, one in the control room and one at the platform. If the interlocks are inoperable,.an additional person will check that "a" and "b" are not violated.

AMENDMENT NO. 187

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.3 EXTENDED CORE AND CONTROL R D DRIVE 4.5.3 ~

EXTENDED RE AND NTROL R D DRIVE MAINTENANCE MAINTENANCE Applies to core reactivity limitations during major core Applies to monitoring during major core alterations.

alterations.

~Ob ec ive: ~Ob ective:

To assure that inadvertent criticality does not result To assure that inadvertent withdrawal of an incorrect when control rods are being removed from the core. control rod does not occur.

Whenever, the reactor is in the refueling condition, Whenever the reactor is in the refuel mode and rod control rods may be withdrawn from the reactor core block interlocks are being bypassed for core provided the following conditions are satisfied: unloading, one licensed operator and a member of the reactor analysis staff will verify that all the fuel from

a. The reactor mode switch shall be locked in the the cell has been removed before the corresponding "Refuel" position. The refueling interlock input control rod is withdrawn.

signal from a withdrawn control rod may be bypassed on a withdrawn control rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core. All other refueling interlocks shall be operable, except those necessary to pull the next control rods.

AMENDMENT NO. 188

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. During core alterations two SRM's shall be operable, one in and one adjacent to any core

.quadrant where fuel or control rods are being moved. Operable SRM's shall have a minimum of 3 counts per second except as specified in d and e below.

c. The SRM's shall be inserted to the normal operating level. Use of special movable dunking type detectors during major core alterations is permissible as long as detector is connected into the normal SRM circuit.

d. Prior to spiral unloading, the SRM's shall have an initial count rate of 3 cps. During spiral unloading, the count rate on the SRM's may drop below 3 cps.

e. During spiral reload, SRM operability will be verified by using a portable external source every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the required amount of fuel is loaded to maintain 3 cps. As an alternative to the above, two fuel assemblies will be loaded in different cells containing control blades around each SRM to obtain the required 3 cps. Until these two assemblies have been loaded, the 3 cps requirement is not necessary.

AMENDMENT NO. 189

BASES FOR 3.5.3 EXTENDED CORE AND CONTROL ROD DRIVE IVIAINTENANCE The intent of this specification is to permit the unloading of a significant portion of the reactor core for such purposes as removal of temporary control curtains, control rod drive maintenance, in-service inspection requirements, examination of the core support plate, etc.

When the refueling interlock input signal from a withdrawn control rod is bypassed, administrative controls will be in effect to prohibit fuel from being loaded into that control cell.

These operations are performed with the mode switch in the "Refuel" position to provide the refueling interlocks normally available during refueling. In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod.

The requirement that the fuel assemblies in the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed insures that withdrawal of another control rod does not result in inadvertent criticality. Each control rod essentially provides reactivity control for the fuel assemblies in the cell associated with the control rod. Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.

The SRM's are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup. Requiring two operable SRM's, one in and one adjacent to any core quadrant where fuel or control rods are being moved, assures adequate monitoring of that quadrant during such alterations. The requirement of 3 counts per second provides assurance that neutron flux is being monitored.

A spiral unloading pattern is one by which the fuel in the outermost cells (four fuel bundles surrounding a control blade) is removed first.

Unloading continues by removing the remaining outermost fuel by cell. The last cell removed will be adjacent to a SRM. Spiral reloading is the reverse of unloading. Spiral unloading and reloading will preclude the creation of flux traps (moderator filled or partially filled cells surrounded on all sides by fuel).

During spiral unloading, the SRM's shall have an initial count rate of 3 cps with all rods fully inserted. The count rate will diminish during fuel removal. After all the fuel is removed from a cell and after withdrawing the corresponding control rod, the refueling interlock will be bypassed on that rod. After withdrawal of that rod, one licensed operator and a member of the reactor analysis staff will verify that the interlock bypassed is on the correct control rod. Once the control rod is withdrawn, it will be valved out of service.

Under this special condition of complete spiral core unloading, it is expected that the count rate of the SRM's will drop below 3 cps before all of the fuel is unloaded. Since there will be no reactivity additions, a lower number of counts will not present a hazard. When all of the fuel has been removed to the spent fuel storage pool, the SRM's will no longer be required. Requiring the SRM's to be operational prior to fuel removal assures that the SRM's are operable and can be relied on even when the count rate may go below 3 cps.

During spiral reload, SRM operability will be verified by using a portable external source every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the required amount of fuel is loaded to maintain 3 cps. As an alternative to the above, two fuel assemblies will be loaded in different cells containing control blades around each SRM to obtain the required 3 cps. Until these two assemblies have been loaded, the 3 cps requirement is not necessary.

AMENDMENT NO. 190

3.6.0 GENERAL REACTOR PLANT A) GENERAL APPLICABILITY Applies to Station process effluents, reactor protection system and emergency power sources.

B) GENERAL OBJECTIVE LIMITINGCONDITIONS FOR OPERATION - To define the lowest functional capability or performance level of the equipment to assure overall Station safety.

SURVEILLANCE REQUIREMENTS - To define the test or inspection required to assure the functional capability or performance level of this equipment.

AMENDMENT NO. 191

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.1 TATI N PROCE EFFLUENT 4.6.1 STATION PRO ES EFFL ENT

a. Effluent release limits are described in Specifi- a. Monitoring the radioactive discharges from Nine cation 3.6.15. Mile Point Unit 1 is described in Specification r 4.6.1 5.

b. The mechanical vacuum pump line shall be capable of automatic isolation by closure of the b. At least once during each operating cycle (prior air-operated valve upstream of the pumps. The to startup), verify automatic securing and signal to initiate isolation shall be from high isolation of the mechanical vacuum pump.

radioactivity (five times normal) in the main.

steam line.

AMENDMENT NO. 192

0 BASES FOR 3.6.1b AND 4.6.1b IVIECHANICALVACUUM PUMP ISOLATION The purpose of isolating the mechanical vacuum pump line is to limit release of activity from the main condenser during a control rod drop accident. During the accident, fission products would be transported from the reactor through the main-steam lines to the main condenser.

The fission product radioactivity would be sensed by the main-steam line radioactivity monitors and initiate isolation.

AMENDMENT NO. 193

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.2 PROTECTIVE INSTRUMENTATION 4.6.2 PROTECTIVE INSTR MENTATI N Applies to the operability of the plant instrumentation Applies to the surveillance of the instrumentation that that performs a safety function. performs a safety function.

~ob ec ive: ~bee ive:

To assure the operability of the instrumentation To verify the operability of protective required for safe operation. instrumentation.

a. The set points, minimum number of trip systems, a. Sensors and instrument channels shall be and minimum number of instrument channels checked, tested and calibrated at least as that must be operable for each position of the frequently as listed in Tables 4.6.2a to 4.6.21.

reactor mode switch shall be as given in Tables 3.6.2a to 3.6.2I.

If the requirements of a table are not met, the actions listed below for the respective type of instrumentation shall be taken.

(1) Instrumentation that initiates scram - control rods shall be inserted, unless there is no fuel in the reactor vessel.

AMENDMENT NO. 194

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) Primary Coolant and Containment Isolation- b. Each trip system shall be tested each time the Isolation valves shall be closed or the valves respective instrument channel is tested.

shall be considered inoperable and Specifica-tions 3.2.7 and 3.3.4 shall be applied. c. At least daily during reactor power operation, the core power distribution shall be checked for (3) Emergency Cooling Initiation or Isolation- Maximum Total Peaking Factor (MTPF) and the The emergency cooling system shall be flow-referenced APRM scram and rod block considered inoperable and Specification signals shall be adjusted, if necessary, as 3.1.3 shall be applied. specified in Figure 2.1.1.

(4) Core Spray Initiation - The core spray system shall be considered inoperable and Specification 3.1.4 shall be applied.

(5) Containment Spray Initiation - The contain-ment spray system shall be considered inoperable and Specification 3.3.7 shall be applied.

(6) Auto Depressurization Initiation - The auto depressurization system shall be considered inoperable and Specification 3.1.5 shall be applied.

(7) Control Rod Withdrawal Block - No control rods shall be withdrawn.

AMENDMENT NO. 195

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (8) Off-Gas and Vacuum Pump Isolation - The respective system shall be isolated or the instrument channel shall be considered

=inoperable and Specification 3.6.1 shall be applied.

(9) Diesel Generator Initiation - The diesel generator shall be considered inoperable and Specification 3.6.3 shall be applied.

(10) Emergency Ventilation Initiation - The emergency ventilation system shall be considered inoperable and Specification 3 4.4 shall be applied.

(11) High Pressure Coolant Injection Initiation-The high pressure coolant injection system shall be considered inoperable and Specifi-cation 3.1.8.c shall be applied.

(12) Control Room Ventilation - The control room ventilation system shall be considered inoperable and Specification 3.4.5 shall be applied.

During operation with a Maximum Total Peaking Factor (MTPF) greater than the design value, either:

AMENDMENT NO. 196

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (1) The APRM scram and rod block settings shall be reduced to the values given by the equations in Specification 2.1.2.a; or (2) The power distribution shall be changed such that the MTPF no longer exceeds the design value.

AMENDMENT NO. 197

TABLE 3.6.2a INSTRUMENTATIONTHAT INITIATES SCRAM Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor IVlode Switch IVlinimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S s ems Tri S stem ~Se Pain erabl CL M

Manual Scram x x (2) High Reactor Pressure 2(o) ~ 1080 psig (p) x x (3) High Drywell Pressure 2(o) < 3.5 psig x (a) (a)

(4) Low Reactor Water Level 2(o) ~ 53 inches x - x (Indicator Scale)

(5) High Water Level Scram 2(o) < 45 gal. (b) x x Discharge Volume AMENDMENT NO. 198

TABLE 3.6.2a (cont'd)

INSTRUMENTATIONTHAT INITIATES SCRAM Limitin Condition for 0 era ion IVlinimum No. of Operable Instrument Reactor IV!ode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S em ~Se Pain er I C

O V)

(6) Main-Steam-Line Isolation Valve 4(h)(o) ( 10 percent valve (c) (c) x Position closure from full open High Radiation Main-Steam-Line 2(o) ~ 5 times normal x x background at rated power Shutdown Position of Reactor (k) x x Mode Switch (9) Neutron Flux (a) IRM (i) Upscale 3(d)(o) ~ 96 percent of (g) (g) (g) full scale AMENDMENT NO. 199

I, TABLE 3.6.2a (cont'd)

INSTRUMENTATIONTHAT INITIATES SCRAIVI Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor IVlode Switch IVlinimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem ~se Pain 0 erable C

O D

V)

(ii) Inoperative 3(d)(o)

(b) APRM (i) Upscale 3(e)(o) Figure 2.1.1 X X (ii) Inoperative 3(e)(o) X X (iii) Downscale 3(e)(o) a 5 percent of (g) (g) (g) full scale (10) Turbine Stop Valve Closure 4(o) a 10% valve closure (11) Generator Load Rejection 2(o)

AMENDMENT NO. 200

TABLE 4.6.2a INSTRUMENTATIONTHAT INITIATES SCRAM Surveillance Re uiremen Instrument Instrument Channel Parameter Sensor Check Channel Tes gali~br Lion (1) Manual Scram None Once per week None (2) High Reactor Pressure None Once per 3 months( ) Once per 3 months "

(3) High Drywall Pressure None Once per 3 months(") Once per 3 months(

(4) Low Reactor. Water Once/day Once per 3 months( ) Once per 3 months( )

Level (5) High Water Level None Once per 3 months Once per 3 months Scram Discharge Volume (6) Main-Steam-Line None Once per 3 months Once per operating cycle Isolation Valve Position (7) High Radiation Main- Once/shift Once per 3 months Once per 3 months Steam Line AMENDMENT NO. 201

l TABLE 4.6.2a (cont'd)

INSTRUMENTATIONTHAT INITIATES SCRAM Surveillance Re uiremen Instrument Instrument Channel

~Parame er Sensor Check Channel Tes ~li r~aion (8) Shutdown Position of None Once during each major None Reactor Mode Switch refueling outage (9) Neutron Flux (a) IRM (i) Upscale (ii) Inoperative (b) APRM (i) Upscale None Once per 3 months Once per week Once per 3 months (ii) Inoperative None Once per 3 months None (iii) Downscale None Once per 3 months Once per week' Once per 3 months (10) Turbine Stop Valve None Once per 3 months Once per operating cycle Closure (11) Generator Load None Once per 3 months Once per 3 months Rejection AMENDMENT NO. 202

NOTES FOR TABLES 3.6.2a and 4.6.2a (a) May be bypassed when necessary for containment inerting.

(b) May be bypassed in the refuel and shutdown positions of the reactor mode switch with a keylock switch.

(c) May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi, or for the purpose of performing reactor coolant system pressure testing and/or control rod scram time testing with the reactor mode switch in the refuel position.

(d) No more than one of the four IRM inputs to each trip system shall be bypassed.

(e) No more than two C or D level LPRM inputs to an APRM shall be bypassed and only four LPRM inputs to an APRM shall be bypassed in order for the APRM to be considered operable. No more than one of the four APRM inputs to each trip system shall be bypassed provided that the APRM in the other instrument channel in the same core quadrant is not bypassed. A Travelling In-Core Probe (TIP) chamber may be used as a substitute APRM input if the TIP is positioned in close proximity to the failed LPRM it is replacing.

Calibrate prior to starting and normal shutdown and thereafter check once per shift and test once per week until no longer required.

(g) IRM's are bypassed when APRM's are onscale. APRM downscale-is bypassed when IRM's are onscale.

(h) Each of the four isolation valves has two limit switches. Each. limit switch provides input to one of two instrument channels in a single trip system.

May be bypassed when reactor power level is below 45%.

Trip upon loss of oil pressure to the acceleration relay.

(k) May be bypassed when placing the reactor mode switch in the SHUTDOWN position and all control rods are fully inserted.

~ ~

Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2a, the primary sensor will be calibrated and tested once per operating cycle.

(m) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during reactor operation when THERMAL POWER a 25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Figure 2.1.1 shall not be included in determining the absolute difference.

AMENDMENT NO. 203

NOTES FOR TABLES 3.6.2a and 4.6.2a (n) Deleted.

(o) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same trip system is monitoring that parameter.

With one channel required by Table 3.6.2a inoperable in one or more Parameters, place the inoperable channel and/or that trip system in the tripped condition" within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With two or more channels required by Table 3.6.2a inoperable in one or more Parameters:

1. Within one hour, verify sufficient channels remain Operable or tripped" to maintain trip capability for the Parameter, and

2. Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, place the inoperable channel(s) in one trip system and/or that trip system"" in the tripped condition", and

3. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, restore the inoperable channels in the other trip system to an Operable status or tripped".

Otherwise, take the ACTION required by Specification 3.6.2a for that Parameter.

An inoperable channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, if the inoperable channel is not restored to Operable status within the required time, the ACTION required by Specification 3.6.2a for the parameter shall be taken.

This ACTION applies to that trip system with the most inoperable channels; if both trip systems have the same number of inoperable channels, the ACTION can be applied to either trip system.

(p) May be bypassed during reactor coolant system pressure testing and/or control rod scram time testing.

AMENDMENT NO. 204

TABLE 3.6.2b INSTRUIVIENTATIONTHAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Limi in Condi ion for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be

~Parame er 0 erable Tri S stems 'Tri S stem ~Se Pain 0 arable C

O CL D

8 M tD PRIMARY COOLANT ISOLATION (Main Steam, Cleanup, and Shutdown)

(1) Low-Low Reactor Water Level 2(f) ) 5 inches X X (Indicator Scale)

Manual X X (2)

MAIN- TEAM-LI'NE ISOLATION (3) High Steam Flow Main-Steam Line 2(f) a 105 psid X X AMENDMENT NO. 205

TABLE 3.6.2b (cont'd)

INSTRUIVIENTATIONTHAT INITIATES PRIMARY COOLANT SYSTEM OR CONTAINIVIENTISOLATION Limitin .'Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch IVlinimum No. Channels per Position in Which of Tripped or Operable Function IVlust Be Parameter 0 erable Tri S stems Tri S sem ~ePoin rable C

O CL V

D I tD IL N (4) High Radiation Main Steam Line 2(f) ( 5 times X X normal background at rated power (5) Low Reactor Pressure 2(f) ) 850 psig (6) Low-Low-Low Condenser Vacuum 2(f) a 7 in. mercury (a) x vacuum (7) High Temperature Main Steam Line 2(f) < 200'F X X Tunnel AMENDMENT NO. 206

TABLE 3.6.2b (cont'd)

INSTRUMENTATIONTHAT INITIATES PRIIVIARY COOLANT SYSTEM OR CONTAINIVIENTISOLATION Limi in Condi ion for eration IVlinimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S s ems Tri S sem ~eP~oin erable C

O D.

D V) M CLEANUP SYSTEM ISOLATION (8) High Area Temperature 2(g) a 190'F X X X X SHUTD WN COOLING SYSTEM ISOLATION (9) High Area Temperature ~ 170'F X X CONTAINMENT ISOLATION (10) Low-Low Reactor Water 2(f) a 5 inches (c) X X (Indicator Scale)

AMENDMENT NO. 207

TABLE 3.6.2b (cont'd)

INSTRUMENTATIONTHAT INITIATES PRIIVIARY COOLANT SYSTEM OR CONTAINIVIENTISOLATION Limi in Condition for 0 era ion Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be

~Parame er 0 erable Tri S stems Tri S stem ~Se Pain rabl C

O a L

C$

CO PD (11) High Drywell Pressure 2(f) a 3.5 psig (c) (b) (b)

(12) Manual X '

AMENDMENT NO. 208

TABLE 4.6.2b INSTR UIVIENTATIONTHAT INITIATES PRIMARY COOLANT SYSTEIVI OR CONTAINMENT ISOLATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration PRIMARY COOLANT ISOLATION (Main Steam, Cleanup and Shutdown)

(1) Low-Low Reactor Once/day Once per 3 months(") Once per 3 months(d)

Water Level (2) Manual Once during each major refueling outage MAIN-STEAM-LINE ISOLATION (3) High Steam Flow Once/day Once per 3 months Once per 3 months Main-Steam Line (4) High Radiation Main- Once/shift Once per 3 months Once per 3 months Steam Line (5) Low Reactor Once/day Once per 3 months( ) Once per 3 months>> )

Pressure AMENDMENT NO. 209

TABLE 4.6.2b (cont'd)

INSTRUMENTATIONTHAT INITIATES PRIIVIARY COOLANT SYSTEM OR CONTAINMENT ISOLATION Surveillance Re uiremen Instrument Instrument Channel Parameter Sensor Check Channel Tes Calibration (6) Low-Low-Low None Once during each major Once during each major Condenser Vacuum refueling outage refueling outage (7) High Temperature None Once during each major Once during each major Main-Steam-Line refueling outage refueling outage Tunnel CLEANUP SYSTEM ISOLATION (8) High Area Once/week Once during each major Once during each major Temperature refueling outage refueling outage SHUTDOWN COOLING SYSTEM ISOLATION (9) High Area Once/week Once during each major Once during each major Temperature refueling outage refueling outage AMENDMENT NO. 210

TABLE 4.6.2b (cont'd)

INSTRUMENTATION THAT INITIATES PRIIVIARY COOLANT SYSTEIVI OR CONTAINMENT ISOLATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration CONTAINMENT ISOLATION (10) Low-Low Reactor Once/day Once per 3 months Once per 3 months Water Level (11) High Drywell Once/day Once per 3 months(d) Once per 3 months Pressure (12) Manual Once during each operating cycle AMENDMENT NO.

NOTES FOR TABLES 3.6.2b and 4.6.2b (a) May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi.

(b) May be bypassed when necessary for containment inerting.

(c) May be bypassed in the shutdown mode whenever the reactor coolant system temperature is less than 215'F.

(d) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2b, the primary sensor will be calibrated and tested once per operating cycle.

(e) Deleted.

(f) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that Parameter.

With the number of Operable Channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for one trip system, either

1. Place the inoperable channel(s) in the tripped condition within

a. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Parameters common to SCRAM Instrumentation, and

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Parameters not common to SCRAM Instrumentation.

of

2. Take the ACTION required by Specification 3.6.2a for that Parameter.

With the number of Operable Channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for both trip systems,

1. Place the inoperable channel(s) in one trip system in the tripped condition within one hour.

'and

2. a. Place the inoperable channel(s) in the remaining trip system in the tripped condition within (1) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Parameters common to SCRAM Instrumentation, and (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Parameters not common to SCRAM Instrumentation.

or

b. take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 212

NOTES FOR TABLES 3.6.2b and 4.6.2b (g) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that Parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels for the Operable Trip System, either

1. Place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

or

2. Take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 213

TABLE 3.6.2c INSTRUMENTATIONTHAT INITIATES OR ISOLATES EMERGENCY COOLING Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem d Set Point 0 erable C

0 a o L Cl M M EMERGENCY COOLING INITIATION 2(e) ~ 1080 psig (b) X X (1) High-Reactor Pressure (2) Low-Low Reactor Water Level 2(e) 5 inches (b) X X (Indicator Scale)

EMER ENCY COOLIN ISOLATION (for each of two systems)

(3) High Steam Flow Emergency 2(a)(f) x 11.5 psid X X Cooling System AMENDMENT NO. 214

TABLE 4.6.2c INSTRUIVlENTATIONTHAT INITIATES OR ISOLATES EMERGENCY COOLING Surveillance Re uirement

=

Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration EMERGENCY COOLING INITIATION (1) High Reactor None Once per 3 months Once per 3 months Pressure (2) Low-Low Reactor Once/day Once per 3 months Once per 3 months Water Level EMERGENCY COOLING ISOLATION (for each of. two systems)

(3) High Steam Flow None Once per 3 months Once per 3 months Emergency Cooling System AMENDMENT NO. 215

0 NOTES FOR TABLES 3.6.2c AND 4.6.2c (a) Each of two differential pressure switches provide inputs to one instrument channel in each trip system.

(b) May be bypassed in the cold shutdown condition.

(c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2c, the primary sensor will be calibrated and tested once per operating cycle.

(d) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

(e) With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:-

1. For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the action required by Specification 3.6.2a for that Parameter.

2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

(f) With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for one trip system, either

1. Place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

of

2. Take the ACTION required by Specification 3.6.2a for that Parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for both trip systems,

1. Place the inoperable channel(s) in one trip system in the tripped condition within one hour and

2. a. Place the. inoperable channel(s) in the remaining trip system in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

or

b. Take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 216

TABLE 3.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem f Set Point 0 erable CL CO M

START CORE SPRAY PUMPS (1) High Drywell Pressure ~ 3.5 psig (d) x (a) (a)

(2) =

Low-Low Reactor Water Level a 5 inches (b) x x x (Indicator Scale)

PEN ORE SPRAY DISCHARGE VALVES (3) Reactor Pressure and either (1) or a 365 psig x x (2) above.

AMENDMENT NO. 217

TABLE 4.6.2d INSTRUMENTATION THAT INITIATES CORE SPRAY Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration START CORE SPRAY PUMPS (1) High Drywell Once/day Once per 3 months( ) Once per 3 months Pressure (2) Low-Low Reactor Once/day Once per 3 months Once per 3 months Water Level OPEN CORE SPRAY DISCHARGE VALVES (3) Reactor Pressure None Once per 3 months( ) Once per 3 months and either (1) or (2) above AMENDMENT NO. 218

NOTES FOR TABLES 3.6.2d AND 4.6.2d (a) May be bypassed when necessary for containment inerting.

(b) May be bypassed when necessary for performing major maintenance as specified in Specification 2.1.1.e.

(c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2d, the primary sensor will be calibrated and tested once per operating cycle.

(d) May be bypassed when necessary for integrated leak rate testing.

(e) The instrumentation that initiates the Core Spray System is not required to be operable, if there is no fuel in the reactor vessel.

(f) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the ACTION required by Specification 3.6.2a for that Parameter.

2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 219

0 TABLE 3.6.2e INSTRUIVIENTATIONTHAT INITIATES CONTAINMENT SPRAY Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem c ~Set Poin 0 erable C

O M

(1) a. High Drywell Pressure < 3.5 psig (a) X X and

b. Low-Low Reactor Water a 5 inches (a) X X Level (Indicator Scale)

AMENDMENT NO. 220

TABLE 4.6.2e INSTRUMENTATION THAT INITIATES CONTAINMENT SPRAY Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration I Once per 3 monthsl I (1) a. High Drywell Once/day Once per 3 monthsI Pressure

b. Low-Low Once/day Once per 3 months " Once per 3 monthsI I Reactor Water Level AMENDMENT NO. 221

NOTES FOR TABLES 3.6.2e AND 4.6.2e (a) May be bypassed in the shutdown mode whenever the reactor coolant temperature is less than 215'F.

(b) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2e, the primary sensor will be calibrated and tested once per operating cycle.

(c) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip system in the tripped condition provided at least one Operable Instrument Channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the ACTION required by Specification 3.6.2a for that Parameter.

2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 222

TABLE 3.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSURIZATION Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem d Set Point 0 erable t:

CL D

I CO M (0 INITIATION (1) a. Low-Low-Low Reactor Water 2(a) 2(a) ~ -10 inches " (b) (b) x Level (Indicator Scale) and

b. High Drywell Pressure 2(a) 2(a) a 3.5 psig (b) (b) x

" greater than (a) means less negative AMENDMENT NO. 223

TABLE 4.6.2f INSTRUMENTATION THAT INITIATES AUTO DEPRESSURIZATION Surveillance Re uiremen Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration INITIATION (1) a. Low-Low-Low None Once per 3 months Once per 3 months Reactor Water and

b. High Drywall Pressure Once/day Once per 3 months Once per 3 months AMENDMENT NO. 224

NOTES FOR TABLES 3.6.2f AND 4.6.2f (a) Both instrument channels in either trip system are required to be energized to initiate auto depressurization. One trip system is powered from power board 102 and the other trip system from power board 103.

(b) May be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation temperature.

(c) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2f,'the primary sensor will be calibrated and tested once per operating cycle.

(d) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one operable channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the ACTION required by Specification 3.6.2a for that Parameter.

2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 225

TABLE 3.6.2g INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWALBLOCK Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor IVlode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem i Set Point 0 erable C

O CL o D L

fD N (1) SRM

a. Detector not in Startup Position 2 2(a)(e)

b. Inoperative 2(a)

c. Upscale 2 2(a) ~ 10 counts/sec (2) IRM

a. Detector not in Startup Position 2 3(b)

b. Inoperative 3(b)

AMENDMENT NO. 226

TABLE 3.6.2g (cont'd)

INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWALBLOCK Limitin Condition for 0 eration IVIinimum No. of Operable Instrument Reactor Mode Switch IVlinimum No. Channels per Position in Which of Tripped or Operable Function IVlust Be Parameter 0 erable Tri S stems Tri S stem i S~et Poin 0 erable C

0 CL D

M U)

c. Downscale 3(b) ~ 5 percent of full X X scale for each scale

d. Upscale 3(b) ~ 88 percent of full X X scale for each scale (3) APRM

a. Inoperative 2(h) 3(c) X X

b. Upscale (Biased by 2(h) 3(c) Figure 2.1.1(h) X X X Recirculation Flow)

c. Downscale 2(h) 3(c) a 2 percent of (d) (d) x full scale AMENDMENT NO. 227

TABLE 3.6.2g (cont'd)

INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWALBLOCK Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function lVlust Be Parameter 0 erable Tri S stems Tri S stem ~Set Poin 0 erable C

O aD iD V

M N (4) Recirculation Flow

a. Comparator Off Normal (6.8%

b. Flow Unit Inoperative

c. Flow Unit Upscale s 103.7%

(5) Refuel Platform and Hoists -2(f)

(6) Mode Switch in Shutdown AMENDMENT NO. 228

TABLE 3.6.2g (cont'd)

INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWALBLOCK Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem Set Point 0 erable C

0

'D CL D

(D V)

(7) Mode Switch in Refuel (Blocks withdrawal of more than 1 rod)

(8) Scram Dump Volume Water Level Scram Bypass AMENDMENT NO. 229

I TABLE 4.6.2g INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWALBLOCK Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1) SRM

a. Detector Not in N/A (g) N/A Startup Position

b. Inoperative N/A (g) N/A

c. Upscale N/A (g) (g)

(2) IRM

a. Detector not in N/A (g) N/A Startup Position

b. Inoperative N/A (g) N/A

c. Downscale N/A (g) (g)

d. Upscale N/A (g) (g)

AMENDMENT NO. 230

TABLE 4.6.2g (cont'd)

INSTRUMENTATIONTHAT INITIATES CONTROL ROD WITHDRAWALBLOCK Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Tes Calibration (3) APRM

a. Inoperative . None Once per 3 months None

b. Upscale (Biased by None Once per 3 months Once per 3 months Recirculation Flow)

c. Downscale None Once per 3 months Once per 3 months (4) Recirculation Flow

a. Comparator Off None Once per 3 months Once per 3 months Normal

b. Flow Unit None Once per 3 months Once per 3 months Inoperative

c. Flow Unit Upscale None Once per 3 months Once per 3 months AMENDMENT NO. 231

TABLE 4.6.2g (cont'd)

INSTRUMENTATION THAT INITIATES CONTROL ROD WITHDRAWALBLOCK Surveillance Re uiremen Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (5) Refuel Platform and (see 4.5.2)

Hoists (6) Mode Switch in Once during each major Shutdown refueling outage (7) Mode Switch in Once during each major Refuel (Blocks refueling outage withdrawal of more than 1 rod)

(8) Scram Dump Once during each major Volume Water Level refueling outage Scram Bypass AMENDMENT NO. 232

0 NOTES FOR TABLES 3.6.2g and 4.6.2g (a) No more than one of the four SRM inputs to the single trip system shall be bypassed.

(b) No more than one of the four IRM inputs to each instrument channel shall be bypassed. These signals may be bypassed when the APRMs are onscale.

(c) No more than one of the four APRM inputs to each instrument channel shall be bypassed provided that the APRM in the other instrument channel in the same core quadrant is not bypassed. No more than two C or D level LPRM inputs to an APRM shall be bypassed and only four LPRM inputs to only one APRM shall be bypassed in order for the APRM to be considered operable. In the Run mode of operation, bypass of two chambers from one radial core location in any one APRM shall cause that APRM to be considered inoperative. A Travelling In-Core Probe (TIP) chamber may be used as a substitute APRM input if the TIP is positioned in close proximity to the failed LPRM it is replacing. If one APRM in a quadrant is bypassed and meets all requirements for operability with the exception of the requirement of at least one operable chamber at each radial location, it may be returned to service and the other APRM in that quadrant may be removed from service for test and/or calibration only if no control rod is withdrawn during the calibration and/or test.

(d) May be bypassed in the startup and refuel positions of the reactor mode switch when the IRMs are onscale.

(e) This function may be bypassed when the count rate is ~ 100 cps.

(f) One sensor provides input to each of two instrument channels. Each instrument channel is in a separate trip system.

(g) Calibrate and/or test prior to startup and normal shutdown. Thereafter, test once per week until no longer required.

(h) The actuation of either or both trip systems will result in a rod block.

(i) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the Trip System in the tripped condition, provided at least one other operable channel in the same Trip System is monitoring that Parameter.

~ j ~

AMENDMENT NO. 233

TABLE 3.6.2h VACUUM PUMP ISOLATION Limitin Condition for 0 eration IVlinimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem b ~Se Point 0 erable C

CL D

CO M MECHANICALVACUUM PUMP High Radiation Main Steam Line ~ 5 times normal X . X background AMENDMENT NO. 234

TABLE 4.6.2h VACUUM PUMP ISOLATION Surveillance Re uiremen Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration MECHANICAL VACUUM PUMP High Radiation Main Steam Once/shift Once per 3 months Once per 3 months Line AMENDMENT NO. 235

NOTES FOR TABLES 3.6.2h and 4.6.2h (a) Deleted.

(b) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the ~

tripped condition provided at least one operable channel in the same Trip System is monitoring that parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement for one trip system, either

1. Place the inoperable channel(s) in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Of

2. Take the ACTION required by Specification 3.6.2a for that Parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument'Channels per Operable Trip System requirement for both trip systems,

1. Place the inoperable channel(s) in one trip system in the tripped condition within one hour.

and

2. a. Place the inoperable channel(s) in the remaining trip system in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

or

b. Take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 236

0 TABLE 3.6.2i DIESEL GENERATOR INITIATION Limitin Condition for 0 eration Reactor IVlode Switch Minimum Position in Which Total No. Channels(1) Channels Function Must Be Parameter of Channels To Trl 0 erable c 0 erable C

O Q.

O D U) M Loss of Power

a. 4.16kV PB 102/103 Emergency Bus Undervoltage (Loss of Voltage) 3 per Bus 2 per Bus 2 per Bus X X

b. 4.16kV PB 102/103 Emergency Bus Undervoltage (Degraded Voltage) 3 per Bus 2 per Bus 2 per Bus X X X (1) -

If one out of three channels becomes inoperable, the inoperable channel will be placed in the trip condition.

AMENDMENT NO. 237

TABLE 3.6.2i (cont'd)

DIESEL GENERATOR INITIATION Limitin Condition for 0 eration Parameter Set Point Inverse Time Undervolta e Rela s Loss of Power Rela Dro out O eratin Time

a. 4.16kV PB 102/103 Emergency Bus ) 3200 volts 0 volts (3.2 seconds Undervolt (Loss of Voltage)

b. 4.16kV PB 102/103 Emergency Bus ~3600 volts 3580 volts 18.5 s 3 seconds Undervoltage (Degraded Voltage)

(a) The operating time indicated in the table is'the time required for the relay to operate its contacts when the voltage is suddenly decreased from operating voltage level values to the voltage level listed in the table above.

AMENDMENT NO. 238

TABLE 4.6.2i DIESEL GENERATOR INITIATION Surveillance Re uirements tnstrument instrument Channel Parameter Sensor Check Channel Test Calibration Loss of Power 80 4.16l<V PB 102/103 Emergency Bus Undervoltage (Loss of Voltage) NA Once per month Once per refueling cycle 4.16l<V PB 102/103 Emergency Bus Undervoltage (Degraded Voltage) NA Once per month Once per refueling cycle (a) The instrument channel test demonstrate the operability of the instrument channel by simulating an undervoltage condition to verify that the tripping logic functions properly.

(b) The instrument channel calibration will demonstrate the operability of the instrument channel by simulating an undervoltage condition to verify that the tripping logic functions properly. In addition, a sensor calibration will be performed to verify the set points listed in Table 3.6.2.i.

(c) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one operable channel in the same Trip System is monitoring that parameter.

AMENDMENT NO. 239

1 TABLE 3.6.2j ElVIERGENCY VENTILATIONINITIATION Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem Set Point 0 erable C

0 D.

I tO V) CO (1) High Radiation Reactor Building 2(d) ( 5mr/hr X X Ventilation Duct (2) High Radiation Refueling Platform a 1000mr/hr (a) (a) (a) (a)

AMENDMENT NO. 240

TABLE 4.6.2j EMERGENCy'ENTILATION INITIATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Test Calibration (1) High Radiation -Once/shift Once during each operating cycle Once per quarter Reactor Building Ventilation Duct (2) High Radiation (b) (c) Once per quarter Refueling Platform AMENDMENT NO.. 241

NOTES FOR TABLES 3.6.2j AND 4.6.2j (a) This function shall be operable any time that irradiated fuel or the irradiated fuel cask is being handled in the reactor building.

(b) Once per shift whenever this function is required to be operable.

(c) Immediately prior to when function is required and once per week thereafter until function is no longer required.

(d) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one Operable Instrument Channel in the same Trip system is monitoring that parameter.

With the number of Operable channels one less than required by the Minimum Number of Operable Instrument Channels for the Operable Trip System, either

1) Place the inoperable channel(s) in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

or

2) Take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 242

TABLE 3.6.2k HIGH PRESSURE COOLANT INJECTION Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem ~set Pain 0 erable C

O U)

(1) Low Reactor Water Level 2(c) ~ 53 inches (a) (a) x (Indicator Scale)

(2) Automatic Turbine Trip (a) (a) x AMENDMENT NO. 243

TABLE 4.6.2k HIGH PRESSURE COOLANT INJECTION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Channel Tes Calibration (1) Low Reactor Water Once per day Once per 3 months( ) Once per 3 months "

Level (2) Automatic Turbine None Once during each operating cycle None Trip AMENDMENT NO. 244

NOTES FOR TABLES 3.6.2k AND 4.6.2k (a) May be bypassed when the reactor pressure is less than 110 psig and the reactor coolant temperature is less than the corresponding saturation temperature.

(b) Only the trip circuit will be calibrated and tested at the frequencies specified in Table 4.6.2k, the primary sensor will be calibrated and tested once per operating cycle.

(c) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillances without placing the Trip System in the tripped condition provided at least one operable channel in the same Trip System is monitoring that parameter.

With the number of Operable channels less than required by the Minimum Number of Operable Instrument Channels per Operable Trip System requirement:

1. For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or take the ACTION required by Specification 3.6.2a for that Parameter.

2. With more than one channel inoperable, take the ACTION required by Specification 3.6.2a for that Parameter.

AMENDMENT NO. 245

TABLE 3.6.2I CONTROL ROOIVI AIR TREATMENT SYSTEIVI INITIATION Limitin Condition for 0 eration Minimum No. of Operable Instrument Reactor Mode Switch Minimum No. Channels per Position in Which of Tripped or Operable Function Must Be Parameter 0 erable Tri S stems Tri S stem ~Set Pain 0 erable O

'O C)

I v C D

M K U) K (1) High Radiation Ventilation Intake ~1000 CPM X X X AMENDMENT NO. 246

L TABLE 4.6.21 CONTROL ROOM AIR TREATIVIENT SYSTEIVI INITIATION Surveillance Re uirement Instrument Instrument Channel Parameter Sensor Check Calibration (1) High Radiation Once/shift Once per quarter Once each operating cycle not Ventilation Intake to exceed 24 months AMENDMENT NO. 247

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to prevent exceeding established limits. In addition, other protective instrumentation is provided to initiate action which mitigates the consequences of accidents or terminates operator error.

The reactor protection system is a dual channel type (Table 3.6.2.a). Each trip system except the manual scram has two independent instrument channels. Operation of either channel will trip the trip system, i.e., the trip logic of the channel is one-out-of-two. A simultaneous trip of both trip systems will cause a reactor scram, i.e., the tripping logic of the. trip systems is two-out-of-two. The tripping logic of the total system is referred to as one-out-of-two taken twice. This system will accommodate any single failure and still perform its intended function and in addition, provide protection against spurious scrams. The reliability of the dual channel system or probability that it will perform its intended function is less than that of a one-out-of-two system and somewhat greater than that of a two-out-of-three system (Section VIII-A.1.0 of the FSAR).

The instrumentation used to initiate action other than scram is generally similar to the reactor protection system. There are usually two trip systems required or available for each function. There are usually two instrument channels for each trip system. Either channel can trip the trip system but both trip systems are required to initiate the respective action. Where only one trip system is provided only one instrument channel is required to trip the trip system. All instrument channels except those for automatic depressurization are normally energized. De-energizing causes a trip. Power to the trip systems for each function is from reactor protection system buses 11 and 12.

The signals for initiating automatic blowdown and rod block differ from other initiating signals in that only one of the two trip systems is required to start blowdown or initiate rod block. Both instrument channels in the trip system must trip to initiate automatic blowdown.

This difference is due to the requirement that automatic depressurization be prevented unless A.C. power is available to the emergency core cooling systems. The instrument channels in'the trip system for automatic depressurization are normally de-energized. In order to cause a trip both instrument channels must be energized. Power to energize the instrument channels is from power boards 102 and 103. If A.C.

power is lost to one power board, one trip system becomes inoperable but the other trip system remains operable and capable of initiating automatic blowdown. If both power boards have lost A.C. power neither trip system can be energized and automatic blowdown is prevented. Only one instrument channel is required to initiated rod block.

AMENDMENT NO. 248

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION Each reactor operating condition has a related reactor mode switch position for the safety system. The instrumentation system operability for each mode switch position is based on the requirements of the related safety system. For example, the specific high drywell pressure trip systems must be tripped or operable any time core spray, containment spray, automatic depressurization or containment isolation functions are required.

In instrumentation systems where two trip systems are required to initiate action, either both trip systems are operable or one is tripped.

Having one trip system already tripped does not decrease the reliability in terms of initiating the desired action. However, the probability of

'spurious actuation is increased. Certain instrument channels or sensor inputs to instrument channels may be bypassed without affecting safe operation. The basis for allowing bypassing of the specified SRM's, IRM's, LPRM's and APRM's is discussed in Volume I (Section Vll-C.1.2)". The high area temperature isolation function for the cleanup system has one trip system. There are three instrument channels; each has four sensor inputs. Only two instrument channels are required since the area covered by any one sensor is also covered by a sensor in one of the other two instrument channels. The shutdown system also has one trip system for high area temperature isolation.

However, since the area of concern is much smaller, only one instrument channel is provided. Four sensors provide input to the channel.

Since the area covered is relatively small only three of the four sensors are required to be operable in order to assure isolation when needed.

Manual initiation is available for scram, reactor isolation and containment isolation. In order to manually initiate other systems, each pump and each valve is independently initiated from the control room. Containment spray raw water cooling is not automatically initiated.

Manual initiation of each pump is required as discussed in 3.3.7 above.

"FSAR; Letter, R.R. Schneider to A. Giambusso, dated November 15, 1973 AMENDMENT NO. 249

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The set points included in the tables are those used in the transient analysis and the accident analysis. The high flow set point for 6

the main steam line is 105 psi differential. This represents a flow of approximately 4.4x10 Ib/hr. The high flow set point for the emergency cooling system supply line is ( 11.5 psi differential. This represents a flow of approximately 9.8x10 Ib/hr at rated conditions.

Normal background for the main steam line radiation monitors is defined as the radiation level which exists in the vicinity of main steam lines after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or more of sustained full rated power. The dose rate at the monitor due to activity from the control rod drop accident of UFSAR Section XV-C.4 or from gross failure of one rod with complete fission product release from the rod would exceed the normal background at the monitor. The automatic initiation signals for the emergency cooling systems have to be sustained for more than 12 seconds to cause opening of the return valves. If the signals last for less than 12 seconds, the emergency cooling system operating will not be automatically initiated.

The high level in the scram discharge volume is provided to assure that there is still sufficient free volume in the discharge system to receive the control rod drives discharge. Following a scram, bypassing is permitted to allow draining of the discharge volume and resetting of the reactor protection system relays. Since all control rods are completely inserted following a scram and since the bypass of this particular scram initiates a control rod block, it is permissible to bypass this scram function. The scram trip associated with the shutdown position of the mode switch can be reset after 10 seconds.

The condenser low vacuum, low-low vacuum and the main steam line isolation valve position signals are bypassed in the startup and refuel positions of the reactor mode switch when the reactor pressure is less than 600 psig. These are bypassed to allow warmup of the main steam lines and a heat sink during startup.

AMENDMENT NO. 250

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The set points on the generator load rejection and turbine stop valve closure scram trips are set to anticipate and minimize the consequences of turbine trip with failure of the turbine bypass system as described in the bases for Specification 2.1.2. Since the severity of the transients is dependent on the reactor operating power level, bypassing of the scrams below the specified power level is permissible.

Although the operator will set the setpoints at the values indicated in Tables 3.6.2.a-1, the actual values of the various set points can differ appreciably from the value the operator is attempting to set. The deviations include inherent instrument error, operator setting error and drift of the set point. These errors are compensated for in the transient analyses by conservatism in the controlling parameter assumptions as discussed in the bases for Specification 2.1.2. The deviations associated with the set points for the safety systems used to mitigate accidents have negligible effect on the initiation of these systems. These safety systems have initiation times which are orders of magnitude greater than the difference in time between reaching the nominal set point and the worst set point due to error. The maximum allowable set point deviations are listed below:

Neutron Flux APRM Scram, a2.3% of rated neutron flux (analytical limit is 120% of rated flux)

APRM Rod Block, a2.3% of rated neutron flux (analytical limit is 110% of rated flux)

IRM, s2.5% of rated neutron flux Recirculation Flow Upscale, a1.6% of rated recirculation flow (analytical limit is 107.1% of rated flow)

Recirculation Flow Comparator, a2.09% of rated recirculation flow (analytical limit is 10% flow differential)

Reactor Pressure, %15.8 psig Containment Pressure s0.053 psig Reactor Water Level, a2.6 inches of water Main Steam Line Isolation Valve Position, a2.5% of stem position Scram Discharge Volume, +0 and -1 gallon Condenser Low Vacuum, a0.5 inches of mercury AMENDMENT NO. 251

'I h

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION High Flow-Main Steam Line, a1 psid High Flow-Emergency Cooling Line, a1 psid High Area Temperature-Main Steam Line, a10'F High Area Temperature-Clean-up and Shutdown, a6'F High Radiation-Main Steam Line, +100% and -50% of set point value High Radiation-Emergency Cooling System Vent, +100% and -50% of set point High Radiation-Reactor Building Vent, +100% and -50% of set point High Radiation-Refueling Platform, +100% and -50% of set point High Radiation-Offgas Line, a50% of set point, (Appendix D)"

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, "Technical Specification improvement Analyses for BWR Reactor Protection System," and MDE-77-0485, "Technical Specification Improvement Analysis for Nine Mile Point Nuclear Station, Unit 1."

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A Suppl2, "Technical Specification Improvement Analyses for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation,"

and with NEDC-31677P-A, "Technical Specification Improvement Analyses for BWR Isolation Actuation Instrumentation." Because of local high radiation, testing instrumentation in the area of the main steam line isolation valves can only be done during periods of Station shutdown. These functions include high area temperature isolation and isolation valve position scram.

NFSAR AMENDMENT NO. 252

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30936P-A, "BWR Owners'roup Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," Parts 1 and 2 and RE-003, "Technical Specification Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for Nine Mile Point Nuclear Station, Unit 1."

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R. D. Binz IV from C. E. Rossi dated July 21, 1992).

Testing of the scram associated with the shutdown position of the mode switch can be done only during periods of Station shutdown since it always involves a scram.

b. The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR is maintained greater than the SLCPR. The trip logic for this function is 1 out of n; e.g., any trip on one of the eight APRM's, eight IRM's or four SRM's will result in a rod block. The minimum instrument channel requirements provide sufficient instrumentation to assure the single failure criteria is met. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A Suppl 1, "Technical Specification Improvement Analyses for BWR Control Rod Block Instrumentation," and with GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R. D. Binz IV from C. E.

Rossi dated July 21, 1992).

The APRM rod block trip is flow biased and prevents a significant reduction in MCPR especially during operation at reduced flow.

The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the SLCPR.

The APRM rod block also provides local protection of the core; i.e., the prevention of critical heat flux in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst case single control rod withdrawal error has been analyzed and the results show that with the specified trip settings rod withdrawal is blocked before the MCPR reaches the SLCPR, thus allowing adequate margin. Below -60% power the worst case withdrawal of a single control rod results in a MCPR > SLCPR without rod block action, thus below this level it is not required.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block action before MCPR approaches the SLCPR.

AMENDMENT NO. 253

BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION A'ownscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and the control rod motion is prevented. The downscale rod blocks are set at 5 percent of full scale for IRM and 2 percent of full scale for APRM (APRM signal is generated by averaging the output signals from eight LPRM flux monitors).

AMENDMENT NO. 254

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.3 EMERGENCY POWER SOURCES 4.6.3 EMERGENCY POWER SOURCES Applies to the operational status of the emergency power Applies to the periodic testing requirements for the sources. emergency power sources.

~Ob ective: ~Ob ective:

To assure the capability of the emergency power sources To assure the operability of the emergency power to provide the power required for emergency equipment sources to provide emergency power required in the in the event of a loss-of-coolant accident. event of a loss-of-coolant accident.

a. For all reactor operating conditions except cold The emergency power systems surveillance will be shutdown, there shall normally be available two performed as indicated below. In addition, components 115 kv external lines, two diesel generator power on which maintenance has been performed will be systems and two battery systems, except as further tested.-

specified in "b," "c," "d," "e," and "h" below.

a. Durin each ma or refuelin o a e - test for automatic startup and pickup of load required for a loss-of-coolant accident.

b. ~Mon hi - manual start and operation at rated load

b. One 115 kv external line may be de-energized provided two diesel-generator power systems are shall be performed for a minimum time of one hour.

operable. If a 115 kv external line is de-energized, Determine the specific gravity of each cell.

that line shall be returned to service within 7 days. Determine the battery voltage.

AMENDMENT NO. 255

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. One diesel-generator power system may be c. ~Weekl - determine the cell voltage and specific inoperable provided two 115 kv external lines are gravity the pilot cells of each battery.

of energized. If a diesel-generator power system becomes inoperable, it shall be returned to an d. Surveillance for star u with an ino rable diesel-operable condition within seven days. In addition, if ~enera or - prior to startup the operable diesel-a diesel-generator power system becomes inoperable generator shall be tested for automatic startup and coincident with a 115 kv line de-energized, that pickup of the load required for a loss-of-coolant diesel-generator power system shall be returned to accident.

an operable condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e. Surveillance for o era ion wi h an ino erable diesel-

d. If a reserve power transformer becomes inoperable, generator - the operable diesel-generator shall be it shall be returned to service within seven days. manually started and operated at rated load for a minimum time of one hour immediately and once per

e. For all reactor operating conditions except startup week thereafter.

and cold shutdown, the following limiting conditions shall be in effect:

(1) One operable diesel-generator power system and one energized 115 kv external line shall be available.'f this condition is not met, normal orderly shutdown will be initiated within one hour and the reactor will be in the cold shutdown condition within ten hours.

AMENDMENT NO. 256

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) If no 115 kv external line is available, both diesel-generator power systems shall be operable with one diesel-generator running. If no 115 kv external line is available after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, normal orderly shutdown will be initiated within one hour and the reactor will be in the cold shutdown condition within ten hours.

f. For all reactor operating conditions except cold shutdown, there shall be a minimum of two day' fuel supply onsite for one diesel-generator or normal orderly shutdown will be initiated within one hour and the reactor will be in the cold shutdown condition within ten hours.

g. When operating with only one diesel-generator, all emergency equipment aligned to the operable diesel-generator shall have no inoperable components.

h. If a battery system becomes inoperable that system shall be returned to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

AMENDMENT NO. 257

BASES FOR 3.6.3 AND 4.6.3 EMERGENCY POWER SOURCES Other than the Station turbine generator, the Station is supplied by four independent sources of a-c power; two 115 kv transmission lines, and two.diesel-generators. Any one of the required power sources will provide the power required for the worst loss-of-coolant accident.

The required loads of 2500 kva and 2750 kva for the loss-of-coolant are calculated in detail in the First Supplement to the FSAR. This loading is greater than that required during a Station shutdown condition. The monthly test run paralleled with the system is based on the manufacturer's recommendation for these units in this type of service. The testing during operating cycle will simulate the accident conditions under which operation of the diesel-generators is required. A detailed tabulation of the equipment comprising the maximum diesel-generator load is given in the answer to Question V-10 of the First Supplement to the FSAR.

As mentioned above, a single diesel-generator is capable of providing the required power to equipment following a major accident. Two fuel oil storage tanks are provided with piping interties to permit supplying either diesel-generator. A two-day supply will provide adequate time to arrange for fuel makeup if needed. The full capacity of both tanks will hold a four-day supply.

It has been demonstrated in Appendix E-I.3.2I" that even with complete d-c loss the reactor can be safely isolated and the emergency cooling system will be operative with makeup water to the emergency cooling system shells maintained manually. Having at least one d-c battery available will permit: automatic makeup to the shells rather than manual, closing of the d-c actuated isolation valve on all lines from the primary system and the suppression chamber, maintenance of electrical switching functions in the Station and providing emergency lighting and communications power.

A battery system shall have a minimum of 106 volts at the battery terminals to be considered operable.

NFSAR AMENDMENT NO. 258

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.4 SHOCK SUPPRESSORS SNUBBERS 4.6.4 SHOCK SUPPRESSORS SNUBBERS A licabilit A licabilit Applies to the operational status of shock suppressors Applies to the periodic testing requirement for shock (snubbers). suppressors (snubbers).

~Ob ective ~Ob ective To assure the capability of the snubbers to: To assure the operability of the snubbers to perform:their intended functions.

Prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, and Allow normal thermal motion during startup and shutdown.

AMENDMENT NO. 259

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT S ecification S ecification

a. During all reactor operating conditions, except cold The following surveillance requirements apply to shutdown, snubbers shall be operable on those snubbers. Snubbers excluded from this inspection systems required to be operable during that program are those installed on nonsafety-related systems particular operating condition except as noted in and then only if their failure or failure of the system on 3.6.4.b, c and.d below. which they are installed, would have no adverse effect on any safety-related system.

Snubbers excluded from this inspection program are those installed on nonsafety-related systems and a.. Visual Ins ection then only if their failure or failure of the system on which they are installed, would have no adverse (i) Visual lns ec ion Fre uenc effect on any safety-related system.

Snubbers shall be visually inspected in

b. With one or more snubbers inoperable, within 72 accordance with the following schedule:

hours replace or restore the inoperable snubber(s) to the operable status or perform an engineering Number of Snubbers evaluation to determine that the components Found Inoperable supported by the snubber(s) were not adversely During Inspection or affected by the inoperability of the snubber(s), i.e. During Inspection Next Required the snubber(s) is (are) not required for system Interval Inspection Interval operability.

0 Refueling period

c. If after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the actions as described in Section 1 12 months ~ 25%

3.6.4b have not been completed, the supported" 2 6 months s 25%

system shall be declared inoperable and the 3,4 124days s 25%

appropriate action statement for that system will be 5,6,7 62 days s 25%

followed. 8 or more '31 days ~ 25%

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

AMENDMENT NO. 260

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

d. If the actions described in 3.6.4.b or c resulted in Snubbers may.be categorized into two types replacement or restoration to the operable status of (mechanical and hydraulic). These may then be the affected snubber(s), perform an engineering classified as "accessible" or "inaccessible" evaluation to determine if the components supported based on accessibility for inspection during by the snubber(s) were adversely affected by the operation. These four groups may be inspected inoperability of the snubber. independently according to the above schedule.

(ii) Visual Ins ection Acce ance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired operability, (2) attachments to the foundation or supporting structure are secure, and (3) in those locations where snubber movement can be manually induced without disconnecting the snubber, that the snubber has freedom of movement and is not frozen up. Snubbers which appear inoperable as a result of visual inspections may be determined operable for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; or (2) the affected snubber is functionally tested in the as found condition and determined operable per Specification 4.6.4.b as applicable.

AMENDMENT NO. 261

0 LllVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. Functional Tes in (i) Functional Tes Fre uenc At least once each refueling cycle, 10% of the total of each type (mechanical or hydraulic, accessible or inaccessible) of snubber in use in the plant shall be functionally tested either in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria of 4.6.4b(ii) an additional 10% of that type of snubber shall be functionally tested.

(ii) Functional Tes Acce an e Re uiremen Hydraulic snubber functiorial test shall verify that:

1. Activation (restraining action) is achieved within the specified range of velocity.

2. Freedom of movement exists in both tension and compression.

Mechanical snubber functional test shall verify that:

1. The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.

AMENDMENT NO. 262

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

2. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

AMENDMENT NO. 263

BASES FOR 3.6.4 AND 4.6.4 SHOCK SUPPRESSORS (SNUBBERS)

Snubbers are required to be operable to ensure that the structural integrity of the reactor coolant system and other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.

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 number of 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.

Hydraulic or mechanical, accessible or inaccessible, snubbers may each be treated as a different entity for the above surveillance programs.

AMENDMENT NO. 264

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.5 Radioac ive Ma erial Sources 4.6.5 Radioac ive Ma erial Sources Applies to the limit on source leakage for sealed or start- Applies to the periodic testing requirements for source up sources. leakage.

~0b'ective: ~ab ective:

To specify the requirements necessary to limit contam- To assure the capability of each source material ination from radioactive source materials. container to limit leakage within allowable limits.

1. The leakage test shall be capable of detecting the Tests for leakage and/or contamination shall be presence of 0.005 microcurie of radioactive material performed by the licensee or by other persons on the test sample. If the test reveals the presence specifically authorized by the Commission or an of 0.005 microcurie or more of removable contam- agreement State, as follows:

ination, it shall immediately be withdrawn from use, decontaminated and repaired or be disposed of in 1. Each sealed source, except start-up sources subject accordance with Commission regulations. Sealed to core flux, containing radioactive material, other sources are exempt from such leak tests when the than hydrogen 3, with a half-life greater than 30 source contains 100 microcuries or less of beta and/ days and in any form other than gas shall be tested or gamma emitting material or 10 microcuries'or less for leakage and/or contamination at intervals not to of alpha emitting material. exceed six months.

2. Results of required leak tests performed on sources, if the tests reveal the presence of 0.005 microcurie or more of removable contamination, shall be reported within 90 days.

AMENDMENT NO. 265

0 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS

3. A complete inventory of radioactive by-product 2. The periodic leak test required does not apply to materials, exceeding the limits set forth in 10 CFR sealed sources that are stored and not being used.

30.71, in sealed sources in possession shall be The sources excepted from this test shall be tested maintained current at all times. for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. In the absence of a certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.

3. Start-up sources shall be leak tested within 31 days prior to being subjected to core flux and following any repair or maintenance.

AMENDMENT NO. 266

BASES FOR 3.6.5 AND 4.6.5 RADIOACTIVE MATERIALSOURCES The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the probable leakage from the source material. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. Quantities of interest to this specification which are exempt from the leakage testing are consistent with the criteria of 10 CFR Parts 30.11-20 and 70.19. Leakage from sources excluded from the requirements of this specification is not likely to represent more than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.

AMENDMENT NO. 267

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIRElVlENT 3.6.11 ACCIDENT MONITORING INSTRUMENTATION 4.6.11 ACCIDENT MONITORING INSTRUMENTATION Applies to the operability of the plant instrumentation Applies to the surveillance of the instrumentation that that performs an accident monitoring function. performs an accident monitoring function.

~Ob'ec ive: ~obec ive:

To assure high reliability of the accident monitoring - To verify the operability of accident monitoring instrumentation. instrumentation.

a. During the power operating condition, the accident Instrument channels shall be tested and calibrated at monitoring instrumentation channels shown in Table least as frequently as listed in Table 4.6.11.

3.6.11-1 shall be operable except as specified in Table 3.6.11-2.

AMENDMENT NO. 268

TABLE 3.6.11-1 ACCIDENT MONITORING INSTRUMENTATION Minimum Number of Operable Parameters Total Number of Channels Sensors or Channels Action See Table 3.6.11-2

1) Relief Valve Position Indication 2/Valve 1/Valve"

2) Safety Valve Position Indication 2/Valve 1/Valve"

3) Reactor Vessel Water Level

4) Drywell Pressure Monitor

5) Suppression Chamber Water Level

6) Containment Hydrogen Monitor

7) Containment High Range Radiation Monitor

8) Suppression Chamber Water Temperature A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance provided at least one Operable channel is monitoring that Parameter.

AMENDMENT NO. 269

TABLE 3.6.1 1-2 A CI DENT VOMIT RING INSTRUMENTATION ACTION TATENlENT ACTION - 1

a. With the number of OPERABLE accident monitoring instrumentation channels 1 less than the total number shown in Table 3.6.11-1, restore to an OPERABLE status during the next cold shutdown when there is access to the drywell ~

b. With the number of OPERABLE accident monitoring instrumentation channels less than the minimum number shown in Table 3.6.11-1, restore the inoperable channel to an OPERABLE status within 30 days or be in at least a HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. The total number of channels shown in Table 3.6.11-1 will be OPERABLE prior to the beginning of each cycle.

ACTION - 2

a. With the number of OPERABLE accident monitoring instrumentation channels less than the total Number of Channels shown in Table 3.6.11-1, restore the inoperable channel(s) to OPERABLE status within seven days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With the number of OPERABLE accident monitoring instrumentation channels less than the minimum Channels OPERABLE requirements of Table 3.6.11-1, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION - 3

a. With the number of OPERABLE channels less than the total Number of Channels shown in Table 3.6.11-1, prepare and submit a Special Report to the Commission 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.

b. With the number of OPERABLE channels less than required by the minimum channels OPERABLE requirements, initiate the pre-planned alternate method of monitoring the appropriate parameter(s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1) either restore the inoperable channel(s) to OPERABLE status within seven days of the event, or

2) prepare and submit a Special Report to the Commission 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.

AMENDMENT NO. 270

TABLE 3.6.11-2 (cont'd)

ACCIDENT MONITORING INSTRUMENTATION ACTION STATEIVIENT ACTION - 4

a. With the number of OPERABLE channels less than the total Number of Channels shown in Table 3.6.11-1, prepare and submit a Special Report to the Commission 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.

b. With the number of OPERABLE channels less than required by the minimum channels OPERABLE requirements, initiate the pre-planned alternate method of monitoring the appropriate parameter(s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1) either restore the inoperable channel(s) to OPERABLE status within seven days of the event, or

2) prepare and submit a Special Report to the Commission 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 system.

If the pre-planned alternate method of monitoring the appropriate parameter(s) is not available, either restore the inoperable channel(s) to OPERABLE status within seven days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

AMENDMENT NO. 271

TABLE 4.6.11 ACCIDENT MONITORING INSTRUMENTATION

. SURVEILLANCE REQUIRElVIENT Parameter Instrument Channel Test Instrumen Channel Calibration (1) Relief valve position indicator (Primary- Once per quarter Once during each major refueling outage Acoustic)

Relief valve position indicator (Backup- Once per quarter Once during each major refueling outage Thermocouple)

(2) Safety valve position indicator (Primary- Once per quarter Once during each major refueling outage Acoustic)

Safety valve position indicator (Backup- Once per quarter Once during each major refueling outage Thermocouple) r (3) Reactor vessel water level Once per quarter Once during each major refueling outage (4) Drywell Pressure Monitor Once per month Once during each major refueling outage (5) Suppression Chamber Water Level Monitor Once per quarter Once during each major refueling outage (6) Containment Hydrogen Monitor Once per month Once per quarter (7) Containment High Range Radiation Monitor Once per month Once during each major refueling outage (8) Suppression Chamber Water Temperature Once per month Once during each major refueling outage AMENDMENT NO. 272

BASES 3.6.11 AND 4.6.11 ACCIDENT MONITORING INSTRUMENTATION Accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," and/or NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980 and NUREG 0661, "Safety Evaluation Report Mark I Containment Long Term Program." ~

The maximum allowable setpoint deviation for the Suppression Chamber Water Level Instrumentation is ~ 1.8 inches.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications,"

as approved by the NRC and documented in the SER (letter to R. D. Binz IV from C. E. Rossi dated July 21, 1992).

AMENDMENT NO. 273

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT 3.6.12 REACTOR PROTECTION SYSTEM AND REACTOR 4.6.12 REACTOR PROTECTION SYSTEM AND REACTOR TRIP SYSTEM POWER SUPPLY MONITORING TRIP SYSTEM POWER SUPPLY MONITORING A licabilit Applicability:

Applies to the operability of instrumentation that Applies to the surveillance of instrumentation that provides protection of the reactor protection system provides protection of the reactor protection system and reactor trip system. and reactor trip system.

~Obec ive: ~Ob ective:

To assure the operability of the instrumentation To verify the operability of protection instrumentation monitoring the power to the reactor protection monitoring the power to the reactor protection and system and reactor trip system. reactor trip buses.

a. Except as specified in specifications b and c a. At least onc ever six mon h below, two protective relay systems shall be Demonstrate operability of the overvoltage, operable for each power supply. undervoltage and underfrequency protective instrumentation by performing an instrument channel test. This instrument channel test will consist of simulating abnormal power conditions by applying from a test source, an overvoltage signal, an undervoltage signal and an under-frequency signal to verify that the tripping logic up to but not including the output contactors functions properly.

AMENDMENT NO. 274

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

b. With one protective relaying system inoperable, b. A least once er refuelin c cle restore the inoperable system to an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or remove the power Demonstrate operability of the overvoltage, supply from service. undervoltage and underfrequency protective instrumentation by performing an instrument

c. With both protective relaying systems inoperable, channel test. This instrument channel test will restore at least one to an operable status within consist of simulating abnormal power conditions 30 minutes or remove the power supply from by applying from a test source an overvoltage service. signal, an undervoltage signal and an underfrequency signal to verify that the tripping logic including the output contactors functions properly at least once. In addition, a sensor calibration will be performed to verify the following setpoints.

i. Overvoltage ~132 volts, ~4 seconds ii. Undervoltage ) 108 volts, <4 seconds iii. Underfrequency ~ 57 hertz, ~2 seconds AMENDMENT NO. 275

BASES FOR 3.6.12 AND 4.6.12 REACTOR PROTECTION SYSTEM AND REACTOR TRIP SYSTEM POWER SUPPLY MONITORING'o eliminate the potential for undetectable single component failure which could adversely affect the operability of the reactor protection system and reactor trip system, protective relaying schemes are installed on Motor Generator Sets 131 and 141, Static Uninterruptible Power Supply Systems 162 and 172, and maintenance bus 130A. This provides for overvoltage, undervoltage and underfrequency protection.

AMENDMENT NO. 276

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.13 REMOTE SHUTDOWN PANELS 4.6.13 REMOTE SHUTDOWN PANELS A licabilit Applies to the operating status of the remote Applies to the periodic testing requirements for the shutdown panels. remote shutdown panels.

~ob ective: ~Ob ec ive:

To assure the capability of the remote shutdown To assure the capability of the remote shutdown panels to provide 1) initiation of the emergency panels to provide 1) initiation of the emergency condensers independent of the main/auxiliary control condensers independent of the main/auxiliary control room 2) control of the motor-operated steam supply room 2) control of the motor-operated steam supply valves independent of the main/auxiliary control room valves independent of the main/auxiliary control room and 3) parameter monitoring outside the control and 3) parameter monitoring outside the control room. room.

a. During power operation and whenever the The remote shutdown panels surveillance shall be reactor coolant temperature is greater than performed as indicated below:

212'F, at least one remote shutdown panel shall be operable. a. Each remote shutdown panel monitoring instrumentation channel shall be demonstrated operable by performance of the operations and frequencies shown in Table 4.6.13-1.

b. Durin each ma or refuelin ou a e

1. Each remote shutdown panel shall be demonstrated to initiate the emergency condensers independent of the main/auxiliary control room.

AMENDMENT NO. 277

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIRElVIENT

b. A remote shutdown panel shall be considered 2. Each remote shutdown panel shall be inoperable if either the emergency condenser demonstrated to open both the rnotor-condensate return valve control switch is operated steam valves.

inoperable, either motor-operated steam supply valve control switch is inoperable, or the number of operable instrumentation channels is less than that required by Table 3.6.13-1.

c. If Specification 3.6.13.a cannot be met, commence an orderly shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

AMENDMENT NO. 278

TABLE 3.6-13-1 REIVIOTE SHUTDOWN PANEL IVIONITORING Limitin Condition for 0 eration INSTRUMENT MINIMUMNUMBER OF OPERABLE CHANNELS Reactor Pressure Reactor Water Level Reactor Water Temperature Torus Water Temperature Drywell Pressure Emergency Condenser Water Level Drywell Temperature "All Rods In" Light AMENDMENT NO. 279

TABLE 4.6.13-1 REMOTE SHUTDOWN PANEL MONITORING Surveillance Re uiremen Parameter Sensor Check Instrument Channel Calibration Reactor Pressure Once per day Once per 3 months Reactor Water Level Once per day Once per 3 rnonthsial Reactor Water Temperature Once per day Once per refueling cycle Torus Water Temperature Once per day Once per refueling cycle Drywell Pressure Once per day Once per 3 months Emergency Condenser Water Level Once per day Once per refueling cycle Drywell Temperature Once per day Once per refueling cycle "All Rods In" Light Once per refueling cycle N/A (a) The indicator located at the remote shutdown panel will be calibrated at the frequency listed in Table 4.6.13-1. Calibration of the remaining channel instrumentation is provided by Specification 4.6.2.

AMENDMENT NO. 280

BASES FOR 3.6.13 AND 4.6.13 REMOTE SHUTDOWN PANELS The remote shutdown panels provide 1) manual initiation of the emergency condensers 2) manual control of the steam supply valves and 3) parameters monitoring independent of the main/auxiliary control room. Two panels are provided, each located in a separate fire area, for added redundancy. Both panels are also in separate fire areas from the main/auxiliary control room. One remote shutdown panel provides the necessary capabilities consistent with 10 CFR 50 Appendix R. Therefore, only one remote shutdown panel is required to be operable.

The electrical design of the panels is such that no single fire can cause loss of both emergency condensers.

Each remote shutdown panel is provided with controls for one emergency condenser loop. The emergency condensers are designed such that automatic initiation is independently assured in the event of a fire 1) in the Reactor Building (principle relay logic located in the auxiliary control room or 2) in the main/auxiliary control room or Turbine Building (redundant relay logic located in the Reactor Building). Each remote shutdown panel also has controls to operate the two motor-operated steam supply valves on its respective emergency condenser loop. A key operated bypass switch is provided to override the automatic isolation signal to these valves. Once the bypass switch is activated, the steam supply valves can be manually controlled from the remote shutdown panels. Since automatic initiation of the emergency condenser is'assured, the remote shutdown panels serve as additional manual controlling stations for the emergency condensers. In addition, certain parameters are monitored at each remote shutdown panel.

The remote shutdown panels are normally de-energized, except for the monitoring instrumentation, which is normally energized. To energize the remaining functions on a remote shutdown panel, a power switch located on each panel must be activated. Once the panels are completely energized, the emergency condenser condensate return valve and steam supply valve controls can be utilized.

AMENDMENT NO. 281

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION 4.6.14 RADIOACTIVE EFFLUENT IN TR MENTATION Applies to the operability of plant instrumentation Applies to the surveillance of instrumentation that that monitors plant effluents. monitors plant effluents.

~Obec ive: ~Ob'ec ive:

To assure the operability of instrumentation to To verify operation of monitoring instrumentation.

monitor the release of radioactive plant effluents.

a. Liquid Effluent a. Liquid Effluent The radioactive liquid effluent monitoring instru- Each radioactive liquid effluent monitoring mentation channels shown in Table 3.6.14-1 instrumentation channel shall be demonstrated shall be operable with their alarm setpoints set to operable by performance of the sensor check, ensure that the limits of Specification 3.6.15.a.1 source check instrument channel calibration and are not exceeded. The alarm setpoints of these channel test operations at the frequencies shown channels shall be determined and adjusted in in Table 4.6.14-1.

accordance with the methodology and para-meters in the Offsite Dose Calculation Manual. Records - Auditable records shall be maintained, in accordance with procedures in the Offsite With a radioactive liquid effluent monitoring Dose Calculation Manual, of all radioactive liquid instrumentation channel alarm setpoint less effluent monitoring instrumentation alarm conservative than a value which will ensure that setpoints. Setpoints and setpoint calculations the limits of 3.6.15.a.1 are met, immediately shall be available for review to ensure that the suspend the release of radioactive liquid effluents limits of Specification 3.6.15.a.1 are met.

monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

AMENDMENT NO. 282

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT With less than the minimum number of radio-active liquid effluent monitoring instrumentation channels operable, take the action shown in Table 3.6.14-1. Restore the instruments to OPERABLE status within 30 days, or outline in the next Semi-Annual Radioactive Effluent Release Report the cause of the inoperability and how the instruments were or will be restored to operable status.

AMENDMENT NO. 283

TABLE 3.6.14-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUNIENTATION Limitin Condition f r0 eration

~lnetrnmen Minimum Channels 0 erable A licabili Gross Radioactivity Monitors A. Liquid Radwaste Effluent 1(c) At all times Line B. Service Water System (d) At all times(')

Effluent Line

2. Flow Rate Measurement Devices A. Liquid Radwaste Effluent At all times Line B. Discharge Canal

3. Tank Level Indicating Devices g A. Outside Liquid Radwaste At all times Storage Tanks

" 'Pumps curves or rated capacity will be utilized to estimate flow.

AMENDMENT NO. 284

NOTES FOR TABLE 3.6.14-1 (a) Provide alarm, but do not provide automatic termination of release.

(b) An operator shall be present in the Radwaste Control Room at all times during a release.

(c) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases may continue provided that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Specification 4.6.15.a, and

2. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving.

Otherwise suspend release of radioactive effluents via this pathway.

(d) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for gamma radioactivity at a lower limit of detection of at least Sx10 microcurie/ml.

(e) During discharge, with the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.

With the number of channels operable less than required by the minimum channels operable requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during liquid additions to the tank.

(g) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents.

(h) deleted Monitoring will be conducted continuously by alternately sampling the reactor building and turbine building service water return lines for approximately 15-minute intervals.

AMENDMENT NO. 285

TABLE 4.6.14-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION S rveillance Re uiremen

~lns rumen Sensor Check Source Checkl I. Channel Tes Channel Calibration Gross Beta or Gamma Radioactivity Monitors

a. Liquid Radwaste Once/day" Once/discharge" Once/3 months Once/year Effluent Line

b. Service Water Once/day Once/month Once/3 months Once/year(b)

Effluent Line

2. Flow Rate Measurement Devices

a. Liquid Radwaste Once/daylcl None None Once/year Effluent Line

b. Discharge Canall I None None None Once/year

3. Tank Level Indicating Deviceslel

a. Outside Liquid Once/day" " None Once/3 months Once/18 months Radwaste Storage Tanks Required prior to removal of blank flange in discharge line and until blank flange is replaced.

During liquid addition to the tank.

AMENDMENT NO. 286

NOTES FOR TABLE 4.6.14-1 (a) The channel test shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1. Instrumentation indicates measured levels above the alarm setpoint.

2. Instrument indicates a downscale failure.

3. Instrument controls not set in operate mode.

(b) The channel calibration shall be performed using one or more reference standards certified by the National Bureau of Standards or using standards that are traceable to the National Bureau of Standards or using actual samples of liquid waste that have been analyzed on a system that has been calibrated with National Bureau of Standard traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement.

(c) Sensor check shall consist of verifying indication of flow during periods of release. Sensor check shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic or batch releases are made.

(d) Pump performance curves or rated data may be used to estimate flow.

(e) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents.

Source check may consist of an installed check source, response to an external source, or (for liquid radwaste monitors) verification within 30 minutes of commencing discharge of monitor response to effluent.

AMENDMENT NO. 287

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

b. Gaseous Process and Effluent b. Gaseous Process and Effluent The radioactive gaseous process and effluent Each radioactive gaseous process and effluent monitoring instrumentation channels shown in monitoring instrumentation channel shall be Table 3.6.14-2 shall be operable with their alarm demonstrated operable by performance of the setpoints set to ensure that the limits of sensor check, source check, instrument channel Specification 3.6.15.b.1 are not exceeded. The calibration and instrument channel test alarm setpoints of these channels shall be operations at the frequencies shown in Table determined and adjusted in accordance with the 4.6.14-2.

methodology and parameters in the Offsite Dose Calculation Manual. Auditable records shall be maintained of the

. calculations made, in accordance with With a radioactive gaseous process and effluent procedures in the Offsite Dose Calculation monitoring instrumentation channel alarm Manual, of radioactive gaseous process and setpoint less conservative than required by the effluent monitoring instrumentation alarm above specification, immediately suspend the setpoints. Setpoints and setpoint calculations release of radioactive gaseous effluents shall be available for review to ensure that the monitored by the affected channel, or declare the limits of Specification 3.6.15.b.1 are met.

channel inoperable, or change the setpoint so it is acceptably conservative.

With less than the minimum number of radio-active gaseous process and effluent monitoring instrumentation channels operable, take the action shown in Table 3.6.14-2. Restore the instruments to OPERABLE status within 30 days or outline in the next Semi-Annual Radioactive Effluent Release Report the cause of the inoperability and how the instruments were or will be restored to operable status.

AMENDMENT NO. 288

TABLE 3.6.14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Limitin Condition for 0 eration Minimum

~lnetrumen Channels 0 erable A licabili Action Stack Effluent Monitoring

a. Noble Gas Activity Monitor (a)

b. Iodine Sampler Cartridge (b)

c. Particulate Sampler Filter (b)

d. Sample Flow Rate (c)

Measuring Device

e. Stack Gas Flow Rate (d)

Measuring Device

2. Main Condenser Offgas Treatment Explosive Gas Monitoring System

a. Hydrogen Monitor (e)

At all times.

During Offgas System Operation.

AMENDMENT NO. 289

TABLE 3.6.14-2 (cont'd)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR MENTATION Minimum

~lnstrnmen Channels 0 erable A lice bili Action Condenser Air Ejector Radio-activity Monitor (Recombiner discharge or air ejector discharge)

a. Noble Gas Activity (g)

b. Offgas System Flow Rate (c)

Measuring Devices

c. Sampler Flow Rate (c)

Measuring Devices

4. Emergency Condenser System

a. Noble Gas Activity Monitor 1 per vent (h)

During operation of the main condenser air ejector During reactor power operating condition AMENDMENT NO. 290

NOTES FOR TABLE 3.6.14-2 (a) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided grab samples are taken once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

/

(b) With the number of channels operable less than required by the minimum channels operable requirements, effluent releases via this pathway may continue provided that samples are continuously collected with auxiliary sampling equipment starting within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of discovery in accordance with the requirements of Table 4.6.15-2.

(c) With the number of channels operable less than required by the minimum channels operable requirements, effluent releases via this pathway may continue provided the flow rate is estimated once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

(d) Stack gas flow rate may be estimated by exhaust fan operating configuration.

(e) With the number of channels operable less than required by the minimum channels operable requirement, operation of the main condenser offgas treatment system may continue provided gas samples are collected and analyzed once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

One monitor on each recombiner. The system is designed to withstand the effects of a hydrogen explosion.

(g) With the number of channels operable less than required by the minimum channels operable requirement, gases from the main condenser offgas treatment system may be released provided:

1. Offgas grab samples are collected and analyzed once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. The stack monitor is operable.

3. Otherwise, be in at least hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(h) With the number of channels operable less than required by the minimum channels operable requirements, steam release via this pathway may commence or continue provided vent pipe radiation dose rates are monitored once per four hours.

AMENDMENT NO. 291

TABLE 4.6.14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT IVIONITORING INSTRUMENT Surveillance Re uirements

~lnstrumen Sensor Check Source Check Channel Tes h nn I Calibre ion

1. Stack Effluent Monitoring System

a. Noble Gas Activity Monitor Once/day ) Once/month Once/3 months(g) Once/year

b. Iodine Sampler Cartridge None None None None

c. Particulate Sampler Filter None None None None

d. Sampler Flow Rate Measuring Once/day None None Once/year Device

e. Stack Gas Flow Rate Measuring Once/day None None Once/year Device

2. Main Condenser Offgas Treatment System Explosive Gas Monitoring system (for system designed to withstand the effects of a hydrogen explosion)

a. Hydrogen Monitor Once/day None Once/month Once/3 months(e)

3. Condenser Air Ejector Radioactivity Monitor (Recombiner Discharge or Air Ejector Discharge)

Once/year( )

a. Noble Gas Activity Monitor Once/day Once/month Once/oper~ting cyclelc

b. Flow Rate Monitor Once/day None None Once/year

c. Sampler Flow Rate Monitor Once/day None None Once/year

4. Emergency Condenser System

a. Noble Gas Activity Monitor Once/day(") Once/month Once/3 months g Once/opegqting cycle(

NO. 'MENDMENT 292

NOTES FOR TABLE 4.6.14-2 (a) At all times.

(b) The channel calibration shall be performed using one or more of the reference standards certified by the National Bureau of Standards, standards that are traceable to the National Bureau of Standards or using actual samples of gaseous effluent that have been analyzed on a system that has been calibrated with National Bureau of Standards traceable'sources. These standards shall permit calibrating the system over its intended range of energy and measurement.

(c) The channel function test shall demonstrate that control room alarm annunciation occurs if either of the following conditions exist:

1) Instrument indicates measured levels above the Hi or Hi Hi alarm setpoint.

2) Instrument indicates a downscale failure.

The channel function test shall also demonstrate that automatic isolation of this pathway occurs if either of the following conditions exist:

1) Instruments indicate two channels above Hi Hi alarm setpoint.

2) Instruments indicate one channel above Hi Hi alarm setpoint and one channel downscale.

(d) During main condenser offgas treatment system operation.

(e) The channel calibration shall include the use of standard gas samples containing a nominal:

1. One volume percent hydrogen, balance nitrogen.

2. Four volume percent hydrogen, balance nitrogen.

(f) During operation of the main condenser air ejector.

(g) The channel test shall produce upscale and downscale annunciation.

(h) During reactor power operating condition AMENDMENT NO. 293

BASES FOR RADIOACTIVE EFFLUENT INSTRUIVIENTATION3.6.14 and 4.6.14 The radioactive liquid and gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid and gaseous effluents during actual or potential releases of liquid and gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the Offsite Dose Calculation Manual to ensure that the alarm/trip will occur prior to exceeding the limits of 10CFR Part 20. This instrumentation also includes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the main condenser offgas treatment system. The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10CFR Part 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to unrestricted areas.

AMENDMENT NO. 294

LllVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.15 RADIOACTIVE EFFLUENTS 4.6.15 RADIOACTIVE EFFLUENTS v ii Applies to the radioactive effluents from the station. Applies to the periodic test and recording requirements of the station process effluents.

~ob'ec ive:

~Ob ec ive:

To assure that radioactive material is not released to the environment in any uncontrolled manner and is within To ascertain that radioactive effluents from the the limits of 10CFR20 and 10CFR50 Appendix I. station are within allowable values of 10CFR20, Appendix B and 10CFR50, Appendix I.

a. Liquid

a. Liquid (1) Concentration (1) Concentration The concentration of radioactive material released in liquid effluents to unrestricted areas Radioactive liquid wastes shall be sampled shall be limited to the concentrations specified and analyzed according to the sampling and in 10CFR Part 20, Appendix B, Table II, analysis program of Table 4.6.15-1.

Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or The results of the radioactivity analyses shall entrained noble gases, the concentration shall be used in accordance with the methodology be limited to 2 x 10 4 microcuries/ml total and parameters in the Offsite Dose Calcu-activity. lation Manual to assure that the concen-trations at the point of release are Should the concentration of radioactive maintained within the limits of Specification material released in liquid effluents to 3.6.15.a. (1).

unrestricted areas exceed the above limits, restore the concentration to within the above limits immediately.

AMENDMENT NO. 295

0 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) Dose (2) Dose The dose or dose commitment to a member of. Cumulative dose contributions from. liquid the public from radioactive materials in liquid effluents for the current calendar quarter and effluents released, from each reactor unit, to the current calendar year shall be determined unrestricted areas (see Figures 5.1-1) shall be in accordance with the methodology and limited: parameters in the Offsite Dose Calculation Manual, prior to each release of a batch of (a) During any calendar quarter to less than liquid waste.

or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and (b) During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ.

With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.3 a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed" actions to be taken to assure 'orrective that subsequent releases will be in compliance with the above limits.

AMENDMENT NO. 296

0 TABLE 4.6.1 5-1 RADIOACTIVE LIQUID WASTE SAIVIPLING AND ANALYSI PROGRAM Surveillance Re uiremen Minimum Sampling Lower Limit of Liquid Release Type Frequency Analysis Frequency Type of Activity Analysis Detection(LLD) (uCi/ml)

A. Batch Waste Tanks Each Batch Each Batch Principal Gamma Sx107 Emitters 1-131 1 x 106 Each Batch Each Batch Dissolved and Entrained 1 x105 Gases (Gamma Emitters)

Monthly H-3 1x105 Each Batch Composite Gross Alpha 1 x107 Quarterly Sr-89, Sr-90 5x108 Each Batch Composite Fe-55 1 x 10-6 B. Service Water System Once/month Once/month Pnnclpal Gammalc) Emltters 5x 107 Effluent I-131 1 x106 Dissolved and Entrained 1 x105 Gases H-3 1 x105 Gross Alpha 1 x 107 Once/qua rterlf) Once/quarter( ) Sr-89, Sr-90 5x 108 Fe-55 1 x106 Completed prior to each release.

AMENDMENT NO. 297

A NOTES FOR TABLE 4.6.15-1 (a) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system which may include radiochemical separation:

LLD = 4.66 S E~V~2.22 x 10 ~Y~exp (-Aht)

Where:

LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 10 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and ht for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

Typical values of E, V, Y and b,t should be used in the calculation.

It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact for a particular measurement.

AMENDMENT NO. 298

NOTES FOR TABLE 4.6.15-1 (b) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representative sampling.

(c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. This list does not mean that only these nuclides are to be considered.

Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semi-Annual Radioactive Effluent Release Report.

(d) If more than one batch is released in a calendar month, only one batch need be sampled and analyzed during that month.

(e) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

(f) If the alarm setpoint of the service water effluent monitor, as determined by the method presented in the Offsite Dose Calculation Manual, is exceeded, the frequency of sampling shall be increased to daily until the condition no longer exists. Frequency of analysis shall be increased to daily for principal gamma emitters (including dissolved and entrained gases) and an incident composite for H-3, gross alpha, Sr-89, Sr-90 and Fe-55.

AMENDMENT NO. 299

0 I

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT

b. Gaseous b. Gaseous (1) Dose Rate (1) Dose Rate The dose rate due to radioactive materials The dose rate due to noble gases in gaseous released in gaseous effluents from the site effluents shall be determined to be within to areas at and beyond the site boundary the limits of Specification 3.6.15 in accor-shall be limited to the following: dance with the methodology and parameters in the Offsite Dose Calculation Manual.

(a) For noble gases: Less than or equal to 500 mrems/year to the total body and The dose rate due to iodine-131, iodine-133,

. less than or equal.to 3000 mrems/year tritium and all radionuclides in particulate to the skin, and form with half lives greater than 8 days in gaseous effluents shall be determined to be (b) For iodine-131, iodine-133, tritium and within the limits of Specification 3.6.15 in all radionuclides in particulate form accordance with methodology and para-with half lives greater than 8 days: meters in the Offsite Dose Calculation Less than or equal to 1500 mrems/year Manual by obtaining representative samples to any organ. and performing analyses in accordance with the sampling and analysis program specified With the dose rate(s) exceeding the above in Table 4.6.15-2.

limits, without delay restore the release rate to within the above limits(s).

AMENDMENT NO. 300

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (2) Air Dose (2) Air Dose The air dose due to noble gases released in Cumulative dose contributions for the gaseous effluents, from each reactor unit, to current calendar quarter and current calendar areas at and beyond the site boundary shall year for noble gases shall be determined be limited to the following: monthly in accordance with the method-ology and parameters in the Offsite Dose (a) During any calendar quarter: Less than Calculation Manual.

or equal to 5 milliroentgen for gamma radiation and less than or equal to 10 mrads for beta radiation and, (b) During any calendar year: Less than or equal to 10 milliroentgen for gamma radiation and less than or equal to 20 mrads for beta radiation.

With the calculated air dose from radio-active noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specifica-tion 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to .

reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

AMENDMENT NO. 301

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (3) Tritium, lodines and Particulates (3) Tritium, lodines and Particulates The dose to a member of the public from Cumulative dose contributions for the iodine-131, iodine-133, tritium and all current calendar quarter and current calendar radionuclides in particulate form with half year for iodine-131, iodine-133, tritium and

~

lives greater than 8 days in gaseous radionuclides in particulate form with half effluents released, from each reactor unit, to lives greater than 8 days shall be determined areas at and beyond the site boundary shall monthly in accordance with the method-be limited to the following: ology and parameters in the Offsite Dose Calculation Manual.

(a) During any calendar quarter: Less than or equal to 7.5 mrems to any organ and, (b) During any calendar year: Less than or equal to 15 mrems to any organ.

With the calculated dose from the release of iodine-131, iodine-'l33, tritium and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

AMENDMENT NO. 302

TABLE 4.6.15-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Surveillance Re uirements Minimum Lower Limit of Gaseous Release Type Sampling Frequency Analysis Frequency Type of Activity Analysis Detection (LLD} (uCi/ml}

A. Containment Purge Each Purge Prior to each release Principal Gamma Emitters 1 x104 Grab Sample Each Purge 'rincipal Gamma Emitters 1 x104 H-3 1 x106 B. Stack Once/Month Once/Month("l Principal Gamma Emitters>> 1x104 Once/Month " Once/Month H-3 1 x106 C. Stack Continuous Once/Week(f} 1-131 1 x 10-12 Charcoal Sample Continuous Once/Week(f} Principal Gamma Emitters 1 X 1p-11 Particulate Sample Continuous(e} Once/Month Gross alpha Sr-89, Sr-90 1 x 1p-11 Composite Particulate Sample Continuous( Noble gas monitor Noble Gases, Gross Gamma 1 x 10-6(g) or Principal Gamma Emitters' AMENDMENT NO. 303

NOTES FOR TABLE 4.6.15-2 (a) The LLD is defined in notation (a) of Table 4.6.15-1.

(b) Purge is defined in Section 1.23.

(c) The principal gamma emitters for. which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, I-131 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semi-Annual Radioactive Effluent Release Report pursuant to Specification 6.9.1.

(d) Sampling and analysis shall also be performed following shutdown, startup or an increase on the recombiner discharge monitor of greater than 50 percent, factoring out increases due to changes in-thermal power level or dilution flow; or when the stack release rate is in excess of 1000 uCi/second and steady-state gaseous release rate increases by 50 percent.

(e) The sample flow rate and the stack flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance-with Specifications 3.6.15.b.(1).(b) and 3.6.15.b.(3).

(f) When the release rate is in excess of 1000 uCi/sec and steady state gaseous release rate increases by 50 percent. The iodine and particulate collection device shall be removed and analyzed to determine the changes in iodine-131 and particulate release rate. The analysis shall be done daily following each change until it is shown that a pattern exists which can be used to predict the release rate; after which it may revert to weekly sampling frequency. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLD's may be increased by a factor of 10.

(g) When RAGEMS is inoperable the LLD for noble gas gross gamma analysis shall be 1 x 10 ".

(h) Tritium grab samples shall be taken weekly from the station ventilation exhaust (stack) when fuel is offloaded until stable tritium release levels can be demonstrated.

AMENDMENT NO. 304

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. Main Condenser c. Main Condenser The gross radioactivity (beta and/or gamma) rate The radioactivity rate of noble gases at the of noble gases measured at the recombiner recombiner discharge shall be continuously discharge shall be limited to less than or equal to monitored in accordance with Table 3.6.14-2.

500,000 uCi/sec. This limit can be raised to 1 Ci/sec. for a period not to exceed 60 days The gross radioactivity (beta and/or gamma) rate provided the offgas treatment system is in of noble gases from the recombiner discharge operation. shall be determined to be within the limits of Specification 3.6.15 at the following frequencies With the gross radioactivity (beta and/or gamma) by performing an isotopic analysis of a represent-rate of noble gases at the recombiner discharge ative sample of gases taken at the recombiner exceeding the above limits, restore the gross discharge:

radioactivity rate to within its limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least Hot Shutdown within the Monthly:

next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following an increase on the

d. Uranium Fuel Cycle recombiner discharge monitor of greater than 50%, factoring out increases due to The annual (calendar year) dose or dose changes in thermal power level and dilution commitment to any member of the public due to flow changes.

releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less d. Uranium Fuel Cycle than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be Cumulative dose contributions from liquid and limited to less than or equal to 75 mrems. gaseous effluents shall be determined in accordance with Specifications 4.6.15.a,(2),

4.6.15.b.(2) and 4.6.16.b.(3) and in accordance with the methodology and parameters in the Offsite Dose Calculation Manual.

AMENDMENT NO. 305

~,

LIMITINGCONDITION FOR OPERATION SURVEILlANCE REQUIREMENT With the calculated doses from the release of Cumulative dose contributions from direct radioactive materials in liquid or gaseous radiation from the reactor units and from effluents exceeding twice the limits of Specifica- radwaste storage tanks shall be determined in tions 3.6.15.a.2(b), 3.6.15.b.2(b) and accordance with the methodology and 3.6.15.b.3(b), calculations shall be made parameters in the Offsite Dose Calculation including direct radiation contributions from the Manual. This requirement is applicable only reactor units and from outside storage tanks to under conditions set forth in Specification determine whether the above listed 40CFR190 3.6.1 5.d.

limits have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10CFR Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report.

AMENDMENT NO. 306

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR 190. Submittal of the report is considered a timely request and a variance is granted until staff action on the request is complete.

AMENDMENT NO. 307

BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVEEFFLUENTS LIQ ID CON ENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to unrestricted areas will be less than the concentration levels specified in 10CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section II.A design objectives of Appendix I, 10CFR Part 50, to a member of the public and (2) the limits of 10CFR Part 20.106 (e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its maximum permissible concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection d i i ' i - ii (LLDs). Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually),

li diih (1968), and Hartwell, J. K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

AMENDMENT NO. 308

BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVEEFFLUENTS

~Li id Dose This specification is provided to implement the requirements of Section II.A, III.A and IV.A of Appendix I, 10CFR Part 50. The L'imiting Conditions for Operation expressed as quarter and annual limits are set at those values found in Section II.A. of Appendix I, in accordance .

with Section IV.A. The Limiting Conditions for Operation provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to unrestricted areas will be kept "as low as is reasonably achievable." There are no drinking water supplies that can be potentially affected by plant operations.

The dose calculation methodology and parameters in the Offsite Dose Calculation Manual implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculation procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the Offsite Dose Calculation Manual for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

AMENDMENT NO. 309

BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVE EFFLUENTS Gaseous Dos Ra This specification is. provided to ensure that the dose at any time at and beyond the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10CFR Part 20 to unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not.result in the exposure of a member of the public in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10CFR Part 20 (10CFR Part 20.106(b)). For members of the public who may at times be within the site boundary, the occupancy of that member of the public will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a member of the public at or beyond the site boundary to less than or equal to 500 mrems/year to the total body or to less than or equal to 3000 mrems/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDsi. Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, ~HASL- 00 (rev(sed annually),

Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Ch m. 40, 586-93 (1968), and Hartwell, J. K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

AMENDMENT NO. 310

BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVE EFFLUENTS Do - Noble Gases This specification is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I, 10CFR Part 50. The Limiting Condition for Operation expressed as quarter and annual limits are set at those values found in Section II.B of Appendix I in accordance with the guidance of Section IV.A. The action statements provid'e the required operating flexibility and at the same time implement the guides set forth in Section IV-A of Appendix I to assure that the releases of radioactive material in gaseous effluents to unrestricted areas will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the Offsite Dose Calculation Manual for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1,109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I, "Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977.

The Offsite Dose Calculation Manual equations provided to determine the air doses at=and beyond the site boundary are based upon the historical average atmospheric conditions.

AMENDMENT NO. 311

BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVEEFFLUENTS Do e - Iodine-131 I in -13 Tri i m and Radi nuclide in Particula e F rm This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10CFR Part 50. The Limiting Conditions for Operation expressed as quarter and annual limits are set at those valves found in Section II.C of Appendix I in accordance with the guidance of Section IV.A. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to unrestricted areas will be kept "as low as is reasonably achievable." The Offsite Dose Calculation Manual calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I be shown by calcula-tional procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The Offsite Dose Calculation Manual calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual'doses based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, iodine-133, tritium and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the site boundary. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man and 4) deposition on the ground with subsequent exposure of man.

Main Condenser Restricting the gross radioactivity rate of noble gases from the main condenser provides assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a very small fraction of the limits of 10CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50. The primary purpose of providing this specification is to limit buildup of fission product activity within the station systems which would result if high fuel leakage were to be permitted over extended periods.

AMENDMENT NO. 312

BASES FOR 3.6.15 AND 4.6.15 RADIOACTIVEEFFLUENTS T al Do e - ranium Fuel le This specification is provided to meet the dose limitations of 40CFR Part 190 that have been incorporated into 10CFR Part 20 by 46FR 182525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrems to the. total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to 4 reactors, it. is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I and if direct radiation doses from the reactor units and outside storage tanks are kept small The Special Report will describe a course of action

~

that should result in the limitation of the annual dose to a member of the public to within the 40CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to a member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contribution from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFR Part 190.11 and 10CFR Part 20 405c, is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10CFR Part 20, as addressed in Specification 3.6.15.a.(1) and 3.6.15.b.(1). An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

AMENDMENT NO. 313

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS 4.6.16 RADIOACTIVE EFFLUENT TREATMENT YSTEMS Applies to the operating status of the liquid, gaseous Applies to the surveillance requirements for the liquid, and solid effluent treatment systems. gaseous and solid effluent treatment systems.

~b'ec ive: ~ob ective:

To assure operability of the liquid, gaseous and solid. To verify operability of the liquid, gaseous and solid effluent treatment system. effluent treatment system.

a. Liquid a. Liquid The liquid radwaste treatment system shall be Doses due to liquid releases to unrestricted areas used to reduce the radioactive materials in liquid shall be projected prior to the release of each wastes prior to their discharge as necessary to batch of liquid radioactive waste in accordance meet the requirements of Specification 3.6.15. with the methodology and parameters in the Offsite Dose Calculation Manual.

b. Gaseous b. Gaseous The gaseous radwaste treatment system shall be. Doses due to gaseous releases to areas at and operable. The gaseous radwaste treatment beyond the site boundary shall be calculated system shall be used to reduce radioactive monthly in accordance with the methodology and materials in gaseous waste prior to their parameters in the Offsite Dose Calculation discharge as necessary to meet the requirements Manual.

of Specification 3.6.15.

AMENDMENT NO. 314

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT With gaseous radwaste from the main condenser c. Solid air ejector system being discharged without treatment for more than 7 days, prepare and The process control program shall be used to submit to the Commission within 30 days, verify the solidification of at least one pursuant to Specification 6.9.3, Special Report representative test specimen from at least every that identifies the inoperable equipment and the tenth batch of each type of wet radioactive reason for its inoperability, actions taken to waste (e.g., filter sludges and evaporator restore the inoperable equipment to OPERABLE bottoms).

status, and a summary description of those actions taken to prevent a recurrence. (1) If any test specimen fails to verify solidifi-cation, the solidification of the batch may

c. Solid then be resumed using the alternative solidification parameters determined by the The solid radwaste system shall be used in process control program.

accordance with a Process Control Program to process wet radioactive wastes to meet shipping (2) If the initial test specimen from a batch of and burial ground requirements. waste fails to verify solidification, the

, process control program shall provide for the With the provisions of the process control collection and testing of representative test program not satisfied, suspend shipments of specimens from each consecutive batch of defectively processed or defectively packaged the same type of wet waste until at least 3 solid radioactive wastes from the site. consecutive initial test specimens demonstrate solidification.

AMENDMENT NO. 315

BASES FOR 3.6.16 AND 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS LiuidRa w eTreamn S sem The requirement that the appropriate portions of this system be used provides assurance that; the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10CFR Part 50 and the design objective given in Section II.D of Appendix I to 10CFR Part 50.

Gaseous Radwa e Trea ment S s em The requirement that this system be used provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10CFR Part 50. Since the capability exists to operate within specification without use of the system, it is conceivable that due to unforeseen circumstances, limited operation without the system may be made sometime during the life of the plant.

Solid Radioac ive Waste This specification implements the requirements of 10CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10CFR part 50.

The process parameters included in establishing the process control program may include, but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents and mixing and curing times.

AMENDMENT NO. 316

LIIVIITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.17 EXPLOSIVE GAS MIXTURE 4.6.17 EXPLOSIVE GAS MIXTURE Applies to the operability of instrumentation to Applies to the surveillance of instrumentation that monitor hydrogen concentration in the main monitors hydrogen concentration in the main condenser off-gas treatment system. condenser off-gas treatment system.

OIIDe~ive: ~Ob'ec ive:

To assure the operability of the hydrogen monitoring To verify operation of monitoring instrumentation.

instrumentation in the main condenser off-gas treatment system.

The concentration of hydrogen in the main condenser The concentration of hydrogen in the main condenser off-gas treatment system shall be limited to 4 percent off-gas treatment system shall be determined to be by volume. within the above limits by continuously monitoring the waste gases in the main condenser off-gas If the concentration of hydrogen in the main treatment system in accordance with Table 3.6.14-2 condenser off-gas treatment system exceeds this of Specification 3.6.14.

limit, restore the concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

AMENDMENT NO. 317

BASES FOR 3.6.17 AND 4.6.17 EXPLOSIVE GAS IVIIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen. Automatic control features are included in the system to prevent the hydrogen concentration from reaching these flammability limits. Maintaining the concentration of hydrogen below flammability limits provides assurance that the releases of radioactive materials, will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10CFR Part 50.

AMENDMENT NO. 318

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.18 MARK I CONTAINMENT 4.6.18 MARK I CONTAINMENT v'i Applies to the venting/purging of the Mark I Applies to the surveillance requirement for venting Containment. and purging of the Mark I Containment when required to be vented/purged through the Emergency

, Ventilation System.

~ob ec ive: ~ob ective:

To assure that the Mark I Containment is vented/ To verify that the Mark I Containment is vented purged so that the limits of specifications 3.6.15.b.1 through the Emergency Ventilation System when and 3.6.15.b.3 are met. required.

The Mark I Containment drywell shall be vented/ The containment drywell shall be determined to be purged through the Emergency Ventilation System aligned for venting/purging through the Emergency unless Specification 3.6.15.b.1 and 3.6.15.b.3 can Ventilation System within four hours prior to start of be met without use of the Emergency Ventilation and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during venting/purging System. of the drywell.

If these requirements are not satisfied, suspend all venting/purging of the drywall.

AMENDMENT NO. 319

BASES FOR 3.6.18 AND 4.6.18 MARK I CONTAINMENT This specification provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10CFR Part 20 for unrestricted areas.

AMENDMENT NO. 320

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.19 LIQUID WASTE HOLDUP TANKS" 4.6.19 LIQUID WASTE HOLDUP TANKS bi i@:

Applies to the quantity of radioactive material that Applies to the surveillance requirements for outdoor may be stored in an outdoor liquid waste holdup tank. liquid waste holdup tanks.

~Ob'ec ive: ~Ob ec ive:

To assure that the quantity of radioactive material To verify the quantity of radioactive material stored in stored in outdoor holdup tanks does not exceed a an outdoor liquid waste holdup tank.

specified level.

The quantity of radioactive material contained in an The quantity of radioactive material contained in each outdoor liquid waste tank shall be limited to less than of the tanks listed in Specification 3.6.19 shall be or equal to 10 curies, excluding tritium and dissolved determined to be within the limit of Specification or entrained noble gases. 3.6.19 by analyzing a representative sample of the tank's contents at least weekly when radioactive With the quantity of radioactive material in any such materials are being added to the tank.

tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank.

Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit and describe the events leading to this condition in the next Semi-Annual Radioactive Effluent Release Report. r "Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to th'e liquid radwaste treatment system.

AMENDMENT NO. 321

BASES FOR 3.6.19 AND 4.6.19 LIQUID HOLDUP TANKS This specification applies to any outdoor tank that is not surrounded by liners, dikes or walls capable of holding the tank contents and that does not have tank overflows and surrounding areas drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks'ontents, the resulting concentrations would be less than the limits of 10CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

AMENDMENT NO. 322

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.20 RADIOLOGICALENVIRONMENTALMONITORING .. '4.6.20 RADIOLOGICALENVIRONMENTALMONITORING PROGRAM PROGRAM Applies to radiological samples of station environs. Applies to the periodic sampling and monitoring requirements of the radiological environmental

~Ob ec ive: monitoring program.

To evaluate the effects of station operations and ~Ob'e ive:

radioactive effluent releases on the environs and to verify the effectiveness of the controls on radioactive To ascertain what effect station operations and material sources. radioactive effluent releases have had upon the environment.

The radiological environmental monitoring program shall be conducted as specified in Table 3.6.20-1. The radiological environmental monitoring samples shall be collected pursuant to Table 3.6.20-1 from With the radiological environmental monitoring the specific locations given in the table and figure(s) program not being conducted as specified in Table in the Offsite Dose Calculation Manual and shall be 3.6.20-1, prepare and submit to the Commission, in analyzed pursuant to the requirements of Table the Annual Radiological Environmental Operating 3.6.20-1 and the detection capabilities required by Report, a description of the reasons for not Table 4.6.20-1 ~

conducting the program as required and the plans for preventing a recurrence.

Deviations are permitted from the required sample schedule if samples are unobtainable due to hazardous conditions, seasonal unavailability, theft, uncooperative residents or to malfunction of automatic sampling equipment. In the event of the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.

AMENDMENT NO. 323

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT With the level of radioactivity (as the result of plant effluents), in an environmental sampling medium exceeding the reporting levels of Table 6.9.3-1 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report pursuant to Specification 6.9.3. The Special Report shall identify the cause(s) for exceeding the limit(s) and define the corrective action(s) to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of Specifications 3.6.15.a.(2),

3.6.15.b.(2) and 3.6.15.b.(3). When more than one of the radionuclides in Table 6.9.3-1 are detected in the sampling medium, this report shall be submitted if concentration 1 limit level (1) limit level (2)

~ ..... a 1 .0 When radionuclides other than those in Table 6.9.3-1 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose.

to an individual is equal to or greater than the calendar year limits of Specification 3.6.15.a.(2),

3.6.15.b,(2) and 3.6.15.b.(3).

AMENDMENT NO. 324

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

.With milk or fruit and/or vegetables no longer available at one or more of the sample locations specified in Table 3.6.20-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the cause of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Semi-Annual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the Offsite Dose Calculation Manual reflecting the new location(s).

AMENDMENT NO. 325

TABLE 3.6.20-1 OPERATIONAL RADIOLOGICALENVIRONMENTALIVIONITORING PROGRAM Limitin Condition for 0 eration Exposure Pathway Sampling and Collection Type of Analysis and and/or Sample Number of;Samples and Locations Frequency Frequency Radioiodine 5 Samples from 5 locations: Continuous sampler operation Radioiodin ni r Particulates with sample collection weekly analyze once/week for

1) 3 Samples from off-site locations in different or as required by dust loading, I-131.

sectors of the highest calculated site average whichever is more frequent D/0 (based on all site licensed reactors) Pa i ula am I r Gross beta radioactivity

2) 1 sample from the vicinity of an established following filter change, year round community having the highest- composite (by location) for calculated site average D/0 (based on all site gamma isotopic analysis( I licensed reactors) once per 3 months, (as a minimum)

3) 1 sample from a control location 10-17 miles distant and in a least prevalent wind direction Direct Radiation 32 stations with two or more dosimeters to be Once per 3 months - Gamma dose once per 3 placed as follows: an inner ring of stations in the months general area of the site boundary and an outer ring in the 4 to 5 mile range from the site with a station in each land based sector." The balance of the stations should be placed in special int'eiest areas such as population centers, nearby residences, schools and in 2 or 3 areas to serve as control stations.

At this distance, 8 wind rose sectors are over Lake Ontario.

AMENDMENT NO. 326

0 TABLE 3.6.20-1 (cont'd)

OPERATIONAL RADIOLOGICALENVIRONIVIENTALMONITORING PROGRAIVI Llmitin Condition for 0 eration Exposure Pathway Sampling and Collection Type of Analysis and and/or Sample Number of Samples and Locations Frequency a Frequency WATERB RNE Surface( ) 1) 1 sample upstream Composite sample over 1 Gamma isotopic analysis

2) 1 sample from the site's downstream cooling month period g once/month. Composite for water intake once per 3 months tritium analysis.

)

Sediment from 1 sample from a downstream area with existing or Twice per year Gamma isotopic analysis(

Shoreline potential recreational value IN E TION Milk 1) Samples from milk sampling locations in 3 Twice per month, April- Gamma isotopic and locations within 3.5 miles distance having the December (samples will be I-131 analysis twice per highest calculated site average D/Q. If there collected in January-March if month when animals are on are none, then 1 sample from milking animals l-131 is detected in November pasture (April-December);

in each of 3 areas 3.5-5.0 miles distant having and December of the preceding once/month at other times the highest calculated site average D/Q (based year) (January-March) if required on all site licensed reactors)

2) 1 sample from a milk sampling location at a control location (9-20 miles distant and in a least prevalent wind direction) " '

AMENDMENT NO. 327

TABLE 3.6.20-1 (cont'd)

OPERATIONAL RADIOLOGICALENVIRONMENTALIVIONITORING PROGRAM Limitin Condition for 0 era ion Exposure Pathway Sampling and Collection Type of Analysis and and/or Sample Number of Samples and Locations Frequency Frequency Fish 1) 1 sample each of two commercially or Twice per year Gamma isotopic analysis recreationally important species in the vicinity on edible portions twice per of a plant discharge area( ) year

2) 1 sample each of the same species from an area at least 5 miles distant from the site.

Food Products 1) Samples of three different kinds of broad leaf Once per year during harvest Gamma isotopic analysis vegetation (such as vegetables) grown nearest season of edible portions (isotopic to each of two different off-site locations of to include I-131 or a highest calculated site average D/0 (based on separate l-131 analysis may all licensed site reactors) be performed) once during the harvest season

2) One sample of each of the similar broad leaf vegetation grown at least 9.3-20 miles distant in a least prevalent wind direction AIVIENDMENTNO. 328

NOTES FOR TABLE 3.6.20-1 (a) It is recognized that, at times, it may not be possible or practical to obtain samples of the media of choice at the most desired location or time. In these instances, suitable alternative media and locations may be chosen for the particular pathway in question and may be substituted. Actual locations (distance and directions) from the site shall be provided in the Annual Radiological Environmental Operating Report. Highest D/Q locations are based on historical meteorological data for all site licensed reactors.

(b) Particulate sample filters should be analyzed for gross beta 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If the gross beta activity in air is greater than 10 times a historical yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

(c) Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the facility.

(d) The purpose of these samples is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites, such as historical control locations which provide valid background data may be substituted.

(e) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purpose of this table, a thermoluminescent dosimeter may be considered to be one phosphor and two or more phosphors in a packet may be considered as two or more dosimeters. Film badges shall not be used for measuring direct radiation.

(f) The "upstream sample" should be taken at a distance beyond significant influence of the discharge. The "downstream sample" should be taken in an area beyond but near the mixing zone, if possible.

(g) Composite samples should be collected with equipment (or equivalent) which is capable of collecting an aliquot at time intervals which are very short (e.g. hourly) relative to the compositing period (e.g. monthly) in order to assure obtaining a representative sample.

(h) In the event commercial or recreational important species. are not available as a result of three attempts, then other species may be utilized as available.

AMENDMENT NO. 329

TABLE 4.6.20-1 DETECTION CAPABILITIES FOR ENVIRONIVIENTALSAIVIPLE ANALYSIS(a, )

LOWER LIMIT OF DETECTION LLD c Surveillance Re uiremen Water Airborne Particulate Fish Milk Food Products Sediment Analysis (pCi/I) or Gases (pCi/m ) (pCi/kg, wet) (pCi/I) (pCi/kg, wet) (pCi/kg, dry) gross beta 0.01 H-3 2000" Mn-54 15 130 Fe-59 30 260 Co-60

'o-58, 15 130 Zn-65 30 260 Zr-95, Nb-95 15 l-131 0.07 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba/La-140 15 If no drinking water pathway exists, a value of 3000 pCi/liter may be used.

If no drinking water pathway exists, a value of 15 pCi/liter may be used.

AMENDMENT NO. 330

NOTES FOR TABLE 4.6.20-1 (a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.d.

(b) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in ANSI N.545 (1975), Section 4.3. Allowable exceptions to ANSI N.545 (1975), Section 4.3 are contained in the Nine Mile Point Unit 1 Offsite Dose Calculation Manual (ODCM).

(c) The LLD is defined, for purposes of these specifications, as the smallest concentration of 'radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

LLD = 4.66 S EV~2. 2~Y+exp (-Ab,t)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, as counts per SI is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, where applicable, A is the radioactive decay constant for the particular radionuclide, and ht for environmental samples is the elapsed time between sample collection, or end of the sample collection period and time of counting.

Typical values of E, V, Y and ht should be used in the calculation.

AMENDMENT NO. 331

NOTES FOR TABLE 4.6.20-1 It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for the particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally, background. fluctuations, unavoidable small sample sizes, the presence of interfering nuclides or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.d.

AMENDMENT NO. 332

BASES 3.6.20 AND 4.6.20 RADIOLOGICALENVIRONMENTALIVIONITORING PLAN The radiological environmental monitoring program required by this specification provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of members of the public resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10CFR Part 50 and thereby supplements the radiological effluent monitoring program. by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.6.20-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually); Currie, L.A., "Limits for Qualitative Detection and Quantitativ'e Determination - Application to Radiochemistry," Anal. Chem 40, 586-93 (1968) and Hartwell, J.K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

AMENDMENT NO. 333

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT 3.6.21 INTERLABORATORYCOMPARISON PROGRAM 4.6.21 INTERLABORATORY COMPARISON PROGRAM Applies to participation in an interlaboratory Applies to testing the validity of measurements on comparison program on environmental sample environmental samples.

analysis.

~Ob ec ive: ~Ob'ective:

To ensure the accuracy of measurements of To verify the accuracy of measurements on radioactive material in environmental samples. radioactive material in environmental samples.

Analyses shall be performed on radioactive materials The Interlaboratory Comparison Program shall be supplied as part of an Interlaboratory Comparison described in the Offsite Dose Calculation Manual. A Program which has been approved by the summary of the results obtained as part of the above Commission. Participation in this program shall required Interlaboratory Comparison Program shall be include media for which environmental samples are included in the Annual Radiological Environmental routinely collected and for which intercomparison Operating Report. Participants in the EPA Cross samples are available. Check Program may provide the EPA program code designation in lieu of providing results.

With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.

AMENDMENT NO. 334

BASES FOR 3.6.21 AND 4.6.21 INTERLABORATORYCOMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring for the purposes of Section IV.B.2 of Appendix I to 10CFR Part 50.

AMENDMENT NO. 335

LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.22 LAND USE CENSUS 4.6.22 LAND USE CENSUS vi Applies to the performance of a land use census in Applies to assuring that current land use is known.

the vicinity of the Nine Mile Point Nuclear Facility.

~ob ec ive: ~ob ective:

To determine the utilization of land within a distance To verify the appropriateness of the environmental of three miles from the Facility. surveillance program.

A land use census shall be conducted and shall The land use census shall be conducted during the identify within a distance of three miles the location growing season at least once per 12 months using in each of the 16 meteorological sectors the nearest that information that will provide the best results, residence and within a distance of three miles the such as conducting a door-to-door survey, aerial location in each of the 16 meteorological sectors of survey or consulting local agriculture authorities. The all milk animals. In lieu of a garden census, specifica- results of the land use census shall be included in the tions for vegetation sampling in Table 3.6.20-1 shall Annual Radiological Environmental Operating Report.

be followed, including analysis of appropriate controls.

With a land use census identifying a milk animal location(s) that represents a calculated D/0 value greater than the D/0 value currently being used in specification 4.6.15.b.3, identify the new location(s) in the next Semi-Annual Radioactive Effluent Release Report.

AMENDMENT NO. 336

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREIVIENT If the D/0 value at a new milk sampling location is significantly greater (50%) than the D/0 value at an existing milk sampling location, add the new location to the radiological environmental monitoring program within 30 days. The sampling location(s) excluding the control station location, having the lowest calculated D/0 may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. Pursuant to Specification 6.9.1.e identify the new location(s) in the next Semi-Annual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the Offsite Dose Calculation Manual reflecting the new location(s).

AMENDMENT NO. 337

BASES FOR 3.6.20 AND 4.6.20 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census. The best survey information such as from a door-to-door survey(s), from an aerial survey or from consulting with local agricultural authorities shall be used.

This census satisfies the requirements of Section IV.B.3 of Appendix I to 10CFR Part 50.

In lieu of a garden census, the significance of the exposure via the garden pathway can be evaluated by the sampling of vegetation as specified in Table 3.6.20-1.

A milk sampling location, as defined in Section 1, requires that at least 10 milking cows are present at a designated milk sample location. It has been found'from past experience, and as a result of conferring with local farmers, that a minimum of 10 milking cows is necessary to guarantee an adequate supply of milk twice per month for analytical purposes. Locations with less than 10 milking cows are usually utilized for breeding purposes which eliminates a stable supply of milk for samples as a result of suckling calves and periods when the adult animals are dry.

AMENDMENT NO. 338

LIMITINGCONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.1 SPECIAL TEST EXCEPTION - SHUTDOWN MARGIN 4.7.1 SPECIAL TEST EXCEPTI N - SHUTDOWN MAR IN DEMONSTRATIONS DEMONSTRATIONS Applicability:

Applies to shutdown margin demonstration in the Applies to periodic inspections required to perform cold shutdown condition. shutdown margin demonstrations in the cold shutdown condition.

~ob'ec ive: ~Ob'ec iv:

~

To assure the capability of the control rod system to To specify the inspections required to perform the control core reactivity. shutdown margin demonstration in the cold shutdown condition.

a. The reactor mode switch may be placed in the startup position to allow more than one control a. Within 30 minutes prior to and at least once per rod to be withdrawn for shutdown margin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the performance of a shutdown demonstration, provided that at least the margin demonstration, verify that:

e following requirements are satisfied.

(1) The source range monitors are operable per (1) The source range monitors are operable in Specification 3.5.1.

the noncoincident condition.

(2) The rod worth minimizer is operable with the (2) The rod worth minimizer is operable per required program per Specification Specification 3.1.1b(3)(b) and is 3.1.1b(3)(b) or a second licensed operator or programmed for the shutdown margin other technically qualified member of the demonstration, or conformance with the unit technical staff is present and verifies shutdown margin demonstration procedure compliance with the shutdown margin is verified by a second licensed operator or demonstration procedure.

other technically qualified member of the unit technical staff.

AMENDMENT NO. 339

LIIVIITINGCONDITIONS FOR OPERATION SURVEILLANCE REQUIRENIENTS (3) The continuous rod withdrawal control shall (3) No core alterations are in progress.

not be used during out-of-sequence movement of the control rods.

(4) No core alterations are in progress.

b. With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the shutdown or refuel position.

AMENDMENT NO. 340

'1 BASES FOR 3.7.1 AND 4.7.1 SHUTDOWN MARGlN DEMONSTRATiON The shutdown margin demonstration has to-be performed prior to power operation. However, the mode switch must be placed in the startup position to allow withdrawal of more than one control rod. Specifications 3.7.1 and 4.7.1 require certain restrictions in order to ensure that an inadvertent criticality does not occur while performing the shutdown margin demonstration, The shutdown margin demonstration will be performed in the cold shutdown condition with the vessel head in place. The shutdown margin demonstration will be performed prior to the reactor coolant system pressure and control rod scram time tests following refueling outages when core alterations are performed. The shutdown margin demonstration is performed using the in-sequence non-critical method.

AMENDMENT NO. 341

5.0 DESIGN FEATURES 5.1 ~Si e The Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant site comprising approximately 1500 acres, is located on the shores of Lake Ontario, about seven miles northeast of Oswego, New York. An exclusion distance of nearly 4000 feet is provided between the Station and the nearest site boundary to the west, a mile to the boundary on the east, and a mile and a half to the southern site boundary (as described in the Sixth Supplement of the FSAR). i Figure 5.1-1 is a Site Boundary Map of Nine Mile Point which allows the identification of gaseous and liquid waste release points. Figure 5.1-1 also defines the unrestricted area within the site boundary that is accessible (except for fenced areas) to member of the public.

5.2 ~Reac or The reactor core consists of no more than 532 fuel assemblies containing enriched uranium dioxide pellets clad in Zircaloy-2.

The core excess reactivity will be controlled by movable control rods and burnable poisons. The core will be cooled by circulation of water internally and external to the pressure vessel through recirculation loops.

5.3 Reactor Vessel The pertinent features of the reactor vessel other than those referred to in the technical specifications are as follows:

Internal Height 63'-10" Internal Diameter 17'-9" Vessel Design Lifetime 40 years Materials of Construction ...

Base Metal SA302B Clad Weld Deposited 308L Electrode AMENDMENT NO. 342

I

~ ~(C) ~(g)

I 0 (d) 0 JV A

E

~CC ~ Prf7MTR ICK

~ ~b) gag~ KWCh NIIC QL!

VENT I STATiOII (c) i MlvatE, L 4INlfNW NIAGARA MOMAWK POWER CO RPORATlON POWER AUTHORlTY STATE OF NEW YORK t

(s)

SCkX~WLCf FiGURE 5.1-1 S!TK BOUNDARIES NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT I AMENDMENT NO. 343

NOTES TO FIGURE 5.1-1 (a) NMP1 Stack (height is 350')

(b) NMP2 Stack (height is 430')

(c) JAFNPP Stack (height is 385')

(d) NMP1 Radioactive Liquid Discharge (Lake Ontario, bottom)

(e) NMP2 Radioactive Liquid Discharge (Lake Ontario, bottom)

. (f) JAFNPP Radioactive Liquid Discharge (Lake Ontario, bottom)

(g) Site Boundary h

(h) Lake Ontario Shoreline

'V Additional Information:

NMP2 Reactor Building Vent is located 187 feet above ground level JAFNPP Reactor and Turbine Building Vents are located 173 feet above ground level

,JAFNPP Radwaste Building Vent is 112 feet above ground level AMENDMENT NO. 344

5.4 CONTAINMENT The containment system consists of a drywell, suppression chamber and a reactor building. The pressure suppression system consists of a drywall with a volume of approximately 243,000 cubic feet and an interconnected suppression chamber with a volume of 209,000 cubic feet. Of this total volume some 180,000 and 120,000 cubic feet of free space are available in the drywell and suppression chamber, respectively.

The pertinent design features not discussed elsewhere in the technical specifications are as follows:

Pressure D well 5 Vents Su resslon Chamber Internal Design Pressure 62 psig 35 psig Internal Design Temperature 310~F 205oF External Design Pressure 2 pslg 1 pslg Material of Construction A-201 and A-212 Grade "B" Firebox Steel made to A-300 requirements.

For long-term post-accident recovery, the pressure suppression system is designed to permit flooding to a level at least six feet above the core.

The reactor building is designed for a maximum in-leakage rate of 100 percent per day at 0.25 inch of water internal vacuum and zero wind speed. Exterior loadings for wind, snow and ice meet all applicable codes. The roof and supporting structures are designed to withstand a loading of 40 psf of snow or ice. The walls and building structure are designed to withstand an external or internal loading of 40 psf which is approximately equivalent to that caused by a wind velocity of 125 mph 30 feet above the ground level.

Pressure relief is provided to prevent damage to the superstructure due to the break of any primary system line in the reactor building.

In this event, blowout panels will fail, relieving pressure..ip the event of a major line rupture.

AMENDMENT NO. 345

5.5 Stora e of Unirradia ed and S en Fuel Unirradiated fuel assemblies will normally be stored in critically safe new fuel storage racks in the reactor building storage vault. Even when flooded with water, the resultant keff is less than 0.95. Fresh fuel may also be stored in shipping containers. The unirradiated fuel storage vault is designed and shall be maintained with a storage capacity limited to no more than 200 fuel assemblies.

The spent fuel storage facility is designed to maintain fuel in a geometry such that k<< is less than 0.95 under conditions of optimum water moderation. The spent fuel storage facility is designed and shall be maintained with a storage capacity limited to no more than 2776 fuel assemblies. Fuel assemblies stored in the 1066 spent fuel storage locations of the non-poison flux trap design are limited to 15.6 grams (3.0 weight percent) of Uranium-235 per axial centimeters of assembly. Fuel assemblies stored in the 1,710 spent fuel storage positions of the poison type which use Boraflex as the neutron absorber are limited to 18.13 grams (3.75 weight percent) of Uranium-235 per axial centimeters of assembly.

Calculations for keff values have been based on methods approved by the Nuclear Regulatory Commission covering special arrays (10CFR70.55).

The reactor building and all contained engineered safeguards are designed for the maximum credible earthquake ground motion with an acceleration of 11 percent of gravity. Dynamic analysis was used to determine the earthquake acceleration, applicable to the various elevations in the reactor building.

AMENDMENT NO. 346

6.0 ADMINISTRATIVECONTROLS-6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Station Shift Supervisor - Nuclear (or during his absence from. the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Executive Vice President - Nuclear shall be re-issued to station personnel on an annual basis.

Onsi e and Offsite Or anization 6.2.1 An onsite and an offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the position for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility and communication shall be established and defined from the highest manage-ment levels through intermediate levels to and including all operating organization positions. Those relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions or in equivalent forms of documentation. The organization charts shall be documented in the Final Safety Analysis Report, and the functional descriptions of departmental responsibilities and relationships and job descriptions for key personnel positions are documented in procedures.

b. The Executive Vice President - Nuclear shall have corporate responsibility for overall plant nuclear safety. The Executive Vice President - Nuclear shall take any measures needed to assure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

c. The Plant Manager shall have responsibility for overall unit operation and shall have control over those resources necessary for safe operation and maintenance of the plant.

AMENDMENT NO. 347

d. The persons responsible for the training, health physics and quality assurance functions may report to an appropriate manager onsite, but shall have direct access to responsible corporate management at a level where

=

action appropriate to the mitigation of training, health physics and quality assurance concerns can be accomplished.

Fa ili S aff 6.2.2 The unit organization shall be subject to the following:

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. During reactor operation, this licensed operator shall be present at the controls of the facility.

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.

A licensed Senior Reactor Operator shall be required in the Control Room during power operations, hot shutdown, and when the emergency plan is activated. This may be the Station Shift Supervisor - Nuclear or the Assistant Station Shift Supervisor - Nuclear or another Senior Reactor Operator during power operations or hot shutdown. When the emergency plan is activated during normal operations or hot shutdown, the Assistant Station Shift Supervisor - Nuclear becomes the Shift Technical Advisor and the Station Shift Supervisor - Nuclear is restricted to the control room until an additional licensed Senior Reactor Operator arrives.

f. A licensed Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling shall be responsible for all movement of new and irradiated fuel within the site boundary. All core alterations shall be directly supervised by a licensed..Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation. All fuel moves within the core shall be directly monitored by a member of the reactor analyst group.

The requirement for a Radiation Protection qualified individual may be less than the minimum requirement for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

AMENDMENT NO. 348

g. DELETED

h. Administrative procedures shall be developed and implemented to limit the working hours of facility staff who perform safety-related functions; e.g., licensed Senior Operators, licensed Operators, health physicists, auxiliary operators and key maintenance personnel.

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the facility is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications on a temporary basis, the following guidelines shall be followed:

1) An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight (excluding shift turnover time).

2) An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period (all excluding shift turnover time).

3) A break of at least 8-hours should be allowed between work periods (including shift turnover time).

4) Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the Plant Manager, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Vice President - Nuclear Generation or designee to assure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

The General Supervisor Operations, Supervisor Operations, Station Shift Supervisor Nuclear and Assistant Station Shift Supervisor Nuclear shall hold senior reactor operator licenses.

AMENDMENT NO. 349 .

TABLE 6.2-1 MINIMUMSHIFT CREW COMPOSITION(1)(6)

Shutdown Operation WIO License Normal Operation "Condition Process Computer Reactor Startups Senior Operator 1(5)

Operator 2(4)

Unlicensed 1 Asst. Station Shift Supervisor (Shift Technical 1(4)

Advisor Function) (Senior Operator License)( )

Notes:

(1) At any one time, more licensed or unlicensed operating people could be present for maintenance, repairs, refuel outages, etc.

(2) Those operating personnel not holding an "Operator" or "Senior Operator" License.

(3) For operation longer than eight hours without process computer.

(4) Hot shutdown condition only. For cold shutdown and refueling conditions, only one senior operator and one operator are required to be on shift.

(5) An additional Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities shall supervise all core alterations.

(6) The Shift Crew Composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

(7) .The Assistant Station Shift Supervisor performs the Shift Technical Advisor function when the emergency plan is activated during normal operations or hot shutdown and shall hold a senior reactor operator license. Normally, the Assistant Station Shift Supervisor is a combined Assistant Station Shift Supervisor/Shift Technical Advisor, however, there may be instances when a shift may be staffed by two Senior Reactor Operators plus a dedicated Shift Technical Advisor.

AMENDMENT NO. 350

6.3 Facilit S aff Qualifica ions 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Manager Radiation Protection who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and the Shift Technical Advisor who shall have a bachelor's degree in a physical science or engineering or a professional engineer license issued by examination and shall have received specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 T~rainin 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Manager Training-and shall meet or exceed the recommendations and requirements of Section 5.5 of ANSI N18.1-1971 and of 10CFR Part 55, and shall include familiarization with relevant industry operational experience.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Manager Training and Supervisor-Fire Protection, Nuclear and shall meet or exceed the requirements of Appendix R to 10CFR50.

d*

6.5.1 a ion 0 er ion R view Committee SORC Function 6.5.1.1 The Station Operations Review Committee shall function to advise the Plant Manager on all matters related to nuclear safety.

Com osition 6.5.1.2 The SORC shall be composed of the:

Chairman: Plant Manager Vice Chairman/Member: Manager Operations Vice Chairman/Member: Manager Technical Support Member: Manager QA Operations

~

Member: Manager Maintenance Member: Manager Chemistry Member: Manager Radiation Protection AMENDMENT NO. 351

A~lamae 6.5.1.3 All alternate members shall be appointed in writing by the SORC Chairman or Vice-Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in SORC activities at any one time.

Mee in Fr uenc 6.5.1.4 The SORC shall meet at least once per calendar month and as convened by the SORC Chairman, Vice-Chairman, or a designated alternate.

~Qer m 6.5.1.5 The quorum of the SORC necessary for the performance of the SORC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman, or a Vice-Chairman, and four members, including alternates.

Re on ibili ies 6.5.1.6 The SORC shall be responsible for:

a. Review of all REPORTABLE EVENTS.

b. Review of unit operations to detect potential safety hazards.

c. Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant Manager or the Safety Review and Audit Board.

d. Investigation of violations of the Technical Specifications and shall prepare and forward a report covering evaluation and recommendations to prevent recurrence to the Vice President - Nuclear Generation and to the Safety Review and Audit Board.

AMENDMENT NO. 352

~Authori 6.5.1.7 The SORC shall:

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

b. Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President - Nuclear Generation and the Safety Review and Audit Board of disagreement between the SORC and the Plant Manager; however, the Plant Manager shall have the responsibility for resolution of such disagreements pursuant to 6.1 1 above.

~

~Record 6.5.1.8 The SORC shall maintain written minutes of each meeting and copies shall be provided to the Vice President-Nuclear Generation and the Safety Review and Audit Board.

6.5.2 Technical Review and Con rol Activities 6.5.2.1 Each procedure and program required by Specification 6.8 and other procedures which affect nuclear safety, and changes thereto, shall be prepared by a qualified individual/organization. Each such procedure, and changes thereto, shall be reviewed by an individual/group other than the individual/group which prepared the procedure, or changes thereto, but who may be from the same organization as the individual/group which prepared the procedure, or changes thereto. Approval of procedures and programs and changes thereto and their safety evaluations, shall be controlled by administrative procedures.

6.5.2.2 Proposed changes to the Technical Specifications shall be prepared by a qualified individual/organization. The preparation of each proposed Technical Specifications change shall be reviewed by an individual/group other than the individual/group which prepared the proposed change, but who may be from the same organization as the individual/group which prepared the proposed change. Proposed changes to the Technical Specifications shall be approved by the Plant Manager.

AMENDMENT NO. 353

6.5.2.3 Proposed modifications to unit structures, systems and components that affect nuclear safety shall be designed by a qualified individual/organization. Each such modification shall be reviewed by an individual/group other than the individual/group which designed the modification, but who may be from the same organization as the individual/

group which designed the modification; Proposed modifications to structures, systems and components and the safety evaluations shall be approved prior to implementation by the Plant Manager; or the Manager Technical Support as previously designated by the Plant Manager.

6.5.2 4 Individuals responsible for reviews'performed in accordance with Specifications 6.5.2.1, 6.5.2.2 and 6.5.2.3 shall be members of the station supervisory staff, previously designated by the Plant Manager to perform such reviews.

Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary.

If deemed necessary such review shall be performed by the appropriate designated station review personnel.

6.5.2.5 Proposed tests and experiments which affect station nuclear safety and are not addressed in the FSAR or Technical Specifications and their safety evaluations shall be reviewed by the Plant Manager, or the Manager Technical Support as previously designated by the Plant Manager.

6.5.2.6 The Plant Manager shall assure the performance of special reviews and investigations, and the preparation and submittal of reports thereon, as requested by the Vice President - Nuclear Generation.

6.5.2.7 The facility security program, and implementing procedures, shall be reviewed at least every 12 months.

Recommended changes shall be approved by the Plant Manager and transmitted to the Vice President - Nuclear Generation and to the Chairman of the Safety Review and Audit Board.

6.5.2.8 The facility emergency plan, and implementing procedures shall be reviewed at least every 12 months.

Recommended changes shall be approved by the Plant Manager and transmitted to the Vice President - Nuclear Generation and to the Chairman of the Safety. Review and Audit Board.

AMENDMENT NO. 354

0 6.5.2.9 The Plant Manager shall assure the performance of a review by a qualified individual/organization of changes to the Radiological Waste Treatment systems.

6.5.2.10 Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President - Nuclear Generation and to the Safety Review and Audit Board.

~ ~

6.5.2.11 Review of changes to the Process Control Program and the Offsite Dose Calculation Manual. Approval of any changes shall be made by the Plant Manager or his designee before implementation of such changes.

6.5.2.12 Reports documenting each of the activities performed under Specifications 6.5.2.1 through 6.5.2.9 shall be maintained. Copies shall be provided to the Vice President - Nuclear Generation and the Safety Review and Audit Board.

6.5.2.13 The Plant Manager shall assure the performance of a review by a qualified individual/organization of the Fire Protection Program and implementing procedures at least every 12 months and submittal of recommended changes to the Safety Review and Audit Board.

6.5.3 Safe Review and Audit Board SRAB Function 6.5.3.1 The Safety Review and Audit Board shall function to provide independent review and audit of designated activities in the areas of:

80 nuclear power plant operations

b. nuclear engineering C. chemistry and radiochemistry

d. metallurgy

e. instrumentation and control

f. radiological safety g mechanical and electrical engineering

h. quality assurance practices (other appropriate fields associated with the unique characteristics of the nuclear power plant)

AMENDMENT NO. 355

Com osi ion 6.5.3.2 The Safety Review and Audit Board shall be composed of the:

Chairman: Staff Engineer or Manager or Vice President Member: Plant Manager or Designee Member: Staff Engineer - Nuclear Member: Staff Engineer - Mechanical or Electrical Member: Consultant (See 6.5.3.4)

~AI erna e 6.5.3.3 Alternate members shall be appointed in writing by the SRAB Chairman to serve on a temporary basis; however, no more than two alternates shall participate in SRAB activities at any one time.

Con~le 6.5.3.4 Consultants shall be utilized as determined by the SRAB Chairman to provide expert advice to the SRAB.

Meetin Fre uenc 6.5.3.5 The SRAB shall meet at least once per six months.

Quorum 6.5.3.6 The quorum of the SRAB necessary for the performance of the SRAB review and audit functions of these Technical Specifications shall consist of not less than a majority of the members, including alternates. The quorum requires the presence of the Chairman or the Chairman's designated alternate and no more than a minority of the quorum shall have line responsibility for operation of the facility.

AMENDMENT NO. 356

Review 6.5.3.7 The SRAB shall review:

a. - The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the 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 50.59, 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 Technical Specifications or operating license.

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

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

g. All REPORTABLE EVENTS.

h. Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems. or components.

i. Reports and meeting minutes of the SORC.

AMENDMENT NO. 357

~Audi s 6.5.3.8 Audits of facility activities shall be performed under the cognizance of the SRAB. These audits shall encompass:

a. The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license conditions at least once per year.

b. The performance, training and qualifications of the entire facility staff at least once per year.

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

d. The performance of all activities required by the Quality Assurance Program to meet the criteria of'Appendix "B", 10CFR50, at least once per two years.

e. The Facility Emergency Plan and implementing procedures at least once every 12 months.

f. The Facility Security Plan and implementing procedures at least once every 12 months.

g. The Facility Fire Protection Program and implementing procedures at least once per two years.

h. Any other area of facility operation considered appropriate by the SRAB or the Vice President - Nuclear Generation.

i. The radiological environmental monitoring program and the results thereof at least once per 12 months.

j. The Offsite Dose Calculation Manual and implementing procedures at least once per 24 months.

k. The Process Control Program'and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months.

AMENDMENT NO. 358

~AUthori 6.5.3.9 The SRAB shall report to and advise the Executive Vice President - Nuclear on those areas of responsibility specified in Section 6.5.3.7 and 6.5.3.8.

Records 6.5.3.10 Records of SRAB activities shall be prepared, approved and distributed as indicated below:

a. Minutes of each SRAB meeting shall be prepared, approved and forwarded to the Executive Vice President-Nuclear within 30 days following each meeting.

b. Reports of reviews encompassed by Section 6.5.3.7 e,f,g and h above, shall be prepared, approved and forwarded to the Executive Vice President - Nuclear within 14 days following completion of the review.

c. Audit repor>s encompassed by Section 6.5.3.8 above, shall be forwarded to the Executive Vice President-Nuclear within 90 days following completion of the review.

AMENDMENT NO. 359

6.6 Re or able 0 rr nce Action 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Sections 50.72 and 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the SORC and the results of this review submitted to the SRAB and the Vice President - Nuclear Generation.

6.7 Safe Limi Viola ion 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The provisions of 10 CFR 50.36(c)(1)(i) shall be complied with immediately.

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Vice President - Nuclear Generation and the SRAB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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

d. The Safety Limit Violation Report shall be submitted to the Commission, within 30 days of the violation, and to the SRAB, and the Vice President - Nuclear Generation within 14 days.

8.8 ~Procedure 6.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the "A" of USAEC Regulatory requirements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix Guide 1.33 except as provided in 6.8.2 and 6.8.3 below.-

a. Written procedures shall be established, implemented, and maintained for activities involving the Fire Protection Program implementation.

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

AMENDMENT NO. 360

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

a. The intent of the 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 and approved by the Plant Manager or designee within 14 days of implementation.

6.9 Re ortin Re iremen In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted in accordance with 10 CFR 50.4.

d>> i i hi~lb b i<<d of an operating license, (2) amendment to the license involving a planned increase power level, (3) installation of fuel has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have that significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days following completion 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.

AMENDMENT NO. 361

b. Annual Occu a ional Ex osure Re ort. A tabulation shall be submitted on an annual basis which includes the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job functions, 1/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe 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 totaling 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.

c. Mon hl 0 eratin Re or . Routine reports of operating statistics and shutdown experience including documentation of challenges to the safety relief valves or safety valves, shall be submitted on a monthly basis, which will include a narrative of operating experience, in accordance with 10 CFR 50.4, no later than the 15th of each month following the calendar month covered by the report.

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

AMENDMENT NO. 362

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d. nnual Radiolo ical Environmen al 0 era in Re or ".

Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1, 1985.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with operational controls as appropriate, and with environmental surveillance reports from the previous 5 years, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3.6.22.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the Table and Figures in the Offsite Dose Calculation Manual, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979..ln the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps"" covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.6.21; discussion of all deviations from the sampling schedule of Table 3.6.20-1; and discussion of all analyses in which the LLD required in Table 4.6.20-1 was not achievable.

A single submittal may be made for a multiple unit station.

One map shall cover stations near the site boundary; a second shall include the more distant stations.

AMENDMENT NO. 363

e. Semi-annual Radioac ive Effluent Release Re or ""

Routine Radioactive Effluent Release Reports covering the operation 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. The period of the first report shall begin on January 1, 1985.

The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1;21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability." This same report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figure 5.1-1) during the report period.

All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The assessment of. radiation doses shall be performed in accordance with the methodology and parameters in the Offsite Dose Calculation Manual.

The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in the Offsite Dose Calculation Manual.

In lieu of submission with the Semi-annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

A single submittal may be made for a multiple unit site. The submittal should combine those sections that are common to all units at the site; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

AMENDMENT NO. 364

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The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:

a. Container volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and,

f. Solidification agent or absorbent (e.g., cement)

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the Process Control Program (PCP) and to the Offsite Dose Calculation Manual (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.6.20.

Changes to the Process Control Program (PCP) shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:

80 Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

c. Documentation of the fact that the change has been reviewed and found acceptable.

AMENDMENT NO. 365

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Changes to the Offsite Dose Calculation Manual (ODCM): Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the Offsite Dose Calculation Manual to be changed, together with appropriate analyses or evaluations justifying the change(s);

b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c. Documentation of the fact that the change has been reviewed and found acceptable.

f. CORE OPERATIN LIMITS REPORT

1. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle for the following:

1) The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.1.7.a and 3.1.7.e.

2) The Kf core flow adjustment factor for Specification 3.1.7.c.

3) The MINIMUMCRITICAL POWER RATIO (MCPR) for Specification 3.1.7.c and 3.1.7.e.

4) The Maximum Total Peaking Factor (MTPF) value for each fuel bundle design utilized for the current fuel cycle for Specifications 2.1.2 and 3.6.2.

5) The LINEAR HEAT GENERATION RATE for Specification 3.1.7.b.

6) The Power/Flow relationship for Specification 3.1.7.d and e.

and shall be documented in the CORE OPERATING LIMITS REPORT.

2. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

AMENDMENT NO. 366

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1) NEDE-24011-P-A "GENERAL ELECTRIC STANDARD APPLICATION FOR REACTOR FUEL" (Latest approved revision).

2) NEDE-30966-P-A "SAFER MODEL FOR EVALUATIONOF LOSS-OF-COOLANT ACCIDENTS FOR JET PUMP AND NON-JET PUMP PLANTS" (Latest Approved Revisions)

Vol I "SAFER LONG TERM INVENTORY MODEL FOR BWR LOSS-OF-COOLANT ACCIDENT ANALYSIS" Vol II "SAFER APPLICATION METHODOLOGY FOR NON-JET PUMP PLANTS"

3) NEDO-20556-P-A "GENERAL ELECTRIC COMPANY ANALYTICALMODEL FOR LOSS-OF-COOLANT ACCIDENT ANALYSIS IN ACCORDANCE WITH 10CFR50 APPENDIX K". (Latest approved revision)

3. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

4. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

6.9.2 Fire Protection Pro ram Re orts Noncompliances with the Fire Protection Program (as described in the Final Safety Analysis Report) that adversely affect the ability to achieve and maintain safe shutdown in the event of a fire shall be reported in accordance with the requirements of 10CFR50.72 and 10CFR50.73.

AMENDMENT NO. 367

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6.9.3 Special Reports Special reports shall be submitted in accordance with 10 CFR 50.4 Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements. of the applicable reference specification:

Reactor Vessel Material Surveillance Specimen Examination, Specification 4.2.2(b) (12 months).

b. Safety Class 1 Inservice Inspection, Specification 4.2.6 (Three months).

c. Safety Class 2 Inservice Inspections, Specification 4.2.6 (Three months) ~

d. Safety Class 3 Inservice Inspections, Specification 4.2.6 (Three months).

e. Primary Containment Leakage Testing, Specification 3.3.3 (Three months).

f. Secondary Containment Leakage Testing, Specification 3.4.1 (Three months).

g. Sealed Source Leakage In Excess Of Limits, Specification 3.6.5.2 (Three months) ~

h. Calculate Dose from Liquid Effluent in Excess of Limits, Specification 3.6.15.a(2)(b) (30 days from the end of the affected calendar quarter).

Calculate Air Dose from Noble Gases Effluent in Excess of Limits, Specification 3.6.15.b(2)(b) (30 days from the end of the affected calendar quarter).

Calculate Dose from 1-131, H-3 and Radioactive Particulates with half lives greater than eight days in Excess of Limits, Specification 3.6.15.b(3)(b) (30 days from the end of the affected calendar quarter).

k. Calculated Doses from Uranium Fuel Cycle Source in Excess of Limits, Specification 3.6.15.d (30 days from the end of the affected calendar year) ~

Inoperable Gaseous Radwaste Treatment System, Specification 3.6.16.b (30 days from the event).

Environm n al R di lo ical Re or . With the level of radioactivity (as the result of plant effluents) in an environmental sampling medium exceeding the reporting level of Table 6.9.3-1, when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within thirty (30) days from the end of the calendar quarter a special report identifying the cause(s) for exceeding the limits, and define the corrective action to be taken.

AMENDMENT NO. 368

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TABLE 6.9.3-1 REPORTING LEVEL FOR RADIOACTIVITYCONCENTRATIONS IN ENVIRONMENTALSAMPLES REPORTING LEVELS Airborne Particulate Food Products Analysis Water (pCi/I) or Gases (pCi/m3) Fish (pCi/kg, wet) Milk (pCi/I) (pCi/kg, wet)

H-3 20,000%

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 'I,000 30,000 Co-60 300 ] 0,000 Zn-65 300 20,000 Zr-95, Nb-95 400 1-131 0.9 100 Cs-134 30 10.0 1,000 60 1,000 Cs-137 50 20.0 2,000 70 2,000 Ba/La-140 200 300 For drinking water samples. This is a 40 CFR 141 value. If no drinking water pathway exists, a value of 30,000 pCi/liter may be used.

If no drinking water pathway exists, a value of 20 pCi/liter may be used.

AMENDMENT NO. 369

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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.

c. REPORTABLE EVENT REPORTS.

d. 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.

h. Records of sealed source leak tests and results.

Records of annual physical inventory of all source material of record 6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a 0 Record 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.

AMENDMENT NO. 370

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e. Records of gaseous and liquid radioactive material released to the environs.

f. Records of transient 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 in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities required by the OA Manual.

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the SORC and the SRAB.

I. Records of analyses required by the'radiological environmental monitoring program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and Quality Assurance records showing that these procedures were followed.

6.11 Radiation Protection Pro ram 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.

6.12 Hi h Radia i n Ar a 6.12.1 In lieu of the "control device" or "alarm signal" required by Paragraph 20.203(c)(2) of 10CFR20, each high radiation area normally accessible" by personnel in which the intensity of radiation is greater than 100 mrem/hr"" but less than 1000

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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 in accordance with site approved procedures. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

AMENDMENT NO. 371

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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 rates in the area have been established and personnel have been made knowledgeable of them.

c. An individual qualified in radiation protection, with a radiatIon dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Manager Radiation Protection or designate in the Radiation Work Permit.

6.12.2 fn addition to the requirements of 6.12.1 areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem"" shall be provided with locked doors to prevent unauthorized entry, and the hard keys or access provided by magnetic keycard shall be maintained under the administrative control of the Station Shift Supervisor or designate on duty and/or the Manager Radiation Protection or designate. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify in accordance with site approved procedures accordingly, the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. ln lieu of the stay time specification of the RWP, continuous surveillance, direct or remote, such as use of closed circuit TV cameras, may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose in excess of 1000 mrem"" that are'ocated within large areas, such as the drywell, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device.

by accessible passage and permanently fixed ladders measurement made at 18" from source of radioactivity AMENDMENT NO. 372

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6.13 Fire Pro ec ion lns ec ion 6.13.1 An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified off-site licensee personnel or an outside fire protection firm.

6.13.2 An inspection and audit by an outside qualified fire consultant shall be performed at intervals no greater than 3 years.

~8 Procedure shall be established, implemented and maintained to meet or exceed the requirements and recommendations of Section 2.1.6.a of NUREG 0578.

Procedures shall be established, implemented and maintained to meet or exceed the requirements and recommendations of Section 2.1.8.c of NUREG 0578.

AMENDMENT NO. 373