ML19323E477
| ML19323E477 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/12/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323E478 | List: |
| References | |
| NUDOCS 8005230588 | |
| Download: ML19323E477 (92) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION g
s, g wasmwaToN, D. C. 20668
%p*.GTj BOSTON EDISON CJMPANY DOCKET NO. 50-P_93_
PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 License No. DPR-35
- 1. 'The Nuclear Regulatory Comission (the Comission) has found that:
A.
The submittals by Boston Edison Company (the licensee) dated May 1,1975; September 1. November 12, November 15,1976; July 20, August 8 August 24, 1977; February 1, March 22,1978; September 27, December 12. December 31, 1979; February 5, March 28, April 3 April 7, April 17. April 24, and April 29, 1980, comply 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 confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendnent 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.
2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachnent to this license amendment, and paragraph 3.B of Facility Operating License No. OPR-35 is hereby amended to read as follows:
l 8 0 05 23 0Nb.
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. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 42, are hereby incorpor-ated 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 COMMISSION Thomas
. Ippolito, Branch Chief Operating Reactors Branch #2 Division of Licensing Attachine~nt: ~ ~ ~ ~
Changes to the Technical Specifications Date of Issuance:
l s
a ATTACHMENT TO LICENSE AMEN (NENT NO. 42 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 i
Revise Appendix A As Follows:
Remove Insert 6
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9 12 12
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13a 13b 16 16 17 17 18 18 26 26 27 27 1
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40 40 43 43 44a 45 45 48 48 50 50 50a 51 51 53a 54 54 55 55 59a 61 61 66a 67 67 71 71 73 73 77 77 91 91 103 103 104 104 106 106 107 107 108 108
o 6 Remove Insert 109 109 124 124 125 125 126 126 127 127 137e 137e 141 1 41 4
l 142 142 146 158 158 158a 158b 158c 1 61 161 165 165 172 172 173 173 1 74 174 174a 174b 177 177 178 178 180 180 1 81 181 1 94 194 194a 195 1 95 197 197 199 199 200 200 205a 205a 205a-1 205a-1 205b 205b 205c 205c 205c-1 205c-2 205c-2 205c-3 205c-3 205c-4 205c-4 205c-5 205c-5 205c-6 205c-6 205d 205d 205e 205e-1 205e-2 205e-3 205e-4 205e-5 205f 206m 206m-
2.'1 LIMITING SAFETT SYSTDf SETTING 1.1 S*AFETT LIMIT r
2.1 FUEL CLADDING INTEGRITY 1.1 PUEL CLinDING INTEGRITY _
_Applicabiliti:
Applicability:
l Applies to trip settings of the Applies to the interrelated instruments and devices which are e
variables associated with fuel provided to prevent the reactor thermal behavior.
system safety limits from being exceeded.
Obiective:
Otiective_:
}
To define the level of the process To establish limits below which variables at which automatic pro-the integrity of the fuel tactive action is initiated to 4
.{
cladding is preserved.
prevent the fuel cladding integrity 1
l safety limits from being exceeded.
_ Specification _:
Specification:
A.
Neutron Flux Scram l
Reactor Pressure > 800 psia and A.
)
Core Flow >10% of Rated _
The limiting safety system trip i
The existence of a m4a4==
settings shall be as specified critical power ratio (MCPR) less below:
than 1.07 mhall constitute vio-Nuetron Flux Trip Settings lation of'the fual cladding 1.
+
A MCPR integrity safety limit.
of 1.07 is hereinafter referred to as the Safsty Limit MCPR.
Setting (Run Mode)_
t Core Thermal Power Limit (Reacto_r i
3.
Pressure 1800 psia and/or Core When the Mode Switch is in the EUN position, the Flow $1021 APRM flux scram trip Nhan the reactor pressure is $800 setting shall be:
psia or core flow is less than or l
equal to 10% of rated, the steady SS. 65W + 55% 2 loop state core thermal power shall not exceed 25% of design thermal power i
Where:
I C.
Power Transient S = Setting in percent of rated thermal The safety limit shall be assumed power (1998 MWt) to be exceeded when scram is known 4
l to have been accomplished by a W = Percent of drive means other than the expected flow to produce scram signal unless analyses a rated core flow
'i demonstrate that the fuel of 69 M lb/hr.
cladding integrity safety defined in Specifi-limits cations 1.1A and 1.1B were not l
exceeded during the actual transient.
/
Amendment No. 42 6
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I 2.1 LIMITING SAFETY SYSTDi SETTING 1.1 SAFETT LDf1T In the event of operation with 6 Whenever the reactor is in the maximian fraction of limiting power D.
cold shutdown condition with density (NFLPD) greater than the irradiated fuel in the reactor fraction of rated power (F'2),
vessel, the water level shall not the setting shall be modified as be less than 12 in. above the top follows:
of the normal active fuel sone.
FDP
~
~
S S (0.65W + 55% )
MrLPo 2 Ioop
fraction of limiting power density where the limiting power density is..
13.4 KW/ft for 8x8 and P8x8R fuel.
l 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.
For no combination of loop recircula-tion flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.
b.
APRM Flux Scram Trip Setting (Refuel or Start and Hot Standby Mode)
When the resctor mode switch is in the RETUEL or STARTUP position, the APRM scraa shall be set at less than or equal to 15% of rated power.
c.
IRM The IRM flux scram setting shall be s120/125 of scale.
B.
APRM Rod Block Trip setting The APRK rod block trip setting shall be:
3RB i 0.65W + 42 2 Loop.
7 Amendment No.42 s
me 6
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11 SAFETT LIMIT 2.1 LIMITING SAFETT STSTM SETTING
- Where, 8R5 = Rod block setting in percent of rated thermal power (1998 MWe)
W = Percent of drive flow required to produce a rated core flow of 69M lb/hr.
In the event of operating with a maximas fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
81B $ (O.65 W + 42%),, )(FLPD,
[
e
- Where, FRP = fraction of rated thermalpower MFLPD= maximum fraction of limiting power density where the limiting power density is 13.410i/f t for 8x8 and pax 8R fuel.
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.
C.
Reactor low water level scram setting shall be) 9 in. on level instruments.
D.
Turbine stop valve closure scram setting shall be f.10 percent vglve closure.
E.
Turbina control valve fast closure setting shall be }_ 150 psig con-trol oil pressure at acceleration relay.
F.
Condenser low vacuum scram setting shall be } 23 in. Hg. vacuum.
G.
Main steam isolation scram setting shall be 110 percent valve clo-sure.
Amendment No.
42 8
2.1 LIMITDIG SAFETY SYSTEM SETIDOG 1.1 SAFETY LDfIT B.
Main steam isolation on nain steam lina low pressure at inlet to turbine valves. Pressure setting shall be 2 880 peig.
I.
Reactor low-low water level initiation of CSCS systems set-ting ah=11 be at or above -49 in.
indicated level.
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Amendment No. 42
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_e-4 s-
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s-s---
i100'
' M
-s-
~4
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' 55 APEN Flow Biased Scram
. -- ; ;;=
s - _. _.. _ s.
. 90 l
- s..-
-.D.i.o.r.n.a.l) *1.2 f
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i 80:
'_-W
_8
..AFEM, Bod _31,ock_.
--7 p-
- @e _.
- Olormal).*1- - - - -
TO.
a
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w
=
- =-
=
s_
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_60::~ 4 82
=
i k
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=
'l for MFLPD greater than FRP,.the interce;ts I'5n :
w w
are varied by the ratio FRP MFLPD u
p _.
D : gOi 2-See Specifications 2.1.A and 2.1.3
.a i
- Ses is the, refuel or startup/ hot sta: thy -
m
=
'2 O :
acdes, the A7?J4 scra= shall be set at 4 m
3C.
15% of desis: ;over D
ret m
20.--
l E 101 z._
0-r 4
20, ho jo 30, 100
- 2: 5223 0 '
.. =Racir.culat. ion Flow (1._f Design) _.
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Figure 2.1.1 AFEM Scram and Rod 31Ec'k' Trip Limiting Safety System Settings
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= _.=g_ -
Amendment No. 42 9
The required input to the statistical model are the uncertainties listed on Table 5-1, Reference 3, the nominal values of the core parameters listed in Table 5.-2, Reference 3, and the relative assembly power distribution shown in Figures 5-1 and 5-1A of Reference 3.
Tables 5-2A and 5-25, Referenca 3, show the R-f actc..
tributions that are input to the statistical model which is used to establish the safsty limit MCPR. The R-factor distrihutions shown are taken near the beginning of the fuel cycle.
The basis for( e uncertainties in the core parameters are given in NEDO 20340 and the basis for the uncertainty in the GEXL correlation is given in NEDO-10958(1). The power distribution is based on a typical 764 assembly core in which the rod pattern vas arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highese power levels. The worst distribution in Pilgrim Nuclear Power Station Unit i during any fuel cycle would not be as severe as the distribution used in the analysis.
B.
Core Thermal Power Limit (Reactor Pressure < 800 pois or Core Flow
< 10% of Rated)
The use of the GEIL correlation is not valid for the critical power calculations at pressures below 800 psig or core flows less than 10% of rated. Therefore, the fuel cladding integrity safety limit is established by other means. This is done by establishing a limiting condition of core thermal power operation with the following basis.
Since the pressure drop in the bypass region is essentially all elevation hes 1 which is 4.56 psi the core pressure drop at low power and all flows will always be greajer than 4.56 pai.
Analyses show that with a flow of 28x10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and driving head will be greater than 28x10ge flow with a 4.56 psi has a value of 3.5 psi.
Thus, the bund lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.
With the design peaking factors,the 3.35 MWt bundle power cor-responds to a core thermal power of more than 50%. Therefore a core thermal power limit of 25% for reactor pressures below 800 psia, or core flow less than 10% is conservative.
x Amendment No. 42 12
- mn
l l
C.
Powe Transions:
Plant safety analyses have shown that the scrams caused by ex-coeding any safety setting will assure that the Safety Limit of Specification 1.1A or 1.1B will not be exceeded. Scram times are checksd periodically to assure the insertion times are
~
adequate. Se thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g.,
scram from neutron flux fo nowing closures of the main turbina stop valves) does not necessarily cause fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is caly accomplished by usans of a backup feature of the plant design. Se concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.
The computer provided with Pilgrim Unit 1 has a sequence annunciation program which win indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur. his program also indicates when the scram setpoint is cleared. This win provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.
D.
Reac
- Water Level (Shutdown Condition)
During periods when the reactor is shutdown, consideration est also be given to water level requirements due to the effect of decay beat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.
The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establish = ant of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will be continuously monitored.
References General Electric Thermal Analysis Basis (GEIAB): Deta, 1.
Correlation and Design Application, General Electric Co.
BWR Systems Department, November 1973 (NEDO.-10958),
2.
Process Computer Performance Evaluation Accuracy, General Electric Company BWR Systems Department, June,1974 (NEDO-20340),
3.
General Electric Boiling Water Rasetor Generic Raiosd Puel Application, NEDE-24011-P.
Amendment No. 42 13
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(DELETED) 1 1
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1 Amendment No. 42 13a
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Amendment No. 42 13b G
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2.1 BASES
The scram trip setting must be adjusted to ensure that the IJiGR transient peak is not increased for any combination of maximum fracts n of limiting power density QtFLPD)and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1 when the NFLPD is greater than the fraction of rated power (FRP).
Analyses of the limiting transients show that no scram adjustment is required to assure MCPR greater than the Safety Limit MCPR when the transient is initiated from MCPR above the operating l
limit MCPR.
For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate. thermal margin between the setpoint and the safety limit, 25 percent of rated. Tha margin is adequate to accommodata anticipated maneuvers associated with power plant startup. Effects of increasing pressure at sero or low void content are minor, cold water from sourt.as available during startup is not unsch colder than that already in the system, temperature coefficients are small, and control rod patterns I
are constrained to be uniform by operating procedures backed up by the rod worth mini =f zer.
Worth of individual rods is very low in a uniform rod pattern.
Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable case of significant power rise. Because the flux distribution associated with uniform rod withdras 4 s does not involve high local peaks, and because several rods mu.. be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally rhe heat flux is in the near equilibrium with the fission rate. In an assumed uniform rod withdravel approach to the scram level, the rate of power rise is no more than five percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before power could exceed the safety limit. The 15% APRM scram r==mina active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 880 psig.
The analysis to support operation at various power and flow re-lationships has considered operation with either one or two re-circulation pumps.
IRM The IBM system consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level l
Amendment No. 42 16 i
l I
+ _ - _, _ _
.... - ~.
5
2.1 3ASES
between that covered by the SIM and the APEM. The 5 decades are covered by the IBM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decada in size. S e IRN scram setting of 120/125 of full scale is active in each range of the IRM. For example, if the instrument were on range 1, the scram setting would be a 120/125 of full scale for that range; likewise, if the instrument were on range 5 the scram,would be 120/125 of full scale on that range. Thus, as the IRM is ranged up to accousodate the incressa in power level, the scram setting is also ranged up.
The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For in-sequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rode that heat flux is in equilibrium with the neutron flux, and an IBM scram would result in a reactor shutdown well before s,y safety limit is exceeded.
9 In order to ensure that the IBM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. B is analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale. H is condition exists at quartar rod density. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak core power limited to one percent of rated power, thus mainemining ICPR above the Safety Limit MCPR. Based on the above analysis, the IBM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and proirides backup protection for the APRM.
B.
APRM Control bd Block Reactor power inval any be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a MCPR less the Safety Limit MCPR. This rod block set point, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal.
The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst came MCPR which could occur during steady-state operation is ar 107%
l of rated thermal power because of the AIRM rod block trip l
Amendment No. 42 17
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_ _ q.--
4 J
2.1 RASES
setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPIN system. As with the APRM scram trip setting, the AFRM rod block trip setting is adjusted downward if the maximum fraction of limiting power density exceeds the fraction of rated
)
power, thus preserving the APRM rod block safety margin.
C.
Reactor Water Low Level Scram Trip Settina (LL1)
Se set point for low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. He results show that scram at this level adequately protects the fuel and the pressure barrier, becauss MCPR l
remains well above the safety limit MCPR in all cases, and system presaura does not reach the safety valve settings. The scram setting is approximately 25 in. below the normal operating range and is thus edequate to avoid spurious scrams.
D.
Turbine Stop Valve Closure Scram Trip Setting The turbine stop valve closure scram anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of $ 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR r-4== above the safety limit MCPR l
even during the worst case transient that assumas the turbine bypass is closed.
E.
Turbine Control Valve Fast Closure Scram Trip Settina The turbine control valve fast closure screa antic 1 pates the pressure, neutron flux, and heat flux increase that c.. ]d result from fast closure of the turbine control valves due to load rejection exceeding the capehility of the bypass valves. The reactor protection system initiates a scram when fast clos,re of the control valves is initiated by the accelerstion relay. This setting and the fact that control valve closure time is approximately twice as long as that for the stop valves means that resulting transients, while similar, are less severe than for stop valve closure. MCPR remains above the safety limit MCPR.
l P.
Main Condenser Tew vacuum Scram Trip Setting To protect the main condenser against overpressure, a loss of condenser vacuum initiates automatic closure of the turbine stop valves and turbine
. bypass valves. To anticipate the transient and automatic scram resulting from the clorure of the turbina stop valves, low condenser vacuum initiates a scram. The low vacuum scram set point is selected to initiate a scram befers the closure of the turbine stop valves is initiated.
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i Amendment No. 42 18
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4.1 SURvZILLANCE REQUIRDElffS_
3.1 LIMITING CONDITION 70E OPERATION REACTOE PROTECTION SYSTEM 3.1 1LEAC10R PRMECTICEI SYSTD(
Applicability:
Applicability:
Applies to the surveillance of Applies to the instrumentatior the instruman':stion and associ-and sesociated devices which stod devices which initiata re-initiate a reactor seras.
actor scram.
Obiective
_Obiective:
To specify the type and frequency To assure the operability of the of surveillsace to be applied to reactor protection system.
the protection instrumentation.
Specifications l
Specification:
Instrumentation systems shall A.
The setpoints, mini = = nuaber of be functionally tested and trip systems, and mininum number calibrated as indicated in of instrument channels that must Tables 4.1.1 and 4.1.2 re-be operable for each position of spectively.
the reactor mode switch shall be as given in Table 3.1.1.
The Daily during reactor power B.
system response times from the operation, the anxisman frac-openinr, of the sensor contact up tion of limiting power density to and including the opening of shall be checked and the scram the tri's actuator contacts shall and APRK Rod Block settings not exceed 50 milli-seconds.
5 ven by equations in 1
1 Specification 2.1. A.1 and 2.1.B shall be calculated if =v4 -
fraction of limiting power
[
density escoeds the fraction of rated power.
l 26 Amendment No. 42
REACTOR PR&rECTION SYSTEM (SCRAM).
TRUNENTATIM ""r=IREMENT Elam. Number Modes in Which Function Operable Inst.
Must Be Operable rh-la per Trip Trip Function Trip Level Setting Refuel (7)
Startup/ Hot Run Actica (1)
(1) System Standby 1
Mode Switch in Shutdous I
I I
A 1
I I
A IRN 3
High Flux 5120/125 of full scala 1
I' (5)
A 3
Inoperative Z
X (5)
A APRM I
2 High Flux (14) (15)
(17)
(17)
I A or B 2
Inoperartve X
I(9)
X A or B 2
Downscale 2 2.5 Indicated on Scale (11)
(11) 1(12)
A or B 2
High Flux (151) 515X of Design Power X
X (16)
A or B 2
High Reactor Pressure
$1085 psig I(10)
X X
A l
2 High Drywell Pressure
$ 2.5 psig I(8)
X(8)
X A
2 Reactor Low Wate,c Level 29 In. Indicated Level X
X X
A 2
High Water J4 vel in Scram Discharge '.ank 139 Callons X(2)
X I
A i
D 2
Turbine Condenser Low h
Vacuun 223 In. Hg Vacuum X(3)
X(3)
I A or C A
2 Main Steam Line High 17X Normal Full Power z
,o Radiation
Background
I X
X A or C b
4 Main Steam Line Isolation Valve Closure 110Z Valve Closure X(3) (6)
X(3) (6)
I(6)
A or C 2
Turb. Cont. Valve Fast 1150 psig control 011 Closure Pressure at Acceleration Relay X(4)
X(4)
X(4)
A or D 4
Turbine Stop Valve closure
$10Z Valve Closure X(4)
X(4)
X(4)
A or D
Two recire. pump operation MFLPD
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l meres FCE TABLE 3.1.1 (Cont'd)
Not required to be operable when the reactor pressure vessel head is not 10.
bolted to the vessel.
The APIBt downscale trip functica is only active when the reactor mode switch 11.
is in run.
The A?RN downscale trip is automatically bypassed when the IBM instrumentation 12.
is operable and not high.
An AFISt will be considered inoperable if there are less than 2 LFRN inputs per level or there is less than 50% of the normal complement of LFEN's to an 13.
AFIN.
W is percent of drive flow required to produce a rated core tiow of 69'H1b/hr.
14.
Trip level setting in percent of design power (1998 left).
15.
See Sectiot.2.1.A.1.
The APEN (151) high flux scram is bypassed when in the run mode.
16.
The AFRN flow biased high flux scran is bypassed when in the refuel or 17.
startup/ bot standby modes.
l l
Amendment No. 42
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29
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41 BASES (Cont'd) l 3
The Maximum Fraction of Limiting Power Density (NFLPD) shall be checked once per day to determine if the AFEN scram requires adjustment. This will normally be done by checking the LPEM readings. Only a small asseber of control rods are moved daily and thus the NFLPD is not expected to change significantly and thus a daily check of the MFLPD is adequate.
The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. This is esapensated for in the APRM system by calibrating every three days using heat balance data and by calibrating individual LPEM's every 1000 effective full power hours using TIP traverse data.
i Asendment No. 42 g
l LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT C.
Control Rod Block Actuation C.
Control Rod Block Actuation l
1.
The limiting conditions of op-Instrumentation shall be func-eration for the instrumenta-tionally tested, calibrated tion that initiates control and checked as indicated in rod block are given in Table Table 4.2.C.
3.2.C.
System logic shall be func-tionally tested as indicated in Table 4.2.C.
2.
The minimum number of operable instrument channels specified
)
in Table 3.2.C for the Rod Block Monitor may be reduced i
by one in one of the trip sys-tema for maintenance and/or testing, provided that this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period.
D.
Radiation Monitoring Systems -
D.
Radiation Monitoring Systems -
Isolation & Initiation Func-Isolation & Initiation Functions i
tions 1.
Steam Air Ejector Off-Gas 1.
Steam Air Ejector Off-Gas System System (a) Except as specified in (b) be-Instrumentation shall be func-low, both steam air ejector tionally tested, calibraced off-gas system radiation moni-and checked as indicated in cors shall be operab's during Table 4.2.D.
reactor power operation.' The trip settings for the monitors System logic shall be func-shall be set at a value not to tionally tested as indicated exceed the equivalent of the in Table 4.2.D.
stack release limit specified in Specification 3.8.B.I.
The time delay setting for closure of the steam air ejector iso-lation valves shall not exceed 15 minutes.
(b) From and after the date that one of the two steam air ejec-tor radiation monitors is made or found to be inoperable, Amendment No. 42 43
EXHIBIT A LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS G.
Recirculation Pump Trip / Alter-G.
Recirculation Pump Trip / Alternate nate Rod Insertion Initiation Rod Insertion The recirculation pump trip Surveillance for instrumentation systes causes a pump trip on which initiates Recirculation a' signal of high reactor Pump Trip and Alternate Rod Inser-pressure or low-low reactor water tion shall be as specified in level when the mode select Table 4.2-G.
switch is in the RUN sode. The alternate rod insertion system provides for initiating control rod insertion whenever the mode switch is in the RUN, STARTUP or SHUTDOWN sode. The limiting con-ditions for operation for the instrumentation are listed in Table 3.2-G.
\\
Amendment No. 42
, - - ~ -
,,,,,-._.--,_,-.-,_.w.
.--,_e--,-,
FNPS TABLE 3.2.A INSTRUMENTATION THAT INITIATES PRIMART CONTAlleENT ISOLATION Minimum # of Operable Instrument h is Per Trip System (1)
Instrument Trip Level Settina Action (2) 2(7)
Reactor Low hter Level E9" indicated level (3)
A and D I
Reactor High Pressure (110 peig D
2 Reactor Low-Low Water Level at or above -49 in.
A indicated level (4) 2 Reactor High Hater Level 4_48" ladicated level (5) 5 2(7)
Righ Drywell Pressure
<2.5 psig A
2 High Radiation Main Steam 4_7 times normal rated B
Line Tunnel full power background
{
2 Law Pressure Main Steam Line
>880 psig (8)
B o
(
2(6)
Righ Flow Main Steam Line (140% of rated steam flow B
o" 2
Main Steam Line Tunnel f
Enhaust Duct High Temperature
[170*F B
2 Turbine Basement Exhaust Duct High Temperature 15 0 L0F B
1 Reactor Cleanup System High Flow
<3001 of rated flow C
I 2
Reactor Cleanup System High Temperacure
< 150 F C
1 4
I PNPS TABLE 3.2.B (Cont'd)
INSTRUMENTAT100i THAT DIITIATES OR C0errROLS THE CORE Alm CDertAllMENT COOLING SYSTDtS Minimum i of Operabie Instrument i
Channels Per Trip System (1)
Trip Function Trip Level Setting Remarka 2
High Dryvell Pressure 12.5 psig 1.
Initiates Core Spray; LPCI; HPCI.
1 1
2.
In conjunction with low-Low Reactor Water lovel, 120 second time delay and LPCI or Core Spray pump running, initiates Auto j
Blowdown (ADS).
J 3.
Initiates starting of Diesel Generators.
1 Reactor Iow Pressure 400 peig i 25 Permissive for Opening Core Spray and LPCI Admission valves.
f 1
Reactor Iow Pressure
<110 psig g,
In conjunction with PCIS signal permits closure of RER (LPCI)
B injection valves.
{
N z
y 1
Reactor Low Pressure 400 psig i 25 In conjunction with low-Iow y
Reactor Water Level initiates N
Core Spray and LPCI.
1 2
Reactor Iow Pressure 500 psig i 25 Prevents actuation of LPCI ao break detection circuit.
(
I PHPS TABLE 3.2.5 (Cont'd)
INSTRIDtENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEMS Minimum # of Operable Instrument g
Channels Per Trip System (1)
Trip Function Trip larvel Setting Remarks 2,
Startup Transformer OV with 1.1 Sec 1.
Trips Startup Transformer loss of Voltage Time Delay to Emergency Bus Breaker. (
3094V with 18 Sec 2.
Locks out automatic closure Time Delay of Startup Transformer to Emergency Bus.
3.
Initiates starting of Diesel Generators in conjunction with loss of auxiliary transformer.
4.
Prevents simultaneous start-ing of CSCS components.
5.
Starts load shedding logic for Diesel Operation in con-junction with (a) Low low E.
ReactorWaterLevelandII-E Reactor Pressure or (b) High R
drywell pressure or (c) Core Standby Cooling System com-2O ponents in service in con-junction with Auxiliary Transformer breaker open.
4
.I o
1.
These trip setpoints define the range of trip settings w
j selected from,the approtiriate relay curve i
~
\\
s.
PHPS TABLE 3.2.5 (Cont'd)
INSTRIMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINHENT COOLING SYSTEMS Minimum # of Operable Instrument Ch nnels Per Trip System (1)
Trip Function Trip Level Settima 2 Remarks 2
Startup Transformer 3745Y + 21 with l.
Trips Startup Transformer Degraded Voltage 9.2 + 0.5 sec.
to Emergency Bus Breaker. (
time' delay" 2.
Locks out automatic closure of Startup Transformer to Emergency Bus.
)
3.
Initiates starting of Diesel Generators in conjunction with loss of auxiliary transformer.
4.
Prevents simultaneous start-ing of CSCS components.
5.
Starts load shedding logic for Diesel Operation in con-junction with (a) low Low f
Reactor Water Level and Low Reactor Pressure or (b) High R.
drywell pressure or (c) Core Standby Cooling System com-ponents in service in con-S junction with Auxiliary 20 Transformer breaker open.
h 2.
Settings, subject to change after installation and tearing of relays, i
~
PNPS TABLE 3.2.B (Cont'd)
INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINNENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trip System (1)
Trip Function Trip Level Setting
~
Remarks 1
RER (LPCI) Trip System NA Monitors availability of power bus power monitor to logic systems.
1 Core Spray Trip System NA Monitors availability of power bus power monitor to logic systems.
1 ADS Trip System bus MA Monitors availability of power power monitor to logic systems and valves.
1 BPCI Trip System bus MA Monitors availability of power power monitor to logic systems.
1 RCIC Trip System bus NA f
Monitors availability of power power monitor to logic systens.
2 Recirculation Pump A d/p
<2 paid Operates RHR (LPCI) break de-tection logic which directs j
2 Recirculation Pump B d/p
<2 paid cooling water into unbroken
~
2 Recirculation Jet Pump 0.5<p<l.5 paid recirculation loop.
O' Riser d/p ^>B
$n
'f 1
Core Spray Sparger t
-1( l.5) psid Alarm to detect core spray Reactor Pressure Vessel sparger pipe break.
C 4/p
PNPS TABLE 3.2.8.1 INSTRtMENTATION THAT MONITORS BUS UNDERVOLTAGE Minimum # of Operable Instrument Channels Per Trip System Function Setting Remarks 1
Emergency Bus (1) 3856 2% with Alerts Operator Undervoltage Annunciation 9.2 0.1 %
to possible degraded Second Time Delay voltage conditions NOTE (1) In the event that the alarm system is determined inoperable, commence logging safety i
related bus voltage every 1/2 hour until such time as the alarm is restored to operable status.
1
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'.4
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1 e
f FNPS
- r TABLE 3.2.C INSTRIBENTATION tilAT INITIATES ROD BIACES Minissas f of Operable Instrsament Tein Level Settima i
i Instrinnent Channels Per Trip Systems (1)
APRM Upecals (Flow (0.6W + 42)
~ FEF" (2) 2
.MFLFO Blased) 2.5 indicated on scala APEM Downscale 2
(0.6W + 42)
FRF *
(2)
Rod Block Monitor JWLPD.
1 (7)
(Flow Bissed)
\\
5/125 of full scale and Block Honitor
'l 1 (7)
Downscale 1RM Downscale (3) 5/125 of full scale 3
IBM Detector not in (8) i 3
Startup Position 1RM Upscale
$108/125 of full scale 3
N SRM Detector not in (4) 2 (5)
Startup Position e
II.
5 SRM Upscale 110 counts /sec.
E1 i,'
2 (5) (6) 2:
,o C
r
r.=
w 1
NOTES FOR TABLE 3.2.C d Selector Switch, For the startup and run positions of the Reactor Mo esystems for each funct i
there shall be two operable or tripped tr p
" mode, and the 1.
The SRM and IM blocks need not be operable in "Run "Startup" mode. If the AFRK and EM rod blocks need not be orporable infirs this condi-ided thatly and daily thereafter; tion may exist for up to seven days prov i
operable system is functionally tested immed ate h
ystem shall be If this condition lasts longer than seven days, t e sIf the firs tems, the tripped.
systems shall be tripped.
d core flow of W is percent of drive flow required to produce a rat power (1998 MWt).
i 2.
t range.
IBM downscale is bypassed when it is on its lowes is 1 100 cys.
3.
This function is bypassed when the count rate 4.
One of the four SRM inputs may be bypassed.
switches are on 5.
This SRM function is bypassed when the IBM range 6.
range 8 or above.
30%.
The trip is bypassed when the reactor power is 1 h is placed in Run.
7.
This function is bypassed when the mode switc 8.
l 55 Amendment No. 42
,m a,
,.e
PNPS TABLE 3.2-G INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP AND ALTERNATE ROD INSERTION Minimum Number of Operable or Tripped Mode Select Instrument Channels Per Trip System (1)
Trip Function Trip Level Setting Requirements (2)
High Reactor Dome 1175 i-15 PSIG A/B Pressure Low-Low Reactor
> 78.5" above the A/B Water Level top of the active fuel NOTES 1.
Minimum number of cperable trip systems shall be two.
1 2.
With the r'aber of OPERABLE trip systems less than that required by note (1),
restore the inoperable trip system to operable status within 14 days or be in at least cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the mode switch positioned in either:
A.
STARTUP, REFUEL, or SHUTDOWN mcde for recirculatien pump trip system.
OR B.
REFUEL mode for alternate rod insertion system.
Amendment No.
42 59a
PNPS TABLE 4.2.B MINDalM TEST AND CALIBRATION FREQUENCY POR CSCS Instrument Channel Instrument Functional Test Calibration Frequency Instrument Check
- 1) Reactor Water Level (1)
Once/3 months once/ day 2)
Drywell Pressure (1)
Once/3' months None
- 3) Reactor Pressure (1)
Once/3' months None
- 4) Auto Sequencing Timers NA Once/ operating cycle None
Pressure Interlock (1)
Once/3 months None 6) a.
Loss of Voltage Monthly Once/ operating cycle Once/12 bra, b.
Degraded Voltage Relays Monthly Once/ operating cycle Once/12 hrs.
- 7) Trip System Bus Power Monitors Once/ operating cycle NA Once/ day
- 8) Recirculation System d/p (1)
Once/3 months once/ day
- 9) Core Spary Spar'ger d/p NA Once/ operating cycle Once/ day
(1)
Once/3 months None
(1)
Once/3 months None 1
i
- 12) Safeguards Area High Temp.
(1)
Once/3 months None
'E.
I
- 13) HPCI and RUIC Steam Lin' Low Pressure (1)
Once/3 months None
- E i
- 14) HPCI Suction Tank Levels (1)
Once/3 months None
- z
,?
,g 15)
Degraded Voltage Alarms Monthly Once/ Operating Cycle Once/12 hrs, i
l l
PNPS Table 4.2-G Minimum Test and Calibration Frequency for ATWS RPT/ARI Instrumentation 1
Instrument Instrument (2)
Instrument Functional (2)
I2)
Channel Test Calibration Check 1.
Reactor High Pressure (1)
Once/ Operating Once/ day Cycle-Transmitter Once/3 months -
Once/ day
/
Trip unit 2.
Reactor Low-Low Water Level (1)
Once/ Operating Once/ day Cycle-Transmitter Once/3 months -
Once/ day Trip unit Amendment No. 42 66a
i NOTES FOR TABLES 4.2.A THROUGH 4.2.G Initially once per month until exposure hours (M 1.
5 with an interval not less than one month nor more than th Functional tests, calibrations and instrument checks are not required when these instruments are not required to be operabl 2.
Calibrations with a required frequency not to exceed once per week.
tripped.
of IRMs and SRMs shall be performed during e Instrument checks shall be performed at least once per day during those periods when the instruments are required to be per week.
This instrumentation is excepted from the functional test definition.
The functional test will consist of injecting a simulated electrical 3.
signal into the measurement channel.
These instrument channels will be calibrated using simulated ele signals once every three months.
Simulated automatic actuation shall be prformed once each operat Where pessible, all logic system functional tests will be 4.
cycle.
performed using the test jacks.
Reactor low water level, high drywell pressure and high radiation o
main steam line tunnel are not included on Table 4.2. A since they 5.
are tested on Table 4.1.2.
The logic system functional tests shall include a 6.
the trip systems.
i l
l Amendment No.
42 67
l
- r r
l 3.2 BASg3 (Cont'd)
The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the Safety Limit McFL The trip logic for this function is 1 out of ni e.g.,
any trip on one of six AFM's, eight IM's, or four SM's will result in a rod block.
Se miniassa instrument channel requirements assure sufficient instru-mentation to assure the single failure criteria is ust.
Se=4=4==
instrument channel requirements for the REM any be reduced by one for maintenance, testing, or calibration. his time period is only 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.
The AFEN rod block function is flow biased and prevents a significant reduction in MCFR, especially during operation at reduced flow. The AFRM provddes gross core protection; i.e.,
limits the gross core power increase from withdravel of control rods in the normal withdrawal He trips are set so that MCFR is mainemined greater than sequence.
the Safety Limit MCPR.
The REN rod block functica provides local protection of the core, for a single rod withdrawal error from a limiting control rod pattern.
The IBM rod block function provides local as well as gross core pro-taction. The scaling arrangement in such that trip setting is less than a factor of 10 above the indicated level.
A downscale indication on an AP M or IBM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrummit will not respond to changes in control rod motion and thus, control rod action is prevented. The downscals trips are set at 2.5 indicated on scale.
The flow comparator and scram discharge voluna high level components have only one logic channel and are not required for safaty.
The refueling interlocks also operate one logic ch=aa=1, and are re-quired for safety only when the mode switch is in the refueling position.
For effective emergency core cooling for small pipe breaks,.the HPCI system mast function since reactor pressure does not decrease rapidly
'3te enough to allow either core spray or LPCI to operate in time.
automatic pressure relief function is provided as a backup to the 71 Amendment No.
42
~ *
- El'**8
- W Whm 9, M WMe*p %
m a _-,-,
--e
,~-,e,- -
-,,-----e e
.~--a-w+--,m-,
3.2 JASES(Cont'd, For each parameter monitored, as listed in Table 3.2.F. there areB two (2) e' 7 1s of instrumentation.
the tw.
1, channels, a near continuous surveillance of instrument Any deviation in readings will initiate performance is available.an early recalibration, thereb/ maintaining t instrument readinas.
The recirculation pump trip / alternate rod insertion sy 25016 (Reference 1) as referenced by the NRC as an acceptable Reference 1 provides both system design (Reference 2) for RPT.
The pump trip is provided descriptions and performance analyses.to minimize reactor pres plant transient coincident with the failure of all control rods to The rapid flow reduction increases core voiding provid scram.
negative reactivity feedback.The recirculation p ap trip is only sensors initiate the trip.
required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam which continues Requiring the to be generated in this unlikely postulated event.
trip to be operable only when in the RUN mode is therefore conserva-The low water level trip function includes a time delay of nine (9) seconds 1 one (1) second to avoid increasing the consequences tive.
This delay has an insignificant effect on of a postulated LOCA.
ATWS consequences.
Alternate rod insertion utilizes the same initiation logic and functions as RPT and provides a diverse means of initiati system to depressurize the scram pilot air header, which in turn reactor scram.
causes all control rods to be inserted.
References NEDO-25016. " Evaluation of Anticipated Transients Witnout Scram for the Monticello Nuclear Generating Plant." September 1.
1976.
NUREG 0460, Volume 3, December 1978.
2.
73 Amendment No.
42
l r
~~
I e
4.2 8ASES (Cont'd) instruments of similar design, a testing interval of once every three months has been found adequate.
f The automatic presrare relief instrumentation can be considered to be a,1 out of 2 logic system and the discussion above applies also.
The instrumentation which is required for the recirculation pump
{
trip and alternate rod insertion systems incorporate analog trans-The i
mitters and are a new, improved line of BWR instrumentation.
calibration frequency is once per operating cycle which is consistent with both the equipment capabilities and the requirements for The calibration 1
similar equipment used by other reactor vendors.
frequency of the trip units is proposed to be quarterly, the same l.tkewise, the test as other similar protective instrumentation.
frequency is specified at monthly like that of other protective j
A sensor check is proposed once per day; this is instrumentation.
considered to be an appropriate frequency, commensurate with the design applications and the fact that the recirculation pump trip and alternate rod insertion systems are backups to existing protective j
instrumentation.
i 77 Amendment No.
42
l
- ~._ ___
'~f 3.3 and 4.3 BASES:
During reactor operation with certain limiting control rod patterns, the withdrawal of a desig-nated single control rod could result in one or more fuel rods with MCFR's less than the Safety Limit MCFR. During use of such patterns, it is
. judged that testing of the REN system prior to withdrawal of such rode to assure its operability will assure that improper withdrawal does not It is the responsibility of the Rasctor occur.
Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperabia control rods in other than limiting patterns.
C.
Scram Insertion Times The control rod system is designed to bring the reac-tor suberitical at a rate fast enough to prevent fuel dausge; i.e., to prevent the McFR from becoming less l
than the Safety Limit MCFR. Analysis of the limiting power transient shove that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above Specification, provide the required protection, and MCFR r==aina greater than the Safety Limit MCFR.
The scram times for all control rods will be deter-mined at the time of each refueling outage. A re-presentative sample of control rods will be screa tested during each cycle as a periodic check against deterioration of the control rod performance.
Amendment No. 42 91 l
-, -. -. -.. - - -.. -. _. - - - - -, _ - - - -.. - -. - - - ~ - -.
-- ------*-w.
e
~
' LIMITI!30 CO*iDITIONS IOR OPERATION SURVEILLANCE REDUTH4 TNT 35 Cone AND C0t!TAffFE'f7 C00 LINO 4.$ CORE AND CONTAINMENT COOLING SYSTD48 8YSTEMS Applicability Applicability Applies to the operational status of Applies to the Surveillance Requirement:
the core and suppression pool cooling of the core and suppression pool cooling subsystems.
subsystems which are required when the corresponding Limiting Condition for ogs-eration is in effect.
Objective Objective h assure the operability of the core To verify the operability of the core and and suppression rool cooling subsystems suppression p>ol cooling subsystens under under all conditions for which this all conditior a for which this cooling ca-cooling espability is an essential re-pability is an essential response to sta-sponse to station abnormalities.
tion abnormalities.
Specification Specification A.
Core Spray and LPCI Subsystems A.
Core Serav and 7,PCI S ibsyste-1.
Both core spray subsystems shall 1.
Core Spray Subsystem Testing.
be operable whenever irradiated fuel is in the vessel and prior to reactor startup from a Cold
.a.
Simulated Once/Operatir.g Condition, except as specified Automatic Cycle in 3 5.A.2 below.
Actuation test.
b.
Pump Operability Once/ month and Once/ cycle from the Alternate Shutdown Panel c.
Motor Operated Once/ month and Valve Operability once/ cycle from the Alternate Shutdoun Panel d.
Pump flow rate Each pump shall deliver at least
~
3600 gpm against a system head corresponding to a reactor vessel pressure of 104 psig, e.
Core Spray Header A p Instrumentation Amendment No. 42 e
103
e
\\.
i LIMITING CONDITIONS POR OPERATION SURVEILLANCE EQUIPMENT 3.5.A Core Soray and LPCI Subsystees 4.5.A core Spray and LPCI subsystems (coat' d)
(cont' d)
Check Once/ day Calibrate Onec/3 months Test once/3 months 2.
From and af ter the date that one 2.
When it is determina.d that one core of the core spray subsystems is spray subsystem is inoperable, made or found to be inoperable the operable core spray subsystem, for any reason, continued reactor the LPCI subsystem and the diesel operation is permissible during generators shall be demonstrated to the succeeding seven days, pro-be operable immediately. The oper-vided that during such seven days able core spray subsystem shall be all active components of the other Besonstrated to be operable daily core spray subsystem and active thereafter.
components of the LPCI subsystem and the diesel generators are op-erable.
3.
The LPCI Subsystems shall be oper-3.
LPCI Subsystem Testing shall be as able whenever irradiated fuel is follows:
in the reactor vessel, and prior to reactor startup from a Cold a.
Simulated Automa-Once/ Operating Condition, except as specified tic Actuation Test Cycle in 3.5. A. 4, 3. 5. A. 5 and 3.5.F.5.
b.
Pump Operability Once/ month and Once/ cycle from the Alternate Shutdown Panel c.
!!otor Operated Once/l'.onth and valve operability once/ cycle from tha Alternace Shutdown Panel d.
Pump Flow Once/3 months Three LPCI pumps sha".1 deliver 14,400 gpm against a system head corresponsing to a vessel pressure of 20 psig Amendment No. 42 104
l l
l i
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.3 Containment coolina subsystem 4.5.3 containment Coolina subsystes 1.
Except as specified in 3.5.B.2, 1.
Contai==aat Cooling subsystem Testing 3.5.3.3, and 3.5.F.3 below, shall 'ae as followst both containment cooling subsystem loops shall be operatie whenever Ites Frequency irradiated fuel is in the reactor Pump & Valve Operability once/3 months vessel and reactor coolant temper-a.
ature is greater than 212'F, and and once/ cycle from the Alternate prior to reactor startup from a Cold Condition.
Shutdown Station b.
Pump Capacity Test Af ter putap Each RBCCU pump shall tsaintenance deliver 1700 gpm at and every 70 ft. TDH. Each SSWS 3 months pump shall deliver 2700 gpa at 55 ft. TDH.
c.
Air test on drywell and once/5 years torus headers and nozzles Amendment No. 42 106
LIMITING CONDITION FOR OPERATION SURVEILTJueCE REQUIREMENT 3.5.3 containment Coolina subsystem 4.5.3 Containment Coolina Subsystem (Cont'd)
(Coat'd) 2.
From and after the date that one 2.
When one containment cooling subsystem contairent cooling subsystem loop loc,p becomes inoperable, the operable is sacs or found to be inoperable subsystem loop and its associated for any reason, continued reactor diesel guarator shall be demonstrated operation is permissible only during to be operable innsediately and the the succeeding seven days unless operable containment cooling subsystma such subsystem loop is sooner made loop daily thereafter.
operable, provided that the other containment cooling subsystem loop, including its associated diesel generator, is operable.
3.
If the requirsments of 3.5.3 can-not be met, an orderly shstdown shall be initiated and the reac-l tor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
)
l C.
HPCI Subsystem C.
HPCI Subsystem 1.
The HPCI Subsystem shall be oper-able whenever there is irradiated 1.
HPCI Subsystem testing shall be per-fuel in the reactor vessel, reactor formed as follows:
pressure is greater than 104 poig, and prior to reactor startup from a.
Simulated Auto-Once/ operating a Cold Condition, except as speci-matic Actuation cycle fied in 3.5.C.2 and 3.5.C.3 below.
Test b.
Pump Operability Once/ month and Once/ cycle from the Alternate Sht.t-down Station c.
Motor Operated Once/ mot.th and Valve Operability Once/ cycle from the Alternate Shutdown Station d.
Flow Rate at Once/3 months 1000 psig e.
Flow Rate at Once/ operating 150 psig cycle i
l Amendment No. 42 107
l 11mTarso CONDITI5N FOR OPEHAT1Um
- ava,s. W - ' h i
~ ~..
1- -
3 5.C HPCISubsystes(Cont'd)
'%.5.C JtPCISubsystem(Cont'd) e The HPCI pump shall deliver at least 4250 spa for a sys-tem head corresponding to a l
reactor pressure of 1000 to 1$0 psig.
'2.
From and after the date that the 2.
When it is determined that the HPCI 1
HPCI Subsystem is made or found to Subsystem is inoperable the RCIC, the be inoperable for any reason, con-LPCI subsystem, both core spray sub-tinued, reactor operation is per-missible only during the succeed-systems, and the ADS subsystem actus-tion logic shall be demonstrated to ing seven days unless such subsys.
be operable immediately. The RCIC tem is sooner made operable, pro.
viding that during such seven system and ADS subsystem logic shall days all active components of the be demonstrated to be operable daily thereafter.
ADS subsystem, the RCIC system, the I.PCI subsystem and both core spray subsystema are operable.
3.
If the requirements of 't.5.C can-not be met, an orderly shutdown shall be initiated and the reac-tor pressure shall be reduced to or below 104 peig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3 5.D _ Reactor Core Isolation Cooling 4 5.D Reactor Core Isolation Cooling (RCIC) Subsystem (RCIC) Subsystem 1.
The RCIC Subsystem shall be oper-1.
able whenever there is irradiated RCIC Subsystem testing shall be per-formed as follows:
fuel in the reactor vessel, the re-actor pressure is greater than 104 s.
Sidsed Auto-Once/ operating pais, and prior to reactor startup natio Actuation from a Cold Condition, except as cycle Test specified in 3 5.D.2 below, b.
Pump Operability Oncs/ month and Once/ cycle from the Alternate Shutdown Station I c.
Motor Operated Once/ month and Valve Operability Once/ cycle fron the Alternate Shutdown Stationi 108 Amendment No.
42
l t
LD4ITING CONDITION FOR OPERATION SURVEILI/J.Cg pg1UIlG24ENTS l
3 5.D Reactor Core Isolation Cooling 4.5.D Reactor Core Isolation cooling (RCIC) Subsystem (Cont'd)
(RCIC) Sub:ystem (Cont'd)
I d.
Flow Rate at once/3 month:
1000 psig e.
Flow Rate at once/ operating l>0 psig cycle The RCIC pump shall deliver at 1 cast LOO sp: for a systc= hcLi corresponding to a reactor pres-sure of 1000 to 150 psig 2.
From and after the date that the 2.
When it is determined that the RCIC RCICS is made or found to be inop-subsystem is inoperable, the hPCIC erable for any reason, continued shall be demonstrated to be operable reactor power operation is permis-immediately and,veekly thereafter.
sible only during the succeeding 1
sevea days provided that during such seven days the HPCIS is oper-1 able.
1 3.
If the requirements of 3.5.D cannot be met, an orderly shutdown shall be initiated and the reactor pres-sure'shall be reduced to or below 104 psig with!.n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.5.E Automatic Depressurization 4.5.E Auto-atic Depressurizatio..
System ( ADS)
System ( ADS)
~
1.
The Automatic Depressurization Sub-1.
Durin6 each operating cycle the system shall be operable whenever following tests shall be perforr.ei there is irradiated fuel in the on the ADS:
reactor vessel and the reactor pres-sure is greater than 104 psig and a.
A sicalated automatic ace.a-ion prior to a startup from a Cold Con-test shall be performed prier to dition, except as specified in startup after each refuel.ng out-3 5.E.2 below.
age.
This test shall also be performed from the Alternate Shutdown Station within the same time frame.
b.
With the reactor at pressure.
each relief valve shall be man-ually opened until a corresponding change in reactor pressure or main turbine bypass valve positions indicate that steam is flowina Amendment No.
42 from the valve 109
9 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 1
l l
3.6.A Thermal and Pressurization 4.6.A Thermal and Pressurization Limitations (Cont'd)
Limitations (Cont'd)
)
neutron flux specimens shall be re-moved at the frequency required by 10 CFR50 Appendix H and tested to experimentally verify or adjust the calculated values of integrated neutron flux that are used to deter-mise the NDTT for Figure 3.6.2.
i 3.
The reactor vessel head bolting 3.
When the reactor vessel head bolt-stude shall not be under cension ing studs are tensioned and the unless the temperature of the reactor is in a Cold Condition, the vessel head flange and the head reactor vessel shall temperature is greater than 500F.
immediately below the head flange shall be permanently recorded.
4.
The pump in an idle recirculation 4.
Prior to and during startup of an loop shall not be started unless idle recirculation loop, the tem-the temperatures of the coolant perature of the reactor coolant within the idle and operating re-in the operating and idle loops circulation loops are within 500F shall be permanently logged.
of each other.
5.
The reactor recirculation pumps 5.
Prior to starting a recirculation shall not be started unless the pump, the reactor coolant temper-coolant tempera;ures between the atures in the dome and in the done and the bottom head drain bottom head drain shall be compared are within 145 F.
and permanently logged.
6.
Thermal-Hydraulic Stability Care thermal power shall not exceed 25% of raced thermal power without forced recirculation.
B.
Coolant Chemistry B.
Coolant Chemistrv 1.
The reactor coolant system radio-1.
a.
A reactor coolanc sample shall activity concentration in water be taken at ". east every 96 shall not exceed 20 microcuries hours and analyzed for radio-of total iodine per al of water.
activity content.
b.
Isotopic analysis of a reactor coolant sample shall be made at least once per month.
2.
The reactor coolant water shall 2.
During startups and at steaming rates not exceed the following limits less than 100,000 pounds per hour, with steaming rates less than a sample of reactor ecolant shall 100,000 pounds per hour, except be taken every four hours and an-as specified in 3.6.B.3:
alyzed for chloride content.
Conductivity.. 2 unho/cm 124 Chloride ion.. 0.1 ppa
- Amendment No. 42
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.3 Coolant Chemistry (Cont'd) 4.6.5 Coolant Chemistry (Cont'd) 3.
For reactor startups and for the J.
a.
With steaming rates of 100,000 first 24 hsurs after placing the pounds per hour or greater, a reactor in the power operating reactor coolant sample shall be condition, the following limits taken at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> shall not be exceeded, and analyzed for chloride ion content.
Conductivity.. 10 unho/cm b.
When all continuous conductivity Chloride ion.. 0.1 ppe monitors are inoperable, a reactor coolant sample shall be taken at lease daily and analyzed for conductivity and chloride ion content.
4.
Except as specified in 3.6.B.3 above, the reactor coolant water shall not exceed the following limits when operating with steam-ing rates greater than or equal to 100,000 pounds per hour.
Conductivity.. 10 umho/cm Chloride ion.. 1.0 ppm 5.
If Specification 3.6.B cannot be met, an orderly shutdown shall be initiated and the reactor shall be in Hot Shutdown within 24 hrs.
and Cold Shutdown within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Coolant Leakage C.
Coolant Leakage 1.
Any time irradiated fuel is in the 1.
Reactor coolant system leakage shall reactor vessel and reactor coolant be checked by the sump and air temperature is above 2120F, reactor sampling system and re:oro6d at coolant leakage into the primary least once per day.
containment from unidentified sources shall not exceed 5 gym.
In additiot, the total reactor coolant system leakage into the primary containment shall not exceed 25 spm.
2.
Both the sump and air sampling sys-taas shall be operable during reac-tor power operation. From and after the date that one of these systems is made or found to be inop-erable for any reason, reactor Amendment No. 42 125 I
I
=
l t,IMITg G CONDITION FOR OPERATION SURVEIf1ANCE REQUIREMENT 3.6.C Coolant Chemistry (Cont'd) 4.6 power operation is permissible only during the succeeding seven days.
3.
If the conditions in't or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown Condi-tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.
Safety and Relief Valves D.
Safety and Relief Valves 1.
During reactor power operating 1.
At least one safety valve and two conditions and prior to reactor relief / safety valves shall be startup from a Cold Condition, or checked or replaced with bench whenever reactor coolant pressure checked valves once per operating is greater than 104 peig and ten-cycle. All valves will be tested perature greater than 340*F, both every two cycles.
safety valves and the safety modes of all relief valves shall be op-The set point of the safety valves arable.
shall be as specified in Specifi-cation 2.2.
2.
At least one of the relief / safety valves shall be disassembled and inspected each refueling outage.
I 1.
If Specification 3.6.D.1 is not met, an orderly shutdown shall be initi-ated and the reactor coolant
~
Amendment No. 42 126
I I
l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 9
3.6.D Safety and Relief Valves (Cont'd) pressure shall be below 104 peig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E.
Jet Pumps E.
Jet Pumps 1.
Whenever the reactor is in the Whenever there is recirculation flow startup or run modes, all jet with the reactor in the startup or pumps shall be operable. If it is run modes, jet pump operability shall determined that a jet pump is be checked daily by verifying that inoperable, en orderly shutdown the following conditions do not oc-shall be initiated and the reactor cur simultaneously:
shall be in a Cold Shutdown Condi-tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1.
The two recirculation loops have a flow imbalance of 15%
or more when the pumps are 1
operated at the same speed.
2.
The indicated value of core flow rate varies from the value derived from 1 cop flow measurements by more than 10%.
3.
The diffuser to lover plenum differential pressure reading on an individual jet pump varies from established' jet pump P characteristics by more than 10%.
F.
Jet Pump Flow Mismatch F.
Jet Pump Flow Mismatch 1.
Whenever both recirculation pumps Recirculation pump speeds shall be are in operation, pump speeds shall checked and logged at least once be maintained within 10% of each per day.
other when power level is greater than 80% and within 15% of each other when power level is less than or equal to 80%.
8:.
, Structural Integrity G.
Structural Integrity 1.
The structural integrity of the Tne nondestructive inspections listed primary system boundary shall in Table 4.6.1 shall be performed as be maintained at the level re-specified. The results obtained from quited by the ASME Boiler and compliance with this specification i
Pressure Vessel Code, Section will be evaluated after 5 years and XI, " Rules for Inservice In-the conclusions of this evaluation spection of Nuclear Power will be reviewed with AEC.
Plant components," 1974 Amendment No. 42
Table 3.6.1 SAFETY REl.ATED SHOCK SUPPRESSORS (SNUBBERS)
Snubber No.
Iocation Elevation Snubber in High Snubbers Snubbers Snubbers Radiation Area Especially Inaccessible Accessible During Shutdown Difficult to During Normal During Normal Remove Operation Operation SS-6-10-1 Feedwater System 42' X (Drywell)
SS-6-10-2 Feedwater System 42' X (Drywell)
SS-6-10-3 Feedwater System 42' X (Drywell)
S S 10-4 Feedwater System 42' X (Drywell)
SS-6-10-5 Feedwater System 42' X (Drywell)
SS-13-3-1 RCIC 38' X (Drywell)
SS-13-3-2 RCIC 38' X (Drywell)
S S-14-3-1 Core Spray 65' X (Drywell)
SS-14-3-2 Core Spray 65' X (Drywell)
SS-14-3-3 Core Spray 65' X (Drywell)
SS-14-3-4 Core Spray 65' X (Drywell)
SS-23-10-1 H.P.C.I.
42' X (Drywell)
SS-23-10-2 H.P.C.I.
42' X (Drywell)
S.-23-3-30 H.P.C.I.
-3' 09" X H.P.C.I. Quadrant S-23-3-31 H.P.C.I.
-3' 09" X H.P.C.1. Quadrant S-23-10-32 H.P.C.I.
-3' 09" X H.P,C.1. Quadrant S-23-10-34 H.P.C.I.
-6' X H.P.C.I. Quadrant g S-23-10-35 H.P.C.I.
- 6' X H.P.C.I. Quadrant S-23-3-36 H.P.C.I.
-3' 09" X H.P.C.I. Quadrant n.
E. S-23-3-37 H.P.C.I.
- 3' 09" X H.P.C.1. Quadrant
'I S-10-3-43 RHR
-3' 06" X RRR Pump Room if, S-10-20-44 RHR
- 3' 06" X RHR Pump Room z S-30-3-45 RBCCW 83'5" X Reactor Building P S-10-10-46 RHR 6"
X Torus Compartment i
b Hodifications to this Table due to changes in high radiation areas should be submitted to the NRC as part of the next license amendment.
w O
BASES:
3.6.B Coolant Chemistry l
The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.
Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the higher limit on chlorides is permitted during POWER OPERATION. During shutdown and refueling oper-ations, the temperature necessary for stress corrosion to occur is not present so high concentrations of chlorides are not considered harmful during these periods.
In the case of BWR's where to additives are used and where neutral PH is maintained, conductivity provides a very good measure of the quality of the reactor water.
Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition caus-ing the change before limiting conditions, with respect to variables i
affecting the boundaries of the reactor coolant, are exceeded. Methods available to the operacor for correcting the off-standard condition include operation of the reactor clean-up system, reducing the input of impurities and placing the reactor in the cold shutdown condition. The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the clean-up system to re-establish the purity of the reactor coolant. During start-up periods, which are in the category of less than 1% reactor power, conductivity may exceed 2 u mho/cm because of tha initial evolution of gases and the initial I
addition of dissolved metals. During this period of time, when the conductivity exceeds 2 u mho/cm (other than short term spikes), samples will be taken to assure that the chloride concentration is less than 0.1 ppm.
)
Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions.
When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivicy meters inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.
Amendment No.
42 141
BASES:
4.6.B Coolant Chemistry The iodine radioactivity will be monitored by reactor water sample analysis.
The total iodine activity would not be expected to change over a period of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
In addition, the trend of the stack off-gas release rate, which is continuously monitored, is an indication of the trend of the iodine activity in the reactor coolant. Since the concentration of radio-activity in the reactor coolant is not continuously measured, coolant sampling would be ineffective as a means to rapidly detect gross fuel element failures.
However, some capability to detect gross fuel element failures is inherent in the radiation monitors in the off-gas system and on the main steam lines.
The conductivity of the reactor coolant is continuously monitored. The samples of the coolant which are taken every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will also be used to determine the chlor-ides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content.
Isotopic analyses to determine major con-tributors to activity can be performed by a gamma scan.
Amendment No. 42 142
(DELETED)
Amendment No. 42 146
~
w.
Lem and 4.7.B Standby
'is TreatmInt System and t
3.7.B. Standby Gas Treatmenti Control Room With Effibfency Air Control %com High Efficiency Air Filtration System Filtration System 1.
Standby Gas Treatment System 1.
. (1.) At least once every 18 a.
Except as specified in a.
3.7.B.l.c below, both trains months, it shall be of the standby gas treatment demonstrated that pressure system cnd the diesel genera-drop across the combined tors requb ed for operation of high efficiency filters such trains shall be operable and charcoal adsorber banks at all times when secondary is less than 8 inches of containment integrity is water at 4000 cfm.
i required or the reactor shall be shutdown.in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
(2.) At least once every 18 months, demonstrate b.
(1.) The results of the in-that the inlet heaters place cold DOP tests on on each train are
)
HEPA filters shall show operable and are capable
>99% DOP removal. The of an output of at least results of halogenated 14 kW. Perform an hydrocarbon tests on instrument functional charcoal adsorber banks test on the humidistats shall show >99% halogenat-controlling the heaters.
ed hydrocarFon removal.
(3.) The tests and analysis of (2.) The results of the Specification 3.7.B.l.b.2 l
laboratory carbon sample shall be perforned at least analysis =Y. i show >95%
once every 18 months or i
methyl fadide removaT at following painting, fire a velocity within 10% of or chemical release in any system design, 0.5 to ventilation zone communicat-1.5 mg/m3 inlet methyl ing with the system while iodide concentration, f the system is operating that.
>70% R.H. and >1900F.
could contaminate the HEPA i
filters or charcoal adsorbers !
c.
From and after the date that one train of the Standby Gas (4.) At least once every 18 Treatment System is made or months, automatic found to be inoperable for initiation of each branch any reason, continued of the standby gas reactor operation or fuel treatment system shall handling is permissible only be demonstrated, with during the succeeding seven Specification 3.7.B.l.d days providing that within satisfied.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and daily thereafter, all active components of the (5.) Each train of the standby other standby gas treatment gas treatment system shall train shall be demonstrated be operated for at least 1
to be operable.
15 minutes per month.
d.
Fans shall operate within (6.) The tests and analysis of
+10% o'.' 4000 cfm.
Specification 3.7.B.1.b. (2) shall be performed after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system Amendment No. 42 158
a 3.7.8 (Continued) 4.7.8 (Con nued) e.
Except as specified in
- b. (1.) Inplace cold DOP testing 3.7.B.l.c, both trains of shall be performed on the the standby gas treatment HEPA filters after each f
system shall be operable completed or partial during fuel handling replacement of the HEPA operations.
If the filter bank and after any system is not operable structural maintenance on fuel movement shall not the HEPA filter system be started (any fuel housing which could affect assembly movement in the HEPA filter bank progress may be completed).
bypass leakage.
(2.) Halogenated hydrocarbon testing shall be performed on the charcoal adsorber bank after each partial or complete replacement of the charcoal adsorber bank or after any struc-tural maintenance on the charcoal adsorber housing which could affect the charcoal adsorber bank bypass leakage.
Amendment No. 42 158A
3.7.8 (Continued) 4.7.8 (Con iued) 2.
Control Room High Efficiency Air 2.
Control Room High Efficiency Air Filtration System Filtration System a.
Except as specified in a.
At least once every 18 Specification 3.7.B.2.c months the pressure drop below, both trains of the across each combined filter Control Room High Efficiency train shall be demonstrated Air Filtration System used to be less than 3 inches of for the processing of inlet water at 1000 cfm.
air to the control room under accident conditions and the b.
(1.) The tests and analysis of diesel generator (s) required Specification 3.7.B.2.b for operation of each shall be perfomed once train of the system shall every 18 months or be operable whenever secondary following painting, fire containment integrity is or chemical release in required and during fuel any ventilation zone handling operations.
communicating with the system. While the systen b.
(1.) The results of the in-is operating.
place cold DOP tests on HEPA filters shall show (2.) Inplace cold DOP testing
>99% DOP removal. The shall be performed after results of the halogenat-each complete or partial ed hydrocarbon tests on replacement of the HEPA charcoal adsorber banks filter bank or after any shall show >99% halogenat-structural maintenance ed hydrocarFon removal on the system housing when test results are which could affect the extrapolated to the HEPA filter bank bypass initiation of the test.
leakage.
(2.) The results of the (3.) Halogenated hydrocarbon laboratory carbon sample analysis shall show >95%
testing shall be performed methyl iodide removaT at after each complete or partial replacement of a velocity within 10% of the charcoal adsorber system design, 0.05 t bank or after any 0.15 mg/m3 inlet methyl structural maintenance iodide concentration, on the system housing
>70% R.H., and >1250F.
which could affect the charcoal adsorber bank c.
From and after the date that bypass leakage.
one train of the Control Room High Efficiency Air Filtration (4.) Each train shall be System is made or found to be operated with the heaters i
incapable of supplying filtered in automatic for at least air to the contml room for any 15 minutes ever.y month-reason, reactor operation or refueling operations are (5.) The test and analysis of permissible only during the Specification 3.7.8.2.b.(2) succeeding 7 days.
If the shall be perfomed after system is not made fully oper-1 every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system aele within 7 days, reactor operation.
Amendment No. 42 i
1588
3.7.8 (Continued) 4.7.B (Continued) shutdown shall be c.
At least once every 18 months initiated and the reactor the following shall be
~
shall be in cold shutdown demonstrated:
within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and irradiated fuel handling (1) Automatic initiation of operations shall be terminated the control room high within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
(Fuel handl-efficiency air filtration ing operations in progress system.
may be completed).
(2) Operability of heaters d.
Fans shall operate within at rated power.
+ 10% of 1000 cfm.
3.
Perform an instrument functional test on the humidistate controlling the heaters.
Amendment N6. 42 158C
TABLE 3.7.1 i
PRIMARY CONTAIMENT ISOLATION VALVES Number of Power Maximum Action se Operated Valves Operating Normal Initiating Group Valve Identification Inboard Outboard Time (sec.)
Position Signal 1
Main Steam Line isolation valves 4
4 3iT 65 0
CC 1
Main steam line drain isolation valves 1
1 30 C
SC l
1**
Reactor Water sample line isolation valves 1
1 10 C
SC 2
Drywell purge supply isola-C(1)
SC tion valves 2
15 0(1)
CC 2
Suppression chamber purge supply isolation valves 2
15 0
CC a
Z 2
Nitrogen purge isolation valve 1
10 0
CC e
2 Nitrogen makeup isolation valve 1
10 C
SC 2
Suppression chamber nitrogen makeup isolation valve 1
10 C
SO 2
Drywell purge exhaust isola-tion valves 2
15 C
SC 2
Drywell exhaust isolation valves 2
10 C
SC g
k 2
Suppression chamber purge g
exhaust isolation valves 2
15 C
SC
- .:[
2 Suppression chamber exhaust isolation valves 2
10 C
SC
NOTES FOR TABLE 3.7.1 (Cont'd) 2.
High reactor vessel pressure l,
3.
High drywell pressure i
i n.
GROUP 4: Isolation valves in the high pressure coolant injection system (HPCI) are closed upon any one of the following signals:
-g:
1.
HPCI steam line high flow lC 2.
High temperature in the vicinity of the HPCI steam line 3.
Low reactor pressure GROUP 5: Isolation valves in the RCIC system are closed upon any one of the following signals:
!:'g 1.
RCIC steam line high flow as f
2.
High temperature in the vicinity of the RCIC steam line 3.
Low reactor pressure l
l GROUP 6: Actions in Group 6 are initiated by any one of the following:
8 1.
Reactor low water level a
2.
Cleanup. area high temperature 3.
Cleanup inlet high flow 1
CThe RHRS shutdown cooling injection isolation valves require a Group 2 signal plus high reactor vessel pressure.
e*The Reactor Water Sample Line Isolation Valves initiate on a Group 1 signal plus high drywell pressure.
1 i
I i
t
BASES
)
3.7.B.1 and 4.7.B.1 - Standby Gas Treatment System The Standbv Gas Treatment System is designed to filter and exhaust the reactor building r rmosphere to the stack during secondary containment isolation conditions. Upon containment isolation, both standby gas treatment fans are designed to start to bring the reactor building pressure negative so that all leakage should be in leakage. After a preset time delay, the standby fan automatically shuts down so the reactor building pressure is maintained approximately % inch of water negative.
Should one system fail to start, the redundant system is designed to start automatically. Each of the two trains has 100% capacity.
High Ef ficiency Particulate Air (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environr.ent and to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radiciodine to the environment. The in-place test results should indicate a system leak j
tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA filter efficiency of at least 99 percent removal of cold DOP particulates.
The lehoratory carbon sample test results should indicate a methyl iodide removal efficiency of at least 95 percent for expected accident conditions.
The specified efficiencies for the charcoal and particulate filters is sufficient to preculde exceeding 10 CFR 100 guid.elines for the accidents analyzed. The analysis of the loss of coolant accident assumed a charcoal adsorber efficiency of 95% and TID 14844 fission product source terms, Hence, installing two banks of adsorbers and filters in each train provides adequate margin. A 14 kW heater maintains relative humidity below 70% in order to ensure the efficient removal of methyl iodide on the impregnated charcoal adsorbers. Considering the relative simplicity of the heating circuit, the test frequency of once per 18 months is adequate to demonstrate operability.
Air flow through the filters and charcoal adsorbers for 15 minutes each month assures operability of the system. Since the system heaters are automatically controlled, the air flowing through the filters and adsorbers will be < 70%
relative humidity and will have the desired drying effect.
Tests of impregnated charcoal identical to that used in the filters indicate that shelf life of five years leads to only minor decreases in methyl iodide removal efficiency. Hence, the frequency of laboratory carbon sample analysis is adequate to demonstrate acceptability. Since adsorbers must be removed to perform this analysis, this frequency also minimizes the system out of service time as a result of surveillance testing. In addition, although the halogenated hydrocarbon testing is basically a leak test, the adsorbers have charcoal of known efficiency and holding capacity for elemental iodine and/or methly iodide, the testing also gives an indication of the relative efficiency of the installed system.
The required Standby, Gas Treatment System flow rate is that flow, less than or equal to 4000 CF1L which is needed to maintain the Reactor Building at a 0.25 inch of water negative pressure under calm wind conditions. This capability is adequately demonstrated during Secondary Containment Leak Rate Testing performed pursuant to Technical Specification 4.7.C.l.c.
~~
172 Amendment No. 42 l
l
...,.a.
-).. mm.n m....
/
i
,The test frequencies a.re udequate to detect equirc.cr.t eleterient.on pricr to significant defects, but the tests are actfrequent enouch.to load the filters The filter or ad:crbers, thus reducing their reserve ca.pacity too quickly.
testing is performed pursuant to appropriate procedures reviewed and 2.; proved by the Operations Review Coszittee pursu: nt to Section o of these Technical d by
, Specifications. De in-place testing of charecal filters is perfernein Measurezants of the constntration upstrec.= and dcyns rean are adsorbers.
The xstio of the inlet and cutlet concentrations gives an overall ande.
A similar precedure indication of the leak tightness of the system.
substituting dicctylphthalate for halogenated hydrocarbon is used so test the EEBL filters.
Pressure drop tests across filter and adsorber banks are performed to detect Considering the pluggLng or leak paths through the filter or adsorber media.
relatively short times the fans will be run for test purposes, plu -ing is 18 months is reasonable.
=14 hly and the test interval of once per System drains and housing gasket doors are designed such that any leakaga This ensures would be inleakage from the Standby Gas Treatment System Room.
that there will be no bypass ot' process ait around the filters or adsorbers.
Only one' of the two Standby Gas Treatment Systems (SBGl'S) is needed to maintain the secondary containment at a O.25 inch of water negative pressure upon containment If one system is found to be inoperable, there is no i==ediate threat to the isolation.
containment system performance and reactor operation or refueling activities may continue while repairs are being sade. In the event one S3GTS is This substantiates inoperable, the redundant system will be tested daily.
' the availability of the operable system and justifies continued reactor or refi_=14y operations.
If both trains of SBGTS are inoperable, the plant is brought to a condition where the SBG2B is not required.
~
Amendment No. 42 e
e e l
y 3 7.3.2.b and 4.7,3.2.b - Control Room Rich Effieleney Air Filtration ::r:ta The Control Room High Efficiency Air Filtration Systes is design =d to filtar intake air for the control room atmosphere during conditions when, normal intake air cay be contaminated. Fonoving ma=ual initiation, the control Room High Efficiency Air Filtration System is desigr.ed to positicn dacpers and start fans which divert the acrsal air flow through charcoal adsorbers before it reaches the control roca.
High Efficiency Particulate Air (HEPA) filters are instaned before the The charcoal charcoal adsorbers to prevent clogging of the icdine adsorbers.
adsorbers are instaued to reduce the potential intake of radiciodice to the control room. A second bank of HEEA filters is instaned downstrea: of the charccal filter.
The in-place test results should indicate a systa= leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and & HEPA The efficiency of at least 99 percent removal of cold,DOP particulates.
laboratory carbon sample test results should indicate a cathyl iodide renovel Tests of efficiency of at least 90 percent for expected accident conditions.
impregnated charcoal identical to that used in the filters indicate that shelf, life of five years leads to only minor decreases in methyl iodide removal efficiency. Hence, the frequency of laboratory carbon sample analysis is adequate to demonstrate acceptability. Since adsorbers =ust be re:oved to perform this analysis, this frequency also minimizes the system out of In addition, although the service time as a result of surveinance testing.
halogenated hydrocarbon testing is basica n y a leak tess, the adsorbers have charcoal of known efficiency and holding capacity for ele ental iodine and/cr methyl iodide, the testing also gives an indication of the relative efficiency of the insta ned system.
Determination of the system pressure drop once per operating cycle provides indication that the HER filters and charcoal adsorbers are not clogged by excessive amounts of foreign matter and that no bypass routes through the filters or adsorbers had developed. Considering the relatively shcrt times the syster.s will be operated for test purpo:es, plugging is ur inely and the test interval of once per operating cycle is reasonable.
The test frequencies are adequate to detect equipment decerioration prior to significant defects, but the tests are not frequent enough to lead the filters or adsorbers, thus reducing their reserve capacity too quickly. The filter testing is performed pursuant to appropriate procedures reviewed and appro'ved by the Operations Review Committee pursuant to Section 6 of these Technical Specifications. The in-place testing of charcoal filters is performed by injecting a halogenated hycrocarbon into the syste:n upstros: of the charcoal adsorbers. Measurements of the concentration upstrems and downstream are made. The ratio of the inlet and ou:les concentrr.: ions gives A similar an overall indication of :he lemk tightness of tha systen.
procedure substituting dioctyl phthalate for halogecaced hydrocarbon is used to test the HE2A filters.
If both treias of the system are found to be inoperanle, there is no Amendment No. 42 174 i
I immediate threat to the control room and reactor operation or fuel handling may continue for a limited period of time while repairs are being made.
If at least one train of the systest cannot be repaired within seven days, the reactor will be brought to a condition where the Control Room High Efficiency Air Filtration System is not required.
Air flow through the filters and charcoal adsorbers for 15 minutes each month assures operability of the system.
Since the system heaters are automatically controlled, the air flowing through the filters and adsorbers will be 1 70%
relative humidity and will have the desired drying effect.
1 Amendment No. 42 174A
1.7.C - Seoendary Costnint.ent The secondary contain=ent is designed to minimize any dround level release of radioactive materials which might result from a serious accidens. me i
reactor bu41diac provides secondary conta4" ent during reactor operation, when the drywell is sealed and in service; the reactor ba41dir3 provides primary containment when the reactor is shutdown and the drywell is open, as during refueling. Because the secondary coctnin=ent is an integrni part of the cocplete containment system, secondary contain:nent is required at all timer that primary contaiarar-t is required as well as duri=g retteling.
Initiating reactor building isolation and operation of the stscdby gas treatment system to saintain at least a 1/l+ inch of water negative pressure within the seccadary con +24= ant provides an adequate test of the cperation of the reactor bn41d4 ng isolation valves, leak tightness of the reactor building and performance of the standby gas treat =ent system. Functior@- testi=g the initiating sensors and associated trip channels demonstrates the cm3mbility for automatic actuation. Performing these tests prior to refn= ling will demonstrate secondary containnent capability prior to the time the primary conta4 vent is opened for refueling. Periodic testing gives sufficiant eenfidmace of reactor W1 A4ng ingggrity and g*ggy ggg treatment system perferunace capability.
174B Amendment No. 42
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRIMENTS 3.8 RADIOACTIVE MATERIALS 4.8 RADIOACTIVE MATERIALS Applicability:
Applicability:
Applies to the controlled release Applies to the periodic test and of radioactive liquids and gases raccrd requirements and sampling from the facility.
and monitoring methods used for facilities affluents.
Objective:
Objective:
To define the limits and conditions To ensure that radioactive liquid for the release of radioactive and gaseous releases from the ef fluents to the environs to assure facility are maintained within that any radioactive releases are ss dhe limits specified by Specifica-
)
low as practicable and would not tions 3.8.A and 3.8.B.
1 result in radiation exposures greater than a few percent cf natural background exposures and, in any event, within the limits of 10 CFR Part 20 for instantaneous release rates.
Specification:
Specification:
A.
Liquid Ef fluents A.
Liquid Effluents l
1.
The instantaneous gross 1.
Facility records shall radioactivity release be maintained of the concentration in liquid endioactive concentrations effluents from the ses-and volume before dilution tion shall not exceed the.
of each batch of liquid values specified in 10 effluent released, and of CFR Part 20, Appendix B, the average dilution flow for unrestricted areas, and length of time over which each discharge occurred.
2.
The release rate of radio-2.
Prior to relemaa of each batch active liquid effluents, of liquid effluent, a sample excluding tritium and noble shall be taken from that batch gases, shall not exceed 10 and analyzed for gross radio-curies during any calendar
' activity (B, 3' ) and a complete quarter, without specific gmama spectrum analysis to dem-approval from the Commission.
onstrate compliance with 3.8.A.
using the circulating water flow rate at the time of discharge.
Amendment No.
42 l
177
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.A Liquid Effluents (Conc'd) 4.8.A Liquid Effluents (Cont'd) 3.
3.
The annual average concentra-tion of tritium prior to dilution in a natural body of water shall not exceed 1 x 10-5 uci/cc.
4.
A monthly proportional composite 4
During release of radioactive vastes, the following condi-liquid waste sample, including an aliquot of each batch released tions shall be met:
during the month, shall be a.
The minimum dilution water analyzed for critium, Sr-89, required to satisfy 3.8.A.1 Sr-90, and gross alpha radio-shall be met.
activity.
b.
The gross activity monitor and recorder on the rad-waste affluent line shall be operable.
c.
The affluent control 5.
At least one representative monitor shall be set to liquid waste batch per ionth alarm and automatically shall be analyzed for dissolved close the waste discharge fission and activated gases, valve prior to exceeding the limits specified in
- 3. 8.A.1 above.
l 6.
d.
Liquid waste activity and l
flow rate shall be contin-i uously monitored and recorded during release.
5.
The equipment installed in the 7.
The liquid effluent radiation liquid radioactive waste system shall be maintained and shall monitor shall be calibrated at least quarterly by means of be operated to process, as a min-a check source and annually inum, all liquids prior to their with a known radioactive source.
discharge when the activity re-Each monitor, as described, shall leased during any calandar quar-also have an instrument channel ter exceeds 1.25 curies.
test monthly and a sensor check 6.
When the release rate of radio-daily.
active liquid effluents, 8.
The status and performance of excluding critium and noble automatic isolation valves and gases, exceed 2.5 curies during discharge tank selection valves any calendar quarter, the l
and results of independent licensee shall notify the Director, Directorate of liquid waste samples shall be l
Licensing within 30 days iden-checked and logged.
tifying the causes and describ-ing the proposed program of action to reduce such release races.
178 l
Amendment No.
42
LIMITING CONDITIONS'FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8.B Airborne Ef fluents (Cont'd) 4.8.B Airborne Effluents (Cont'd)
When the release rate exceeds 0.05 a.
Within one month after the C1/see for a period of greater date of initial criticality than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the licensee shall of the reactor.
notify the Director, Directorate of Licensing, in writing within b.
At least monthly thereafter.
10 days, c.
Following each refueling, process change or other occurrence which could alter the mixture of radionuclides.
4 During release of gaseous 4.
The release rate of critium vastes, the following condi-in the gaseous effluents shall tions shall be met:
be determined on the basis of a representative sample collected a.
The gress activity monitor, and analyzed for tritium at the iodine activity monitor least quarterly, and particulate activity monitor shall be operable.
b.
The minimum air flow shall be maintained, c.
Isolation devices capable of limiting gaseous release rates from the main stack to g
within the values specified in 3.8.B.1 above shall be operable.
5.
One reactor building exhaust vent 5.
Samples of offgas effluents shall and one plant stack monitoring be taken at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> system shall be operable, and the and a ratio of long-lived to off-gas radiation monitors shall short-lived radioactivity deter-be operable or operating whenever mined. When these samples s team pressure is available to the air ejectors.
If these indicate a change in this ratio requirements are not satisfied, of greater than 20% from the a normal orderly shutdown shall ratio established by the previous be initiated witnin one hour, monthly isotopic analysis, a new and the reactor shall be in the isotopic analysis shall be hot shutdown condition within perfo rmed.
10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> in the case of the stack monitor or 10 days in the :sse of 6.
Facility records of iodine and the building vent monitor.
particulate releases with half lives greater than eight days 6.
The containment shall not be shall be maintained on the basis purged except through the stand-of all filter cartridges counted, by gas treatment system while These filters shall be analyzed the reactor is in the EUN mode.
for I-131 (charcoal), gross radio-activity (B,[ ) and a complete i
(
gamma spectrum (particulate).
These filters shall be analyzed Amendment No.
42 weekly when the iodine or par-ticulate release rate is less 180
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I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS I
\\
3.8.B Airborne Effluents (Cont'd) 4.8 B Airborne Effluents (Cont'd) l than the annual average release rate given in 3.8.B.2 above, other-i l
wise the cartridges shall be removed and analyzed daily until a steady release level has been estab-I lished.
7.
A weekly charcoal filter from each j
release point shall be analyzed for I-133 and I-135 at least quarterly.
l 8.
Aweekly particulate filter from l
aach release point shall be analyzed j
for gross alpha radioactivity at j
least quarterly. _ A composite of a months' filters from each release point shall be analyzed for Sr-89 and*Sr-90 at least quarterly.
9.
When the average daily gross
(
radioactivity release rate equals or exceeds that given in 3.8.B.3 or increases by 50%
over the previous day, the iodine and particulate cartridge shall be analyzed to, determine the release rata increase for iodines and particulates.
10.
All waste gas monitors shall be calibrated at least quarterly by means of a built-in check source and annually with a known radioactive source. Each monitor I
shall have an instrument channel test at least monthly and sensor check at lease daily.
- 11. At least annually, automatic initiation and closure of vaste gas system shall be verified.
(
Amendment No.
42 181
LIMITI O CONDITIONS FOR OPD ATION SURVEILLANCE REQUIRDENTS l
t 3 9 AUXILIARY ELECMICAL SYSTM
%.9 AUXILIARY ELECMICAL SYSTM Applicability:
Almlicability:
Applies to the auxiliary electrical Applies to the periodic testing re-power system.
quirements of the auxiliary electri-1 ca1.ystems.
~
Objective:
' Objective:
To assure an adequate supply of elec-Verify the operability of the trical power for operation of those
.uwm=q elcetrical system.
systems required for safety.
Specification:
Specification:
A.
Auxiliary Electrical muipment A.
Auxiliary Electrical muipment Surveillance The reactor shall not be made critical 1.
Diesel Generators
)
unless all of the following conditions are satisfied:
a.
Each diesel generator shall be manually
)
started and loaded once each month to
)
At least one offsite' transmission line demonstrate operational readiness. The and the startup transformer are avail-test shall continue for at least a able and capable "of automatically one hour period at rated load.
supplying auxiliary power to the emer-gency buses.
During the monthly generator test the diesel generator starting air com1res-2.
An additional source of offsite power sor shall be checked for operation and consisting of one of the following:
its ability to recharge air receivers.
The operation of the diesel fuel oil a.
A transmission line and shutdown trans-transfer pumps shall be demonstrated, former capable of supplying power to and the diesel starting time to reach the emergency 4160 volt buses, rated voltage and frequency shall be loEged.
b.
The main transformer and unit auxiliary transformer available and capable of Also, once per operating cycle the diesel supplying power to the emergency 4160 generator shall be manually started and volt buses.
loaded from the Alternate Shutdown Statiai 3
Both diesel generators shall be oper-able. Each diesel generator shall have
- b. Once per operating cycle the condition under which the diesel generator is a minimum of 19,800 gallons of diesel required will be simulated and a test fuel on site.
conducted to demonstrate that it will start and accept the emergency load within the specified time sequence.
The results shall be logged.
Amendment No. 42 194 e
iW.
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L'IMITING CONDITIONS FOR OPgRATION SURVEILLANCE REQUIREMENTS 1.
Verifying de-energization of the emergency buses and load shedding from the emergency buses.
2.
Verifying the diesel starts from ambient condition on the auto-start signal energizes the emergency buses with permanently connected loads, energizes the auto-connected emergency loads through the load sequence and operates for i 5 minutes while its generator is loaded with the emergency loads.
The results shall be logged.
C.
Once per operating cycle with the diesel loaded per 4.9.A.1.b verify that on diesel generator trip sec-ondary (off-site) a-c power is auto-natically connected to the emergency service buses and emergency loads are energized through the load sequencer in the same manner as described in 4.9.A.1.b.1.
The results shall be logged.
2.
Secondary,Off-Site Power A.
A test will be performed once per operating cycle to verify that the shutdown transformer breakers will close on to the safety related buses within 12 to 14 seconds.
l Amendment No. 42 194A
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9.A Auxiliary Electrical Equipment 4.9.A Auxiliary Electrical Equipment Surveillance 4.
4160 volt buses A5 and A6 are ener-c.
Once a month the quantity of diesel gized and the associated 480 volt fuel available shall be logged, buses are energized.
d.
Once a month a sample of diesel 5.
The station and switchyard 125 and fuel shall be checked for quality 250 volt batteries are operable.
in accordance with ASTM D270-1975.
Each battery shall have an operable The quality shall be within the battery charger.
acceptable limits specified in Table 1 of ASTM D975-77 and logged.
6.
Emergency Bus under Voltage Annunciation System is operable.
2.
Station and Switchyard Batteries Emergency buses A5 & A6 shall not
- a. Every week the specific gravity, be operated below 3745 during the voltage and temperature of the normal operation.
pilot cell and overall battery voltage shall be measured and logged.
- b. Every three months the measurements shall be made of voltage of each cell to nearest 0.1 volt, specific gravity of each cell, and temperature of every fifth cell. These measurements shall be logged.
- c. Once each operating cycle, the stated batteries shall be subjected to a rated load discharge test.
The spe-cific gravity and voltage of each cell shall be determined after the discharge and logged.
B.
_ Operation with Inoperable Equipment 3.
Emergency Under Voltage Annunciation System Whenever the reactor is in Run Mode or startup M de with the reactor not in a
- a. Once each operating cycle, calibrate o
Cold condition, the availability of the alarn sensor, electric power shall be as specified in 3.9.B.1, 3.9.B.2, 3.9.B.3, 3.9.B.4
- b. Once each 31 days perform a channel and 3.9.B.5.
functional test on the alarm system.
1.
From and after the date that incoming
- c. In the event the alarm system is power is not availble from the start-determined inoperable under 3.b above, up or shutdown transformer, continued commence logging safety related bus voltage every 30 minutes until such time as the alarm is restored to operable status.
1 i
Amendment No. 42 195
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9.B Operation with Inoperable Equipar;nt l
l following conditions are satisfied and the AEC is notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the occurrence and the plans for restoration of the inoperable components:
a.
The startup transormer and both offsite 345 kV transmission lines are available and capable of automati-cally supp1ying a,uxiliary power to the emergency 4160 volt buses, b.
A transmission line and associated shutdown transformer are available
)
and capable of automatically sup-l plying auxiliary power to the emergency 4160 volt buses, j
5.
From and after the date that one of the 125 or 250 volt battery systems is made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding three days within electrical safety considerations, provided repair work is initiated in the most expeditious manner to return the failed component to an operable state, and Specification 3.5.F is satisfied. The AEC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the situation, the precautions to be taken during this period and the plans to return the failed component to an operable state.
6.
With energency bus voltage below 3745 during normal operation, transfer the safety related buses to the diesel generators and be in at least Hot shutdown within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and in Cold shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Amendment No. 42 197
l RASES (Continued) 3.9 can be used for either 125 volt battery, (2) a 250 volt d-c back-up battery charger is supplied. Thus, on loss of normal battery charger, the back-up charger can be used. The 125 volt battery system shall have a minimum of 105 volts at the battery terminals to be considered operable. The 250 volt battery system shall have a minimum of 210 volts at the battary terminals to be considered operable.
l Automatic second level undervoltage protection is inatalled on the startup transformer and is available when safety related loads are being supplied from this source. During normal operation, the unit auxiliary transformer supplies i
safety related buses. Automatic second level undervoltage protection is not installed on the unit auxiliary transformer. Safety related loads are in use during normal operation and thus are not provided with automatic second level undervoltage protection. BECo is conducting new grid studies with the intent of providing automatic second level undervoltage protection.* We expect these studies to be completed and any necessary modifications to be installed prior to startup from reload 5.
During Cycle 5, the Safety Bus Degraded Voltage Alarm System will be relied upon in conjunction with operator action, to pre-clude operation with a degraded bus voltage condition.
l
- for the unit auxiliary transformer and the startup transformer.
Amendment No. 42 199
D
'w B ASES:
4.9 s_.
The monthly test of the diesel generator is conducted to check for equipment failures and deterioration. Testing is conducted up to equilibrium operating conditions to demonstrate proper operation at these conditions.
The diesel generator will be manually started, connected to the bus and load picked up.
The diesel generator should be loaded to a least 75% of rated load to pre-vent fouling of the engine. It is expected that the diesel generator will be run for one to two hours. Diesel generator experience at other generating stations indicates that the testing frequency is adequate and provides a high reliability of operation should the system be required.
Each diesel generator has one air compressor and two air receivers for start-ing, one air compressor and three receiver tanks for turbo-charger assist in starting and loading. It is expected that the air compressors will run only infrequently. During the monthly check of the diesel generator, one receiver in each set of receivers will be drawn down below the point at which the cor-responding compressor automatically starts to check operation and the ability of the compressors to recharge the receivers.
The diesel generator fuel consumption rate at full load is approximately 193 gallons per hour. Thus, the monthly load test of the diesel generators will test the operation and ability of the fuel oil transfer pumps to refill the day tank and will eneck the operation of these pumps from the emergency source.
g j
The test of the diesel generator during the refueling outaga will be more com-prehensive in that it sill functionally test the system; 1.e.,
it will check diesel generator startrig and closure of diesel generator breaker and sequenc-ing of load on the diesel generator. The diesel generator will be started by simulation of a loss-of-coolant accident.
In conjunction with this an under-voltage condition will be imposed to simulate a loss of off-site power. The timing sequence will be checked to assure that the diesel generators can operate the core spray pumps at rated power within thirty seconds and the LPCI pumps at rated power within forty-three seconds. Additionally, with the Diesel Generator operating as described above the capability of supplying power the the Emergency Bus will be further substantiated. This will be accomplished by tripping the Die'sel Generator Breaker and verifying that secondary offsite power is connected to the Emergency Bus and emergency loads are energized through the load sequencer.
Periodic tests between refueling outages verify the ability of the diesel generator to run at full load and the core and containment cooling pumps to deliver full flow.
Periodic testing of the various components, plus a functional test once-a-cycle, is sufficient to maintain adequate reliability.
Although station batteries will deteriorate with time, utility experience in-dicates there is almost no possibility of precipitous failure. The type of surveillance described in this specification is that which has been demon-strated over the years to provide an indication of a cell becoming irregular or unserviceable long before it becomes a failure. In addition, the checks described also provide adequate indication that the batteries have the speci-fied ampere hour capability.
Amendment No. 42 200
1 SURVIILLANCE REQUIRDtENTS LIMITDIG CONDITIONS FOR OPERATION 4.11 REACTOR YUIL ASSDtBLY 3.11 REACTOR FUEL ASSEMBLT Applicability Applicability l
The Limiting Conditions for Operation The surveillance Requirements associatsd with the fuel rods apply apply to the parameters which to those parameters which monitor the the fuel rod operating condi-
)
tions.
fuel rod operating conditions.
h tive Obiective The Objective ci the Limiting Condi-The Objective of the Surveil-tions for Opr.ation is to assure the lance Requirements is to performance of the fuel rods.
specify the type and frequency of surveillance to be applied to the fuel rods.
1 Specifications Specif'estions A.
Aveuse Planar Linear Best A.
Average Planar Linear Best lianeration Kate (APIRGR)
I
_E eration Rate (AFLHGR)
The AP13GR for each type of During power operation with both fuel as a function of average i
recirculation pumps operating, the planar arposure shall be i
APLEGR for each type of fuel as a determined daily during function of average planar exposure reactor operation at 2 25%
shall not exceed the applicable rated thermal power.
limiting value shown in Figures 3.11-1 through 3.11-5.
The top curves are applicable for core flow greater than or equal to 90% of rated core flow. When core flow is less than 90% of rated core flow, the lower curves shall be limiting. If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated o. thin 15 minutes to re-store operation to within the prescribed limits.
If the APLHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition with-in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is with'.n the prescribed limits.
Amendment No. 42 205A
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=
LIMITING CONDIT10ttS FOR OPERATION SURVEILLANCE REQUIRDENTS 3.
Linear Beat Generation Rate (LEGR) 3.
Linear Heat Generation Rate (LHCR)
During reactor power operation The LEGR as a function of core I
the linear heat generation rate (LEGR) height shall be checked daily of any rod in any fuel assembly at during reactor operation at 225%
ar.y arial location shall not exceed rated thermal power.
13.41ar/f t for 8x8 and P8x8R fuel.
If at any tima during operation it is determined by normal surveillance that the limiting value for LEGE is being exceeded, action shall be initiated within 15 minutes tu restorn operation to within the prescribed limits.
If the LEGR is not returned to within the prescribed limits within two (2) bours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
i Surveillanca and corresponding action shall continue until reactor operation is within the prescribed limits.
Amendment No. 42 205A-1
l
~.
j I
SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATIOR l
Minissa Critical Power Ratio (MCFR)
C Minimum Critical Power Ratio (MCFR1 C.
MCFR shall be determined daily During power operation MCFR shall be2 during reactor power operation at If at 1.35 for 8x8 and F8x8R fuel.
> 25% rated thermal power and any tima during operation it is deter-following any change in power mined by normal surveillance that the level or distribution that would limiting value for MCFR is being ex-cause operation with a limiting coeded, action shall be initiated control rod pattern as described within 15 minutes to restore operation in the bases for Specification If to within the prescribed limits.
3.3.3.5.
the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and j
corresponding action shall continue until reactor operation is within the prescribed limits.
For core iknee other than rated the MCPR shall be 2 1.35 for 8x8 and l
P8x8Rfuel times Kg, where Kg is as shown in Figure 3.11-8.
As an alternative method providing equivalent thermal-hydraulic protec-tion at core flows other than rated, the calculated MCPR may be divided by Kg, where Kg is as shown in Figure 3.11-8.
l Amendment No.
42 205B l
f l
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.EAEE8-3.11A Averene Planar Linear Best Generation Rate (AFuGQ_
This specifications assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50. Appendix K.
De peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat geniaration rate of all the rods of a fuel, assembly at any axial location and is only dependent, secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a UlGR for the highest powered rod which is equal to or less than the design' UlGR.
This LEGR times 1.02 is used in the heat-up code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factors. The limiting value for APulGR is this LEGR of the highest powered rod divided by its local peaking factor.
De calculational procedure used to establish the APulGR limit for each fuel type is based on a less-of-coolant accident analysis.
l The emergency core cooling system (ECCS) evaluation models which l
are employed to determine the effects of the loss of coolant accident (LOCA) in accordance with 10CFR50 and Appendix K are discussed in Reference 1.
The models are identified as MMB, SCAT, SAFE, REFLOOD, and SASTE. 'The MMB Code calculates the short term bicwdown response and core flow, which are input into the SCAT code to calculate blowdown beat transfer coefficients.
The SAFE code is used to determine longer term system response and flows from the various ECC systems. libere appropriate, the output of SAFE is used in the REFLOOD code to calculate liquid The results of these codes are used in the CBASTE code levels.
to calculate fuel clad temperatures and W== average planar linear heat generation rates OtArulGR) for each fuel type.
The significant plant input parameters are given in Reference 2.
MAPLHGR's for the present fuel types were calculated by the above The curves in Figures procedure and are included in Reference 3.3.11-1 through 3.11-5 These multipliers Reference 3 by factors given in Reference 4.
were developed assuming no core spray heat transfer credit in the LOCA analysis.
Amendment No. 42 205C
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Amendment No. 42 205C-1
REFERENCES 1.
General Electric BWR Generic Reload Fuel Application, NEDE-24011-P.
2.
Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDO-21696, August 1977.
3.
" Supplemental Itaload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1 Reload 4", NEDO-24224 November 1979.
4.
" Supplement 1 to Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1 Reload 4" NEDO-24224-1 March 1980.
l I
i Amendment No.
42 205C-2
N3N.5 3.11C MINDalH CRITTM POWER RATIO 00CFR)
Operatian Limit McFR por any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that tha resulting MCFR does not decrease below the safety Limit MCFR at any time during the transient assuming instrument trip setting given in Specification 2.1.
The required operating limit McFR at steady state conditions in Specification 3.11.C was chosen conservatively at a value higher than McFR's of past analysis with the objective of establishing an operating limit McFR which is fuel type and cycle independant.
The differenes between the specified Operating Limit MCFR in Specification 3.11C and the safety Limit MCFR in Specification 1.3A defines the largest reduction in critical power ratic (CFR) permitted during any anticipated abnormal operating transient.
To ensure that this reduction is not exceeded, the most limiting transients are analised for each reload and fuel type (8x8 and P8x8R) to determine that transient which yields the largest value l
ef A CFL This value, when added en the safsty Limit MCPR unst be less than the min 4== operating limit MCFR's of Specification 3.11.C.
The result of this evaluation is documented in the
" Supplemental Esload Licensing submittal" for the current reload.
The evaluation of a given transient begins with the system gt parameters shown in Tables 5-4, 5-6 and 5-8 of MIDg-24011-P Supplemented by reload unique inputs given in the current Supplemental Esload Licensing Submittal. These values are input to a GE core dynamic behavior transient computer program described in NEDO-10802(2). Also, the void reactivity coefficients that were input to the transiong..t,.dational procedure are based on a new uthod of calculation termed MEV which provides a better agreement between the calculated so4 plant instrumes. power distributions. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting handle with the single chan t transient thermal hydraulic SCAT code described in NEDE-20566(3).
The principal result of this evaluation is the reduction in EPR caused by the transient.
l Amendment No. 42 205C-3 e
O eg
- t.-
eg gspe wee eee st-s e ese
- 4-m.
meo m
f Two codes are used to analyse the rod withdrawal error transient, The first code simulates the three dimensional BWR core melaar and thermal-hydraulic characteristics. Uefag this code a limiting control rod pattern is determined; the following assumptions are included in this determination:
(1) The core ia operating at hil power in the zenon-free condition.
(2) The highest worth control rod is sseumed to be fully inserted.
(3) The analysis is performed for the most reactive point in the cycle.
(4) The control rods are assumed to be the worst possible pattern without exceeding thermal limits.
(5) A bundle in the vicinity of the highest worth control rod is assumed to be operating at the.xutimum allowable linear heat generation rate.
(5) A bundle in the vicinity of the highest worth control M is assumed to be operating,the mini== allowable critical power ratio.
The three-dimensional BWR code then sim lates the core s aponse to a
the control rod withdrawal error. The second code calculates the Rod Block Monitor response to the rod withdrawal error. This code simulates the Rod Block Monitor under selected failure conditions (LPRM) for the core rod use (calculated by the 3-dimensional BWR simulation code) for the control rod withdrawal.
/.mendment No.
42 205C-4 i
The analysis of the rod withdrawal error for Pilgt*n Unit 1 considers the continuous withdrawal of the==w4==
worth control rod at its===i== drive speed from the reactor. A summary of the analytical methods used to determine the nuclear characteristics is given in Section 5.2.1.5 of NEDE.-240u-P.
MCFR LIMITS FOR CORE FIDWS OTHER THAN RATED The purpose of the Kg factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCPR is the product of the operating limit EPR and the Kg factor. Specifically, the Kg factor provides the required thermal margin to protect against a flow in-crease transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speed up caused by a actor-generator speed control failure.
For operation in the automatic flow control mode, the E, factors assure that the operating limit NCPR given in Specification 3. u C will not be violated should the most limiting transient occur at less than rated flow. In the manual flow control mode, the Kg factors assure that the Safety Limit MCPR win not be violated for the sama postulated transient event.
f actor curves shown in Figure 3.11-8(4) wre developM The Kf generica n y which are applicable to all BWR/2, BWR/3, and BWR/4 The Kg f actors were derived using the flow control reactors.
line corresponding to rated thermal power at rated core flow.
For the manual flow control mode, the Kg factors were calculated such that at the maximum flow state (as limited by the ' pump scoop tube set point) and the corresponding core power (altmg the raced flow control line), the limiting bundle's relative power was adjusted t:ntil tne MCPR was slightly above the Safety Limit.
Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different core flove. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR determines the Eg.
For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was establi N d such that the MCPR was equal to the operating limit MCPR at rated power and flow.
l Amendment No. 42 205C-5
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e
+ s I4) are conservative for the The K, factors shown in Figure 3.11-S Pilgrim Unit 1 operation because the operating limit MCPR given in Specification 3.11C is greater than the original 1.20 operating limit MCPR used for the generic derivation ef Eg.
MINDEM CRITICAL PWER RATIO WCPR) - NMG WN 6.11.C At core thermal power levels less than or equal to 25%, the reactor will be operating at =in1== recirculation pump speed For all and the moderator void content will be very small.
designated control rod patterns which may be employed at this point, operating plant asperience indicated that the resulting MCPR value is in excess of requirements by a considerableW increase would only place operation in a more conservative mo margin.
j a MCPR evaluation will be ande at 25% thermal power level with relative to MCPR.
The MCPR margin will thus sini== recirculation pump speed.be demonstrated such that future MC The daily re-power level will be shown to be unnecessary.
quirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been.significant power or control rod changes.
The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCFR will be known j
following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.
)
Amendment No. 42 205C-6 n -,
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REFERENCES General nectric BUR Generic Raioed Fuel Application, MEDs-24011-F.
1.
R. 3. Linford, Analytical Nothods of Plant Transient Evaluations 2.
for the GE EWR, Febnary 1973 (NEDO-10802).
General Electric Company Analytical Model for lose-of-Coolant 3.
Analysis in Accordance with 10 CFE 50, Appendia K. NELE-20566 (Draft) AuInst 1974.
Letter from J. E. Mouard, Booten Edison Company to D. L. Ziemana 4.
UEMEC, dated October 31, 1975.
l Amendment No. 42 205D j
(DELETED)
Anendnent No. 42 205E
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F IGURE 3.11-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE VERSOS PLANAR AVERAGE EXPOSURE FUEL TYPE BD B 21s t 7
6
$g m
DC F-COR.E FLOW 44 m
(E 13.4 3 907c itATED aL9 $1 11.1 W$
'M Io.t w
2 10.5 w w l *i
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>d 10 1o.8 96 p
24 gg 3W sn 2Z S.G gg CoRG Flow < 9o7. RATEIl 44 2s s
a 8.1 6
0 5,000 1Q000 1!!p00 20,000 25.000 30,000
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PLANAR AVERAGE EXPOSURE (MWft)
Amendment No. 42 205E-1
F IGURE 3.11-E MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE VERS 05 PLAN AR AVERAGE EXPOSURE FUEL TYPE BDB 21'3 H 7
0 11 CCRE FLOW
[
>f 90% RA7CD gq<
m 2E
'll.4 3Z 88 3 11 3 11
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,0.5 80*5 I#
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s s
8.7 8
o 5,000 1Qooo 14000 20,000 25,000 30,000
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PLANAR AVERAGE EXPOSURE (MWgt)
Amendment No. 42 205E-2
F IGURE 3.11-3 l
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE VERSOS PLANAR AVERAGE EXPOSURE
__ _.F_LI E L.T Y P E..BD B 2 s t ___
7 O
CORE Flow D907o RATED jg, W
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44 r
ZE
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CORE Flow 4 90'7 RATEI) e' m
44 9.5 1
2mz 9
9 2.
3 6
0 5,000 1Q000 14000 20,000 25.000 30,000
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PLANAR AVERAGE EXPOSURE (MWft)
Amendment No. 42 205E-3
F IGURE 3.11-4 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE VERSOS PLANAR AVERAGE EXPOSURE
-. FUEL TYPE PeIiRB 2G S L 7
O 11 m
$k CORE Flow ><9c% RATED ze 3z
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to.e m z
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B O
5,000 1QOOO 1!!p00 20000 25,000 30000
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PLANAR AVERAGE EXPOSURE (MWft)
Amendment No. 42 205E-4 I
F IG UR E 3.11-f MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE VERSD5 PLANAR AVERAGE EXPOSURE FUEL TYPE P8DRB282 7
6 11 m
E l-4 CORE Flow & "Jo7, KATED yg
$z t!.o I. I w>
lo S 10.9 Og 4m Ez C
to.4 4
10.1 ro.3 10 sc.z ge z e
3y io.o 1
96 Co R6 Flow 4. C o7e *R ATit J
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- 9. 4 f4z 9
3 a
O 5f)OO 1QOOO 14000 20,000 24000 3g000
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PLANAR AVERAGE EXPOSURE (MWgt)
Amendment No. 42 205E-5
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l Amenament No. 42 205F i
1
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.y 5.0 MMOR DESIGN FEATURES 5.1 SITE FEATURES Filgria Nuclear Power Station is located on the Western Shore of Cape Cod Bay in the Youn of Plymouth, Flymouth County, Massachu-The site is located at approximately 41'51' north latitude setts.
and 70835' west lonsitude on the Manoner quadrangle, Massachusetts, Flymouth County 7.5 Minute Serias (topographie) amp issued by U.S.
Geological Survey. UTM coordinates are 19-46446N-3692E.
The reactor (center line) is located approximately 1800 feet from the nearest property h===dacy.
5.2 REACTOR A.
The core shall consist of not more than 580 fuel assemblies of 8x8 153 fuel rods) and F8x81 (62 fuel rods).
B.
The reactor core shall contain 145 cruciform-shaped control rods. The control asterial shall be boron carbide powder (3 C) compacted to approximately 70% of theoretical density.
4 5.3 RRACTOR VESSEL The reactor vessel shall be as described in Table 4.2.1 of the FSAR.
The applicable design codes shall be as described in Table 4.2.1 of i
the FSAR.
l 5.4 ColtrAIlemlrr 1
A.
The principal design parameters for the primary contat.nment shall be as given in Table 5.2.1 of the FSAR. The applicable design codes shall be as described in Section 12.2.2.8 of the FSAR.
B.
The secondary contsizusent shall be as described fa Section 5.3.2 of the FSAR.
C.
Fenetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in 5xtion 5.2.3.4 of the FSAR.
l Amendment No.
42 l
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