Text

Proposed Tech Specs,Correcting Typos & Editorial Errors
	ML20034H195
Person / Time
Site:	FitzPatrick
Issue date:	03/09/1993
From:	; POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	;
Shared Package
ML20034H189	List: JPN-93-012, Forwards Application for Amend to License DPR-59 to Make Miscellaneous Administrative Changes Correcting Typos & Editorial Errors; ML20034H191; ML20034H195;
References
	NUDOCS 9303160152
	Download: ML20034H195 (82)
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' Attachment I to'JPN-93-012 '-

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' PROPOSED TECHNICAL SPECIFICATION CHANGES '

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MISCELLANEOUS ADMIN [EIB4IjyJ_CBANfaEE i

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New York Power Authority r

i JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No; 50-333 DPR-59 l

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JAFNPP 1

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Pace 1.0 Definitions 1

LIMIT!NG SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 Fuel Cladding Integrity 2.1 7

1.2 Reactor Coolant System 2.2 27 SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.0 General 4.0 30 3.1 Reactor Protection System 4.1 30f 3.2 Instrumentation 4.2 49 A.

Primary Containment isolation Functions A.

49 B.

Core and Containment Cooling Systems -

B.

50 Initiation and Control C.

Control Rod Block Actuation C.

50 D.

Radiation Monitoring Systems -Isolation D.

50 and initiation Functions E.

Drywell Leak Detection E.

54 F.

DELETED F.

54 l'

G.

Recirculation Pump Trip G.

54 H.

Accident Monitoring Instrumentation H.

54 l.

4kV Emergency Bus Undervoltage Trip 54 3.3 Reactivity Control 4.3 88 A.

Reactivity Limitations A.

88 B.

Control Rods B.

91 C.

Scram insertion Times C.

95 D.

Reactivity Anomalies D.

96 3.4 Standby Liquid Control System 4.4 105 A.

Normal Operation A.

105 B.

Operation With inoperable Components B.

106 C.

Sodium Pentaborate Solution C.

107 1

3.5 Core and Containment Cooling Systems 4.5 112 A.

Core Spray and LPC! Systems A.

112 B.

Containment Cooling Mode of the RHR B.

115 System C.

HPCI System C.

117 D.

Abtomatic Depressurization System (ADS)

D.

119 E.

Reactor Core isolation Cooling (RCIC)

E.

121 System Amendment No.

2,1 0,1 4,1 3,

JAFNPP LIST OF TABLES Teble Iltle Eana 3.1-1 Reactor Protection System (Scram) Instrumentation Requirement 41 3.1-2 Reactor Protection System Instrumentation Response Times 43a 4.1-1 Reactor Protection System (Scram) instrument Functional Tests 44 4.1-2 Reactor Protection System (Scram) instrument Calibration 46 3.2-1 Instrumentation that initiates Primary Containment isolation 64 3.2-2 Instrumentation that initiates or Controls the Core and Containment 66 Cooling Systems 3.2-3 instrumentation that initiates Control Rod Blocks 72 3.2-4 (DELETED) 74 3.2-5 Instrumentation that Monitors Leakage Detection inside the Drywell 75 3.2-6 (DELETED) 76 3.2-7 Instrumentation that initiates Recirculation Pump Trip 77 3.2-8 Accident Monitoring instrumentation 77a 3.2-9 Primary Containment isolation System Actuation instrumentation 770 Response Times i

4.2-1 Minimum Test and Calibration Frequency for PCIS 78 4.2-2 Minimum Test and Calibration Frequency for Core and Containment 79 Cooling Systems l

4.2-3 Minimum Test and Calibration Frequency for Control Rod Blocks 81 Actuation

?

4.2-4 (DELETED) 82 4.2-5 Minimum Test and Calibration Frequency for Drywell Leak Detection 83 4.2-6 (DELETED) 4.2-7 Minimum Test and Calibration Frequency for Recirculation Pump Trip 85 endment No. f,9,1 0,1 1,1 3,

JAFNPP 1.1 (cont'd) 2.1 (cont'd) b.

APRM Flux Scram Trio Settina (Refuel or Start B.

Core Thermal Power Limit (Reactor Pressure s785 osia)

& Hot Standbv Mode)

When the reactor pressure is $785 psig or core flow is APRM - The APRM flux scram setting shall be less than or equal to 10% of rated, the core thermal s15 percent of rated neutron flux with the power shall not exceed 25 percent of rated thermal Reactor Mode Switch in Startup/ Hot Standby power.

or Refuel.

C.

Power Transient c.

6PRM Flux Scram Trio Settinas (Run Mode)

To ensure that the Safety Limit established in Specification (1)

Flow Referenced Neutron Flux Scram Trip 1.1.A and 1.1.8 is not exceeded, each required scram Setting shall be initiated by its expected scram signal. The Safety Limit shall be assumed to be exceeded when scram is When the Mode Switch is in the RUN accomplished by a means other than the expected scram position, the APRM flow referenced flux signal.

scram trip setting shall be less than or equal to the limit specified in Table 3.1-1 This setting shall be adjusted during single loop operation when required by Specification 3.5.J.

For no combination of recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 117% of rated thermal power.

Amendment No.14,

,7,

,1 4, 1 2,

JAFNPP 2.1 BASES (Cont'd) l B.

Not Used C.

Referenceji 1.

(Deleted) 2.

" General Electric Standard Application for Reactor Fuel",

NEDE 24011-P-A (Approved revision number applicable at time that reload fuel analyses are performed).

3.

(Deleted) 4.

FitzPatrick Nuclear Power Plant Single-Loop Operation, NEDO-24281, August,1980.

l Amendment No. If%,4,, 9,1 2,

(Next page is 23)

~.

JAFNPP 1.2 and 2.2 BASES The reactor coolant pressure boundary integrity is an important ANSI Code permits pressure transients up to 20 percent over the barrier in the prevention of uncontrolled release of fission products.

design pressure (120% x 1,150 = 1,380 psig). The safety limit l

It is essential that the integrity of this boundary be protected by pressure of 1,375 psig is referenced to the lowest elevation of the establishing a pressure limit to be observed for all operating Reactor Coolant System.

l conditions and whenever there is irradiated fuel in the reactor vessel.

The current reload analysis shows that the main steam isolation valve closure transient, with flux scram, is the most severe event The pressure safety limit of 1,325 psig as measured by the vessel resulting directly in a reactor coolant system pressure increase. The steam space pressure indicator is equivalent to 1,375 psig at the.

reactor vessel pressure code limit of 1,375 psig, given in FSAR lowest elevation of the Reactor Coolant System. The 1,375 psig Section 4.2, is above the peak pressure produced by the event value is derived from the design pressures of the reactor pressure above. Thus, the pressure safety limit (1,375 psig) is well above vessel and reactor coolant system piping. The respective design the peak pressure that can result from reasonably expected pressures are 1250 psig at 575 F for the reactor vessel,1148 psig overpressure transients. (See current reload analysis for the curve at 568"F for the recirculation suction piping and 1274 psig at 575 produced by this analysis.) Reactor pressure is continuously for the discharge piping. The pressure safety limit was chosen as indicated in the control room during operation, the lower of the pressure transients permitted by the applicable design codes: 1965 ASME Boiler and Pressure Vessel Code, A safety limit is applied to the Residual Heat Removal System Section ill for pressure vessel and 1969 ANSI B31.1 Code for the (RHRS) when it is operating in the shutdown cooling mode. When reactor coolant system piping. The ASME Boiler and Pressure operating in the shutdown cooling mode, the RHRS is included in Vessel Code permits pressure transients up to 10 percent over the reactor coolant system.

l design pressure (110% x 1,250 = 1,375 psig) and the The numerical distribution of safety / relief valve setpoints shown in 2.2.1.B (2 @ 1090 psi,2 @ 1105 psi,7 @ 1140 psi) is justified by analyses described in the General Electric report NEDO-24129-1, Supplement 1, and assures that the structural acceptance criteria set forth in the Mark i Containment Short Term Program are satisfied.

Amendment No.

,1 4,

JAFNPP 3.1 BASES l A.

The reactor protection system automatically initiates a The outputs of the subchannels are combined in a 1 reactor scram to:

out of 2 logic; i.e., an input signal on either one or-both of the subchannels will cause a trip system trip.

1.

Preserve the integrity of the fuel cladding.

The outputs of the trip systems are arranged so that a trip on both systems is required to produce a 2.

Preserve the integrity of the Reactor Coolant reactor scram.

System.

This system meets the intent of IEEE-279 (1971) for 3.

Minimize the energy which must be absorbed Nuclear Power Plant Protection Systems. The following a loss of coolant accident, and prevent system has a reliability greater than that of a 2 out inadvertent criticality.

of 3 system and somewhat less than that of a 1 out of 2 system.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system With the exception of the average power range to perform its intended function even during periods when monitor (APRM) channel the intermediate range instrument channels may be out of service because of monitor (IRM) channels, the scram discharge volume, maintenance. When necessary, one channel may be made the main steam isolation valve closure and the inoperable for brief intervals to conduct required functional turbine stop valve closure, each subchannel has one tests and calibrations.

instrument channel. When the minimum condition for operation on the number of operable instrument The Reactor Protection System is of the dual channel type channels per untripped protection trip system is met (Reference subsection 7.2 FSAR). The System is made up or if it cannot be met and the affected protection trip of two independent trip systems, each having two system is placed in a tripped condition, the subchannels of tripping devices. Each subchannel has an effectiveness of the protection system is preserved.

input from at least one instrument channel which monitors a critical parameter.

' Three APRM instrument channels are provided for each protection trip system. APRM's A and E operate contacts in one subchannel and APRM's C and E operate contacts in the other -

Amendment No. f,

JAFNPP 3.2 (cont'd) 4.2 (cont'd)

E.

Drvwell Leak Detection E.

Drvwell Leak Detection The limiting conditions of operation for the instrumentation Instrumentation shall be calibrated and checked as indicated that monitors drywell leak detection are given in Table in Table 4.2-5.

l 3.2-5.

F.

(Deleted)

F.

(Deleted)

G.

Recirculation Pumo Trio G.

Recirculation Pumo Trio The limiting conditions for operation for the instrumentation Instrumentation shall be functionally tested and calibrated as that trip (s) the recirculation pumps as a means of limiting indicated in Table 4.2-7.

the consequences of a failure to scram during an anticipated transient are given in Table 3.2-7.

System logic shall be functionally tested as indicated in Table 4.2-7.

H.

Accident Monitorina Instrumentation H.

Accident Monitorina Instrumentation The limiting conditions for operation of the instrumentation that provides accident monitoring are given in Table 3.2-8.

Instrumentation shall be demonstrated operable by performance of a channel check and channel calibration as 1.

4kv Emeraency Bus Undervoltaae Trio indicated in Table 4.2-8.

The limiting conditions for operation for the instrumentation that prevents damage to electrical equipment or circuits as a result of either a degraded or loss-of-voltage condition on the emergency electrical buses are given in Table 3.2-2.

Amendment No. If6, If0, If1, 54

JAFNPP TABLE 4.2-1 MINIMUM TEST AND CAllBRATION FREQUENCY FOR PCIS Instrument Channel (8)

Instrument Functional Test Calibration Frequency Instrument Check (4) 1)

Reactor High Pressure (1)

Once/3 months None (Shutdown Cooling Permissive) 2)

Reactor Low-Low-Low Water Level (1)(5)

(15)

Once/ day 3)

Main Steam High Temp.

(1)(5)

(15)

Once/ day 4)

Main Steam High Flow (1)(5)

(15)

Once/ day 5)

Main Steam Low Pressure (1)(5)

(15)

Once/ day 6)

Reactor Water Cleanup High Temp.

(1)

Once/3 months None 7)

Condenser Low Vacuum (1)(5)

(15)

Once/ day

' Logic System Functional Test (7) (9)

Frequency 1)

Main Steam Line Isolation Valves Once/6 months

.l Main Steam Line Drain Valves Reactor Water Sample Valves 2)

RHR -Isolation Valve Control Once/6 months Shutdown Cooling Valves 3)

Reactor Water Cleanup Isolation Once/6 months 4)

Drywell isolation Valves Once/6 months TIP Withdrawal Atmospheric Control Valves 5)

. Standby Gas Treatment System Once/6 months Reactor Building isolation NOTE: See notes following Table 4.2-5.

Amendment No.

,1 6,1 1,1 2,

-r JAFNPP.

TABLE -4.2-2

~

MINIMUM TEST AND CAllBRATION FREQUENCY FOR CORE AND CONTAINMENT COOLING SYSTEMS Instrument Channel Instrument Functional Test '

Calibration Frequency -

Instrument Check (4)..

l1 1)

Reactor Water Level.

(1)(5)

(15)

Once/ day 2a)

.Drywell Pressure (non-ATTS)

-(1)

Once/3 months

'None 2b)

Drywell. Pressure (ATTS)

(1)(5)

-(15)

Once/ day -

3a)

Reactor Pressure (non-ATTS)

(1)

Once/3 months None 3b)

Reactor Pressure (ATTS)

(1)(5)

(15)

Once/ day.

4).

Auto Sequencing Timers None Once/ operating cycle None 5)

- ADS - LPCI or CS Pump Disch.

(1)

Once/3 months None 6)

Trip System Bus Power Monitors (1)

None None.

8)

-Core Spray Sparger d/p (1)

- Once/3 months

. Once/ day.

9)

Steam Line High Flow (HPCI & RCIC)

(1)(5)

(15)

Once/ day -

10)

Steam Line/ Area High Temp.(HPCI & RCIC)

(1)(5).

(15)

Once/ day 12)

HPCI & RCIC Steam Line Low Pressure (1)(5)

(15)

Once/ day.

13)

HPCI & RCIC Suction Source Levels (1)

Once/3 months None 14) 4kV Emergency Bus Under-Voltage Once/ operating cycle Once/ operating' cycle None (Loss-of-Voltage, Degraded Voltage LOCA and non-LOCA) Relays and Timers.

15) ~

HPCI & RCIC Exhaust Diaphragm (1)

Once/3 months

.None' Pressure High 17)

LPCl/ Cross Connect Valve Position Once/ operating' cycle None L None' NOTE:

See notes following Table 4.2-5.

Amendment No.1f,

,f,'1f6,If,If0,1 1

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l JAFNPP 3.5 (cont'd) 4.5 (cont'd) condition, that pump shall be considered inoperable for 2.

Following any period where the LPCI subsystems or core purposes of satisfying Specifications 3.5.A 3.5.C, and 3.5.E.

spray subsystems have not been maintained in a filled condition; the discharge piping of the affected subsystem H.

Averaae Planar Linear Heat Generation Rate (APLHGR) shall be vented from the high point of the system and water flow observed.

During power operation, the APLHGR for each type of fuel as a function of axiallocation and average planar exposure shall 3.

Whenever the HPCI or RCIC System is lined up to take be within limits based on applicable APLHGR limit values suction from the condensate storage tank, the discharge which have been approved for the respective fuel and lattice piping of the HPCI or RCIC shall be vented from the high types. These values are specified in the Core Operating Limits point of the system, and water flow observed on a l

Report. If at anytime during reactor power operation greater monthly basis, than 25% of rated power it is determined that the limiting value for APLHGR is being exceeded, action shall then be 4.

The level switches located on the Core Spray and RHR initiated within 15 minutes to restore operation to within the System discharge piping high points which monitor these prescribed limits. If the APLHGR is not returned to within the lines to insure they are full shall be functionally tested prescribed limits within two (2) hours, an orderly reactor each month.

power reduction shall be commenced immediately. The reactor power shall be reduced to less than 25% of rated H. Averace Planar Linear Heat Generation Rate (APLQ power within the next four hours, or until the APLHGR is returned to within the prescribed limits.

The APLHGR for each type of fuel as a function of average planar exposure shall be determined daily during reactor operation at 2:25% rated thermal power.

,7f,,,19,1f7,1f2,1 Amendment No.

4,1 2,

JAFNPP 3.6 LIMITING CONDITIONS FOR OPERATION 4.6 SURVEILLANCE REQUIREMENTS

~

3.6 REACTOR COOLANT SYSTEM 4.6 REACTOR COOLANT SYSTEM Acolicability:

Acolicability:

Applies to the operating status of the Reactor Coolant System.

Applies to the periodic examination and testing requirements for the

- Reactor Coolant System.

Objective:

Obiective:

To assure the integrity and safe operation of the Reactor Coolant To determine the condition of the Reactor Coolant System and the System.

operation of the safety devices related to it.

Soecification:

A.

Pressurization and Thermal Limita A.

Pressurization and Thermal Limits 1.

Reactor Vessel Head Stud Tensioning 1.

Ronctor Vessel Head Stud Tensioning The reactor vessel head bolting studs shall not be under When in the cold condition, the reactor vessel head tension unless the temperatures of the reactor vessel

'lange and the reactor vessel flange temperatures shall be flange and the reactor head fienge are at least 90 F.

recorded:

a.

Every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactcr vessel head flange is s120 F and the studs are tensioned.

b.

Every 30 minutes when the reactor vessei head flange is s100 F and the studs are tensioned.

c.

Within 30 minutes prior to and every 30 minutes during tensioning of reactor vessel head botting studs.

2.

in-Service Hydrostatic and Leak Tests 2.

In-Service Hydrostatic and Leak Tests During in-service hydrostatic or leak testing the Reactor During hydrostatic and leak testing the Reactor Coolant Coolant System pressure and temperature shall be on or System pressure and temperature shall be recorded every to the right of curve A shown in Figure 3.6-1 Part 1,2, 30 minutes until two consecutive temperature readings or 3 and the maximum temperature change during any are within 5 F of each other.

. one hour period shall be:

, If3,1f8, Amendment No.1 136

JAFNPP 3.6 (cont'd) 4.6 (cont'd) 7.

Reactor Vessel Flux Monitoring The reactor vessel Flux Monitoring Surveillance Program complies with the intent of the May,1983 revision to 10 CFR 50, Appendices G and H. The next flux monitoring surveillance capsule shall be removed after 15 effective full power years (EFPYs) and the test procedures and reporting requirements shall meet the requirements of ASTM E 185-82.

B.

Deleted B.

Deleted C.

Coofant Chemistry C. Cociant Chemistry

1. The reactor coolant system radioactivity

1. a. A sample of reactor coolant shall be taken at least concentration in water shall not exceed the every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and analyzed for gross gamma activity.

[

equilibrium value of 3.1 pCi/gm of dose equivalent I-131. This limit may be exceeded, following a power

b. Isotopic analysis of a sample of reactor coolant shall be l

transient, for a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . During this made at least once/ month.

iodine activity transient the iodine concentrations shall not exceed the equilibrium limits by more than a

c. A sample of reactor coolant shall be taken prior to factor of 10 whenever the main steamline isolation startup and at 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months intervals during startup and l

valves are open. The reactor shall not be operated analyzed for gross gamma activity.

more than 5 percent of its annual power operation under this exception to the equilibrium limits. If the

d. During plant steady state operation end following an iodine concentration exceeds the equilibrium limit by offgas activity increase (at the Steam Jet Air Ejectors) more than a factor of 10, the reactor shall be placed of 10,000 pCi/sec within a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period or a power l

l in a cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

level change of 2:20 percent of full rated power /hr reactor coolant samples shall be taken and analyzed for gross gamma activity. At least three samples will be takers at 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months intervals. These sampling requirements l

may be omitted whenever the equilibrium I-131 concentration in the reactor coolant is less than 0.007 pCi/ml.

Amendment No.1 9, 139

JAFNPP 4.6 (cont'd) e.

If the gross activity counts made in accordance with a, c, and d above indicate a totaliodine concentration in excess of 0.007 yCi/ml, a quantative determination shall be made for 1-131 and 1-133.

2.

The reactor coolant water shall not exceed the following 2.

During startups and at steaming rates below 100,000 limits with steaming rates less than 100,000 lb/hr except Ib/hr, and when the conductivity of the reactor coolant as specified in 3.6.C.3:

exceeds 2 pmhos/cm, a sample of reactor coolant shall be Conductivity 2 mho/cm taken every 4 hr and analyzed for conductivity and Chloride ion 0.1 ppm chloride content.

3.

For reactor startups the maximum value for conductivity 3.

a.

With steaming rates greater than or equal to shall not exceed 10 mho/cm and the maximum value 100,000 lb/hr, a reactor coolant sample shall be for chloride ion concentration shall not exceed 0.1 ppm, taken at least every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and whenever the l

l for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after placing the reactor in the continuous conductivity monitors indicate abnormal power operating condition. During reactor shutdowns, conductivity (other than short-term spikes), and specification 3.6.C.4 will apply.

analyzed for conductivity and chloride ion content.

b.

When the continuous conductivity monitor is inoperable, a reactor coolant sample shall be taken at least daily and analyzed for conductivity and chloride ion content.

f, Amendment No.

140

~.

JAFNPP 4.6 (cont'd)

~

3.6 (cont'd)

F.

Structural Inteority F.

Structural Intearity 1.

Nondestructive inspections shall be performed on the The structural integrity of the Reactor Coolant System shall be ASME Boiler and Pressure Vessel Code Class 1,2 and 3 maintained at the level required by the original acceptance components and supports in accordance with the standards throughout the life of the Plant.

requirements of the weld and support inservice inspection program. This inservice inspection program is based on an NRC approved edition of, and addenda to,Section XI of the ASME Boiler and Pressure Vessel Code which is in effect 12 months or less prior to the beginning of the inspection interval.

2.

An augmented inservice inspection program is required for those high stressed circumferential piping joints in the main steam and feedwater lines larger than 4 inches in diameter, where no restraint against pipe whip is provided.

The augmented in-service inspection program shall consist of 100 percent inspection of these we!ds per inspection interval.

3.

An Inservice inspection Program for piping identified in the NRC Generic Letter 88-01 shall be implemented in accordance with NRC staff positions on schedules, l

methods, personnel, and sample expansion included in this l

Generic Letter, or in acordance with alternate measures approved by the NRC staff.

i l

l G. Jet Pumos G. Jet Pumos l

Whenever the reactor is in the startup/ hot standby or run Whenever there is recirculation flow with the reactor in the modes, all jet pumps shall be operable. If it is determined that a startup/ hot standby or run modes, jet pump operability shall be j

jet pump is inoperable, the reactor shall be placed in a cold checked daily by verifying that the following conditions do not j

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

occur simultaneously:

Amendment No. f, if4,1f0, 144 l -

JAFNPP 3.6 and 4.6 BASES (cont'd)

B.

Deleted annunciating at appropriate concentration levels such that sampling for isotopic analysis can be initiated. The design details of such a system must be submitted for evaluation and C.

Coolant Chemistry accepted by the Commission prior to its implementation and incorporation in these Technical Specifications.

A radioactivity concentration limit of 20 pCi/mi total iodine can be reached if the gaseous effluents are near the limit as Since the concentration of radioactivity in the reactor coolant is set forth in Radiological Effluent Technical Specification not continuously measured, coolant sampling would be Section 3.2.a if there is a failure or a prolonged shutdown of ineffective as a means to rapidly detect gross fuel element the cleanup demineralizer.

failures. However, some capability to detect gross fuel element failures is inherent in the radiation monitors in the offgas system in the event of a steam line rupture outside the drywell, with and on the main steam lines.

this coolant activity level, the resultant radiological dose at the site boundary would be 33 rem to the thyroid, under adverse Materials in the Reactor Coolant System are primarily 304 meteorological conditions assuming no more than 3.1 Ci/gm stainless steel and Zircaloy fuel cladding. The reactor water of dose equivalent 1-131. The reactor water sample will be chemistry limits are established to prevent damage to these used to assure that the limit of Specification 3.6.C is not materials. Limits are placed on chloride concentration and exceeded. The total radioactive iodine activity would not be conductivity. The most important limit is that placed on l

expected to change rapidly over a period of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , in chloride concentration to prevent stress corrosion cracking of addition, the trend of the stack offgas release rate, which is the stainless steel. The attached graph, Fig. 4.6-1, i!!ustrates continuously monitored, is a good indicator of the trend of the the results of tests on stressed 304 stainless steel specimens, iodine activity in the reactor coolant. Also during reactor Failures occurred at concentrations above the curve; no failures startups and large power changes which could affect iodine occurred at concentrations below the curve. According to the levels, samples of reactor coolant shall be analyzed to insure data, allowable chloride concentrations could be set several iodine concentrations are below allowable levels. Analysis is orders of magnitude above the established limit, at the oxygen required whenever the 1-131 concentration is within a factor concentration (0.2-0.3 ppm) experienced during power of 100 of its allowable equilibrium value. The necessity for operation. Zircaloy does not exhibit similar stress corrosion continued sampling following power and offgas transients will failures.

be reviewed within 2 years of initial plant startup.

However, there are various conditions under which the The surveillance requirements 4.6.C.1 may be satisfied by a dissolved oxygen content of the reactor coolant water could be continuous monitoring system capable of determining the total higher than 0.2-0.3 ppm, such as refueling, reactor startup, and iodine concentration in the coolant on a real time basis, and hot standby. During these periods with steaming rates less Amendment No.1 9, 149

JAFNPP 3.6 and 4.6 BASES (cont'd) than 100,000 lb/hr, a more restrictive limit of 0.1 ppm has During startup periods, which are in the category of less than been established to assure the chloride-oxygen combinations 100,000 lb/hr, conductivity may exceed 2 mho/cm because of of Fig. 4.6-1 are not exceeded. At steaming rates of at least the initial evolution of gases and the initial evolution of gases 100,000 lb/hr, boiling occurs causing deaeration of the reactor and the initial addition of dissolved metals. During this period of water, thus maintaining oxygen concentration at low levels.

time, when the conductivity exceeds 2 mho/cm (other than short-term spikes), samples will be taken to assure the chloride When conductivity is in its proper normal range, pH and concentration is less than 0.1 ppm.

chloride and other impurities affecting conductivity must also be within their normal ranges. When and if conductivity The conductivity of the reactor coolant is continuously becomes abnormal, then chloride measurements are made to monitored. The samples of the coolant which are taken every determine whether or not they are also out of their normal 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months will serve as a reference for calibration of these l

operating values. This is not necessarily the case.

monitors and is considered adequate to assure accurate Conductivity could be high due to the presence of a neutral readings of the monitors. If conductivity is within its normal salt; e.g., Na SO, which would not have an effect on pH or range, chlorides and other impurities will also be within their 2

chloride. In such a case, high conductivity alone is not a normal ranges. The reactor coolant samples will also be used to cause for shutdown. In some types of water-cooled reactors, determine the chlorides. Therefore, the sampling frequency is conductivities are,in fact, high due to purposeful addition of considered adequate to detect long-term changes in the chloride additives. In the case of BWR's, however, where no additives ion content. Isotopic analyses of the reactor coolant required are used and where neutral pH is maintained, conductivity by Specification 4.6.C.1 may be performed by a gamma scan.

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 D. Coolant Leakaoe condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor Allowable leakage rates of coolant from the Reactor Coolant coolant, are exceeded. Methods available to the operator for System have been based on the predicted and experimentally correcting the condition include operation of the Reactor observed behavior of cracks in pipes and on the ability to make Cleanup System, reducing the input of impurities and placing up Reactor Coolant System leakage in the event of loss of the reactor in the cold shutdown condition. The major benefit off-site a-c power. The normally expected background leakage of cold shutdown is to reduce the temperature dependent due to equipment design and the detection capability for corrosion rates and provide time for the Reactor Water determining system Cleanup System to reestablish the purity of the reactor coolant.

If9, Amendment No.

150

. ~. - -.

JAFNPP l

3.6 and 4.6 BASES (cont'd) loakage were also considered in establishing the limits. The The capacity of the drywell sump pumps is 100 gpm, and the behavior of cracks in piping systems has been experimentally capacity of the drywell equipment drain tank pumps is also 100 and analytically investigated as part of the USAEC-sponsored gpm. Removal of 50 gpm from either of these sumps can be Reactor Primary Coolant System Rupture Study (the Pipe accomplished with considerable margin.

Rupture Study). Work utilizing the data obtained in this study indicates that leakage from a crack can be detected before the The performance of the Reactor Coolant Leakage Detection crack grows to a dangerous or critical size by mechanically or System will be evaluated during the first 5 years of plant l

thermally induced cyclic loading, or stress corrosion cracking operation, and the conclusions of this evaluation will be or some other mechanism characterized by gradual crack reported to the NRC.

growth. This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the It is estimated that the main steam line tunnel leakage detectors probability is small that imperfections or cracks acsociated are capable of detecting a leak on the order of 3,500 lb/hr. The with such leakage would grow rapidly. However, the system performance will be evaluated during the first 5 years of l

establishment of allowable unidentified leakage greater than plant operation, and the conclusions of the evaluation will be that given in 3.6.D, on the Puis of the data presently reported to the NRC.

available would be premature because of uncertainties associated with the data. For leakage of the order of 5 gpm The reactor coolant leakage detection systems consist of the as specified in 3.6.D, the experimental and analytical data drywell sump monitoring system and the drywell continuous suggest a reasonable margin of safety such that leakage of atmosphere monitoring system. The drywell continuous this magnitude would not result from a crack approaching the atmosphere monitoring system utilizes a three-channel monitor critical size for rapid propagation. Leakage less than the to provide information on particulate, iodine and noble gas magnitude specified can be detected reasonably in a matter of activities in the drywell atmosphere. Two independent and a few hours utilizing the available leakage detection schemes, redundant systems are provided to perform this function. This and if the origin cannot be determined in a reasonably short system supplements the drywell sump monitoring system in time, the Plant should be shut down to allow further detecting abnormal leakage that could occur from the reactor investigation and corrective action.

Coolant system. In the event that the drywell continuous atmosphere monitoring system is inoperable, grab sample will be taken on a periodic basis to monitor drywell activity.

Amendment No. f, 151

JAFNPP

~

3.7 LIMITING CONDITIONS FOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Acolicability:

Acolicability:

Applies to the operating status of the primary and secondary Applies to the primary and secondary containment integrity.

containment systems.

Obiective:

To assure the integrity of the primary and secondary containment To verify the integrity of the primary and secondary containment l.

systems.

Soecification:

A.

Primary Containmerit A. Primary Containment 1.

The volume and temperature of the water in the torus 1.

The torus water level and temperature shall be monitored shall be maintained within the following limits whenever as specified in Table 4.2-8. The accessible interior the reactor is critical or whenever the reactor coolant surfaces of the drywell and above the water line of the temperature is greater than 212'F and irradiated fuel is torus shall be inspected at each refueling outage for in the reactor vessel:

evidence of deterioration. Whenever there is indication of relief valve operation or testing which adds heat to the a.

Maximum vent submergence level of 53 inches.

suppression pool, the pool temperature shall be continually monitored and also observed and logged every 5 minutes b.

Minimum vent submergence level of 51.5 inches, until the heat addition is terminated. Whenever there is indication of relief valve operation with the temperature of The torus water level may be outside the above the suppression pool reaching 160"F or more and the limits for a maximum of four (4) hours during primary coolant system pressure greater than 200 psig, an required operability testing of HPCI, RCIC, RHR, external visual examination of the torus shall be CS, and the Drywell-Torus Vacuum System.

conducted before resuming power operation.

c.

Maximum water temperature (1)

During normal power operation maximum water temperature shall be 95 F.

8,1f1, Amendment No.

,1 165

JAFNPP 4.7 (cont'd)

Type A test shall be performed at each plant shutdown for refueling or approximately every 18 months, whichever occurs first, until two consecutive Type A.

tests meet the acceptance criteria.

l

b. Type B tests (Localleak rate testing of containment penetrations)

(1.) All preoperational and periodic Type B tests shall be performed by local pneumatic pressurization of the containment penetrations, either individually or in groups, at a pressure not less than Pa, and the gas flow to maintain Pa shall be measured.

(2.) Acceptance criteria The combined leakage rate of all penetrations and valves subject to Type B and C tests shall be less than O.60 La, with the exception of the valves sealed with fluid from a seal system.

Amendment No. If5, 170

JAFNPP 4.7 (cont'd)

(5)

Type C test.

Type C tests shall be performed during each reactor shutdown for refueling but in no case at intervals greater than two years.

l (6)

Other leak rate tests specified in Section 4.7d shall be performed during each reactor shutdown for refueling but in no case at intervals greater than two years.

f.

Containment modification l

Any major modification, replacement of a component which is part of the primary reactor containment boundary, or resealing a seal-welded door, performed after the preoperational leakage rate test shall be followed by either a Type A, Type B, or Type C test, as applicable, for the area affected by the modification. The measured leakage from this test shall be included in the test report. The acceptance criteria as appropriate, shall be met. Minor modifications, replacements, or rescaling of seal-welded doors, performed directly prior to the conduct of a scheduled Type A test do not require a separate test.

4,1f0, Amendment No.

,1 5,1

JAFNPP 3.7 BASES A. Primary Containment The integrity of the primary containment and operation of the The pressure suppression pool water provides the heat sink for Emergency Core Cooling Systems in combination limit the the Reactor Coolant System energy release following a offsite doses to values less than those specified in 10 CFR 100 postulated rupture of the system. The pressure suppression in the event of a break in the Reactor Coolant System piping.

chamber water volume must absorb the associated decay and Thus, containment integrity is required whenever the potential structural sensible heat released during reactor coolant system for violation of the Reactor Coolant System integrity exists.

blowdown from 1,020 psig.

Concem about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception to the Since all of the gases in the drywell are purged into the pressure requirement to maintain primary containment integrity is suppression chamber air space during a loss of coolant allowed during core loading and during low power physics accident, the pressure resulting from isothermal compression testing when ready access to the reactor vesselis required.

plus the vapor pressure of the liquid must not exceed 56 psig, There will be no pressure on the system at this time, which will the suppression chamber design pressure. The design volume greatly reduce the chances of a pipe break. The reactor may be of the suppression chamber (water and air) was obtained cy taken critical during this period, however, restrictive operating considering that the total volume of reactor coolant to be procedures and operation of the RWM in accordance with condensed is discharged to the suppression chamber and that Specification 3.3.B.3 minimize the probability of an accident the drywell volume is purged to the suppression chamber occurring. Procedures in conjunction with the Rod Worth (updated FSAR Section 5.2).

l Minimizer Technical Specifications limit individual control worth such that the drop of any in-sequence control rod would not result in a peak fuel enthalpy greater than 280 calories /gm. In the unlikely event that an excursion did occur, the reactor building and Standby Gas Treatment System, which shall be operational during this time, offers a sufficient barrier to keep offsite doses well within 10 CFR 100.

Amendment No. If,15,

.187

JAFNPP 3.7 BASES (cont'd) complete containment system, secondary containment is be replaced whenever significant changes in filter officiency required at all times that primary containment is required as occur. Tests (11) of impregnated charcoal identical to that used well as during refueling.

in the filters indicated that shelf life up to 5 yr leads to only minor decreases in methyl iodine removal efficiency.

The Standby Gas Treatment System is designed to filter and exhaust the reactor building atmosphere to the main stack The 99 percent efficiency of the charcoal and particulate filters during secondary containment isolation conditions with a is sufficient to prevent exceeding 10CFR100 guidelines for the minimum release of radioactive materials from the reactor accidents analyzed. The analysis of the loss-of-coolant accident building to the environs. Both standby gas treatment fans are assumed a charcoal filter efficiency of 90 percent, and TID designed to automatically start upon containment isolation; 14844 fission product source term. Hence, requiring 99 however, only one fan is required to maintain the reactor percent efficiency for both the charcoal and particulate filters building pressure at approximately a negative 1/4 in. water provides adequate margin. A heater maintains relative humidity gage pressure; allleakage should be in-leakage. Each of the below 70 percent in order to assure the efficient removal of two fans has 100 percent capacity..lf one Standby Gas methyl iodine on the impregnated charcoal filters.

Treatment System circuit is inoperable, the other circuit must l

be verified operable daily. This substantiates the availability The operability of the Standby Gas Treatment System (SGTS) of the operable circuit and results in no added risk: thus, must be assured if a design basis loss of coolant accident reactor operation or refueling operation can continue. If (LOCA) occurs while the containment is being purged or vented neither circuit is operable, the Plant is brought to a condition through the SGTS. Flow from containment to the SGTS is via 6 where the system is not required.

inch Valve N"mber 27MOV-121. Since the maximum flow through tne 6 inch line following a design basis LOCA is within While only a small amount of particulates is released form the the desi;,n capabilities of the SGTS, use of the 6 inch line Pressure Suppression Chamber System as a result of the assures the operability of the SGTS.

loss-of-coolant accident, high-efficiency particulate filters are specified to minimize potential particulate release to the D. Primary Containment Isolation Valves environment and to prevent clogging of the iodine filter. The high-efficiency filters have an efficiency greater than 99 Double isolation valves are provided on lines penetrating the percent for particulate matter larger than 0.3 micron. The primary containment and open to the free space minimum iodine removal efficiency is 99 percent. Filter banks will Amendment No. If4, 191

JAFNPP

- 3.9 (cont'd) 4.9 (cont'd)-

.3.

From and after the time that one of the Emergency 3.

The emergency diesel generator system instrumentation -

Diesel Generator Systems is made or found to be shall be checked during the monthly generator test,-

inoperable, continued reactor operation is permissible for a period not to exceed 7 days provided that the two

. incoming power sources are available and that the remaining Diesel Generator System is operable. At the '

l end of the 7 day period, the reactor shall be placed in a cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless the affected diesel generator system is made operable sooner.

4 4.

When both Emergency Diesel Generator. Systems are 4.

Once each operating' cycle, the conditions under which the --

made or found to be inoperable, a reactor shutdown-Emergency Diesel Generator System is required will be shall be initiated within two hours and the reactor placed simulated to demonstrate that the pair of diesel generators'.

in a cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation of will start, accelerate, force parallel, and accept the.

shutdown.

emergency loads in the prescribed sequence.

5.

.Once within one hour and at least once per twenty-four hours thereafter while the reactor is being operated in accordance with Specifications 3.9.B.1, 3.9.B.2/ or 3.9.B.3 the availability of the operable Emergency Diesel Generators shall be demonstrated by manual starting and' force paralleling where applicable.-

O l

f

'217

-......... _... -. =., - - -.... -. - -. -. -. -

,...-w.-.-a.-..-_----

=. - -. -

JAFNPP-3.9 (cont'd)

F.

LPCI MOV Indeoendent Power Supolies

1. ~ Reactor shall not be made critical unless both independent power supplies, including the batteries, inverters and chargers and their associated buses (MCC-155 and MCC-165) are in service, except as specified below.

2. - During power operation, if one independent power supply becomes unavailable, repairs shall be made immediately and continued reactor operation is permissible for a period not to exceed 7 days unless the unavailable train is made ope 9ble sooner. From and after the date one of the independent power supplies is made or found to be inoperable for any reason, the following would apply:

a.

The other independent power supply including its charger, inverter, battery and associated bus is operable.

b.

Pilot cell voltage, specific gravity and temperature and overall battery voltage are measured immediately and weekly thereafter for the operable independent power supply battery.

c.

The inoperable independent power supply shall be isolated from its associated LPCI VOV bus, and this l

bus will be manually switched to its alternate power source.

Amendment No.

JAFNPP 3.9 BASES (cont'd)

C. Diesel Fuel E.

Batterv System l

Minimum on-site fuel oil requirements are based on operation of 125 v DC power is supplied from two plant batteries each sized the emergency diesel generator systems at rated load for 7 to supply the required equipment at design power following a

days, loss-of-coolant accident with a concurrent loss of normal and reserve power. Each battery is provided with a charger sized to Additional diesel fuel can be delivered to the site within 48 maintain the battery in a fully charged state while supplying hours.

normal operating loads.

If one of the Emergency Diesel Generator Systems is not F.

LPCI MOV Indeoendent Power Sucolies l

operable, the plant shall be permitted to run for 7 days provided both sources of reserve power are operational. This is based on There are two LPCI MOV Independent Power Supplies each the following:

consisting of a charger, rectifier, inverter and battery. Each independent power supply charger-rectifier is normally fed from 1.

The operable Emergency Diesel Generator System is capable the emergency A-C power supply system to maintain the of carrying sufficient engineered safeguards and emergency battery in a fully charged state. In the event of a LOCA each core cooling system equipment to cover allloss-of-coolant independent power supply is automatically isolated from the accidents.

Emergency A-C power system. The battery and inverter have sufficient capacity to power the MOV's essential to the 2.

The reserve (offsite) power is highly reliable, operation of the LPCI System. An alternate power source is l

provided for each LPCI MOV bus whereby in the event its l

D.

Not Used independent power supply is out of service, the LPCI MOV bus may be energized directly from the Emergency A-C Pcwer System.

Amendment No. f,1f4, t

., =. -.,

..m.

- - ~

JAFNPP 3.9 BASES (cont d) l G.

Reactor Protection System Power Sucolies Each of two RPS divisions may be supplied power from it's respective RPS MG set or from an alternate source which derives power from the same electrical division. The MG sets and siternate sources for both divisions are provided with redundant, seismic qualified, class 1E electrical protection assemblies between the power source and the RPS bus. Any abnormal output type failure in either of the MG sets or alternate sources (if in service) would result in a trip of one or both of the electrical protection assemblies producing a half scram on that RPS division and retaining full scram capability in the other RPS division.

Limiting operating conditions in Section 3.9.G provide a high degree of assurance that RPS buses are protected as described above.

Amendment No. -7,

224a

~..

JAFNPP 4.9 BASES (cont'd) l D.

Not Used l

E.

Battery System Measurements and electrical tests are conducted at specified intervals to provide indication of cell condition and to determine the discharge capability of the batteries.

Performance and service tests are conducted in accordance with the recommendations of IEEE 450-1987.

l F.

LPCI MOV Indcoendent Power Sunoiv Measurement and electrical tests are conducted at specified intervals to provide indication of cell condition, to determine the discharge capability of the battery. Performance and service tests are conducted in accordance with the recommendations of IEEE 450-1987.

l l

G.

Reactor Protection Power Sucolies i

i Functional tests of the electrical protection assemblies are l

conducted once each six (6) months utilizing a built-in test I

device and once per operating cycle by performing an instrument caSbration which verifies operation within the limits of Section 4.9.G.

l l

l Amendment No. f,7f, If4,1f7,.

226

e JAFNPP l. A.

Hioh Pressure Water Fire Protection Systqm (Cont'd)

A.

Hioh Pressure Water Fire Protection System (Cont'd)-

3.

If 1. above cannot be fulfilled, place the reactor in Hot 11gm Frecuency Standby within six (6) hours and in Cold Shutdown within the following thirty (30) hours.

h. Fire pump diesel engine Once/ Month by verifying the fuel storage tank contains at least 172 gallons of fuel.

i. Diesel fuel from each Once/ Quarter tank obtained in accordance with ASTM-D270-65 is within the acceptabla limits for quality as per the following:

Flash Point

'F 125'F min.

Pour Point

'F 10'F max.

Water & Sediment 0.05% max.

Ash 0.01 % max.

Distillation 90% Point 540 min.

Viscosity (SSU) @ 100 F 40 max.

I Sulfur 1 % max.

Copper Strip Corrosion No. 3 max.

Cetane #

35 min.

j. Fire pump diesel engine Once/18 months by inspection during shut down l

in accordance with procedures l

prepared in conjunction with manufacturers recommendations and verifying the diesel, starts from ambient conditions on the auto start signal and operates for 2:20 minutes while loaded with the fire pump.

,1 f4, Amendment No.

244c.

JAFNPP 2.

If the fire protection systems smoke and/or heat detectors in Tables 3.12.1 and 3.12.2 cannot be restored to an operable status within 14 days, a written report to the Commission outlining the action taken, the cause of inoperability and plans and schedule for restoring the detectors to an operable status shall be prepared and submitted within 30 days.

F.

Fire Barrier Penetration Seals ~

F.

Fire Barrier Penetration Seals 1.

All fire barrier penetrations, including cable penetration 1.

All fire barrier penetration seals for each protected area barriers, fire doors and fire dampers, in fire zone shall be visually inspected once/1.5 years to verify boundaries protecting safety related areas shall be functionalintegrity. For those fire barrier-penetrations functional.

that are not in the as-designed condition, an evaluation shall be performed to show that the modification has not degraded the fire rating of the fire barrier penetration.

l 2.

With one or more of the required fire barrier penetrations 2.

Any repair of fire barrier penetration seals shall be non-functional, within one hour establish a continuous fire followed by a visual actspection.

watch on at least one side of the affected penetration or verify the operability of fire detectors on at least one side of the non-functional fire barrier and establish an hourly-fire watch patrol. Restore the non-functional fire barrier penetration (s) to functional status within 7 days or, in lieu of any other report required by Specification 6.9.A, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.B within 30 days outlining the action taken, the cause of the non-functional penetration and plans and schedule for restoring the fire barrier penetration (s) to functional status.

l Amendment No. 3f, If6, If7, 244g

JAFNPP W

3.12 and 4.12 BASES The Fire Protection System specifications provide pre-established B. Safety related equipment areas protected by water spray or minimum levels of operability to assure adequate fire protection sprinklers are listed in Table 3.12.1. Whenever any of the during any operating condition including a design basis accident or protected areas, spray or sprinklers are inoperable continuous safe shutdown earthquake.

fire detection and backup fire protection equipment is available in the area where the water spray and/or sprinkler A. The high pressure water fire protection system is supplied by protection was lost.

redundant vertical turbine pumps, one diesel driven and one electric motor driven, each design rated 2500 gpm at 125 peig Performance of the tests and inspections listed in Table discharge pressure. Both pumps take suction from the plant 4.12.1 will prevent and detect nozzle blockage or breakage intake cooling water structures from Lake Ontario. The high and verify header integrity to ensure operability.

pressure water fire protection header is normally maintained at greater than.115 psig by a pressure maintenance subsystem. If C. The carbon dioxide systems provide total flood protection for pressure decreases, the fire pumps are automatically started by eight different safety related areas of the plant from either a their initiation logic to maintain the fire protection system 3 ton or 10 ton storage unit as indicated in Table 3.12.2.

header pressure. Each pump, together with its manual and Both CO storage units are equipped with mechanical 2

automatic initiation logic combined makes up a redundant high refrigeration units to maintain the storage tank content at pressure water fire pump.

O F with a resultant pressure of 300 psig. Automatic smoke and heat detectors are provided in the CO protected areas 2

A third fire pump, diesel-driven, has been installed and is set to and initiation is automatic and/or manual as indicated in automatically actuate upon decreasing pressure after the Table 3.12.2. For any area in which the CO, protection is actuation of the first two fire pumps. No credit is taken for this made or found to be inoperable, continuous fire detection is pump in any analyses and the requirements of Technical available and one or more large wheeled CO fire 2

Specifications 3.12 and 4.12 do not apply.

extinguisher is also available for each area in which protection was lost.

Pressure Maintenance subsystem checks, valve position checks, system flushes and comprehensive pump and system flow Weekly checks of storage tank pressure and level verify and/or performance tests including logic and starting subsystem proper operation of the tank refrigeration units and tests provide for the early detection and correction of availability of sufficient volume of CO to extinguish a fire in 2

component failures thus ensuring high levels of operability.

any of the protected areas.

Amendment No. f, If2,1f 6, 244h

JAFNPP 5.5.B Bases The sperit fuel pool and high density fuel storage racks are Class I structures designed to store up to 2,797 fuel bundles. The storage racks are designed to maintain a subcritical configuration having a multiplication factor (k,,,)

less than 0.95 for all possible operational and abnormal conditions. The nuclear criticality analyses for the Spent Fuel Racks (References 1 and 3) conclude that fresh fuel bundles with 3.3 w/o U-235 meet the 0.95 k,,, limit. This design basis bundle was reanalyzed to determine its infinite lattice multiplication factor, k, when in a reactor core geometry (Reference 2). This k.,was obtained under conservative calculational assumptions and reduced by 2.33 times the standard deviation in the calculation resulting in the Technical Specification limit of 1.36.

References:

1) Increased Spent Feel Storage Modification, Stone &

Webster Engineering Corporation, Boston, Mass. March 15,1978.

2) General Electric letter, P. Van Dieman to G. Rorke, l

FitzPatrick Fuel Storage K-infinity Conversion, Revision 1, dated July 10,1986.

3) increased Storage Capacity for FitzPatrick Spent Fuel Pool, Holtec International, Mount Laurel, New Jersey, February,1989.

l Amendment No. If1, If5, 246a

i JAFNPP 2.

An SRO or an SRO with a license limited to fuel handling shall directly supervise all Core Alterations. This person shall have no other duties during this time; 3.

A fire brigade of five (5) or more members shall be maintained on site at all times. This excludes two (2) members of the minimum shift crew necessary for safe shutdown and any personnel required for other essential functions during a fire emergency; 4.

In the event of illness or unexpected absence, up to two (2) hours is allowed to restore the shift crew or fire brigade to the minimum complement.

5.

The Operations Manager, Assistant Operations Manager, Shift Supervisor and Assistant Shift Supervisor shall hold a SRO license and the Senior Nuclear Operator and the Nuclear Control Operator shall hold a RO license or an SRO license.

6.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g., senior reactor operators, health physicists. auxiliary operators, and maintenance personnel who are working on safety-related systems.

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 plant 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 modifications, on a temporary basis, the following guidelines shall be followed:

a.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time, b.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven day period, all excluding shift tumover time.

c.

A break of at least eight hours should be allowed between work periods, including shift turnover time.

i d.

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 Resident Manager or the General Manager - Operations, 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 Resident Manager or his designee to assure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

Amendment No. 4f,1 f1, If0,1f 7,1f 8, 247a

JAFNPP 6.3 PLANT STAFF QUALIFICATIONS 6.3.1 The minimum qualifications with regard to educational background and experience for plant staff positions shown in FSAR Figure 13.2-7 shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions; except for the Radiological and Environmental Services Manager who shall meet or exceed the

. l qualifications of Regulatory Guide 1.8, September 1975.

6.3.2 The Shift Technical Advisor (STA) shall meet or exceed the minimum requirements of either Option 1 (Combined SRO/STA Position) or Option 2 (Continued use of STA Position), as defined in the Commission Policy Statement on Engineering Expertise on Shift, published in the October 28,1985 Federal Register (50 FR 43621). When invoking Option 1, the STA role may be filled by the Shift Supervisor or Assistant Shift Supervisor. (1) 6.3.3 Any deviations will be justified to the NRC prior to an individual's filling of one of these positions.

NOTE:

(1)

The 13 individuals who hold SRO licenses, and have completed the FitzPatrick Advanced Technical Training Program prior to the issuance of License Amendment 111, shall be considered qualified as dual-role SRO/ STAS.

6.4 RETRAINING AND REPLACEMENT TRAINING A training program shall be maintained under the direction of the Training Manager to assure overall proficiency of the plant staff organization. It shall consist of both retraining and replacement training and shall meet or exceed the minimum requirements of Section 5.5 of ANSI N18.1-1971.

The retraining program shall not exceed periods two years in length with a curriculum designed to meet or exceed the requalification requirements of 10 CFR 55.59. In addition, fire brigade training shall meet or exceed the I

requirements of NFPA 27-1975, except for Fire Brigade training sessions which shall be held at least quarterly. The effective date for implementation of fire brigade training is March 17,1978.

6.5 REVIEW AND AUDIT Two separate groups for plant operations have been constituted. One of these, the Plant Operating Review Committee (PORC), is an onsite review group. The other is an independent review and audit group, the offsite Safety Review Committee (SRC).

g y

y g

g 1

JAFNPP

7.0 REFERENCES

(9)

C.H. Robbins, " Tests of a Full Scale 1/48 Segment of (1)

E. Janssen, " Multi-Rod Burnout at Low Pressure," ASME the Humbolt Bay Pressure Suppression Containment,"

Paper 62-HT-26, August 1962.

GEAP-3596, November 17,1960.

(2)

K.M. Backer, " Burnout Conditions for Flow of Boiling (10)

" Nuclear Safety Program Annual Progress Report for Water in Vertical Rod Clusters," AE-74 (Stockholm, Period Ending December 31,1966, Progress Report Sweden), May 1962.

for Period Ending December 31,1966, ORNL-4071."

(3)

- FSAR Section 11.2.2.

(11)

Section 5.2 of the FSAR.

(4)

FSAR Section 4.4.3.

(12)

TID 20583, " Leakage Characteristics of Steel Containment Vessel and the Analysis of Leakage Rate (5) 1.M. Jacobs, " Reliability of Engineered Safety Features as Determinations."

a Function of Testing Frequency," Nuclear Safety, Vol.

9, No. 4, July-August 1968, pp 310-312.

(13)

Technical Safety Guide, " Reactor Containment Leakage Testing and Surveillance Requirements,"

(6)

Benjamin Epstein, Albert Shiff, UCRL-50451, improving USAEC, Division of Safety Standards, Revised Draft, Availability and Readiness of Field Equipment Through December 15,1966.

Periodic inspection, July 16,1968, p.10, Equation (24),

Lawrence Radiation Laboratory.

(14)

Section 14.6 of the FSAR.

(7) 1.M. Jacobs and P.W. Mariott, APED Guidelines for (15)

ASME Boiler and Pressure Vessel Code, Nuclear Determining Safe Test Intervals and Repair Times for Vessels, Section 111. Maximum allowable internal Engineered Safeguards - April 1969.

pressure is 62 psig.

(8)

Bodega Bay Preliminary Hazards Report, Appendix 1, (16) 10' CFR 50.54, Appendix J, " Reactor Containment l

Docket 50-205, December 28,1962.

Testing Requirements."

(17) 10 CFR 50, Appendix J, February 13,1973.

l Amerviment No.

285

~ -.

. 1 to JPN 93-012 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES MISCELLANEOUS ADMINISTRATIVE CHANGES (JPTS-92-001) 1.

DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to the James A. FitzPatrick Technical Specifications are administrative and are addressed below.

Minor changes in format, such as type font, margins or hyphenation, are not described in this submittal. These changes are typographical in nature and do not affect the content of the Technical Specifications.

1.

Pace i. TABLE OF CONTENTS in Specification 3.2.F, replace the title " Surveillance information Readouts" with the title " DELETED."

2.

Paae v. LIST OF TABLES In the title of Table 4.2-2, replace the word " System" with the word " Systems."

3.

Pace 8. Soecification 2.1 in the Specification outline numbering pattern, delete Section numbers "A." and' "1."

4.

Paae 20. Bases 2.1

a. Add a "B." to indicate Section 2.1.B and insert the text "Not Used."

b. Add Amendment number "14," to the Amendment list at the bottom left corner of the page.

5.

Epro 29. Bases 1.2 and 2.2 in the second paragraph, replace the parentheticals "(110% x 1,250 - 1,375 psig)" and "(120% x 1,150 - 1,380 psig)" with the parentheticals "(110% x 1,250 = 1,375 psig)" and "(120% x 1,150 = 1,380 psig)."

i 6.

Pace 32. Bases 3.1 in the first paragraph, add an "A.".to indicate Section 3.1.A.

7.

Paae 54. Soecification 4.2.E in the first sentence, add a period to the end of the sentence.

. 1 to JPN-93-012 SAFETY EVALUATION Page 2 of 13 8.

Paae 78. Table 4.2-1 in the Logic System Functional Test, replace " Main Steam Line Isolation valves" with " Main Steam Line isolation Valves."

9.

Paae 79. Table 4.2-2 in the'first column, remove the note "(8)" from the " Instrument Channel" heading.

10. Pace 123. Soecification 3.5.H in the third sentence replace the phrase "If anytime during" with the phrase "If at anytime during."

11. Epae 136. Soecification 3.6.A.2 Move Specification 3.6.A.2 to line up with Specification 4.6.A.2.

12. Paae 139. Soecification 3.6.C.1 and 4.6.C.1

a. In Specification 3.6.C.1 for the second and last sentences, replace the abbreviation *hr" with the word " hours" in two locatict ;.

b. In Specification 4.6.C.1.a replace the abbreviation "hr" with the word

" hours."

c. In Specification 4.6.C.1.c replace the abbreviation "hr" with the word i

" hour. "

d. In Specification 4.6.C.1.d for the first and second sentences, replace the abbreviation "hr" with the word " hour" in two locations.

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13. Pace 140. Soecification 3.6.C.3 and 4.6.C.3 I

a. In Specification 3.6.C.3 replace the abbreviation "hr" with the word " hours."

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b. In Specification 4.6.C.3.a replace the abbreviation "hr" with the word

" hours."

14. Pace 144. Soecification 3.6.G Move Specification 3.6.G to line up with Specification 4.6.G.

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. 1 to JPN-93-012 SAFETY EVALUATION Page 3 of 13

15. Paae 149. Bases 3.6.C and 4.6.C in the second paragraph replace the abbreviation "hr" with the word " hours."

16. Paae 150. Bases 3.6.C and 4.6.C in the fourth paragraph replace the abbreviation "hr" with the word " hours."

17. Paae 151. Bases 3.6.D and 4.6.D

a. In the Section outline numbering pattern replace the phrase "3.6 4.6 BASES (cont'd)" with the phrase "3.6 and 4.6 BASES (cont'd)."

b. In the third and fourth paragraphs replace the abbreviation "yr" with the word " years."

18. Pace 165. Soecification 4.7 in the Objective, delete the "," in the phrase " primary, and secondary."

19. Paae 170. Soecification 4.7.A.2.a.(10.)

In the last sentence, delete the " * " and the referenced note which reads:

In accordance with an exemption from 10 CFR 50 Appendix J, a Type A test need not be performed during the 1988 refueling outage."

20. Paae 174. Soecification 4.7.A.2

a. In Specification 4.7.A.2.e.(5) replace the word " year" with the word

" years."

b. In Specification 4.7.A.2.f delete the " * " and the referenced note which reads:

"* In accordance with an exemption from 10 CFR 50 Appendix J, - a Type A, B, or C test is not required for:

1 The replacement of the HPCI turbine exhaust line block valve (23-HPl-11) during the 1988 outage; or 2.

The repair of the Core Spray test return line weld 10-14-884A during the 1989 maintenance outage."

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Attachment ll to JPN-93-012 SAFETY EVALUATION Page 4 of 13

21. Paae 187, Bases 3.7.A in the third paragraph, replace the parenthetical "(Section 5.2)" with the parenthetical (updated FSAR Section 5.2)."

22. Paae 191. Bases 3.7.B and 3.7.C in the second paragraph, replace the phrase "be tested daily" with the phrase "be verified operable daily."

23. Pace 217. Soecification 3.9.B.3 In the first paragraph replace the phrase "7-day" with the phrase "7 day."

24. Paae 222b. Soecification 3.9.F.2.c Replace the word " maintenance" with the word " alternate."

25. Paaes 224 and 224a. Bases 3.9

a. Add a "D." to indicate Bases Section 3.9.D and insert the text "Not Used."

b. In Bases Section 3.9.E, replace the phrase "A maintenance power source" with the phrase "An alternate power source."

c. Renumber Bases Sections "D.", *E." and "F." as "E.", "F." and "G.",

respectively.

26. Paae 226, Bases 4.9

a. Add a "D." to indicate Bases Section 4.9.D and insert the text "Not Used."

b. Renumber Bases Sections "D.", "E." and "F." to read "E.", "F." and "G.",

respectively.

27. Paae 244c. Snecification 3.12. A.1.d.3 In the title to Section 3.12.A, replace the word " Waster" with the word "Wat er."

28. Pace 2440. Soecification 4.12.F.1 in the last sentence delete the * * " and the referenced note which reads:

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"' The current surveillance interval for visually inspecting fire

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I Attachment il to JPN-93-012 SAFETY EVALUATION Page 5 of 13 barrier penetration seals is extended until May 15,1992. This t

is a onetime extension, effective only for this inspection interval. The next surveillance intervc! began September 27, 1991."

29. Paae 244h. Bases 3.12. A and 4.12. A Delete the second sentence which reads:

"Both pumps take suction from the plant insure."

30. Paae 246a. Bases 5.5.B 1

in the third sentence, replace the werd " analysis" with the word " analyses" and the word " concludes" with the word " conclude."

31. Paae 247a. Soecification 6.2.2.2

a. Replace the first sentence:

"An SRO or SRO with a license limited to fuel handling shall directly supervise all Core Alternations" with the sentence:

"An SRO or an SRO with a license limited to fuel handling shall directly supervise all Core Alterations."

l

b. Delete the second sentence, which reads:

"This person shall directly supervise all Core Alterations."

32. Paae 248. Soecifications 6.0

a. In Specification 6.3.1, replace the word "Radioligical" with the word

" Radiological."

b. In Specification 6.4 replace the phrase "10 CFR 55, Appendix A" with the phrase "10 CFR 55.59."

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33. Pace 285. Soecification 7.0

a. In Reference 16, replace the phrase "10CFR50.54" with the phrase "10 CFR 50.54" and add a quotation mark in front of the word " Reactor."

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, 1 to JPN 93-012 SAFETY EVALUATION Page 6 of 13

b. In Reference 17, replace the phrase "10CFR50" with the phrase "10 CFR 50."

i ll.

PURPOSE OF THE PROPOSED CHANGES This application makes miscellaneous administrative changes (e.g., correcting editorial errors, typographical errors and specification renumeration corrections) to the James A. FitzPatrick Technical Specifications. The proposed changes will clarify and improve the quality of the Technical Specifications. The purpose of each change identified in Section I (the numbers in Sections I and li correspond) is as follows:

1.

The proposed change updates the table of contents to reflect Amendment 181 (References 1 and 2) which deleted Specifications 3.2.F and 4.2.F.

2.

The proposed change corrects a typographical error making the title in the list consistent with the' actual title.

3.

The proposed change removes unnecessary specification outline numbers to be consistent with the pattern used everywhere else in the Technical Specifications.

4.

The first proposed change modifies the outline numbering pattern in the Bases Section to be consistent with the associated LCO and Surveillance Requirements. The second proposed change adds Amendment number "14" to the page Amendment number listing. This Amendment number was inadvertently omitted when the page was revised as part of Amendment 49 (References 3 and 4).

5.

The proposed changes correct two typographical errors in the ASME Boiler &

Pressure Vessel Code and ANSI Code pressure transient allowance equations, i

The proposed changes replace the " " symbols in two locations with the l

appropriate " =" symbols.

6.

The proposed change modifies the outline numbering pattern in Bases Section 3.1 to be consistent with the associated specification outline numbering pattern.

7.

The proposed change corrects a punctuation error by adding a period to the end of the sentence.

t 8.

The proposed change corrects a typographical error by capitalizing the "v" in valves for internal consistency.

. 1 to JPN-93-012 SAFETY EVALUATION -

Page 7 of 13 9.

The proposed change corrects a typographical error by removing reference to Note "(8)" from Table 4.2-2. Note (8) clarifies that surveillance testing for reactor low water level, high drywell pressure and high radiation main steam line tunnel Instruments are not included on Table 4.2-1 but on Table 4.1-2.

This note was inadvertently added to Table 4.2-2 as part of Amendment 160 (References 5 and 6).

10. The proposed change corrects a typographical error by adding the word "at" to the sentence. This word was inadvertently omitted as part of Amendment 134 (References 7 and 8). The change restores the sentence to its original form.

11. The proposed change relocates Specification 3.6.A.2 to line up with Specification 4.6.A.2. The change is made to clarify association between the LCO and the surveillance (the change relocates it next to Surveillance Requirement 4.6.A.2) and to be consistent witn Technical Specification format between LCO and Surveillance Requirements.

12. The proposed changes make editorial corrections by replacing the abbreviation "hr" with the word " hour" or " hours" as applicable.

13. The proposed changes make editorial corrections by replacing the abbreviation "hr" with the word " hours."

14. The proposed change relocates Specification 3.6.G to line up with Specification 4.6.G. The change is made to clarify association between the LCO and the surveillance (the change relocatesit next to Surveillance Requirement 4.6.G) and to be consistent with Technical Specification format between LCO and Surveillance Requirements.

15. The proposed change rr.aices an editorial correction by replacing the abbreviation "hr" with the word " hours."

16. The proposed change makes an editorial correction by replacing the abbreviation "br" with the word " hours."

17. The proposed change makes an editorial correction by replacing the abbreviation "yr" with the word " years." The change also revises the section heading to be consistent with other section headings.

18. The proposed change corrects an editorial error by removing an unnecessary comma from Specification 4.7.

19. The proposed change removes a past exemption from 10 CFR 50 Appendix J.

The exemption, added by Amendment 125 (References 9 and 10), eliminated

L

. 1 to JPN-93-012 SAFETY EVALUATION Page 8 of 13 a requirement to conduct Type A primary containment integrity leak rate test

- during the 1988 refueling outage. Since this exemption is no longer in effect, the exemption is removed.

20. The proposed change makes an editorial correction by revising the word

" year" to reflect the correct plural form of " years." The proposed change also removes two past exemptions from 10 CFR 50 Appendix J.' The exemptions, added by Amendments 125 and 140 (References 11 and 12) eliminated a requirement to conduct Type A, B, or C leak rate tests for two plant modifications. Since these exemptions are no longer in effect, these exemptions are removed.

21. The proposed change clarifies a reference by indicating that the current design Bases is contained in the James A. FitzPatrick updated Final Safety Analysis Report (FSAR).

22. This change makes the Bases Section consistent with a prior change tu the Surveillance Requirement. Amendment 148 (References 13 and 14) replaced the word " demonstrate" with the word " verify" where necessary to eliminate redundant and unnecessary surveillance tests performed to satisfy overlapping requirements. Bases Section 3.7.B and 3.7.C should have been changed at that time along with the Amendment 148 changes. This proposed change will correct this omission.

23. 'The proposed change makes an editorial correction for consistency within the Technical Specifications.

24. The proposed change revises the name of the Low Pressure Coolant injection j

(LPCI) " maintenance power source" to " alternate power source." The name.

l change was made as part of a recent plant modification to the LPCI Motor

~

Operated Valve (MOV) power source circuitry. The modification provided a, j

control scheme enabling the plant operators, from the control room, to isolate i

the LPCI valve bus independent power supplies and connect a maintenance

-l bypass (renamed alternate feed) from another safety related emergency Motor.

Control Center (MCC) in the same safety division to the. valve bus. The j

modification gives operators full control over the power sources for the LPCI valve bus in the event the reactor building becomes restricted due to postulated post-accident radiation dose levels. The modification had no effect.

on the Technical Specifications except for this name change. Due to the nature of the change (i.e., renaming the power source) this change is j

considered administrative.

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25. The proposed chang'es revises the outline numbering pattern of Bases Section

)

I 3.9 to be consistent with the associated LCO Specifications of 3.9 by adding.

a section and indicating

hat it is not being used. The proposed change also.-

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p t 1 to JPN-93-012 SAFETY EVALUATION Page 9 of 13 renumbers other Bases Sections to be consistent with the associated LCO Section outline numbering pattern. The change which renarnes the

" maintenance power source" to " alternate power source" is made to be i

consistent with the changes of item 24.

26. The proposed changes revise the outline numbering pattern of Bases Section 4.9 to be consistent with the associated Surveillance Requirements of Section 4.9 in the same manner as the revisions to Section 3.9 in item 25. The proposed changes add a section that is indicated as not being used and renumber the Bases Sections to be consistent with the Surveillance Section

[

outline numbering pattern.

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27. The proposed change corrects the spelling of the word " Water."

28. The proposed change removes a past one time extension to the fire barrier penetration seal visual inspection interval The extension, added by Amendment 177 (References 15 and 16), ended on May 15,1992. Since this extension is no longer in effect, the extension is removed.

29. The proposed change corrects a typographical error introduced by Amendment 176 (Reference 17 and 18). The error inadvertently duplicated part of a following sentence. This change removes the duplication.

30. The proposed changes make two editorial corrections by revising words to reflect the correct plural and singular forms.

31. The proposed changes make three editorial corrections. In the first sentence the word."an" is added and the spelling of the word " Alterations" is corrected.

l The second sentence is removed since it is duplicate to the first sentence.

32. The proposed change corrects the spelling of the word " Radiological" and revises a reference to reflect a change in the location of regulations from 10 CFR 55, Appendix A to 10 CFR 55.59, effective May 26,1987 (Reference 19). The regulation change incorporated the licensed operator requalification requirements into 10 CFR 55.59 and subsequently deleted 10 CFR Part 55,.

Appendix A. This change corrects the reference to Title 10 Code of Federal Regulations.

33. The proposed changes make editorial corrections by revising reference to the Code of Federal Regulations, by adding proper spacing. The changes also add a missing quotation mark.

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Attachment il to JPN-93-012 SAFETY EVALUATION Page 10 of 13 111.

SAFETY IMPLICATIONS OF THE PROPO$fy CHANGES The proposed changes to the James A. FitzPatrick Technical Specifications will not affect plant safety or operations. The proposed changes will correct editorial and typographical errors as well as remove past exceptions to Specifications. These changes will clarify and improve the quality of the Technical Specifications. The nature of each change assures that no safety implications are associated with these changes. The proposed changes involve no limiting conditions for operation, surveillance requirements, setpoint or safety limit changes, nor do they affect the environmental monitoring program. The proposed changes do not change any system or subsystem and will not alter the conclusions of either the updated FSAR 7*

or the SER.

IV.

EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:

1.

involve a significant increase in the probability or consequences of an accident previously evaluated.

The intent of the proposed changes is to clarify and improve quality of the Technical Specifications. The proposed changes will correct editorial and typographical errors as well as remove past exceptions to Specifications.

These changes will clarify and improve the quality of the Technical Specifications. The changes by their nature have no affect on previously evaluated accidents. There are no setpoint changes, safety limit changes, surveillance requirement changes or limiting conditions for operation changes.

These changes have no affect on plant safety or operations.

2.

create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes involve no plant modifications, changes to surveillance test methods or frequencies, changes to operating procedures or relaxation of any LCO. The proposed changes are administrative in nature and involve such changes as editorial corrections, typographical corrections and removal of past exceptions to Specifications. These proposed changes clarify and improve the quality of the Technical Specifications and by their nature cannot create the possibility of a new or different kind of accident.

b

. 1 to JPN-93-012 SAFETY EVALUATION Page 11 of 13 3.

involve a significant reduction in a margin of safety.

The proposed changes are administrative in nature and will clarify and improve quality in the Technical Specifications. The proposed changes will correct editorial and typographical errors as well as rernove past exceptions to Specifications. These changes, by their nature, can have no affect on the margin of safety. These changes do not change any setpoint or safety limit changes regarding isolation or alarms. The proposed changes do not affect the environmental monitoring program. These changes do not affect the plants safety systems.

V.

IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not adversely affect the ALARA or Fire Protection Programs at the FitzPatrick plant, nor will the changes affect the environment. This application for an amendment makes miscellaneous administrative changes and can have no affect on these programs or the environment.

VI. CONCLUSION The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they:

1.

will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; 2.

will not increase the possibility of an accident or malfunction of a type different from any previously evaluated in the Safety Analysis Report; and 3.

will not reduce the margin of safety as defined in the basis for any technical specification; 1

The changes involve no significant hazards consideration, as defined in 10 CFR 50.92.

Vll. REFERENCES References relied upon to prepare the Technical Specification change request:

1.

NYPA Letter, R.E. Beedle to B.C. McCabe dated May 30,1992 (JPN-92-042).

Attachment il to JPN-93-012 SAFETY EVALUATION Page 12 of 13 Submittal for Amendment 181 to the Technical Specifications.

2.

NRC Letter, B.C. McCabe to R.E. Beedle dated May 14,1992 ([[::JAF-92-139|JAF-92-139]]),

Transmits Amendment 181 to the Technical Specifications.

3.

NYPA Letter, G.T. Berry to T.A. Ippolito dated March 4,1980 (JPN-80-015).

Submittal for Amendment 49 to the Technical Specifications.

4.

NRC Letter, T.A. Ippolito to G.T. Berry dated July 11,1980 ([[::JAF-80-142|JAF-80-142]]).

Transmits Amendment 49 to the Technical Specifications.

5.

NYPA Letter, J.C. Brons to D.E. LaBarge dated January 1,1990 (JPN-90-003).

Submittal for Amendment 160 to the Technical Specifications.

6.

NRC Letter, D.E. LaBarge to J.C. Brons dated May 18,1990 ([[::JAF-90-162|JAF-90-162]]).

Transmits Amendment 160 to the Technical Specifications.

7.

NYPA Letter, J.C. Brons to D.E. LaBarge dated May 24,1989 (JPN-89-030).

Submittal for Amendment 134 to the Technical Specifications.

8.

NRC Letter, D.E. LaBarge to J.C. Brons dated January 24,1990 ([[::JAF-90-028|JAF-90-028]]).

Transmits Technical Specification Replacement Pages (Re-issues page 123 as of Amendment 134).

9.

NYPA Letter, J.C. Brons to D.E. LaBarge dated November 9,1988 (JPN 060). Submittal for Amendment 125 to the Technical Specifications.

10. NRC Letter, D.E. LaBarge to J.C. Brons dated February 17,1989 (JAF [

066). Transmits Amendment 125 to the Technical Specifications.

11. NYPA Letter, J.C. Brons to D.E. LaBarge dated September 28,1989 (JPN ;

062). Submittal for Amendment 140 to tho Technical Specifications.

12. NRC Letter, D.E. LaBarge to J.C. Brons dated October 4,1989 ([[::JAF-89-347|JAF-89-347]]).

j Transmits Amendment 140 to the Technical Specifications.

13. NYPA Letter, J.C. Brons to D.E. LaBarge dated May 31,1989 (JPN-89-034).

Submittal for_ Amendment 148 to the Technical Specifications.

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14. NRC Letter, D.E. LaBarge to J.C. Brons dated December 26,1989 (JAF 002). Transmits Amendment 148 to the Technical Specifications.

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15. NYPA Letter, R.E. Beedle to B.C. McCabe dated December 19,1991 (JPN-91--

i 069). Submittal for Amendment 177 to the Technical Specifications.

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Attachment il to JPN-93-012 SAFETY EVALUATION Page 13 of 13

16. NRC Letter, B.C. McCabe to R.E. Becdle dated February 10,1992 (JAF 036). Transmits Amendment 177 to the Technical Specifications.

17. NYPA Letter, R.E. Beedle to B.C. McCabe dated December 19,1991 (JPN 068). Submittal for Amendment 176 to the Technical Specifications.

18. NRC Letter, B.C. McCabe to R.E. Beedle dated January 16,1992 (JAF 015). Transmits Amendment 176 to the Technical Specifications.

19. Federal Register dated March 25,1987 (52 FR 9460).

References reviewed but not specifically referenced:

i 1.

James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report Section 5.2, through Revision 5, dated January 1992.

2.

James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER),

f dated November 20,1972, and Supplements.

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. 11 to JPN-93-012 '

i PROPOSED TECHNICAL SPECIFICATION CHANGES MISCELLANEOUS ADMINISTRATIVE CHANGES ^

ll MARKUP OF TECHNICAL SPECIFICATION PAGES il I

(JPTS-92-001) l

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-1 1'!

- New York Power Authority i

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

[

f-

,a

JAFNPP o

TECHNICAL SPECIFICATIONS q

TABLE OF CONTENTS

.,~

1.0 Definitions UMITING SAFETY SAFETY UM!TS SYSTEM St:n iLNGS r

1.1 Fuel Cladding integrity 2.1 7

1.2 Reactor Codant System 22 27 SURVEILLANCE UMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.0 General 4.0 30 3.1 Reactor Protection System 4.1 30f 32 instrumentation 42 49 A.

Primary Containtnent isdation Functions b

A.

49 B.

Core and Containment Cooling Systems - Initiation and B.

50 l

Controf C.

Control Rod Block Actuation kg C.

50 D.

Radiation Monitoring Systems -! solation and initiation D.

50 Functions g

n E.

Drywell Leak Detection E.

54 Q1 G. h(Surveillance information Headouts F.

F.

54 Recirculat>on Pump Trip G.

54 H.

Accident Monitoring instrumentation H.

54 -

I.

4kV Emergency Bus Undervottage Trip 54 3.3 Reactivity Control 4.3 88 A.

Reactivity Umrtations A.

88 B.

Cor: trol Rods B.

91 C.

Scraminsertion Times C.

95 D.

Reactivity Anomalies D.

96 3.4 -

Standby Uqtsd ControlSystem 4.4 105 A.

Nortrei Operation A.

105 B.

Operation With inoperable Components B.

106 C.

Sodium Pentaborate Solution C.

107 3.5 Core and Containment Coohng Systems 4.5 112 A.

Core Spray and LPCISystems A.

112 D.

Containment Cooling Mode of the RHR System B.

115 C.

HPCISystem C.

117 D.

Automatic Depressurization System (ADS)

D.

119 E.

Reactor Core isolation Cooling (RCIC) System E.

121 Amendment No.M 136,ladi,)d i

JAFNPP e.,

USTOF TABLES Table Tale

.Pge, 3.1 1 Reactor Protection System (Scram) instrumentation Requirement 41 0

3.1 2 Reactor Protection System Instrumentation Response Tirnos 43a 4.1 1 Reactor Protection System (Scram) instrument Functonal Tests 44 H

4.1 2 Reactor Protection System (Scram) instrument Calibration 46 32-1 instrumentation that INtiates Primary Containment isolaton 64 32 2 Instrumentation that initiates or Controls the Core and Containment 66 Cooling Systems 3.2 3 Instrumentation that initiates Control Rod Blocks 72 324 (DELETED) 74 32-5 instrumentation that Monitors Laakage Detection inside the Drywell 75 32-6 (DELETED) 76 Q

32-7 instrumentation that irstiates Recircutation Pump Trip 77 32 8 Accident Moretoring instrumentation 77a O

32 9 Primary Containment 1 solation System Actuation instrumentation t

77e Response Times

{

42-1 Minimum Test and Calibration Frequency for PCIS 78 C

42-2 Minimum Test and Calibration Frequency for Core and Containment 79 l

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3 Minimum Test and Calibration Frequency for Control Rod Blocks 81 Actuation 424 (DEtETED) 82 42 5 Minimum Test and Calibration Frequency for Drywell Leak Detection 83 42-6 (DELETED) 42 7 Minimum Test and Calibration Frequency for Recirculation Pump Trip 85 Amendment No. p6.p!I, ig,1g, f V

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JAFNPP 1.1 (cont'd) 2.1 (cont'd)

A.

1.

b.

APRM Flux Scram Trip Setting (Refuel or Start & Hot B.

Core Thormal Power Umit (Reactor Pressure $785 psig)

Standby Mode)

When the reactor pressure is $785 psig or core flow is less than APRM - The APRM flux scram setting shall be $15 or equal to 10% of rated, the core thermal power shall not

. percent of rated neutron flux with the Reactor Mode exceed 25 percent of rated thermal power.

Switch in Startup/ Hot Standby or Refuel.

C.

_Pwer Transient c.

APRM Rux Scram Trip Settings (Run Mode)

To ensure that the Safety Umit established in Specification 1.1.A ard 1.1.B is not exceeded, each required scram shall be initiated (1)

Row Refomnced Neutron Rux Scram Tn.p by its expected scram signal. The Safety Umit shall be assumed 08E"9 to be exceeded when scram is accomplished by a means other When the Mode Switch is in the RUN position, than the expected scram signal.

the APRM flow referenced flux scram trip setting shall be less than or equal to the limit specified in Table 3.1-1. This setting shall be e

adjusted during sir $ loop operation when required by Specification 3.5J.

For no combination of recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 117%

cf rated thermal power.

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I Amendment No.1/M,4K 72,96,134 8

($

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8

(

JAFNPP 2.1 BASES (Cont'd)

C.

References 1.

(Dolotod) 2.

" General Electric Standard Application for Reactor Fuel *,

NEDE 24011-P-A (Approved revision number applicable at time that roload fuel analyses are performed).

Wl 3.

(Delotod) 4.

FitzPatrick Nucisar Power Plant Singlo-Loop Operation, NEDO-24281, August,1900.

9

{,y Amendment No go,Jii4,Jid (Next page is 23)

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t JAFNPP 1.2 and 2.2 BASES Ch I

The reactor coolant pressure boundary integrity is ANSICodepermitspressuretransientsujpto20 percent important barrier in the prevention of uncon-over the design pressure (120% a 1,150 0- 1,380 psig),

en trolled release of fission products.

It is The safety limit pressure of 1,375 psi is referenced essential that the integrity of this boundary be to the lowest elevation of the Reactor Coolant System.

protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The current reload analysis shows that the main steam The pressure safety limit of 1,325 psig as measured isolation valve closure transient, with fluz scram, is by the vessel steam space pressure indicator is the most severe event resulting directly in a reactor squivalent to 1,375 psig at the lowest elevation of coolant system pressure increase. The reactor vessel the Reactor Coolant System. The 1,375 psig value pressure code limit of 1,375 psig, given in FSAR j

is derived from the design pressures of the reactor Section 4.2, is above the peak pressure produced by pressure vessel and reactor coolant system piping.

the event above. Thus, the pressure safety limit The respective design pressures are 1250 psig at (1,375 psig) is well above the peak pressure that can 575*F for the reactor vessel,,Il48 psig at 568"F result from reasonably espected overpressure tran-L ter the recirculation suction piping and 1274 psig sients.

(See current reload analysis for the curve at 575* for the discharge piping. The pressure produced by this analysis.) Reactor )ressure is safety limit was chosen as the lower of the continuously indicated in the control room during pressure transients permitted by the applicable operation.

design codes:

1965 ASME Boiler and Pressure Vessel Code,Section III for pressure vessel and 1969 ANSI A safety limit is applied to the Residual Heat B31.1 Code for the reactor coolant system piping.

Removal System (RIIRS) when it is operating in the The ASME Doller and Pressure Vessel Code permits shutdown cooling mode.

When operatlag in the shut-pressure transients up to 10 percent over design down coollag mode, the RHRS is included in the pressure (110% a 1,250()1,375 psig) and the reactor coolant system.

The numerical distribution of safety / relief valve set-points shown in 2.2.1.B (2 0 1090 psi, 2 0 1105 psi, 7 e 1140 psi) is justifled by analyses described in 0;

the General Electric report HEDO-24129-1, Supplement 1, and assures that the structural acceptance criteria set forth in the Mark I Containment Short Term Program are satisfied.

Amendment No.

4e 4

29

... ~. _.

_ _ _ = =

y l

~

(Tl JAt1h>p 3.1 UASES k,The reactor protection system automatically initiates.

The outputs of tho.sut Iannels are combined a teactor sc r am t*o :

In a 1 out of 2 logics i.e., an input' signal on either one or both of the subchannels will caune trip system trip. The outputs of the trip 1.

Preserve the integrity of t.he fuel cladding.

a

-ayntems are arranged so that a trip on both 2.

Preserve the integrity of the Iteactor Coolant systems is required to produce a reactor scram.

System.

Thi's system rnects the intent of IEEE-279 (1971) 3.

itinimize the energy which must be absorbed

.for riuclear Power Plant Protection Systems.. The following a loss of coolant accident, and system has a reliability greater'than that of a prevent inadvertent cri ticality.

2 out of 3 system and'somewhat less than that of a 1 out of 2 system.,

This specification provides the limiting conditions f or operation necessary to preserve the ability With the exception of the average power range of the system to perforu its intended function monitor (APRM) channel the intermediate range O

even during periods when instrument channels may monitor (IRM) channels, the scram discharge volume, be out of service because of maintenance. When the main steam isolation valve closurd and the necessary, one channel may be made inot,ierable for turbine stop valve closure, each subchannel has brief intervals to conduct required functional one instrument channel. When the minimum Le*;t.s and calibrations.

condition for operation on the numb,er of operable instrument channals p'er untripped protection The Itcactor Protec'tlon System is of the dual channel trip system is net or if it cannot be met and the type (lieference subsection 7.2 l'SAa).

The System affected protection trip system is placed in a is made up of two independent Ltip systems, each tripped condition, the effectiveness of the having two subchannels of tripping devices.Each protection system is preserved.

subchannel has an input from at 1 cast ono Instrument channel which nonitors a critical parameter.

Three APRM instrument channels are provided for-each protection trip system. APlui's A and E operate contacts in one subchannel and APRM's 8

C and E operate contacts in the other t

4 Amenementm.g.g,g

)

\\

~

JAFNPP 32 (cont'd) 42 (cont'd)

E.

Drywell Leak Detection E.

Drywell Leak Detection The limiting conditions of opera 6on for Die instrumentation that instrumentation shall be calibrated and checked as indicated in q

monitors drywell leak detecuon are given in Table 32-5.

Table 42.

O l)

F.

(Deleted)

F.

(Deleted)

G.

Recirculation PumpTrip G.

Recirculation Pump Trip f

The limiting conditions for operation for the instrumentation that Instrumentation shall be funcuonally tested and calibrated as D

trip (s) the recirculation pumps as a means of limiting the indicated in Table 42 7.

consequences of a failure to scram during an anticipated System logic shall be functionalty tested as indicated in Table transient are given in Table 32-7.

42-7.

H.

Accident Monitoring Instrumentation H.

Accident Monitoring Instrumentation The limiting conditions for operation of the instrumentation that Instrumentation shall be demonstrated operable by performance provides accident monitoring are given in Table 32-8.

of a channel check and channel calibration as indicated in Table 1.

4kv Emergency Bus Undentoitage Trip 42-8.

The limiting conditions for operation for the instrumentauon that prevents damage to electrical equipment or circuits as a result of either a degraded or loss-of-voltage condiuon on the emergency electrical buses are given in Table 32-2.

i 1

Amendment No. W, W [

54

I JAFNPP TABLE 42-1 MINIMUM TEST AND CAUBRATION FREQUENCY FOR PCIS instrument Channel (B)

Instrument Functional Test Calibration Frequency instrurnent Check (4)

1) Reactor High Pressure (1)

Once/3 months None (Shutdown Cooling Permissive) t

2) Reactor low-Low-Low Water Level \\

(1)(5)

(15) a w / day

3) Main Steam High Temp.

(1)(5)

(15)

Once/ day

4) Main Steam High Flow (1)(5)

(15)

Once/ day

5) Main Steam Low Pressure (1)(5)

(15)

Once/ day

6) Reactor Water Cleanup High Temp.

(1)

Once/3 months None

7) Condenser Low Vacuum (1)(5)

(15)

Once/ day Logic System Functional Test (7) (9)

Frequency 1)

Main Steam UneisolationCyaivesQ Once/6 months -

Main Steam Une Drain Valves Reactor Water Sample Valves n

2)

RHR -Isolation Valve Control Myg3 Once/6 months Shutdown Cooling Vaives Iy 3)

Reactor Water Cleanup Isolation Once/6 months I

Q 4)

Drywellisolation Valves Once/6 months TlP Withdrawal Atmospheric ControlValves 5)

Standby Gas Treatment System Once/6 months Reactor Building isolation NOTE:

See notes following Table 42-5.

Amendment No. ;#, ps, Ip6,J6T,[

78

~

((

JAFNPP

' TABLE 4.2-2 MINIMUM TEST AND CAUBRATION FREQUENCY FOR CORE AND CONTAINMENT COOUNG SYSTEMS

(

instrument Functional Test Calibration Frequency instrument Check (4) instrument Channelg 1)

Reactor Water Level (1)(5)

(15)

Once/ day.

2a)

Drywell Pressure (non-ATTS)

(1)

Once/3 months None 2b)

Drywell Pressure (ATTS)

(1)(5)

(15)

Once/ day 3a)

Reactor Pressure (non-ATTS)

(1)

Once/3 months None 3b)

Reactor Pressure (ATTS)

(1)(5)

(15)

Once/ day 4)

Auto Sequencing Timers None Once/ operating cycle None

- 5)

ADS - LPCI or CS Pump Disch.

(1)

Once/3 months None 6).

Trip System Bus Power Monitors (1)

None None 8)

Core Spray Sparger d/p (1)

Once/3 months Once/ day 9)

Steam Une High Flow (HPCI & RCIC)

(1)(5)

(15)

Once/ day 10)

Steam Une/ Area High Temp- (HPCI & RCIC)

(t)(5)

- (15)

Once/ day

- 12)

HPCI & RCIC Steam Line Low Pressure (1)(5)

(15) -

Once/ day 13)

HPCI & RCIC Suction Source Levels (1)

Once/3 months None" 14) 4kV Emergency Bus Under-Voltage Once/ operating cycle Once/ operating cycle None-

. (Loss-of-Voltage, Degraded Voltage LOCA and non-LOCA) Relays and Timers.

15)

HPCI & RCIC Exhaust Diaphragm (1)

Once/3 months None

' Pressure High

- 17)

LPCl/ Cross Connect Valve Position '

Once/ operating cycle None

.None; NOTE:

See notes following Table 4.2-5.

i l

t Amendment No. - 14,48,56, pli,196, )2d, }S6, J

n

.. _ _ _. ~..,

._._._.-..,,.m c

h

~

JAFNPP 3.5 (cont'd) 4.5 (cont'd) condition, that pump shall be considered inoperable for 2.

Following any period where the LPCI subsystems or core purposes of satisfying Specifications 3.5.A,3.5.C, and spray subsystems have not been maintained in a filled 3.5.E.

condition; the discharge piping of the affected subsystem shall be vented from the high point of the system and H.

Average Planar unear Heat Generation Rate (APWGR) water flow observed.

During power operation, the APLHGR for each type of fuel as a 3.

Whenever the HPCI or RCIC System is lined up to take Qk function of axiallocation and average planar exposure shall be suction from the condensate storage tank, the discharge within limits based on applicable APWGR limit values which piping of the HPCI or RCIC shall be vented from the high have been approved for the respective fuel and lattice types.

point of the system, and water flow observed on a monthly i

values are specified in the Cofe Operating Umits Report, basis.

If ime during reactor power operation greater than 25% of 4.

The level switches located on the Core Spray and RHR r ed power it is determined that the limiting value for APWGR is being exceeded, action shall then be initiated within 15 minutes System discharge piping high points which monitor these to restore operation to within the prescribed limits. If the lines to insure they are full shall be functionally tested each APWGR is not returned to within the prescribed limits within two (2) hours, an orderly reactor power reduction shall be H.

Average Planar unear Heat Generation Rate (APLHGR) commenced immediately. The reactor power shall be reduced to less than 25% of rated power within the next four hours, or The AF%HGR for each type of fuel as a function of average until the APWGR is returned to within the prescribed limits.

planar exposure shall be determined daily during reactor operation at > 25% rated thermal power.

I Amendment No. 48,64',7A,86, So,189,11(, IM,134 [z -

1 123

n

]

(

%)

JAFNPP 3.6 UMITING CONDITIONS FOR OPERATION 4.6 SURVEILLANCE REQUIREMENTS 3.6 REACTOR COOLANT SYSTEM 4.6 REACTOR COOLANT SYSTEM Applicability:

Applicability:

Applies to the operating status of the Reactor Coolant Systern.

W t

ex W nat tes% r$ranents la the Objectivo:

gg To assure the integrity and safe operation of the Reactor Coolant System.

To deterrnine the condition of the Reactor Coolant System and the operation of the safety devices related to it.

Specification:

A.

Pressurization and Thermal Umits A.

Pressurization and Thermal Umits 1.

Reactor Vessel Head Stud Tensioning 1.

Reactor Vessel Head Stud Tensioning The mactor vessel head bolting studs shall not be under When in the cold condition, the reactor vessel head flange tension unless the temperatures of the reactor vessel and the reactor vessel flange temperatures shall be flange and the reactor head flange are at least 90 F.

reca ded:

Every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor vessel head flange a.

is < 120"F and the studs are tensioned.

b.

Every 30 ranutes when the reactor vessel head flange is < 100"F and the studs are tensioned.

2.

In-Service Hydrostatic and Laak Tests Within 30 rninutes prior to and every 30 rnnutes c.

Q During in-service hydrostatk:ior leak testing the Reactor I

p!

I.

Coolant System pressure and temperature shall be on or L

to the right of curve A shown in Figure 3.6-1 Part 1,2, or 3 P 2.

IrvService Hydrostatic and Leak Tests and the maximum temperature change during any one hour period shall be:

Dunng hydrostatic and leak testing the Reactor Coolant System pressure and temperature shall be recorded every 30 minutes until two consecutive temperature readings aro within 5*F of each otner.

Amendment No. *4, trJ, 1[

i3e

h JAFNPP 3 G (cont'd) 4.6 (cont'd)

O 7.

Reactor Vessel Flux Monitoring i

I The reactor vessel Flux Monitoring Surveillance Program cu ips with the Intent of the May,1983 r6 vision to 10 CFR 50, Appendices G and H. The next flux monitoring surveillance capsule shall be removed after 15 effective fug power years (EFPYs) and the test-procedures and reporting requirements shall meet the requirements' of ASTM E 185-82.

I iB.

OcIcted B.

Deleted I

uY' S, kurs k

g

[

C.

Coolant Chemistry C.

Coolant Chemistry.

l I.

The reactor coolant system radoactivity cei curitration in 1.

a.

A reactor coolant shall be taken at least water shall not exceed the equilibrium value of 3.1'pCl/gm every and analyzed for gross gamma activity._

of dose equivalent I-131. This limit may be exceeded

b.

Isotopic of a semple.of reactor coolant shall following a power transient, for-a maximum of 48 lodine '(

g g g gog During this iodine activity transient ' the concentrations shall not exceed the equilibrium limits by

c. -

A sample of reactor coolant shall be taken prior to -

more than a factor of 10 wi,ei~;c-the main steemline

%i startup and W i@ intervals during startup-and isolation valves are open The - reactor 'shall not ' be ?

,houir analyzed for y-T

... activity.

operated more than 5 percent of its W m d.

During plant' steady stese operellon and fogowing an operation under this exception to the equilibrium.llmits.'If M

dfges @ h (W h:

Steam JW Air the lodino concentration exceeds the equilibrium limit by W 164% wNhin a more than a factor of 10. the reactor shall be placed in a a

W ct m M rated cold condition within 24 g

.g g g and analyzed for groes gamme activity. At!ieset

.1 three sampies we be taken m 4 $omitted whe 1ntervals. These how S

_,g,g, %,

ma Ty the equNibrium I-131 concentratEii in the ren@

coolant is less than 0.007'pCl/ml.

- Amendment No. I 9 q

139 r\\Our

(

4.6 (cont'd) If JAFNPP c.

the gross activity counts made in accordance with a,

c, and d above indicate a

total iodine concentration in execcs of i

l 0.00.7 p Ci/:nl, a quantative determi-.: tion sn211 he made for I-131 and I--133.

2.

During startups and at stca:ning 2.

The reactor coolant water shall rates below 100,000 lb/hr.,

and not exceed the following limits when the conductivity of the stith stcaning rates less than reactor coolant exceeds 100,,000 lb/hr c:: cept

.as 2

mhos/cm, a sampic o.f r,cactor specified in 3.6.C.3:

coolant shall be taken, every It hr and analyzed, for Conductivity 2 s.mho/cm conductivity and chlorido u

Chloride ion 0.1 ppm content.

3.

For reactor' startups the 3.

a.

With steamitig rates mari2num value for conductivity greater than or* equal to chall not c:<ceed 10 p.mbo/cm 100,000 lb

/hr, a reactor a r.d the ma::1:r.um value for coolant sample shall be chloride ion concentration taken at least every 96 shall not exceed 0.1 ppm, for and whenever.

the the first 211 M after placing r>

c'ontinuous conductivity p

f the reactor /\\ is tbc power U inonitors indicate cbnormal L T operating condition. Durin9 conductivity (other than nOur5 reactor shutdov.'ns, specification short-tonn npikes),

and 3.G.C.4 vill apply..

analyzed for conductivity b

and chloride ion content.

h

^

b.

When the continuous i

conductivity monitor is fl0urs inoperable, a

reactor coolant samplo shall be taken at least daily and analyzed for conductivity and-chloride ion content.

Amendment No.

7 140

t()

(

/

JAFNPP 3.6 (cont'd) 4.6 (cont'd)

F.

Structural Integrity F.

Structural Integrity The structuralintegrity of the Reactor Coolant System shall bo rnalntained at the lovel required by the original acceptanco

" " " U" m-starxiards throughout the life of the Plant.

I' 2 l

c rw.er48 and supports in ercms.w with the l

requirements of the weld and support inservice inspection program. This inservice inspection program is based on an NRC approved edition of, and addenda to,Section XI of the ASME Boiler and Pressure Vessel Code which is in offect 12 months or less prior to the beginning of the inspection interval.

2.

An augmented inservice inspection program is required for those high stressed circumferentief piping joints in the main steam and feedwater lines larger than 4 loches in diameter, where no restra!nt against pipe whip is provided.

The augmented Irvservice inspection program shall consist of 100 percent inspection of these welds per inspection interval.

n G.

Jet Pumps 3.

An inservice irir,pectice Program for piping identified in the Whenever tho reactor is in the startup/ hot standby or run NRC Generic Letter 88-01 shall be implemented in modos, all jot pumps shall be operablo. If it is dotormined that a accudarici with NRC staff positions on schedules, jot pump is inoperable, the reactor shall be placed in a cold methods, personnel, and sample expansion included in condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

this Generic Latter, or in acordance with attemato measures approved by the NRC staff.

y C

G.

Jet Pumps C

Whenever there is recirculailon flow with the reactor in the startup/ hot standby or run modes, Jet pump operability shall be chocked daily by verifying that the following conditions do not occur simultaneously:

e g

144

(Q

'\\

~,

xs JAFNPP 3 6 and 4 6 BASES (cont'd)

~

r9

[0~

Deleted a

a ng at appmpnaM conconkaNn Ms M M sampling for isotopic analysis can be initiated. The design details of such a system must be submitted for ovaluation and C.

Coolant Chemistry accepted by the Commission prior to its implorribntation ar-d A radioactivity concentration limit of 20 pCi/mi total iodine can k

l be reached if the gaseous effluents are near the limit as set Since the concentration of radioactivity in the reactor coolant is g l

forth in Radiological Efiiuent Technical Specification Section not continuously measured, coolant sampling would be v 3.2.a if there is a failuto or a prolonged shutdown of the inoffective as e means to rapidly detect gross fuel elemont cleanup domineralizer.

failures. However, some capability to detoct gross fuel element failures is inherent in the radiation monitors in the offgas in the event of a steam lino rupture outsido the drywell, with this system and on the main steam lines.

coolant activity level, the resultant radiological dose at the site boundary would be 33 rem to the thyroid, under adverso Materials in the Reactor Coolant System are primarily 304 meteorological conditions assuming no more than 3.1 pCl/gm stainless steel and Zircatoy fuel cladding. The reactor water of doso equivalent i-131. Tho reactor water samplo will to used chemistry limits are established to prevent damage to those l

10 assure that the limit of Specification 3.6.C is not exceeded materials. Umits are placed on chlorido concentration and The total radioactivo iodino activity would not be expected t_o conductivity. The most important limit is that placed on change rapidly over a period of 96 Qw.n addition, the trend of chlorido concentration to provent stress corrosion cracking of the stack offgas release rate, whichTs'bontinuously monitored, i the stainless stool. The attached graph, Fig. 4.6-1, illustrates is a good indicator of the trend of the iodine activity in the DOurs, the resutts of tests on stressed 304 stainless steel specimens.

reactor coolant. Also during reactor startups and large power Failures occurred at concentrations above the curve; no changes which could affect iodine levels, samples of reactor failures occurred at concentrations below the curve. According coolant shall be analyzed to insure lodino concentrations are to the data, allowable chloride concentrations could be set below allowablo levels. Analysis is required whenever the 1-131 several orders of magnitude above the estabilshed limit, at the concentration is within a factor of 100 of its allowable oxygen concentration (0.2 0.3 ppm) experienced during power equilibrium value. The necessity for continued sampling operation. Zircaloy does not exhibit similar stress corrosion following power and offgas transients will be reviewed within 2 fai!ures, years of initial plant startup.

However, there are various conditions under which the The surveillance requirements 4.6.C.1 may be satisfied by a dissolved oxygen content of the reactor coolant water could be continuous monitoring system capable of determining the total higher than 0.2-0.3 ppm, such as refueling, reactor startup, and iodine concentration in the coolant on a real timo basis, and hot standby. During these periods with steaming rates less 1/9

}

nmendmeni no.

149

(E) u' JAFNPP 3 G and 4.6 BASES (cont'd) than 100,000 lb/hr, a more restrictive limit of 0.1 ppm has been startup periods, which are in the category of loss than 100,000 established to assure the chloride-oxygen combinations of Fig.

Ib/hr, conductivity may exceed 2 pmho/cm because of the initial 4.6-1 are not exceeded. At steaming rates of at least 100,000 Ib/hr, boiling occurs causing doacration of the reactor water, ovolution of gases and the initial evolution of gases,and the initial thus maintaining oxygen concentration at low lovels.

addition of dissolved metals. During this period of timo, when the conductivity exceeds 2 prnho/cm (other than short-term spikes), samples will be taken to assure the chloride When conductivity is in its proper normal range, pH and chlorido concentration is loss than 0.1 ppm.

and other impurities affecting conductivity must a8so be within their normal ranges. When and if conductivity becomes The conductivity of the reactor coolant is continuously abnormal, then chlorido measurements are mado to determino whether or not they are also out of their nonnat operating values.

monitored. The sampics of the coolant wtilch are taken every 9G hr 'll serve as a reference for calibration of these monitors and This is not necessarily the casa. Conductivity could be high due h is considered adequate to assure accurato readings of the to the presence et a neutral salt; e.g., Na SO, which would not monitors. If conductivity is within its normal range, chlorides and g

4 have an effect on pH or chlorido. In such a caso, high other impurities will also be within their normal ranges. The conductiv:'y alone is not a cause for shutdown in some types of reactor coolant samples will also be used to determino the water-cooled reactors, conductivities are, in fact, high due to chloridos. Therefore, the sampling frequency is considered purposeful addition of additivos. In the case of BWR's, however, W

where no additives are used and wtwre neutral pH is maintained, adequato to detect long term changes in the chlorido ion g

conductivity provides a very good measure of the quality of the content. Isotopic analysos of the reactor coolant required by reactor water. Significant changes therein provide the operator Specification 4.6.C.1 may be performed by a gamma scan.

with a waming mechanism so he can investigate and remedy the bouf5 condition causing the change before limiting conditions, with D.

Coolant Leakage respect to variables affecting the boundaries of the reactor coolant, are exceeded. Methods available to the operator for correcting the condition include operation of the Reactor Allowable leakage rates of coolant from the Reactor Coolant System have been based on the predicted and experimentally Cicanup System, reducing the input of impurities and placing the observed behavior of cracks in pipes and on the ability to make reactor in the cold shutdown condition. The major benefit of up Reactor Coolant System leakage in the event of loss of off-cold shutdown is to reduce the temperature dependent sito a c power. The reird expected background leakage duo i

corrosion rates and provide timo for the Reactor Water Cleanup System to reestablish the purity of the reactor coolant. During to equipment design and the detection capability for determining system Amendment No.

150

s 3.6 4.6IIASES(cont %I)

JAf1ttP 3

leakage were also conaldered in than the rapgnitudo specified can be detectnd roanonably in a saatter of a ODO establishing the limits.

The hahavlor of cract.s in piping systeins few hours utilizing the available has been crperimentally and I cakagit detection schemos, and if analytically investigated as part of the origin cannnt be determined in a the usM.C-ciencored Iteactor Prlinary reason.shly abort timo, the Plant tahoulal lea chiit down t o allow further Casolant Sys t ete Itu pt.itr e St.indy (the 1 ipe lenpturo Studyl. 1;ork ut.111 ming i nvo::t ig.st lo: and correctivo action.

the sta t a obtained 1:n ihls atudy I sid i cators tie.it. leakages f rtwa a crack The capacity of the drywell sump c.in less sic tuct end ler f te ro the crack pumpre in 100 rJp:n, and the capacity grows to a slange rours or critical of the drywell equipreent drain tank ni:o by mechanically or thermally pumpn in also 100 gpm.

Ite:soval.of I nd.:ce:I cyclic loading, or otrears So gple f ror's either of theco su:nps cueronlon cracking or so:mo other can leu accoxplinhed ulth con-r ete;ha n t s:a cha racterr land by gradual siderable margin.

s at k growtli.

~ This evidence tuygosts Lliat for leakage cosecubat The perfoniaance of the Itcactor Coolant 1.c ak agio Detection Systea gro. ster than the limit specifLed for will be evaluctOd during t.he first 5 unlelentlfled

leahano, the pro-t,ahllity is sinall that linperfectiosis ye of plant operation, and the onclusloas of this ovaluation will or cracks "ansociated with such gy$

19 atago unseld grow rapidly.

I he rerorted to the liitC.

Ilowove r,

the estatelishaput of allowable unidentitlud levakago It is estimnted that the realn stea n quenter then that glven in 3.6.D line tunnel leakage detectors are on the hasin of the data presently ca p.ahlo of detecting a Icak on the avallahlo uculd 1:o pre:saturo becanto orator of 3,500 lb/hr.

The systea of uncertaint les arnociated wit.h the p9et.oriaance u 1 hn evaluated diaring d.:t a.

I'or Inahago of t he ordne of 5 the firnt_5 of plant operation,

)

Spa as

pecified in 3. 6.li, the C GT-5 a n,1 the nelusions of the exper i: ::enta l and analytical data esaluation

'be reported to the n J990ut a reasonable margin of IntC.

nefoty su :h that leakage of'this The eactor coolant leakage detection systems s.agnitude would not result front a crack areproaching the critical nLzo consist of the drywull sump toonitoring systein f or rapid preng.agation. henkage less and the drywell continuous atmosphere snonttoring sy n tese. The drywell continimous alsnosphere inosiltoring syntesi utilizes a three-channot sionttor to provido information on particulato, lodino and noble gas activit.las in the drywell atsansphere.

Two Indttpendent and rodovulani syntems are pro-vided to perforse this function. This systein' siepptomonto tho drywoll sus.p monitoring syntese in detecting abnoriaal leakago that could occur f rove the reactor coolant uyntese.

In the event that the dcywell contismous atsposphere inonitoring 151 Amendment 110

, y,,,,, 3, 3,,p,,,,,9 y,,3,,,pg, util he taken em a sw rlottic hasis to snonttor dryngil activity.

JAFNPP 3.7 UMITING CONDITIONS FOR OPERATION 4.7 SURVEll1ANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Applicability:

Applicability:

Applies to the operating status of the primary and secondary Applies to the primary and secondary containment integrity.

containment systems.

Objective:

To assure the integrity of the primary and secondary cordainment To verify the integrity of the pr secondary containment systems.

systems.

Specification:

A Primary Containment A.

Primary Containment O

nl 1.

The volume and temperature of the water in the torus shall 1.

The torus water level and temperature shall be monitored be maintained within the following limits whenever the as specified in Table 4.2-8. The accessible interior reactor is critical or whenever the reactor coolard surfaces of the drywell and above the water line of the temperature is greater than 212 F and irradiated fuel is in torus shall be inspected at each refueling outage for the reactor vessel:

evidence of deterioration. Wtwnever there is indication of relief valve operation or testing which adds heat to the a.

Maximum vent submergence level of 53 inches.

suppression pool, the pool temperature shall be b.

Minimum vent submergence level of 51.5 inches.

conyy MtaM W also hM MW my 5 minutes until the heat addition is terminated. Whenever l

The torus water level may be outside the above there is indication of relief valve operation with the limits for a maximum of four (4) hours during temperature of the suppression pool reaching 100"F or required operability testing of HPCI, RCIC, RHR, CS.

more and the primary coolant system pressure greater l

and the Drywell-Torus Vacuum System.

than 200 psig, an extemal visual examination of the torus l

sW be cWucted befse rewng pows opwation.

g c.

Maximum water temperature i

O (1)

During normal power operation maximum O#

water temperature shall be 95'F.

Amendment No. 36,jfili.

165

I JAFNPP 4.7 (cont'd)

Type A test shall be performed at each plant shutdown for refueling or approximately every*

18 months, whichever occurs first, until two

've Type A tests meet the acceptance O criteria.

l L

b. Type B tests (Local leak rate testing of containment 9

penetrations)

(1.)

All preoperational and periodic Type B tests shall be performed by local pneumatic pressurization of the containment penetrations, either individually or in groups, at a pressure not less than Pa, and the gas flow to maintain Pa shall be measured.

(2.) Acceptancecriteria The combined leakage rate of all penetrations and valves subject to Type B and C tests shall be less than 0.60 La, with the exception of the valves sealed with fluid from a seal system.

In accordance with an exemption from 10 CFR 50 3

Appendix J, a Type A test need not be performed during the 1988 refueling outage.

1[5 Amendment No.

170 i

-+

m

~

JArsePP

~

4J(corfd)

Type Cleet.

Type C tests shaR he performed dunng eam raar*w shutdoum lor sma% but in rm case atinierveis greeder then two D

M Otter leek rete tests O in Secton 4.7d shat be pedormed during each reser shutdown for refueEng but in no. case at intervals geester men two years.

O 1.

Containment mogrmainn%

\\

Any mayor mareerneson, reploosment of a componema which is part d the pnmery rear *v (l containment boundary, or ramaaling a meet-weided \\/

door, pertormed siter the preoperstand leakage rate test sheB be tasowed by elmer a Type A Type B, O

or Type C tout, as appucable, for the area asocsed by the mnresemainn. The rneemured leakeGe prom this test shsE be irwertant in the test report. The ar prance altaria as appropriego, shes be met.

unor % :=rdanamente, or reeeeang or seel-weidad doors, performed drectly prior to the conduct of a schedidad Type A heet do not rapara a seperate test.

In accordance wlBi an amermphon from 10 CFR 50 Appendix J, a Type A, B, or C test is not required for: 1. The rapisoament d Wie HPCI turbine exhaust line block valve (2MPI-11) during the 19tNI outage; or 2. The sepair of the Core Spnny test retum line weld 10-144184A during the 19ENDrneintenanceadary Amendrnent No. g)#,1)tf, p(,

174

...~

t

[I j

JAFNPP 4

3.7 BASES

<f.

A. PrimaryContainment

^

\\

r The integrity of the pimary containment and operation'of the n

Core CooEng W in combinadon Nmit h The pressure suppreselon pool water provides the host sink for I

D ** '"*'9Y "I'

1 l'

offsite doses to values less then those speclRed in 10 CFR 100 in the event of a break in the Reactor 'W System piping-

@ # " U" F***""

r a

l lThus, containment integrity is required whenever the potential chamber water volume must absorb the associated decay and ;

a sor vioseson of the Reactor Cooient System integrity adets.

structural eensible heat reisesed during reactor cooient p Concem about such a vlosellon esdots whenever the reactor is blowdown from 1,0a0 peig-1 1

critical and abovw atmospheric pressure. - An 'sucepuon to the Since all of the genes in the dyweN are purged._into the -

requirement -to maintain primary containment integrity is -

preneure suppression chamber. air spoos during a. lose of.

a aNowed during core ioeding and during, low power phyelos coolant accident, the prosaure resulung from:loothermal teenne when ready access to the mector veneel is required.

w.v.:::'

plus the vapor pressure of the liquid must not:

There wlN be no pressure on the erstwn at this ame, which we.

- 58 peig, the suppression chamber design pressure.

greatly reduce the chances of a pipe break. The reactor may The design volume of the suppreselon chamber (water and air) be taken crescal during this period, however, restriceve was obtained by considering that the total volume of reactor operating procedures ' and operadon - of - the i RWM, in.

coolant to be condensed is discharged to the suppression accordance wth W 3.3.B.3 minimize tio probabluty chamber and. that : the drywell volume is-purged.to ' the of an accident occuntng. Procedures in conjunction with the-suppreselon chamber (Section 5.2).

Rod Wbrth lenimizer. Technical W Nmit indvidual control worth such that the drop of any in sequence control rod would not reedt in a peak fuel enthalpy greater then 280 calories /gm. In the unilhely event that en excurolon dd occur,-

)

the reactor bulkung and Staruby Gas Treatment System, which shall be operational during this time, offers a eunicient barrier to

-U qF..

keep offsite doses well within 10 CFR 100..

o Amendment No. )d,

l-187.-

1-s 3

m m

JAFNPP 3.7 BASES (cont'd) -

complete contamment system, secondary containment is replaced whenever significant changes in filter efficiency occur.

required at all temos that gutmery containment is required as Tests (11) of impregnated charcoal identical to that used in the wed as durin0 refueErg.

finers indicate that shelf life up to 5 yr leads to only minor lho Stan@y Gas Treatment System is' designed to Riter and heeses b W W rem val W.

exhaust the reactor bukSng semosphere to tie main stack

. The 90 percent efficiency of the charcoal and particulate filters durin0 secondary containment lealetion concRelons wilh a is suf5cient to prove.W exceeding 10CFR100 guidelines for the minimum reisees of rarmaardve meterless from the reactor accedents analyzed. The analysis of the loss 4 coolant buktng to the erwirone. Bott steney gas treatment fans are accident assumed a charcoal filter efficiency of 90 percent, a doengned to automouceSy start upon contelnment innin8&nn; particulate filter effeciency of 90 percent, and TIO 14844 Assion however, only one ten is required to mainteln the reactor..

bulkSnD preneure at approelmstely a negative 1/4 in, water

. product source term. Hence, requiring 99 percent efficsoncy for both the charcoal and partsculate filters provides adequate gage preneure; ad leaka0s shouki be in4eekage. Each of the margin. A heater mairtains relative humidity below 70 percest -

N,) Treatment Svetem circuit is inoperable, the other circuit must in order to assure the efficient removal of methyl iodine on the two tens has 100 percent cepecity. ' If one Staney Gas V

impregnated charcoalfilters.

Oe Ms

m avsbabiky W the operable circuit and reeutts in no added rielq thus, reactor The operatmisty of the Standby Gas Treatment System (SGTS) '

8'

"'"**"U"""

must be assured W a design basis loss of cooled accident b operable,em N h W to a 6 h #m (LOCA) occurs whde the containment is bemg purged or system W W vented through the SGTS. Flow from containment to the SGTS is via 6 inch Wdve Number 27MOV-121. Smos the maximum While only a ames amount of per#culates is rolessed kwm the _

flow through the 6 inch line following a design basis LOCA is Pressure Suppreselon Chamber System as a reetdt of the loss.

within the design 5-i72 of the SGTS, use of the 6 inch line ~

of.cooient accident,4. "Ny particadete Stars are -

manures the operability of the SGTS.

specNied to minimize potendal panictdate rolesse to em wwironment and to prevent ciogging of the iodine star. The D-Primary Containment isolation Valves high-efRciency 91ters haws an ediciency greater then 99 percent Double isolation valves are provided on lines penetrating the h

for particulate matter larger than 0.3 micron. <The rninimum primary containment and open to the free spece nodine removal eNiciency is 90 percent. Fiber banks wiu be be.Neid'c.kOfCNehh.

Amendment No. 1 191

m

' ()

h

[

' JAFNPP 1

3.9 (cont'd)'

4.9 (cont'd)'-

fl 3.

Frorn and after the time that one of the Emergency Diesel 3.

The emergency diesel generator system instrurnentation Generator Systems is. made or found to be inoperable,'

shall be checked during the monthly generator test.

. continued reactor operation is permisseie for a period not-

' to exceed 7 days provided that the two incoming power sourcesL are ;available 'and that' the i remaining Generator System is operable. At the end'of the period,~ the reactor shall _be placed in La cold condition

]dD

,l

-within 24 ' hours,1 uniees the aNected; diesel generator systemis made operable sooner.

When both Emergency Diesel Generator Systems.'are 4.

. Once each operating cycle, the conditions under which the '

4.

made or found to be inoperable, a reactor shutdown shall Emergency Diesel Generator System is. required will bc be initiated within two hours and the reactor placed in a-simulated to d=T,cawtrate that the pair of diesel generators

. cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation of shutdown.

will stN,. accelerate, force parallel, and ' accept the -

- emergency lords in the prescribed sequence.'

5.

Once within one hour and at least once per twenty. lour hours thereafter while the reactor is being. operated in f accordance with Specifications 3.9.B.1,3.9.B.2, or 3.9.B.3.-'

the availability of. the operable Emergency Diesel Generators shall be demonstrated by manual starting and force paralleling where applicable. '

N-

-4 I

i l

l t

i i

l l

-' Amendment No ;16 78 217

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JN4FP 3.9 (cont'd)

O F.

Iici tiW Inwhperybnt ther Sipplies 1.

Ivractor shall not 1x2 trab critical unlcr.s Loth ind: pendent tuer supplies, includliv3 the batteries, inverters and chargers and their associated txises (IE-155 and tE-%5) atu in thrvice, exwpt as siccified beloa.

i I>arlsr3 t ~.nr operation, if one 2.

1 ubpo:xli:nt [nur surply boccues unavailable, repairs shall be '

nude intnediately and contitard reactorhuration is permissible for a tvirio4 tot to excxxx17 days unicas the unaypilable train is nvuh oper--

able nooper. Frun and af tur the date ore of Ihe in3cpencknt powr supplies t

is nnde!;br foussi to be inoperable for any reason, the following would alV1yt t

ti a.

'Ihe pther irvlependent power mqvly inc3pling its diar,Jer, inverter, battery and associated bus is qwipble.

b.

Pil t cell voltage, specific gravity and tenperature and overall battery voltage are treasured inundi-alc)y and weekly thereaf ter for the qxspdale indercraient power sulply batpery.

t

'Ihe{ingxtrable independent power c.

styply shall be isolated frun its associated IICI MN bus, and this Qhg gk I

Ints will bq uunually switched to its Maniptenancu)to.ur s suroa.

O 222b Amendment 1

g i

y

s JAFNPP E.

3.9 BASES (cont'd)

C.

Diesel Fuel D.

anwy Systen1 Minimum on-sito fuel oil requirements are based on operation 125 v DC power is supplied from two plant batteries each sized of the emergency diesel generator systems at rated load for 7 to supply the required equipment at design power following a days.

loss-of-coolant accident with a concurrent loss of normal and reservo power. Each battery is provided with a charger sized b

Additional diesel fuel can be delivered to the sito within 48 to maintain the battery in a fully charged stato while supplying hours.

normaloperatingloads.

If one of the Emergency Diosol Generator Systems is not E.

LPCI MOV Independent Power Supplies operable, the plant shall be permitted to run for 7 days provided both sources of reservo power are operational. This Thero are two LPCI MOV Independent Power Supplies each is based on Mollowing:

consisting of a charger, rectifier, inverter and battery. Each independent power supply charger-rectifier is normally fed from 1.

The operable Emergency Diesel Generator System is the emergency A-C power supply system to maintain the capable of carrying sufficient engineered safeguards and battery in a fully charged stato. In the event of a LOCA cach emergency coro cooling system equipment to cover all independent power supply is automatically isolated from the loss of-coolant accidents.

Emmgency AC pows system. W kuey M inMw have sufficient canncity to power, the MOV's_ essential to the 2.

The reservo (offsite) power is highly reliable.

operation of the LPCI SystemT(4 maintenanceDwer source is provided for each LPCI MOV, bus whereby in the event its independent power supply is f service, the LPCI MOV bus may energized directly fr the Emergency A-C Power k

An cMernake.

Amendment No. % 1 4

~4

~

c.

7 m-.

o.k),.

s

JAFNPp t '.. s,

. ~

3.9 BASES (cont'd)

P F.

Iteactor P,rotection System Power Supp1_ica nach of two HPS divisions may be supplied power from it's respective RPS HG set or from an alternate source which derives power from the same electrical division.

The HG sets and alternate sources for both divisions are provided with redundante seismlo qualified,. class IE electrical protection assemblics between the power cource and the RPS bus.. Any'ahnormal output type failure in either of the HG sets or alternate sourcen (if in service) would result in a trip of one'or both of the electrical protection assemblies producing a half scram on that HPS d1 Vision and retalning full ocram capability'in the other HPS division. -

Limiting operating conditions in section 3.9.6 provido a high degree of anhutance that RPd bused are protected as desdelbod s) above.

C

?

O e

'g AmendmentNo..fk

^

224a

<E). Na Used JAFNPP 4.9 BASES (cont'd)

{,

D.

ticry System Measurements and electrical tests are conducted at specified intervals to provido indication of cell condition and to determine the discharge capability of the batteries. Performance and service tests are conducted in accordance with the recommendations of IEEE 450-1987.

V{ q E.

LPCI MOV Independent Power Supply Measurement and electrical tests are conducted at specified

. intervals to provido indication of cell condition, to determino the

- dischargo tapability of the battery. Performanco and service tests are conducted in accordance with the recommendations of IEEE 450-1907.

F.

Reactor Protection Power Supplies Functional tests of the electrical protection assemblics are g!

conducted onco each six (6) months utilizing a built-in test device and once per operating cyclo by performing an instrument calibration which verifies operation within the limits i

of Section 4.9.G.

Amendment No.p,7,1[

l 7

[

m

G O

[

JAFNPP A.

High Pressure aster to Protection System (Cont'd) essum Water Rm RoWon %stm WW v

Il 1. be cannot be fulfilled, place the reactor in Hot hem 3.

wncy Standby within six (6) hours and in Cold Shutdown within the following thirty (30) hours.

' h.

Fire pump diesel engine Once/ Month by verifying the fuel storage tank contains at least 172 gallons of fuel.

] g "i 1.

Dieselfuel from each Once/ Quads W G TC.T tank obtained in accordance with ASTM-D27045 is within the acceptable limits for quality as per the following:

Flash Point *F 1257 min.

Pour Point "F 10"F max.

Water & Sediment 0.05% max.

Ash 0.01% max.

Distillation 90% Point 540 min.

Viscosity (SSU) @ 100*F 40 max.

O Sulfur 1% max.

Copper Strip Corrosion No. 3 max.

I Cetano #

35 min.

j.

Fire pump dicsci engine Once/18 months by inspection during shut down in accordance with procedures prepared in conjunction with manufacturers recommendations and verifying the diesci, starts from ambient conditions on the auto start signal and operates for >20 minutes while loaded with the fire pump.

Amendment No. p( [

244c

~.

j

..-l l

i Py 1

bl6 1

1 1p I a ll 1

il!{!!!

ll ilk

~

oa a

IlElikllil0!

7 i

W!!y 1

ll!l1l'!!!!$

3 i

e

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(M m

s l

l JAFNPP 3.12 and 4.12 BASES The Firo Protection System specifications provido pre-establishod B.

Safety related equipment areas protected by water spray or minimum levels of operability to assuro adequate fire protection during sprinkfors are listed in Table 3.12.1. Whenever any of the any operating condition including a design basis accident or safo shutdown earthquake.

protected areas, spray or sprinklers are inoperable continuous fire detection and backup fire protection equipment is available in the area where the water spray and/or sprinkler protection l

A.

The high pressure water fire protection system is supplied by was lost.

redundant vertical turbine pumps, one diesel driven and one electric motor driven, each desian rated 2500 com at 125 psig discharge pressure. @oth pumps take_ suction from the plant]

Performanco of the tosts and inspections listed in Table 4.12.1 insureDBoth pumps take suction from the plant intake cooling will prevent and detect nozzio blockage or breakage and verity p; water structures from Lake Ontario. The high pressure water fire header intogrity to ensure operability.

protection header is normally maintained at greater than 115 C.

The carbon dioxide systems provido total flood protection for psig by a pressure maintenance subsystem. If pressure eight different safety related areas of the plant from either a 3 ton decreases, the fire pumps are automatically started by their or 10 ton storage unit as indicated in Table 3.12.2. Both CO initiation logic to maintain the fire protection system header 2

prossure. Each pump, together with its manual and automatic storago units are equipped with mechanical refrigeration units to maintain the storage tank content at 0*F with a resultant pressure initiation logic combined makes up a redundant high pressure of 300 psig. Automatic smoke and heat detectors are provided water fire pump.

In the CO2 protected areas and initiation is automatic and/or manual as indicated in Tabio 3.12.2. For any area in which the A third fire pump, diesel-driven, has been installed and is set to CO2 protection is made or found to bo inoperable, continuous automatically actuate upon decreasing pressure after the fire detection is available and one or more large whooled CO actuation of the first two fire pumps. No credit is taken for this fire extinguisher is also available for each area in which 2

pump in any analyses and the requirements of Technical protection was lost.

Specifications 3.12 and 4.12 do not apply.

Pressure Maintenance subsystem checks, valve position checks, Weekly checks of storage tank pressure and level verify proper system flushes and comprehensive pump and system flow operation of the tank refrigeration units and availability of sufficient volume of CO to extinguish a fire in any of the 2

and/or performance tests including logic and starting subsystem protected areas.

tests provide for the early detection and correction of component failures thus ensuring high levels of operability.

Amendment No.

(~

(}

, }.

JAFNPP analyses 4

15.B Bases

=

The spent fuel pool high density fuel storage racks are ClassI structures signed to store up to 2,797 fuel bundies.

The storage racks ere designed to maintain a subcritical configuration havi a multiplication factor (kg ) less than 0.95 for all possi operational and abnormal conditions. The nuclear h

criticalitygor the Spent Fuel Racks (References 1 and 3) p>uvuo, mat fresh fuel bundles with 3.3 w/o U-235 meet the 0.95 k limit. This design basis buncRe was reanalyzed to Q{

determine its infinite lattice multiplication factor, k,, when in a g

reactor core geometry (Reference 2). This k, was obtained under conserva'.tve calculational assumptions and reduced by 2.33 times the standard deviation in the calculation resulting in the Technical Spocification limit of 1.36.

References:

1) increased Spent Fuel Storage Modification, Stone &

Webster Engineering Corporation, Boston, Mass. March

}

15,1978.

2)

General Electric letter, P. Van Dieman to G. Rorke, FitzPatrick Fuel Storage K-infinity Conversion, Revision 1, dated July 10,190G.

3)

Increased Storage Capacity for FitzPatrick Spent Fuel

)

Pool, Holtec International Mount Laurel, New Jersey, February,1909.

Cont \\ude, mendment No. J0f, 246a

t JAFNPP g.

On O

o 2.

An SRO or RO with a license limited to' fuel 3,,. shall directlylsupervise all Core Afternatio Chrs person shall directly supervisc e.oiu Alterations]This person sha!!

have no otner duties during this time; Mk&O LOUS 3.

A fire brigade of five (5) or more members shall be maintained on site at all times. This 1

excludes two (2) members of the minimum shift crew necessary for safe shutdown and any personnel required for other essential functions during a fire emergency; 4.

In the event of illness or unexpected absence, up to two (2) hours is allowed to restore the shift crew or fire brigade to the minimum complement.

p 5.

The Operations Manager, Assistant Operations Manager, Shift Supervisor and Aredars ! l Shift Supervisor shall hold a SRO license and the Senior Nuclear Operator and the Nuclear V Control Operator shall hold a RO license or an SRO license.

y 6.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; s.g., senior reactor operators, health physicists, auxiliary operators, and maintenance personnel who are working on safety-related systems.

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 plantis operating.

However, in the event that unforeseen problems require substantial amounts of overtime to

.2 be used or during extended periods of shutdown for refueling, major maintenance or M

major modifications, on a temporary basis, the following guidelines shall be followed:

a.

An individual should not be permrtted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excludng sNft tumover time.

b.

An indtvidual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven day period, all excluding sNf! tumover time.

c.

A break of at least eight hours should be allowed between work periods, including shift turrx:rver time.

d.

Except during _ extended shutdown periods, the use of overtime should be considered on an indmdual basis and not for the entire staff on a sNft.

Any deviabon from the above guidelines shall be authorized by the Resident Manager or p

. the General Manager - Operations, or Ngher levels of management, in accordance with Li estabhshed procedures and with documentabon of the basis for granting the deviation.

M Controls shall be included in the procedures such that indmdual overtime shall be I

reviewed montNy by the Resident Manager or Ns designee to assure that excessive hours.

have not been assigned. Routine deviation from the above guidelines is not authorized.

f Y

Amendment No. p,1f1,1, If7, 18 247a B

JAFNPP

~

GC 0 OjC L

6.3 Pl. ANT STAFF OUAUFICATIONS 6.3.1 The minimum quahfications with regard to educational background and experience for I

t plant staff positions shown in FSAR Figure 13.2-7 shall meet or exceed the minimump q quahfications of ANSI N18.11971 for comparable positions; except for the@adiolioicRi and Environmental Services Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

j 6.3.2 The Shift Technical Advisor (STA) shall meet or exceed the minimum requirements of either Option 1 (Combined SRO/STA Position) or Option 2 (Continued use of STA Position), as defined in the Commission Policy Statement on Engineering Expertise on Shift, published in the October 28,1985 Federal Register (50 FR 43621). When invoking Option 1, the STA role may be fiiled by the Shift Supervisor or Assistant Shift Supervisor. (1) 6.3.3 Any deviations will be justified to the NRC prior to an individua!'s filling of one of these positions.

NOTE:

(1)

The 13 individuals who hold SRO licenses, and have completed the FitzPatrick Advanced Technical Training Program prior to the issuance of Ucense Amendment 111, shall be considered qualified as dual-role SRO/ STAS.

6.4 RETRAINING AND REPLACEMENT TRAINING q

T (Ed A training program shall be maintained under the direction of the Training Manager to I l l

assure overall p*oficiency of the plant staff organization. !! shall consist of both retraining and replacement training and shall meet or exceed the minimum requirements of Section 5.5 of ANSI N18.1 1971.

i The retraining program shall not exceed periods two years in lenath with a curriculum f/

designed to meet or exceed the requalification requirements of(O CFR 55, Appendix Al in addition, fire brigade training shali meet or exceed the requirements of NFPA 27-h 1975, except for Fire Brigade training sessions which sha!! be held at least quarterly.

The effective date for implementation of fire brigade training is !Aarch 17,1978.

6.5 REVIEW AND AUDIT i

Two separate groups for plant operations have been constituted. One of these, the Plant Operating Renew Committee (PORC), is an onsite rev ew group. The other is an independent review and audit group, the offsite Safety Renew Committee (SRC).

\\o C W 65.5 %

_=, _ /

Amendment No. f, f, f,9f, if1, lh, if7, If8 248

(

.. m.

p f

5

=

D

~

S 4

[

- JAFNFI QDya '

.,G 7.0

[

(9)

Coll.

Robbins,

" Tests of a Full REFEkENCES 5

Scale 1/48 Segment of the !!umbolt N1)

E.

Janssen,

" Multi-Rod Burnout at g

Bay Pressure Suppression low Pressure," ASME Paper 62-HT-26, J

Containment," GEAP-3596, November

,A August 1962.

g 17, 1960.

. s rf 2 K.M.

Backer,. " Burnout Conditions l

(10) " Nuclear Safety Program Annual

. h) for. Flow of Bolling Water in Verti-g.

Progress Report for Period Ending cal Rod Clusters,"

AE-74

.A.

December 31, 1966, Progress Report

- A (Stockholm, Sweden), May.31962.

for Period Ending December 31, 1966, ORNL-4071."

-(3)

FSAR Section.11.2.2.

4 (11) Section 5.2 df the FSAR.

$8)

FSAR Section 4.4.3.

D l

y (12) TID 20583, " Leakage Characteristics

,(5)

I.M.-

Jacobs, " Reliability of Engi-of Steel Containment Vessel and the

{

neered Safety Features as a Func-

.t-nations."

Analysis of Leakage Rate Determi-tion of Testing Frequency," Nuclear s

Safety, Vol. 9, No.

4, July-August

^-

1968, pp 310-312.

(13) Technical Safety Guide,

" Reactor 4

Containment Leakage Testing und (6)

Benjamin Epstein, Albert Shiff, Surveillance Requirements," USAEC, UCRL-50451, Improving Availability Division of Safety Standards, and Readiness of Field Equipnent Revised Draft, December 15, 1966.

Through Periodic. Inspection,. July 16,

1968, p.

10, Equation '(24),

(14)-Section 14.6 of the FSAR.

Lawrence hadiation. Laboratory.

(15) ASME Boiler and. Pressure Vessel (7)

I.M.

Jacobs and P.W. Mariott, APED Code, Nuclear Vessels,Section III.

Guidelines for Determining:

Safe Maximum allowable internal pressure Test Intervals and Repair Times for is 62 psig.

Engineered.Safcguards - April 1969

-- M (16)

DCFR5_0.5d Appendix J,4Reactor _ Con-(8)

Bodega Bay Preliminary. !!azards Re--

A tainmentE Testing Require!EUnts. F port, Appendix--

1, Docket 50-205, December 28, 1962.

(17)

OCFR5 Appendix J, February-13, j

.1973.

Aoesa % %

4

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ML20034H195

Contents

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References:

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References:

7.0 REFERENCES

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