ML20197F127
| ML20197F127 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/06/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Power Authority of the State of New York |
| Shared Package | |
| ML20197F131 | List: |
| References | |
| DPR-59-A-098 NUDOCS 8605150438 | |
| Download: ML20197F127 (40) | |
Text
_ _ _ _ - - -
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/
o UNITED STATES
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NUCLEAR REGULATORY COMMISSION n
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WASHINGTON, D. C. 20555
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POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 98 License No. DPR-59 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Power Authority of the State of New York (the licensee) dated October 11, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 rulas and regulations of the Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
i 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license 1
amendment and paragraph 2.C(2) of Facility Operating License No.
DPR-59 is hereby amended to read as follows:
8605150438 860506 PDR ADOCK 05000333 P
t
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1 1 (2) Technical Specifications The Technical. Specifications contained in Appendices A and B, as revised through Amendment No. 98, 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.
FOR THE NUCLEAR REGULATORY COMISSION 0
M D niel
. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: May 6, 1986
' t s
I ATTACHMENT'TO LICENSE AMENDMENT NO. 98 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET N0. 50-333 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Pages 11 vii 7
8 8a*
9 10 10a*
12 14 15 20 23 31 4
41 43a 72 74 123 124a 124tf 124c*
130 131 134 135g 135h 1351 i
145 145a 145b-3 145c i
145d 145e 145f j
145g 155 156 4
- Page added
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JAFNPP f
i TABLE OF CONTENTS (cent'd) j
.P331 I
F.
Minimum Emergency Core Cooling System F.
122 l
Availability G.
Maintenance of Filled Discharge Pipe G.
122 H.
Average Planar Linear Heat Generation H.
123 Rate (APLHCR)
I.
Linear Heat Generation Rate (LHGR)
I.
124 J.
Thermal Hydraulic Stability J.
124a SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.6 Reactor Coolant System 4.6 A.
Thermal Limitations A.
136 B.
Pressurization Temperature B.
137 C.
Coolant Chemistry C.
139 D.
Coolant Leakage D.
141 E.
Safety and Safety / Relief Valves E.
142a F.
Structural Integrity F.
144 G.
Jet Pumps C.
144 l
H.
DELETED I.
Shock Suppressors (Snubbers)
I.
145b i
3.7 Containment Systems 4.7 165 A.
165 B.
Standby Gas Treatment System B.
181 C.
184 D.
Primary Containment Isolation Valves D.
185 3.8 Miscellaneous Radioactive Material Sources 4.8 214 3.9 Auxiliary Electrical Systems 4.9 215 A.
Normal and Reserve AC Power Systems A.
215 B.
Emergency AC Power System B.
216 C.
Diesel Fuel C.
218 D.
Diesel-Generator Operability D.
220 E.
Station Batteries E.
221 F.
LPCI NOV Independent Power Supplies F.
222a 3.10 Core Alterations 4.10 227 A.
Refueling Interlocks A.
227 B.
Core Monitoring B.
230 C.
Spent Fuel, Storage Pool Water Level C.
231 D.
Control Rod and Control Rod Drive Maintenance D.
231 3.11 Additional Safety Related Plant Capabilities 4.11 237 A.
Main Control Room Ventilation A.
237' B.
Crescent Area Ventilation B.
239 C.
Battery Room Ventilation C.
239 AmendmentNo.[
, 98, 11
7 JAFNPP LIST OF FIGURES Flaure Title Eage_
3.1-1
- Manual Flow Control 47a 47b 3.1-2 Operating Limit MCPR versus 4.1-1 Graphic Aid in the Selection of an Adequate Interval 48 Between Tests 4.2-1 Test Interval vs. Probability of Sys' ten Unavailability 87 3.4-1 Sodium Pentaborate Solution of System Volume-Concentration 110 Requirements 3.4-2 Saturation Temperature of Sodium Pentaborate Solution 111 3.5-1 Thermal Power and Core Flow Limits of Specifications l
3.5.J.1 and 3.5.J.2 134 s
3.5-6 (Deleted) 135d 3.5-7 (Deleted) 135e 3.5-8 (Deleted) 135f 3.5-9 MAPLHGR Versus Planar Average Exposure Reload 4, P8DRB784L 135g 3.5-10 MAPLHGR Versus Planar Average Exposure Reloads 4 & 5, P8DRB299 135h 3.5-11 MAPLHGR Versus Planar Average Exposure Reload 6 BP8DRB299 1351 3.6-1 Reactor Vessel Yhermal Pressurization Limitations 163 4.6-1 Chloride Stress Corrosion Test Results at 500*F 164 6.1-1 Management organization Chart 259 6.2-1 Plant Staff Organization 260 Amendment No. [ f[, [, [ J, /,ftI, 98,
/
vii
JAFNPP 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTRGRITY Applicability:
Appilcability:
The Safety Limits established to preserve the fuel The Limiting Safety System settings apply to trip cladding integrity apply to those variables which settings of the instruments and devices which are monitor the fuel thermal behavior.
provided to prevent the fuel cladding integrity Safety Limits from being exceeded.
objective:
objective:
The objective of the Safety Limits is to establish The objective of the Limiting Safety System Settings limits below which the integrity of the fuel cladding is to define the level of the process variables at is preserved, which automatic protective action le initiated to prevent the fuel cladding integritj Safety Limits from being exceeded.
Specifications:
Specifications:
A.
Reactor Pressure > 785 psim and Core Flow > 10%
A.
Trip Settints of Rated The limiting safety system trip settings shall be The existence of a minimuts critical power ratio as specified below:
(MCPR) less than 1.07 shall constitute violation of the fuel cladding integelty safety limit.
1.
Neutron Flux Trip Settinas l
hereafter called the Safety Limit. An MCPR Limit l
of 1.08 shall apply during single-loop operation.
a.
IRN - The IRN flux scram setting shall be set at 1120/125 of full scale.
l Amendment No. J 4 g, g, pf, 98, 7
JAFNPP 1.1 (cont'd) 2.1 (cont'd)
A.1.b.
APRM Flux scram Trip Settina (Refuel or Start & Hot Standby Mode)
APRM The APRM flux scram setting shall be _< 15 percent of rated neutron flux with the _Reactor Mode Switch in Startup/ Hot Standby or Refuel.
B.
Oore Thermal Power Limit (Reactor Pressure 1785 psix)
(1) Flow Referenced Neutron Flux Scram Trip Setting i
When the reactor pressure is _< 785 psig or core flow is less than 10% of rated,_ the core thermal When the Mode Switch is in the RUN power shall not exceed 25 percent of rated position, the APRM flow referenced flux thermal power, scram trip setting shall be:
C.
Power Transient S 10.66 W + 54% for two loop operation or:
To ensure that the Safety Limit established in S 1(0.66 W + 54% - 0.666W) for single j
Specification 1.1.A and 1.1.B is not exceeded, loop operation each required scram shall be initiated by its where:
j expected scram signal.
The Safety Limit shall be assumed to be exceeded when scram is accomplished S=
Setting in percent of rated by a means other than the expected scram signal.
thermal power (2436 MWT)
W=
Recirculation flow in percent of i
rated 1
6W= Difference between two loop and single loop effective delve flow at the same core flow.
( AW = 0 for j
two loop operation. A W for single loop operation is to be determined upon implementation of single loop j
operation.)
)
i Amendment No. / g, g, f[, 98, 8
1 i
JAFNPP j
2.1 (cont'd)
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.
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Amendment No. 98, 8a j
4
JAFNPP 2.1 (cont'd) 1.1 (cont'd)
D.
Reactor Water Level (Not or Cold Shutdown In the event of operation with a maximum fraction' of limiting power density (NFLPD) greater than Conditions).
the fraction of cated power (FRP), the setting Whenever the reactor is in the shutdown condition shall be modified as follows:
with irradiated fuel in the reactor vessel, the water level shall not be less than that corres-S _S(0.66 W + 54%)(FRF/MFLPD) ponding to 18 inches above the Top of Active Fuel for two loop operation or, when it is seated in the core.
,_(0.66 W + 54% - 0.666W)(FRP/MFLPD)
S for single loop operation.
Where:
fraction of cated thermal power FRP =
(2436 MWt) maximum fraction of limiting power MFLPD
=
density where the limiting power density is 13.4 KW/ft.
The ratio of FRP to MFLPD shall be set equal to
~
1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
(2) Fixed High Neutron Flux Scram Trip Setting When the Mode Switch is in the RUN position, the APRM fixed high flux scram trip setting shall be:
._120% Power S
Amendment No. }[ [, [, p/, ) [, 98, 9
JAFNPP l
~
1.1 (cont'd) 2.1 (cont'd) i i
A.1.d APRM Rod Block settina i
The APRM Rod block trip setting shall be:
i S <_(0.66 W + 42%)
4 i
for two loop operation or, i
S 1(0.66 W + 42% - 0.66aW)
J for single loop operation.
where:
a
.S = Rod block setting in percent of thermal 1
power (2346 MWt).
I
]
W=
Loop recirculation flow rate in percent i
of rated.
i i
j A W=
Difference between two loop and single loop effective drive flow at the same j
core flow.
l 1
In the event of operation with a maximum f
fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
}
S 1(0.66 W + 42%)(FRP/MFLPD) i i
for two loop operation or, I
S 1 (0.66 W + 42% - 0.666W)(FRP/MFLPD) l for single loop operation.
where:
i 1
fraction of rated thermal power i
=
(2436 MWt) 1 f
AmendmentNo.[.[,f,[,f[,[,98, d
10 1
1 JAFNPP 2.1 (cont'd)
MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
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Amendment No. 98, 10a
JAFNPP J
1 i
1.1 BASgS 1.1 FUEL CLADDINC INTEGRITY elevated clad temperature and the possibility of The fuel cladding integrity limit is set such that no clad failure.
However, the existence of critical
[
i calculated fuel damage would occur as a result of an 4
j abnormal operational transient.
Because fuel damage power, or boiling transition, is not a directly t
step-back approach is observable parameter in an operating reactor.
}
is not directly observable, Therefore, the margin to boiling transition. is a
used to establish a Safety Limit such that the mini-mum critical power ratio (MCPR) is no less than 1.07.
calculated from plant operating parameters such l
as core power, core flow, feedwater temperature, I
MCPR > 1.07 represents a conservative margin relative and core power distribution. The margin for each to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical fuel assembly is characterized by the critical I
barriers which separate radioactive materials from power ratio (CPR) which is the ratio of the j
the environs.
The integrity of this cladding barrier bundle power which would produce onset of transi-i i
is related to its relative freedom from perforations tion boiling divided by the actual bundle power.
The minimum value of this ratio for any bundle in j
or cracking.
Although some corrosion or use related cracking may occur during the life of the cladding, the core is the minimum critical power ratio fission product migration from this source is incro-(MCPR).
It is assumed that the plant operation is controlled to the nominal protective setpoints j
mentally cumulative and continuously measurable.
i Fuel cladding perforations, however, can result from via the instrumented variable, i.e.,
the oper-i thermal stresses which occur from reactor operation sting domain.
The current load line limit l
significantly above design conditions and the protec-analysis contains the current operating domain j
tion system safety settings.
While fission product map.
The safety 1.init (MCPR of 1.07) has migration from cladding perforation is just as sufficient conservatism to assure that in the l
measurable as that from use related cracking, the event of an abnormal operational transient l
thermally caused cladding perforations signal a
initiated from the MCPR operating conditions in threshold, beyond which still greater thermal specification 3.1.B. more than 99.91 of the fuel s
stresses may cause gross rather than incremental rods in the core are expected to avoid boiling h ',
cladding deterioration.
Therefore, the fuel cladding transition.
The MCPR fuel cladding safety limit l
is increased by 0.01 for single-loop operation as l
Safety Limit is defined with margin to the conditions discussed in Reference 2.
The margin between which would produce onset of transition bo111ag, (MCPR l
of 1.0).
These conditions represent a significant MCPR of 1.0 (onset of transition boiling) and the l
departure from the condition intended by design for safety Limit is derived from a detailed statisti-cal analysis considering all of the uncertainties planned operation.
in monitoring the core operating state including A.
Reactor Pressure > 785 pela and Core Flow > 101.
the uncertainty in the boiling transition corre-lation as described in Reference 1.
The uncer-l of Rated tanties employed in deriving the Safety Limit are i
I onset of transition boiling results in a decrease
]
in heat transfer from the clad and, therefore, f
Amendment No. [, JJI, [ k, [ [, 98, 12
~
JAFNPP i
1.1 BASES (Cont'd)
C.
Power Transient i
Plant safety analyses have shown that the scrans Safety Limit at 18 in, above the top of the fuel caused by exceeding any safety system setting will provides adequate margin.
This level will be assure that the Safety Limit of 1.1.A or 1.1.B continuously monitored whenever the recirculation will r.ot be exceeded.
Scram times are checked pumps are not operating.
4 periodically to assure the insertion times are adequate.
The thermal power transient resulting E.
References j
when a scram is accomplished other than by the expected scram signal (e.g.,
scram from neutron 1.
General Electric BWR Thermal Analysis Basis flux following closure of the main turbine stop (CETAB)
- Data, Correlation and Design valves) does not necessarily cause fuel damage.
Application, NEDO 10958 and NEDE 10958.
However, for this specification a Safety Limit t
l violation will be assumed when a scram is only 2.
FitzPatrick Nuclear Power Plant Single-Loop accomplished by means of a backup feature of the operation, NEDO 24281 August 1980.
l plant design.
The concept of not approaching a J
Safety Limit provided scram signals are opersble 3.
Generic Reload Fuel Appilcation, NEDE 24011 -
1 is supported by the extensive plant safety P-A and Appendices.
l analysis.
1 1
D.
Reactor Water Level (Hot or Cold shutdown Condition) 4 j
During periods when the reactor is shut down, 4
i consideration must also be given to water level requirements due to.the effect of decay heat.
If j
reactor water level should drop below the top of the active fuel during this time, the ability to j.
cool the core is reduced.
This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.
The 1
core will be cooled sufficiently to prevent elad l
molting should the water level be reduced to
]
two-thirds the core height. Establishment of the i
i Amendment No.
98, 14 l
1
JAFNPP i
1 BASgS 2.1 FUEL CLADDING INTEGRITY i
The abnormal operational transients applicable to The most limiting transients have been analyzed operation of the FitzPatrick Unit have been ana-to determine which result in the largest reduc-t j
lysed throughout the spectrum of planned operating tion in CRITICAL POWER RATIO.
The type of tran-conditions up to the thermal power condition 2535 sients evaluated were increase in pressure and MWt. The analyses were based upon plant operation power, positive reactivity insertion, and coolant in accordance with the operating map given in the temperature decrease.
The limiting transient current load line limit analysis.
In addition, yields the largest delta MCPR. When added to the j
2436 is the licensed maximum power level of Fits-Safety Limit, the required operating limit MCPR Patrick, and this represents the maximum steady-of Specift:stion 3.1.B is obtained.
4 i
state power which shall not knowingly be exceeded.
I The oraluation of a given transient begins with
'l The transient analyses performed for each reload the system initial parameters shown in the cur-are given in Reference 2.
Models and model rent reload analysis and Reference 2 that are conservatism are also described in this input to a core dynamic behavior transient com.-
reference.
As discussed in Reference 4 the core puter program described in References 1 - and 3.
I wide transient analysis for one recirculation The output of these programs along with the pump operation is conservatively bounded by initial MCPR form the input for the further i
two-loop operation analysis, and the flow-analyses of the thermally limited bundle with a j
dependent rod block and scram setpoint equations single channel transi,ent thermal hydraulic code.
i are adjusted for one-pump operation.
The principal result of. the evaluation is the reduction in MCPR caused by the transient, j
Fuel cladding integrity is assured by the oper-l sting limit MCPR's far steady state conditions given in Specification 3.1.B.
These operating
{
limit MCPR's are derived from the established fuel cladding integrity Safety Limit, and an analysis of abnormal operational transients.
For j
any abnormal operating transient analysis evalu-1 ation with the initial condition of the reactor being at the steady state operating limit, it is l
required that the resulting MCPR does not decrease l
below the Safety Limit MCPR at any time during the transient.
Amendment No. f(, f(, J 4 98, 15
_=
.~
j JAFNPP j
2.1 BASES (Cont'd)
C. References
- 1. Linford, R.B.,
" Analytical MeEhods of Plant Transient Evaluations for the General Electric Boiling Water Reactor". NEDO-10802 Feb., 1973
- 2. " General Electric Fuel Application".
NEDE 24011-P-A (Approved revision number appilcable I
at time that reload fuel analyses are per-formed).
i
- 3. " Qualification of the One-Dimensional Core Transient Mooel for Boiling Water Reactors",
NEDO-24154, October, 1978.
- 4. FitzPatrick Nuclear Power Plant Single-Loop Operation NEDO-24281, August, 1980.
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AmendmentNo.p(,[,98, 20 j
(Next page is 23)
JAFNPP 3
t 3.1 (CONTINUED)
MCPR Operating Limit for incremental C.
NCPR shall be determined daily during reactor Cycle Core Averste Exposure power operation at >,,_ 25% of rated thermal powsr and following any change in power level or dis-At RBM Hi-trip BOC to EOC-2GWD/t to EOC-1CWD/t tribution that would cause operation with a level settina EOC-2GWD/t EOC-1GWD/t to EOC limiting control rod pattern as described in the 3
bases for Specification 3.3.8.5.
S =.66W + 39%
1.24 1.29 1.31 D.
When it is determined that a channel has failed 5 =.66W + 40%
1.27 1.29 1.31 in the unsafe condition, the other RPS channels i
that monitor the same variable shall be function-
)
S =.66W + 41%
1.27 1.29 1.31 ally tested immediately before the trip system containing the failure is tripped.
The trip l
3 =.66W + 42%
1.29 1.29 1.31 system containing the unsafe failure may be i
placed in the untripped condition during the l
5 =.66W + 43%
1.30 1.30 1.31 period in which surveillance testing is being performed on the other RPS channels.
S =.66W + 44%
1.34 1.34 1.34 E.
Verification of the limits set forth in speci-
)
During single loop operation, the operating listit fication 3.1.5 shall be performed as follows:
1 NCPR shall be increased by 0.01 from that in the I
table above to - reflect the increase in safety limit 1.
The average scram time to notch position 38
)
MCPR.
(See Specification 1.1.A) shall be:
AVEI B i
f 2.
The average scram time to notch position 38 is determined as follows:
(
l n
n N1 i
Ni
=
g VE A
i 1-1 i=1 j) number of surveillance tests where:
a =
performed to date in the cycle. Ni = number j
]
of active rods measured in l
l i,
f Amendment No. g, ) [ J[, [ 98, 31
)
i
JAFNPP TABLE 3.1-1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No.
Modes in Which Total Number of Operable Trip Level FuncH a= **g s t be of Instrument Instrument Trip Function Setting Operable Channels Pro.
Action Channels vided by Design (1) per Trip Refuel Startup Run for Both Trip System (1)
(6)
Systems (16) 1 Mode Switch in I
I I
1 Mode Switch A
Shutdown (4 Sections) 1 Manual Scram I
X 2 Instrument A
Channels 3
IRM High Flux 1120/125of I
I 8 Instrument A
full scale Channels 3
IRM Inoperative I
I 8 Instrument A
Channels 2
APRM Neutron Flux-115% Power I
I 6 Instrument A
Startup(15)
Channels 2
APRM Flow Referenced S_< (0.66W+54%) (FRP/MFLPD)
I 6 Instrument A or b Neutron Flux (Not to Channels exceed 117%) (12)(13)
(14)(17) 2 APRM Fixed High 1120% Power 1
6 Instrument A or B Neutron Flux (14)
Channels 2
APRM Inoperative (10)
I 1
1 6 Instrument A or B Channels d
AmendmentNo.)#,J,[,
[, 98,
JAFNPP TABLE 3.1-1 (c*nt'd)
I
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REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT l
NOTES OF TABLE 3.1-1 (cont'd) l 14.
The APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 l
seconds. The APRM fixed high neutron flux singal does not incorporate the time constant, but responds directly to instantaneous neutron flux.
i 15.
This Average Power Range Monitor scram function is fixed point and is increased when the reactor mode switch is place in the Run position.
f 16.* During the proposed Hydrogen Addition Test, the normal background radiation level will increase by approximately a factor of 5 for peak hydrogen concentration. Therefore, prior to performance of the test, the Main Steam Line Radiation Monitor Trip Level Setpoint will be raised to.I three times the increased radiation levels. The test will be conducted at power levels > 80% of normal rated power. During controlled power reduction, the setpoint will be readjusted prior to going below 20% rated power without the setpoint change, control rod withdrawal will be prohibited until the necessary trip setpoint adjustment is made.
l 17.
This APRM Flow Referenced Scram setting is applicable to two loop operation. For one loop operation this setting becomes S $ (0.66W+54%-0.666W)(FRP/MFLPD) where AW = Difference between two-loop and single-loop effective drive flow at the same core flow.
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- This specification is in effect only during Operating Cycle 7.
i AmendmentNo.fEI.[.f,98, d
43a 4
JAFNPP TABLE 3.2-3 INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS Minimum no.
Total Number of of Operable Instrument Instrument Trip Level Setting Instrument Channels Action Channels Per Provided by Design for Both Channels Trio System l2 APRM Upscale (Flow Blased) si(0.66W+42%)(FRP/MFLPD)(11) 6 Inst. Channels (1) l 2
APRM Upscale (Start-up Mode) 112%
6 Inst. Channels (1) i 2
APRM Downscale 12.5indicatedon 6 Inst. Channels (1) seale g 1 (6)
Rod Block Monitor S 10.66W+K (8)(12) 2 Inst. Channels (1) l (Flow Biased) j 1 (6)
Rod Block Monitor 12.5indicatedon 2 Inst. Channelt (1)
(Downseale) seale j
3 IRN Downscale (2) 12%offullscale 8 Inst. Channels (1) i 3
IRN Detector not in (7) 8 Inst. Channels (1)
Start-up Position 3
IRM Upscale 186.4% cf full scale 8 Inst. Channels (1) l 2 (4)
SRM Det'ector not in (3) 4 Inst. Channels (1)
Start-up Position 2 (4) (5)
SRM Upscale 1 105 counts /sec 4 Inst. Channels (1) i 1
Scram Discharge Instrument 126.0 gallons per 2 Inst. Channels (9) (10) l Volume High Water Level instrument volume j
NOTES FOR TABLE 3.2-3 J
1.
For the Start-up and Run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function. The SRM and IRM block need not be operable in run mode, and 1
Amendsent No. )(. g,[, 98,
JAFNPP l
TABLE 3.2-3 (Cont'd)
I INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS NOTES FOR TABLE 3.2-3 1
- 11. This is the APRM Rod Block line for two loop operation. For single loop operation this line is S $,(0.66W+42%-0.66A W)(FRP/MFLPD).
AW = Difference between two-loop and single loop effective drive
~
flow at the same core flow.
i
- 12. This is the RBM Rod Block line for two loop operation. For single loop operation this line is S < (0.66W+K-0.66AW) where:
AU = Difference between two-loop and single-loop effective drive flow at the same core flow, i
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Amendment No. 28, 43, 93, 98, 2
74
O JAFNPp TABLE 3.7-3 (Cont'd)
INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS NOTES FOR TABLE 3.2-3
- 11. This is the APRM Rod Block line for two loop operation. For single loop operation this line is S < (0.66W+42bO.66 A W)(FRP/MFLPD).
AW = Difference between two-loop and single loop effective drive flow at the same core flow.
2 l
- 12. This is the RBM Rod Block line for two loop operation. For single loop operation this line is S < (0.66W+K-0.66 AW) where:
i AW=
Difference between two-loop and single-loop effective drive
.l flow at the same core flow.
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l Amendment No. 28, 43, 93, 98, 74
I f
.1AFNpP l
3.5 (cont'd) 4.5 (cont'd) l condition, that pump shall be considered inoper-2.
Following any period where the LPCI subsys-able for purposes satisfying Specifications tems or core spray subsystems have not been i
3.5.A 3.5.C. and 3.5.E.
required to be
- operable, the discharge piping of the inoperable system shall be
{
H.
Averate Planar Linear Heat Generation Rate vented from the high point prior to the 1
(APLHCR)_
return of the system to service.
1 l
The APLHCR for each type of fuel as a function of 3.
Whenever the NPCI, RCIC, or Core Spray System j
average planar exposure shall not exceed the is lined up to take suction from the condon-
]
limiting value shown in Figures 3.5-9 through sate storage tank, the discharge piping of l
3.5-11 for two loop operation.
For single loop the NPCI, RCIC, and Core Spray shall be operation these values are reduced by multiplying vented from the high point of the system, l
by 0.84 (see Specifiestion 3.5.K.
Reference 1).
and water flow observed on a monthly basis.
i If anytime during reactor power operation greater i
than 25% of rated power it is determined that the 4.
The level switches located on the Core Spray l
limiting value for APLHCR is being exceeded, and RNR System discharge piping high points action shall then be initiated within 15 minutes which monitor these lines to insure they are j
to restore operation to within the prescribed full shall be functionally tested each month.
3 limits.
If the APLHCR is not returned to within j
the prescribed limits within two (2) hours, an H.
Averste Planar Linear Heat Generation Rate l
orderly reactor power reduction shall be (APLHCR)
{
commenced issnediately.
The reactor power shall
~
be reduced to less than 25% or rated power within The APLHCR for each type of fuel as a function of the next four hours, or until the APLHCR is average planar exposure shall be determined daily j
returned to within the prescribed limits.
during reactor operation at } 25% rated thermal l
power.
1 l
4
.i I
i j
1 f
Amendment No. pdf, p4, J, g, 98,
/
123
t i
JAFNPP l
3.5 (cont'd) 4.5 (cont'd) l; i
J.
Thermal Hydraulic Stability J.
Thermal Hydraulic Stability 1.
Whenever the reactor is in the startup or 1.
Establish baseline APRM and LPRM neutrop flux
)
run
- modes, two Reactor Coolant System noise values within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of enterlag the 1
recirculation loops shall be in operation, region for which monitoring is required with:
unless baselining has been performed since the last refueling outage.
Detector levels A j
- a. Total core flow greater than or and C of one LPRM string per core octant plus 1
equal to 45 percent of cated, or detectors A and C of one LPRM string in the l
center of the core should be monitored.
- b. Thermal power less than or equal to the limit specified in Figure 3.5-1 (Line A),
i l
except as specified in Specifications 3.5.J.2 and 3.5.J.3.
)
2.
With two Reactor Coolant System recircula-tion loops in operation and total core flow 4
l 1ess than 45 percent of rated, and thermal power greater than the limit specified in Figure 3.5-1 (Line A); or with one Reactor Coolant System loop. operating and thermal l
power greater than the limit specified in Figure 3.5-1 (Line A):
~
l
I j
- 1. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after reaching steady-j state within the regions of Figure 3.5-1 where monitoring is required, and at least once per 8
hours thereafter; and j
i 4
AmendmentNo.[,98, j
124a 4
i
JAFMPP 3.5 (cont'd)
- 2. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after completing an increase in thermal power of 5 percent or more of rated thermal power.
- b. If the APRM and LPRM neutron fluz noise levels are greater than 5 percent and greater than three times their established baseline noise
- levels, initiate corrective action within 15 I
minutes to restore the noise levels to
~
within the required limits within 2
hours, by increasing core flow and/or reducing thermal power.
3.
If during single-loop operation, core thermal power is greater than the limit defined by line A of Figure 3.5-1, and core flow is less than 39 percent, issnediately initiate corrective action to restore core thermal power and/or core flow to within the
- limits, specified -in Figure 3.5-1, by increasing core flow and/or initiating an orderly reduction of core thermal power by inserting control rods.
4.
The requirements applicatie to single-loop operation in Specifications 1.1.A.
2.1.A.
1 3.1.A, 3.1.5, 3.2.C and 3.5.H shall be in effect within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the removal of one recirculation loop from service, or the reactor shall be placed in the hot shutdown condition.
j Amendment No. 98, 124b 1
JAFNPP 3.5 (cont'd) 5.
During resumption of two-loop operation following a period of single-loop operation, the discharge valve of the low-speed pump shall not be opened unless the speed of the faster pump is less than 50 percent of it's rated speed.
With no Reactor Coolant System Recirculation 6.
loop in service, the reactor shall be placed in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Amendment No. 98, 124c I
e w
JAFNpP 1
3.5 BASg3 (cont'd)
(
i requirements for the emergency diesel generators.
generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50
}
C.
Maintenance of Filled Discherme pipe Appendix K limit.
The limiting value for AFLHCR l is shown in Figure 3.5-9 through 3.5-11.
The l
If the discharge piping of the core spray LpCI, reduction factor for single loop operation for i
RCIC, and HpCI are not filled, a water hammer can the above curves is 0.84.
The derivation of this develop in this piping when the pump (s) are factor can be found in specification 3.5.K.
started.
To minimize damage to the discharge Reference 1.
piping and to ensure added margin in the opera-tion of these systems, this technical specifica-I.
Linear Heat Ceneration Rate (LHCE) tion requires the discharge lines to be filled whenever the system is required to be operable.
This specification assures that the linear heat i
If a discharge pipe is not filled, the pumps the generation rate in any rod is less than the supply that line must be assumed to be inoperable design linear heat generation.
l for technical specification purposes.
- However,
)
if a water hammer were to occur, the system would The LHCR shall be checked daily during reactor i
still perform its design function.
operation at 1 25% rated thermal power to deter-
[
mine if fuel burnsp. or control rod movement, has i
H.
Averate planar Linear Heat Ceneration Rate caused changes in power distribution.
For LHCR j
(APLHCR) to be a limiting value below 25% rated thermal l
power, the ratio of local LHCR to average LHCR j
This specification assures that the peak cladding would have to be greater than 10 which is pre-i temperature following the postulated design basis cluded by a considerable margin when employing loss-of-coolant accident will not exceed the limit any permissible control rod pattern.
4 specified in 10 CFR 50 Appendiz K.
l 4
f The peak cladding temperature following a
postulated loss-of-coolant accident is primarily i.
a function of the average heat generation rate of all the rods of a fuel assembly at any axial l
I location and is only dependent secondarily on the rod to rod power distribution within an assembly.
i since expected local variations in power distri-i bution within a fuel assembly affect the calcu-lated peak clad temperature by less than i 20*F j
4 l
relative to the peak temperature for a typical l
d fuel design, the limit on the average linear heat j
AmendmentNo.g,7g,g,98, j
130 I
i l
1
JAFNpP I
?
3.5 BASES (cont'd) l J.
Thermal Hydraulle Stab 111tr I,
Opagation la certain regions of the power vs.
flow curve have been identified as having a high potential for thermal hydraulic instability i
(Figure 3.5-1).
These regions are located in the l
high power / low flow area of the curve and can be l
l encountered during
- startup, shutdown, rod sequence exchange or recirculation pump trip.
i operation in these regions is associated with higher than normal neutron flux noise levels.
l Increased awareness of LPRM and APRM signal noise j
when operating in these regions will identify lastability and allow operator action to correct the problem.
The neutron flux noise level, thermal power and core flow limits are prescribed la accordance with the recommendations of General l
Electric Service Information Letter No.
- 380, l
Revision 1
"BWR Core Thermal Hydraulic
]
Stability", dated February 10, 1984.
1 j
Requiring the discharge valve of the lower speed j
loop to remain closed untti the speed of the i
faster pump is below 50 percent of its cated speed provides assurance when going from one to j
two pump operation that excessive vibration of i
the jet pump risers will not occur.
j K.
References 1.
"Fitzpatrick Nuclear Power plant Single-Loop 4
operation", NEDO-24281. August 1980.
l l'
AmendmentNo./,f,98, 131
Figuro 3.5-1 Thermal Power and Core Flow Limits of Specifications 3.5.J.1, 3.5.J.2 and 3.5.J.3 70 1
Stability Stability Monitor 4.ng Monitoring (APRM and LPRM) Required Stability Monitoring (AP m and For Single Loop (APRM and LPRM) Required lpm)
Operation 60 - During Two-Loop Operation Pequired O
~
During Line A 8
Single and I
'IWo-Icop P
50..
Operation y
Single-Loop Operation u
Prohibited
$e 1
i 40 g
d
'~
g 6
35 l
l 30 1
R
<a i
g 5
i Stability Mor!itoring Not Required rag 20 i
g us 1
1 I
.I 10 l
I I
1 0
ii ii iiiiii i
i 8
30 40 45 50 60 70 t
+
CORE FLOW (PERCENT RATED)
AmendmentNo.[,)d,JVI,f4 134 98, l
JAFNPP Figure 3.5-9 13-Reload 4 P8DRB284H y
gg 12-zu k 3 25
.S a:
E 11 -
ua N-o?
E E 10 -
$c E$
93 Eo 9-i 5e r
g g
g i
u a
5 5
5 10 15 20 25 30 35 40 Planar Average Exposure (GWD/t)
Maxilaum Average Planar Linear Heat Generation Rate (MAPLHGR)
Versus Planar Average Exposure I
For single-loop operation, these MAPLHGR References NEDO-21662-2 values are multiplied by 0.84.
(As amended August 1981 )
Amendment No. X, 98, J35g
JAFNPP Pigure 3.5-10 13" Reloads 4 & 5 P8DRB299 a
12-gg
- R W3 8~
-a
'E 11-wa Ek O*
- J ea
?Q 10-te ES 82 Ro
+C 9
N$
x i
e i
a s
s 8
5 10 15 20 25 30 35 40 i
Planar Average Exposure (GWD/t)
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
Versus Planar Average Exposure I
For single-loop operation, these MAPLHGR References NEDO-21662-2 values are multiplied by 0.84.
(As amended August 1981 )
I Amendment No. M g, 98, 135h
JAFNPP Figure 3.5-11 13-Reload 6 BP8DRB299 N
12-ed uk 25 S a:
E 11-N'$
E-aB EM 10 "
k$v Sa 28 9~
58 x
y e
i e
a i
a 5
10 15 20 25 30 35 40 Planar Average Exposure (GWD/t)
Maxi = = Average Planar Linear Heat Generation Rate (MAPLHGR)
Versus Planar Average Exposure
Reference:
For single-loop operation, these MAPLHGR values are multiplied by 0.84.
(As amended December 1984)
Amendment No. Jef 98, 1351
JAFNPP 4.6 (cont'd) 1.
The two recirculation loops have a flow imbalance of 15 percent or more when the pumps are operated at the same speed.
2.
The indicated value of core flow rate varies from the value derived from loop flow measurements by more than 10 percent.
3.
The diffuser to lower plenum differential pressure reading on an individual jet pump varies from the average of all jet pump differential pressures by more than 10 percent.
d A.
Whenever the reactor is in the startup/ hot standby or run modes, and there is one loop recirculation flow, jet pump operability shall be i
verified as follows.
a.
Baseline readings will be taken and operating characteristics for the following parameters established:
1' 1.
Jet Pump Loop Flow and Recirculation Pump Speed for the operating loop.
2, Individual Jet Pump percent differential pressures for all jet pumps.
l b.
Initially, and daily thereafter, jet pump operability will be verified by aduring that the following do not occur sinuitaneously:
AmendmentNo.)[,98, 145 i
1 i
JAFNPP 4.6 (cont'd)
- 1. The ratio of jet pump loop flow to recirculation pump speed for the operating loop does not vary from the initially established value by more than 10 percent.
- 2. The ratio of individual jet pump percent differential pressure to the loop's average jet pump percent differential pressure does not vary from the initially i
established value by more than 20 percent.
M l
I I
I l
1 l
i Amendment No. %, /, 98, I
145a j
i i
i
(
JAFNPP l
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT l
3.6 (cont'd) 4.6 (cont'd) 3.6.I Shock Suppressors (Snubbers) 4.6.I shock Suppressors (Snubbers)
Applicability Appilcability Applies to the operational status of the Applies to the periodic testing requirement 3
shock suppressors (snubbers).
for the shock suppressors (snubbers).
Objective Objective i
l To assure the capability of the snubbers to:
To assure the capability of the snubbers to perform their intended functions.
prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, and I
Allow normal thermal motion during startup i
and shutdown.
Specification Specification Each snubber shall be demonstrated operable by performance of the following augmented
- 1. During all modes of operation except Cold inservice inspection program.
Shutdown and Refueling, all snubbers which l
are required to protect the primary
- 1. Snubbers shall be visually inspected in i
coolant system or any other. safety related accordance with the following schedule:
l system or component shall be operable.
During Cold Shutdown or Refueling mode of No. Inoperable Snubbers Subsequent Visual operation, only those snubbers shall be per Inspection period Inspection period I
operable which are on systems that are required to be operable in these modes.
0 18 months i 25%
1 12 months 25%
j
- 2. With one or more snubbers inoperable.
2 6 months i 25%
l within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during normal operation, 3,4 124 days i 25%
j or within 7 days during Cold Shutdown or 5,6,7 62 days i 25%
l Refueling mode of operation for systems 8 or more 31 days i 25%
- The inspection interval may not be extended 4
more than one step at the time, j
Amendment No. M 98, j
145b s
b b
~
JAFNpp r
3.6 (cont'd) 4.6 (cont'd) which are required to be operable in these
- The snubbers may be categorized into two i
modes, complete one of the following:
groups:
Those accessible and those t
inaccessible during reactor operation.
gach
- a. replace or restore the inoperable group may be inspected independently in snubber (s) to operable status or, accordance with the above schedule.
b declare the supported system inoperable
- 2. Visual inspection shall verify (1) that and follow the appropriate limiting there are no visible indications of damage a
j condition for operation statement for or impaired OPgEABILITY, (2) attachments to l
that system or, the foundation or supporting structure are l
secure, and (3) in those locations where l
- c. perform an engineering evaluation to snubber movements can be manually induced j
demonstrate the inoperable snubber is without disconnecting the snubber, that the 1
j uncocessary to assure operability of.the snubber has freedom of movement and is not i
system or to meet the design criteria of frozen up.
Snubbers which appear l
the system, and remove the snubber from inoperable as a
result of visual l
the system.
inspections may be determined OPgRABLg for the purpose of establishing the next visual
- 3. With one or more snubbers found inoperable, inspection interval, providing that (1) the within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> perform a visual inspection cause of the rejection is clearly i
of the supported component (s) associated established and remedied for that with the inoperable snubber (s) and document particular snubber and for other snubbers the results.
For all modes of operation that may be generically susceptible; and i
except Cold Shutdown and Refuellag, within (2) the affected snubber is functionally j
14 days complete an engineering evaluation tested in the as found condition and
}
as per Specification 4.6.I.6 to ensure that determined OPgEABLg per Specifications j
the inoperable snubber (s) has not adversely 4.6.I.7 or 4.6.I.8 as applicable.
i affected the supported component (s).
For Hydraulic snubbers which have lost Cold Shutdown or Refueling
- mode, this sufficient fluid to potentially cause evaluation shall be completed within 30 uncovering of the fluid reservoir-to-t i
days.
snubber valve assembly port or bottoming of j
the fluid reservoir piston with the snubber i
1 Amendment No. )(, g, 98, i
t 145c 4
I
JAFNPP 3.6 (cont'd) 4.6 (cont'd) in the fully extended position shall be functionally tested to determine operability.
- 3. Once each operating cycle, 10% of each type of snubbers shall be functionally tested for operability, either in place or in a bench test.
For each unit and subsequent unit that does not meet the requirements of 4.6.I.7 or 4.6.I.8, an additional 10% of that type cf snubber shall be functionally tested untti no more failures are found, or all units have been tested.
- 4. The representative sample selected for functionally testing shall include the various configurations, operating environments and the range of size and capacity of snubbers.
At least 25% of the snubbers in the representative sample rhell include snubbers, from the following three
]
categories:
- a. The first snubber away from reactor vessel nozzle.
- b. Snubbers within 5
feet of heavy equipment (valve, pump, turbine, motor, t
etc.).
- c. Snubbers within 10 feet of the discharge 1
from a safety relief valve.
l 1
4 0
Amendment No. g, 98, t
145d
JAFWpp 4.6 (cont'd) 3.6 (cont'd)
In addition to the regular semple, snubbers which failed the previous functional test shall be rotested during the next test period.
If a
spare snubber has been installed in place of a failed snubber, then both the failed snubber (if it is repaired and installed in anothee position) and the spare snubber shall be retested.
Test results of these snubbers may not be included for the re-sampling.
- 5. If any snubber selected for functional J
testing either f ails to lockup or f ails to
- move, i.e.
is frozen in place, the cause will be evaluated and if due to manufacturer or design deficiency, snubbers of the same design subject to the same defect shall be functionally tested.
This testing requirement shall be independent of the requirements stated above for snubbers not meeting the functional tes'. acceptance criteria.
- 6. For the snubber (s) found inoperable, an engineering evaluation shall be performed on the components which are supported by the snubber (s).
The purpose of this engineering evaluation shall be to determine if the components supported by the snubber (s) were adversely affected by the inoperability of the snubber (s) in order to ensure that the supported components remain capable of meeting the designed service requirements.
AmendmentNo.[,98, I
145e l l
.l
JAFNpP 3.6 (cont'd) 4.6 (cont'd)
- 7. The hydraulic snubber functional test shall verify that; i
- a. Activation (restraining action) is i
achieved within the specified range of I
velocity or acceleration in both tension and compression.
- b. Snubber bleed, or release rate, where required, is within the specified range in compression or tension.
For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.
- 8. The mechanical soubbee functional test shall verify that:
I
- a. The force that initiates free movement i
of the snubber rod in either tension or compression is less than the specified j
maximum drag force.
Drag force shall i
not have increased more than 50% since the last functional test.
- b. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension l
and compression.
s 1
i I
1 Amendment No. pf 98, g
i 145f l l
i 1
4 JAFNpP 3.6 (cont'd) 4.6 (cont'd) 1 1
- c. Snubber release rate, where required, is i
within the specified range in compression or tension.
For snubbers i
specifically required not to displace 1
under continuous load, the ability of 1
the snubber to withstand load without
)
displacement shall be verified.
- 9. Snubber Service Life Monitoring 1
A record of the service. life of each
- snubber, whose failure could adversely j
affect the primary coolant or other safety-related system, the date at which the designated service life commences, and j
the insta11st'.on and maintenance records on
{
which the de,signated service life is based i
shall be maintained as required by j
specification 6.10.B.13.
j At least once per operating cycle, the f
installation and maintenance records for each snubber, whose failure could adversely affect the primary coolant or other safety
}
related system, shall be reviewed to verify 1
that the indicated service life has not i
been exceeded or will not be exceeded prior to the next sheeduled snubber service life l
review.
If the indicated service life will be exceeded prior to the next scheduled 1
snubber service life review, the snubber
,i service life shall be reevaluated or the anubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next schedule service life review.
This reevaluation, replacement or reconditioning shall be indicated in the records.
l Amendment No. M 98, s
j 145gl 1
i
JAFWpp 3.6 and 4.6 BASES (cont'd) would provide a leakage path past the core thus reducing the core flow rate.
The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 j
percent to 6 percent) in the total coes flow measured.
This decrease, together with the loop flow increase, would result in a lack of corre-lation between measured and derived core flow rate.
Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
i A nozzle-riser system failure c;Old also generate the coincident failure of a jet pump body; however, the converse is not true.
The lack of J
any substantial stress in the jet pump body makes initial nozzle-i failure impossible without an l
riser system failure.
I i
Surveillance tests are performed to verify jet pump operability.
Significant changes in either:
l (1) the relationship between loop flow and recirculation pump speed, or (2) individual jet l
pump differential pressure compared to average jet pump differential pressure, are used to l
detect degraded jet pump performance.
l r.
i i
1
)
I 1
AmendmentNo.[,98, i
155 l:
4 JAFNpP 3.6 and 4.6 BASES (cont'd) 4 l
l H.
(DELETED) i followed.
As an alternative to snubber repair or l
replacement an engineering evaluation may be 1
I.
Shock Suppressors performed:
to demonstrate that the inoperable j
snubber is unnecessary to assure operability of i
Snubbers are designed to prevent unrestrained pipe the system or to meet the design criteria of the notion under dynamic loads as might occur during system; and, to remove the snubber from the an earthquake or severe transient, while allowing system.
With one or more snubbers found I
normal thermal motion during startup and shutdown.
inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> a visual inspection l
The consequence of an inoperable snubber is an shall be performed on the supported component (s) i increase in the probability of structural damage associated with the inoperable snubber (s) and the i
l to piping as a result of a seismic or other event results shall be documented.
For all modes of l
l initiating dynamic loads.
It is therefore operation except Cold Shutdown and Refuellag, l
required that all snubbers required to protect within 14 days an engineering evaluation shall be i
the primary coolant system or any other safety performed to ensure tht the inoperable snubber (s) i system or component be operable during reactor has not adversely affected the supported l
operation.
Snubbers excluded from this component (s).
For Cold Shutdown or refueling j
inspection program are those installed on mode, this evaluation shall be completed within 3
non-safety related system and then only if their 30 days.
A period of 7 days has been selected j
failure or failure of the system on which they for repair or replaconent of the inoperable j
are installed would have no adverse effect on any snubber during cold shutdown or refueling mode of I
safety-related system.
Because the snubber operation becuase in these modes the relative I
protection is required only during low probability of structural damage to the piping l
}
probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (for systems would be lower due to lower values of l
normal operation) or 7 days (for cold shutdown or total stresses on the piping systems.
In case a refueling mode of operation) is allowed for shutdown is required, the allowance of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> repairs or replacement of the snubber prior to to reach a cold shutdown condition will permit an taking any other action.
Following the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> orderly shutdown consistent with standard j
(or 7 day) period, the supported system must be operating procedures.
j declared inoperable and the Limiting condition of l
Operation statement for the supported system i
l 2
4 Amendment No. [. [ 98, l
156 1
.