ML20045F999
| ML20045F999 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/29/1993 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20045G000 | List: |
| References | |
| NUDOCS 9307090279 | |
| Download: ML20045F999 (36) | |
Text
i ato O
t UNITED STATES i
T j
NUCLEAR REGULATORY COMMISSION C
WASHINGTON, D.C. 20E0001
%...../
POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET N0. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.190 License No. DPR-59 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Power Authority of the State of New York (the licensee) dated March 9, 1993, complies with-the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of tt.is amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical 4
Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-59 is hereby amended to read as follows:
l 9307090279 930629 PDR ADOCK 05000333 p
PDR g.
. t (2) Technical Scecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.190, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION Y O' W
Robert A. Capra, Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 29, 1993 1
4
ATTACHMENT TO LICENSE AMENDMENT NO.190 FACILITY OPERATING LICENSE NO.'DPR-59
~
DOCKET NO. 50-333 Revise Appendix A as follows:
Remove Paaes Insert Paaes i
i v
v i
8 8
20 20 29 29 32 32 54 54 78 78 79 79 123 123 136 136 i
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139 139 140 140 144 144 149 149 150 150 151 151 165 165 i
170 170 174 174 187 187 3
191 191 217 217 222b 222b 224
-224 224a 224a l
226 226 244c 244c 244g 244g 244h 244h 246a 246a 247a 247a 248
-248 285 285
'i V
-I 1
.s.
JAFNFP TECHNICAL SPECIFICATIOrif TABLE OF CONTENTS
?.il2ft 1.0 Definitions 1
LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 Fuel Cladding integrity 2.1 7
l 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.
. Radiatica 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 1.
4kV dmergency 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 3.5 Core and Containment Cooling Systems 4.5
- 112 A.
Core Spray and LPCI Systems A.
112 B.
Containment Cooling Mode of the RHR B.
115 System C.
HPCI System C.
117 l
D.
Automatic Depressurization System (ADS)
D.
119 E.
Reactor Core Isolation Cooling (RCIC)
E.
121
-j System j
Amendment No. f2,1 0,1 4,1I3,190 i
m
JAFNPP LIST OF TABLES Table Iit!n Eitqe 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 77e Response Times i
4.2-1 Minimum Test and Calibration Frequency for PCIS 78 4.2-2 Minimurn Test and Calibration Frequency for Core and Containment 79 Cooling Systems l
4 4.2-3 Minimum Test and Calibration Frequency for Control Rod Blocks 81 Actuation 4.2-4 (DELETED) 62 4.2-5 Minimum Test and Calibration Frequency for Drywell Leak Detection 83 4.2-6 (DELETED) j 4.2-7 Minimum Test and Calibration Frequency for Recirculation Pump Trip 85 Amendment No.
,10,1 1, 1 3,190
I JAFNPP
~
1.1 (cont'd) 2.1 (cont'd)
I b.
APRM Flux Scram Trio Settina IRefuel or StaII B.
Core Thermal Power Limit (Reactor Pressure s785 osial
& Hot Standby Mode)
When the reactor pressure is s785 psig or core flow is APRM - The APRM flux scram setting shall be l
less than or equal to 10% of rated, the core thermal s 15 percent of rated neutron flux with the I
power shall not exceed 25 percent of rated thermal Reactor Mode Switch in Startup/ Hot Standby power.
or Refuel.
l C.
Power Transient c.
APRM Flux Scram Trio Settinos (Run Modg)
To ensure that the Safety Limit established in Specification (1)
Flow Referenced Neutron Flux Scram Trip 1.1.A and 1.1.B 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 l
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 I
~
rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 117% of rated thermal power.
Amendment No. 1
,7,
,1 4, 1 2,190
JAFNPP 2.1 BASES (Cont'd) l B.
Not Used
~
C.
References 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 Singlo-Loop Operation, NEDO-24281, August,1980.
l l
Amendment No. If,4,
,1 2,
190 (Next page is 23) l E...
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 ossential 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.
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 peessure 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 teactor 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 et 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 til 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 psil 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 l Containment Short Term Program are satisfied.
Amendment No.
, 1 4,
190
JAFNPP 3.1 BASES l A.
The reactor protection system automatically initiates a The outputs of the subchannels are combmed 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 offectiveness of the protection system is preserved.
input from at least one instrument channel which monitors a critical parameter.
Three APRM iristrument 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. 1
.5, 190
9 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 Emeroency Bus Undervoltaae Trio indicated in Table 4.2-8.
The limiting conditions for operation for the instrumontation 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.
If6, If0, If1, 190 Amendment No.
54
I JAFNPP
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TABLE 4.2-1 MINIMUM TEST AND cal.lBRATION 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 4
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 2,190
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) l 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)
(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 None NOTE:
See notes following Table 4.2-5.
, If6,1
,1 0, 1 1, 190 Amendment No.1
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 exiallocation 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. Averano Planar Linear Heat Generation Rate (APLHGR) 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,,f,If9,If7,If2,If4,1 2,190 Amendment No.
m.
JAFNPP 3.6 LIMITING CONDITIONS FOR OPERATION 4.6 SURVEILLANCE REQUIREMENTS 3.6 REACTOR COOLANT SYSTEM 4.6 REACTOR COOLANT SYSTEM Wicability:
Acolicability:
Applies to the operating status of the Reactor Coolant System.
Applies to the periodic examination and testing requirements for the Reactor Coolant System.
Obiective:
Objective:
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:
Soecification:
A.
hessurization and Thermal Limits A.
Pressurization and Thermal Limits 1.
Reactor Vessel Head Stud Tensioning 1.
Reactor 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 flange and the reactor vessel flange temperatures shall be flange and the reactor head flange are at least 90*F.
recorded:
I a.
Every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor vessel head flange is s 120*F and the studs are tensioned.
b.
Every 30 minutes when the reactor vessel head flange is s 100*F and the studs are tensioned.
c.
Within 30 minutes prior to and every 30 minutes during tensioning of reactor vessel head bolting 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 le.... 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, l
one hour period shall be:
1[3, If8,190 Amendment No.1
JAFNPF 3.6 (cont'd) 41 (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.
Coolant Chemistry C. Coolant 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 overy 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and analyzed for gross gamma activity.
l equilibrium value of 3.1 Ci/gm of dose equivalent 1-131. 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 and following an i, line 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period or a power l
l in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
level change of 220 percent of full rated power /hr reactor coolant samples shall be taken and analyzed for gross gamma activity. At least three samples will be taken at 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> intervals. These sampling requirements l
may be omitted whenever the equilibrium 1-131 concentration in the reactor coolant is less than 0.007 pCi/ml.
Amendment No. I 9, 190 139
JAFNPP 4.6 (cont'd) e.
If the gross activity counts made in accordance with a, c, and d abovo indicato a total iodino concentration in excess of 0.007 pCi/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 samplo of reactor coolant shall be Conductivity 2 pmho/cm taken overy 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 shcIl not exceed 10 pmho/cm and the maximum value 100,000 lb/hr, a reactor coolant samplo shall be for chloride ion concentration shall not exceed 0.1 ppm, taken at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and whenever tho l
l for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after placing the reactor in the continuous conductivity monitors indicato 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 chlorido ion content.
i 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.
l f,190 Amendment No.
140
JAFNPP 4.6 (cont'd) 3.6 (cont'd)
F.
Structural Intearity F.
Structural Intearity 1.
Nondestructive inspections shall be performed on the The structuralintegrity of the Reactor Coolant System shall be ASME Boiler and Pre ure Vessel Code Class 1,2 and 2 maintained at the level required by the original acceptance components and supgarts 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 those welds 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.
methods, personnel, and sample expansion included in this Generic Letter, or in acordance with alternate measures approved by the NRC staff.
G. Jet Pumos G. Jet Pumos Whenever the reactor is in the startup/ hot standby or run Whenever there is recirculation flow with the reactor in the If it s determined that a startup/ hot standby or run modes, jet pump operability shall be i
modes, all jet pumps :; hall be oper able.
jet pump is inoperaF e, the reacto" shall be placed in a cold checked daily by verifying that the following conditions do not condition within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.
occur simultaneously:
Amendment No. f, If4, If0, 190 144
e 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 st,bmitted 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 Ci/ml 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 domineralizer.
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 adverso Materials in the Reactor Coolant System are primarily 304 meteorological conditions assuming no more than 3.1 pCi/gm stainless steel and Zircaloy fuel cladding. The reactor water of dose equivalent I-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 chlorido concentration and exceeded. The total radioactive iodine activity would not be conductivity. The most important limit is that placed on
[
expected to change rapidly over a period of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. In chloride concen' ration 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, illustrates 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 l
startups and large power changes which could affect iodine occurred at concentrations below the curve. According to the l
levels, samples of reactor coolant shall be analyzed to insure data, allowable chlorido 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 l
of 100 of its allowable equilibrium value. The necessity for operation. Zircaloy does not exhibit similar stress corrosion i
continued sampling following power and offgas transients will failures.
be reviewed within 2 years of initial plant startup.
l However, there are various conditions under which the The surveillance requirements 4.6.C.1 may be satisfied by 3 dissolved oxygen content of the reactor coolant water could be l
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 l
Amendment No. 1 9,190 149
JAFNPP 3.6 and 4.6 BASES (cont'd) than 100,000 lb/hr, a more restrictive limit of O 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 pmho/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 doaeration 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 chlonde 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 <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 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 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. Theref ore, the sampling frequency is conductivities ate, 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 Leakaae 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.
Amendment No. If 9,190 150
JAFNPP i
l 3.6 and 4.6 6ASES (cont'd) leakage 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 &ywell 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 l
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.
l 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 i
probability is small that imperfections or cracks associated are capable of detecting a leak on the order of 3,500 lb/hr. The with such leaksge 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 basis of the data presently reported to the NRC.
available wotid 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 rize for rapid propagation. Leakage less than the to provide information on particulate, iodine and noble gas magr Lade soecified 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 l
time, the Plant should be shut down to allow further detocting abnormal leakage that could occur from the reactor l
inv9stigation and corrective action.
coolant system. In the event that the drywell continuous atmosphere monitoring system is inoparable, grab sample will be taken on a periodic basis to monitor drywell activity.
l l
i Amendment No. f 190 151
e JAFNPP 3.7 LIMITING CONDITIONS FOR OPERATIQH 4.1 SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Aonlicabilitv:
A_foficability:
Applies to the operating status of the primary and secondary Applies to the primary and secondar< containment integrity.
containment systems.
Obiective-Obiective:
To assure tw integrity of the primary and secondary containment To verify the integr"y of the primary and secondary containment l
systems.
systems.
St,ecificatbn:
Soecification:
A.
Primary Containment A. Primary Containment 1.
The volume and temperature of the water in the torus 1.
The torus water level and temperature shcil be monitored shall be maintained within the following limits whenever as specified in Table 4.2-8. The accessible interior
'he 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 fuelis torus shall be inspected at esch 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 i
s.
Maximum vent submergence level of 53 inches.
suppression pool, the pool temperature shall be continually l-monitored and also observed and logged every 5 minutes l
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 operab;lity 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.
Amendment No. f.1 8,Ifl.190 165
1 a
4 JAfNPP 4.7 (cont'd)
Type A test shall be performed at each plant shutdown for refueling or approximately overy 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 end periodic Type B tests shall be performed by local pneumatic pressurization of the containment penetrations, eitner 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 the9xception of the valves sealed with fluid from a l
seal system.
I 1
I i
l l
l l
l l
l Amendment No.1 5,190 l
170 l
.s.
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 refooling 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, j
shall be met. Minor modifications, replacement <-
or resealing of seal-welded doors, performed directly prior to the conduct of a scheduled Type
[
A test do not require a separate test.
i i
S
. ~..
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-..,..., -, ~.... ---,. -..,. -.. ~
r. --
.-...--,.,~....L.,,--..----...m._
~... ~ a _.a.,
JAFNPP 3.7 DASIS A. Erimarv Containment The integrity of the primary containment and operation of the The pressure suppressicn 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 Systun integrity exists.
blowdown from 1,020 psig.
Concern 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 wh9n ready access to the reactor vessel is 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 by 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, If5,190 187
4 JAFNPP 3.7 BASES (cont'd) complete containment system, secondary containment is be replaced whenever significant changes in filter efficiency required at all times that primary cor.tainment is required as occur. Tests (11) of impregnated charcoal identical to that used I
well as during refueling.
in the filters indicated that sholf 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 TlO 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 officiency 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, if one Standby Gas methyl iodine on the impregnated charcoal filters.
Treatment System circuit is inoperable, the other circuit must j
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 neithee frcuit 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 Number 27MOV-121. Since the maximum flow through the 6 inch line following a design basis LOCA is within l
While only a small amount of particulates is released from the the design 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-coobnt 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 officiency is 99 percent. Filter banks will Amandment No. 1[4,190 191
6 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless the affected diesel generator system is made operable sooner.
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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiation of will start, accolerate, 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 3.9.8.3 the availability of the operable Emergency Diesel Generators shall be demonstrated by manual starting and force paralleling where applicable.
t Amendment No. f. f, If1, If3,190
o JAFNPP 3.9 (cont'd)
F.
LPCI MOV Indeoendent Power Sucolies 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 operable 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 MOV bus, and this l
bus will be manually switched to its alternate power source.
Amendment No. f,190 222b
~-
h 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 sito 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 shsil 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 suppte is out of service, the LPCI MOV bus may be energized directly from the Emergency A-C Power System.
Amendment No. f, If4,190 224
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 alternate 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 isoth of the electricel protection assemblies producing a half scram on that RPS division and retaining full scram capability in tte 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. 7f.' 190 224a
JAFNPP 4.9 BASES (con't) 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 Indeoendent Power Sucolv Measurement and electrical tests are conducted at specified intervals to provide indication of cell condition, to determine the discharge capability of the battery.
Parformance and service tests are conducted in accordance with the recommendations of IEEE 450-1987.
l G.
Reactor Protection Power Suonfies Functional tests of the electrical protection assemblies are conducted at specified intervals utilizing a built-in test device and once per operating cycle by performing an l
instrument calibration which verifies operation within the l
limits of Section 4.9.G.
l l
l l
l l
Amendment No. [%p,[, J,86, 190 226
s JAFNPP l
A.
Hiah Pressure Water Fire Protection System (Cont'd)
A.
Hioh Pressure Water Fire Protection _Eyggm (Cont'd) 3.
If 1. above cannot be fulfilled, place the reactor in Hot item Frecuency Standby within six (G) 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 acceptable limits for quality as per the following:
Flash Point
- F 125"F min.
Pour Point
'F 10cF max.
Water & Sediment 0.05% max.
Ash 0.01 % max.
Distillation 90% Point 540 min.
Viscosity (SSU) @ 100 F 40 max.
Sulfur 1% max.
Copper Strip Conrosion No. 3 max.
Cetane #
35 min.
- j. Fire pump diesel engine Once/18 months by inspection during shut down in accordance with procedures 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.
Amendment No. 3f, If4,190 244c
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JAFNPP 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 psig Performance of the tests and inspections listed in Table I
discharge pressure. Both pumps take suction from the plant 4.12.1 will prevent and detect nozzle binckage 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 eithat 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 manuel and Both CO, storage units are equipped with mechanical automatic initiation logic combined makes up a redundant high refrigeration units to maintain the storage tank content at pressure water fire pump.
0"F with a resultant pressure of 300 psig. Automatic smoke and heat detectors are provided in the CO, protected areas 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 requiremants of Technical available and one or more large wheeled CO, fire 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 inc.luding logic and starting subsystem proper operation of the tank refrigeration units and tests provide for the early detection and correction of availability of suf ficient volume of CO, to extinguish a fire in component failures thus ensuring high levels of operability.
any of the protected areas.
Amendment No. f, If2, If 6, 190 244h
W JAFNPP 5.5.8 Bases The spent 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 O.95 for all possible operational and abnormal I
conditions. The nuclear criticality analyses for the Spent Fuel Racks (References i 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 Fuel Storage Modification, Stone &
l 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 e
1, dated July 10,1986.
- 3) Increased Storage Capacity for FitzPatrick Spent Fuel Pool, Holtec International, Mount i.aurel, New Jersey, February,1989.
i Amendment No. If1, If 5, 190 246a
4 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 fol! owed:
a.
An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.
b.
An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period, all excluding shift turnover time.
c.
A break of at least eight hours should be allowed between work periods, including shift turnover time.
d.
Except during extended shutdown periods, the use of overtime should be considered on an individual basis and no! 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. f, I f1, If0, If 7, If 8,190 247a l
i I
I
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 Po! icy 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. (11 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.
l 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 l
requirements of NFPA 27 1975, except for Fire Brigade training sossions 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).
i Amendment No. 2f,, Sf, f, Ifl, If4,1 7,1 f 8,190 l
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4 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 31,1960, 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 Containrmnt Leakage Testing and Surveillance Requirements,"
(6)
Benjamin Epstein, Albert Shiff, UCRL-50451, improving USAEC, Division of Safety Standards, Revised Draf t, 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.
l (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 Ill. Maximum allowable internal Engineered Safeguards - April 1969.
pressure is 62 psig.
l (8)
Bodega Bay Preliminary Hazards Report, Appendix 1,-
(16) 10 CFR 50.54, Appendix J, " Reactor Containment l
l Docket 50-205, December 28,1962.
Testing Requirements."
l (17) 10 CFR 50, Appendix J, February 13,1973.
l Amendment No. 190 j
285
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