ML20094A150
| ML20094A150 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/11/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Tennessee Valley Authority |
| Shared Package | |
| ML20094A159 | List: |
| References | |
| DPR-68-A-070 NUDOCS 8408030125 | |
| Download: ML20094A150 (23) | |
Text
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION i
E WASHINGTON, D. C. 20555 x...../
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, llNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. DPR-68 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated January 23, 1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; p-C.
There is reasonable assu'rance (i) that the activities authorized by this amendment can be conducted without endarigering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission!s regulations; 1
l D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of th~e Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-68 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 70, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
8408030125 840711 PDR ADOCK 05000296 P
PDR I
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2-3.
This license amendment is affective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Spec *ifications Date of Issuance: July 11, 1984
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ATTACHMENT TO LICENSE AMENDMENT NO. 70 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET N0. 50-296 Revise Appendix A as follows:
1.
Remove the following pages and replace with the identically numbered vii 165 viii 176 9
178 12 18Eb 18 19 20 21 22 23 24 28 32
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36 2
The marginal lines on each page indicate the revised area.
3.
Add the following new pages:
35A 182c
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- 4. 2 E Minimum Test and Calitraticn Trequency f or Drywell Leak Detection Instrumentation ion
- 4. 2. r Minimum Test and Calibration Frequency for surveillance Instrumentation 102 4.2.G surveillance Ptquire ents for Control Room Isolation Instrumentation 10J 4.2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation 10e 4.2.J seismic Monitoring Instrument surveillance Requirements 105 4.6.A Reactor Coolant system Inservice Inspection 209 3.$ I M bid
. Average Planar Exposure 181, 182 s
182a, 182b 3.7.A Primary Containment Isolation Valves 262
, 3. 7. 3 Testable Penetrations with Double 0-Ring Seals 268 3.7.C Testable Penetrations' with Testable Bellows 269
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3.7.D Primary Containment Testable Isolation valves 3.7. E suppression Chamber Influent Lines stop-Check Globe Valve Leakage Rates 279 3.7.7 Check valves on suppression Chamber Influent Lines 280 3.7.G Check Valves on Drywell Influent Lines 281 3.7.H Testable Electrical Penetrations 2'83
- 4. 8. A Radioactive Liquid Waste sampling.and Analysis 310 4.8.8 Radioactive Gaseous Waste sampling and Analysis 311 6.3.A Protection Factors for Respirators 373 6.8.A Minimum shif t Crew Requirements 390 9
vii kendment No.
70 1
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e LIST OF ILLUSTRATIONS f.lguJ.s Tim M
2.1-1 APRM Flow Reference Scram and APRM Rod Block settin~gs 14 2.1-2 APRM Flow Bias Scram delationship to Normal Operating Conditions 25 4.1-1 Graphic Aid in the -Selection of an Adequate Interval Between Tecto 48 4.2-1 System Unavailability 117 3.4-1 Sodium Pentaborate Solution volume Concentrati.'n Requirements 141
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'3. e-2 Sodium Pentaborate Solution Temperature Requirements 142 3.5.K-1 MCPR Limits 182c
- 3. 5. 2 Kp Factor vs. Percent Core Flow 183
- 3. 6-1 Temperature-Pressure Limitationo 207
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3,6-2 Change in charpy V Temperature vs.
Neutron Exposure 208 6.1-1 TVA Of fice of Power Organization for Operation of Nuclear Power Plants 391 6.1-2 Functional organization 392 6.2-1 Review end Audit Function 393 6.3-1 In-Plant Fire Program Organization 394
'viii M
AmendmentNo.[,.HI,70.
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e SAFETY LIMIT LIMITING SAFETY SYSTEM EETTING 1.1 FUEL CLADDING I'NTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability Applicability Applies to the interrelated Applies to trip settings of the variables associated with fuel instruments and devices which thermal behavior, are provided to prevent the reactor system safety limits from being exceeded.
obi etive obiective To define the level of the To establish limits which process variab)as at which ensure the integrity of the automatic protective action is fuel cladding.
initiated to prevent the fuel cladding integrity safety limit from being exceeded.
Specification Speci fica tion s
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The limiting safety system A.
Thermal Power Limits settings shall be as specified below:
- 1. Reactor Pressure > 800 psia and Core Flow > 105 A. Neutron Flux Trip Settings of Rated.
1.
APRM Flux Scram Trip When the reactor pressure Setting (Run Mode) is greater than 800 psia, (Flow Biased) the existence of a minimum a.
When the Mode Switch critical power ratio is in the RUN (MCPR) less than 1.07 position, the APRM shall constitute violation flux scram trip of the fuel cladding setting shall be:
integrity safety limit.
SS (0.66W + 545) where:
S=
Setting in per-Cent of rated thermal power a
(3293 MWt) i i
9 Amendment NO. M % 70 i
hAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1 FUEL CLADDING INTEGRITY
- e. Fixcd High Neutron Flux Sera =
Trip Setting - When the mode switch is in the RUN position, the APRM fixed high finx scram trip setting shall be:
5:,120% power
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2.
Reactor Pressure jL 800 PSIA or Core 2.
APRM and IRM Trip Settings Flow jL 10% of Rated (Startup and Hot Standby Modes).
When the reactor pressure is j:,800.
- a. APRM - When the reactor mode PSIA or core flow is { 10%'of rated, 4
the core thermal power shall not switch is in the STARTUP position, exceed 823 MWt (-25% of rated thermal the APRM scram shall be set.at power).
less than or equal to 15% of rated power.
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l AmendmentNo.[
70 12 l
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2.1
.BAggg:
LIMIIING SAFETY SYSTEM S ETTINGS RELATED TO FUEL
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CLAQDING INTEGRITY The ab.7ormal operational transients app'licable to operation of the - Browns Ferry Nuclear Plant have been analyzed throughout the spectrum of planned operating condit.;ons up to the design thermal power condition of 3440 MWt.
The s.viyses were based upon plant operation in accordance with the operating map given in Figure s'
3.7-1 of the FSAR.
In addition, 3293 MWt is the licensed maximum power. level of. Browns Ferry Nuclear Plant, and this. represents the maximum. steady-state power which shall not knowingly be exceeded.
Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coef ficient, control rod scram worth, scram delay time, peaking f a c, tors, and axial power shapes.. These f actors are selected conservatively with respect to their effect on the applicalbe transient results as determined by the current analysis model.
-This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic performance.
Results obtained from a General Electric boiling water reactor have been compared with predictions made by the model.
The comparisions and results are summarized in Ref erence 1.
1 The void reactivity coefficierit and the scram worth are described in detail in reference 1.
The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical specifications as further.
i described in Reference 1.The effect of scram worth, scram delay t*,e and l
l rod insertion rate all conservatively applied, are of greatest significance in the early portion of the negative reactivity inse rtion '
. The rapid insertion of negative reactivity is assured by the time requirements for.51 and 20% ' insertion.
By the time the rods are 60% inserted, approximately four dollars of negative reactivity has been inserted which strongly turns the transient, and accomplishes the desired ef fect.
The times for 50% and 901 j
incertion are given to assure proper completion of the expected
)
perf ormance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.
For analyces of the thermal consequences of the transients a MCPR of ***
is conservatively assumed to exist prior to initiation of the transients.
This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.
- see section 3.5.K.
x 18 Amendment No.
70
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2.1 BASES In summary 1.
The licensed maximum power level is 3,293 MWt.
2.
Analyses of transients employ adequately conservative values of the controlling reactor parameters.
3 The abnormal operational transients were analyzed to a power level of 3440 MWt.
4.
The analytical procedures now used result in a more logical answer than the alternative methed of assuming a higher starting power in conjunction with the expected values for the parameters.
The bases fur individual set points are discussed below:
A.
Neutron Flux Scram 1.
APRM Flow-Biased High Flux Scram Trip Setting (Run Mode)
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated power (3293 MWt).
Because fission chambers provide the basic input signals, the APRM system responds directly to core average neutron flux.
During transients, the instantaneous fuel surface heat flux is l
less than the instantaneous neutron flux by an amount depending upon the duretion of the transient and the fuel time cons tant. For this reason, the flow-biased scram APRM flux signal is passed through a filtering network with a time constant which is representative of the fuel time constant.
As a result of this filtering, APRM flow-biased scram will occur only if the neutron flux signal is in excess of the i
setpoint and of sufficient time duration to overcome the fuel time constant and result in an average fuel surface heat flux which is equivalent to the neutron flux trip setpoint. This setpoint is variable up to 120% of rated power based on
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recirculation drive flow according to the equations given in section 2.1.A.1 and the graph in figure 2.1.2.
For the purpose of licensing transient analysis, neutron flux scram is assumed to occur at 1205 of rated power. Therefore, the flow 1
biased provides additional margin to the thermal limits for slow transients such as loss of feedwater heating. No safe.ty credit is taken for flow-biased scrams.
19 Amendment No.
, 70
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r s
e The scram trip setting must be adjusted to ensure that the LHCR transient peak it
'.ot combination of CHTLED andTRT.
increased for any scram setting is adjusted in accorcance with the f ormula '
The An Speci fication 2.1.A.1, when the cMr;p; exceeds TFT.
I Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > 107 transient is initiated from MCPR > ***
when the 2.
APRM riur Scram Trie settir.o Standy model_
tref uel or Start & Rot at low pressure,For operation in the startup mode wr.ile the reacto the APRM scram setting of 15 percent of r is setpoint and the safety limit, rated power provides adequate 25 percent of rated. The associated with power plantmargin is adequate to accomodate a increasing pressure at' sero or Jow startup.
Effects of minor, cold water from sources ava:lacle durznq startup void content are is net much colder than that temperature coefficients are small, and controlalready in the system, pat terns are constrained to be unif orm by operating rod procedures DacAed up by the red worth minimiter and t Rod sequence control system.
he is very low in a uniform rod pattern. Worth of individual rods possible sources of reactivity input, Thus, all of rod withdrawal is the most unif orm control Because the flux distribution associatedprobable caus power rise.
with uniform rod withdrawals does not peaks, and because several rods must involve high local power by a cignificant be moved to change rate of power rise is very slow. percentage of rated power, the flux is in near equilibrium with the fission rateGenerally, the heat an assumed uniform rod withdrawal aperoach In of rated power per minute. level, the rate of power rise is no m to the scram cent could exceed the safety limtt.more enan adequate to assure ower scram remains active until the mode switch is placed i The 15 percent APRM the RUN po siti on.
j Dressure is creater than 800 psig.This swrtch occurs when reactor n
1 3.
3.B.5,. I.}u_2,,,S,er ar Ir_ip a
Set ti na
' The 2RM System consists of 8 chamberg, reactor. protection system logic channels. in each of the e
The IRM is a See Section 3.5.K.
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20 Amendment i o.
,,Fs, 70 1
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1 5-decade instrument which covers the range of power level between tha covered by the SRM and the APRM.
by means of a range switch, and the 5 decades are broken do ranges, each being one-half of a decade in size.
of 120 divisions is active in each range of the IRM.The IRM scram setting the instrument was For example, if for that range; likewise, if the instrument was on range 5on rang setting would be 120 divisions on that range.
the scram Thus, as the IRM is ranged up to accommodate the increase in power level, the scram is also ranged up.
A scram at 120 divisions on the IRM instruments remains in effect as long as the reactor is in the startup mode
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APRM 15-percent scram will prevent higher power operation witho t b i The in the run mode.
occur both locally and over the entire core.The IRM scram provides prot u
e ng The most significant control rod withdrawal. sources of reactivity change during the power increase ar of change of power is slow enough, due to the physical limit withdrawing control rods, of neutron flux and an IRM scram would result in a reactor shut
,before any safety limit is exceeded.
well This analysis included starting the accident at va yzed.
The most severe case involves an initial condition in which t r levels.
is just suberitical and the IRM system is e reactor condition exists at quarter rod density.
not yet on scale. This in paragraph 7.5.5.4 of the FSAR.
Quarter rod density is illustrated this analysis by assuming that Additional conservatism was taken in the IRM channel closest red is bypassed.
The results of this analysis show thatto the withdrawn maintaining'MCPR above 1.07. scrammed and peak power limited to one perc
>-~
, thus provides protection against Based on the above analysis, tla IRM continuous withdrawal of control rods in sequence. local control rod with 4.
Fixed High Neutron Flux Scram Trip using heat balance data taken during steady-state c a
rated percent of rated power (3293 FNt).
reads in Licensing analyses have demonstrated that with a neutron flux.
flux scram of 120% of rated power, none of the abnormal op on stantial margin from fuel damage. transients analyzed violate the erational B.
APRM Control Rod Block Reactor the recirculation flow ra te.)ower 1cvel may be varied by moving The APRM system provides a control rodcontrol ro block to prevent rod withdrawal beyond Amendnent No. J, 70 21
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a givan point at constant recirculation flow rate, and thus 5V l
to protect against the condition of a MCPR less than 1.07 Thin roel block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power. level to excess values due to control rod withdrawal.
The flow variable trip setting provides substantial margin from fuel de u.ge, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the saf ety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during the steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting.
The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM nystem.
As with the APRM scram trip se tting, the APRM. rod block trip setting is adjusted downward if the CMFLPD exceeds FRP thus
- preserving tne APRM rod block safety margin.
C.
Reactor Water Iow Level Scram and Isolation (Except Main Steamlines)
The set point for the low level scram is above the bottom of the separator skirt.
This level has been used in transient analyses dealing with coolant inventcry decrease.
The l
results reported in FSAR cubsection 14.5 show that scram and isolation of all process, lines (except main steam) at this level adequately protects the fuel and the pressure barrier,
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because MCPR is greater than 1.07 in all cases, and system
$A pressure does not reach the safety valve settings.
The scram setting is approximately 31 inches below the normal operating range and is thus adequate to avoid spurious scrams.
D.
Turbine Stop Valve Closure Scram The turbine stop valve closure trip anticipates the pressure-neutron flux and heat flux increases that would result from closure of the stop valves.
With a trip setting of 10% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal marpns are maintained even during the worst case transient that assumes the turbine bypass valves remain clused. (Reference 2).
L.
Turbine Control Valve Fast Closure or Turbine Trip Scram l
Turbine control valve fast closure or tu'rbine trip scram anticipates the pressure, neutron flux, and heat flux increase that could *.esu>t from control valve fast closure due to load rejection or control tcalve closure due to tutbine trip; each without bypass valve capability. The reactor protection system initiates a scram in Icss than 30 mill;.econds after the start of control valve f ast closure due to load rejection or control valve closure due to turbine trip. T,his scram is achieved by rapidly reducina hydratilic conern1 f
22 Amendment No.), J8', fd, 70 l
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oil pressure at the main turbine control valve actuator dise dump valves.
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This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system.
This trip setting, a nominally f,01' greater closure time and a different valve characteristic from that of the turbine stop valve. combine to produce transients very similar to that for the stop valve.
in References 1 and 2.
Relevant transient analyses are discusred This scram is bypassed when turbine steam flev is below 30% of rated, by the turbine first state pressure.
as measured F.
Main condenser Low vacuum Scram To protect the main concenser 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 closure of the turbine stop valves, low condenser vacuum initiates a The low vaccum scram set point is selected to scram.
initiate a scram before the closure of the turbine stop valves is initiated.
G. & H.
Main Steam Line Isolation on Low Pressure and Main Steam Line Isolation Scram
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The low pressure isolation of the main steam lines at psig was provided to protect against rapid reactor 650 depressurizi. tion and the resulting rapid cooldown o f the vessel.
Advantage is taken of the scram feature that when the main steam line isolation valves are closed, occurs provide f or reactor shutdown so that high power operation at to low reactor pressure dces not occur, thus providing protect 4on for the fuel cladding integrity safety limit.
Operation of the reactor at requires that the reactor mode switch be in the STARTUPpreccures lower than 85 23
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U position, where protection of the. fuel cladding int egrity saf ety limit is provided by the IRM and APRM high neutron flux scrams..Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability o; the fuel cladding integrity saf ety limit. In addition, the saolation yalve closure scram anticipates the pressure and flux transients that occur during nor. mal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve clostre, neutron flux does not increase.
I.J.&K.
Reactor low water level set point for initiation of H PCI and RC TC, closino main steam isolation valves, and startino LPCI and core spray cu.m ns These systems maintain adequate coolant inventory and provide core cooling with the objective of preventino excessive clad tempe ratu re s. The design of these systems to adequately perf orm the intended function is based on the specified low level scram set point and initiation set points.
Transient l analyses reported in Section 14 of the TSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.
L.
References "BWR Transient Anlaysis Model Utilizing $he RETRAN 1.
Program," TVA~fR81-01-A.
2.
Generic Reload Fuel Application. Licensing Topical Report NEDE 24011-P-A and Addenda.
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t Amendment No.
70 em
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1.2 bAEEE REACTOR COOLAt@ SYSTEM INTEGRITY The saf ety limits for the rqactor coolant system pressure have been selected such that they are below pressures at which it can be shown that the integrity of the system is not endangered.
enough such that no foreseeable circumstances can cause thehoweve system pressure to rise over these limits.
The pressure safety limits are arbitrarily selected to be the lowest transient overpressures allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
The design pressure. (1,250 psig) of the reactor vessel is established such that, when the 10 percent allowance (1; '
psi) allowed by the ASME Boiler and Pressure Vessel Code Section III for pressure transients is added to the design established. pressure, a transient pressure limit of 1,375 psig is Correspondingly, the design pressure (1,148 psig for suction and 1,326 psig for discharge) of the reactor recirculation system piping are such that, when the 20 percent allowance (230 and 265 psi) allowed by USAS Piping Code, for pressure. transients are added to the design pressures, Section B31.1 transient pressure limits of
~~
1,378 and 1,591 peiq are established.
power operation is established atThus, the pressure safety Aimit applicable to s%
1,375 peig (the lowest transient overpressure allowed by the pertinent codes)
Code, Section B31.1. Boiler and Pressure Vecsel Code,Section III, and USAS P ASME The current cycle's safety analysis concerning the most operational transient resulting directly in a reactor coolant system severe abnormal l
pressure increase is given in the submittal fnr the current cycle.
reinad licend nr The reactor vessel pressure code limit of the saf ety analysis report 1,375 psig given in subsection 4.2 of produced by the overpressure transient described above.is well above the pea Thus, the pressure safety limit is well above the peak pressure applicable to power operation reasonably expected overpressure transients.that can result due to within the reactor coolant system than for the reactorHighe vessel.
design which assures that,These increased design pressures create a consisten vessel does not exceed 1,375 psig, the pressures within theif the press piping cannot exceed their respective transient
. limits due to static and pump heads.
pressure 28
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Amendment No. 70
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TABLE 3.1. A
- ct R EActoR Protection SYSTEM (SCRAM) INSTdUMENTATION EEQUIREMENT 3r+
Min. No.
y of 9perable Modes in which Function Z
inst.
Must Be Operable risa nnel s Shut-Startup/Eot Per Trip y
l System til (23) Trip runction Trip level Settina do"D Pefuel 171 Standby EuD Actionf11 y
1 Mode Switch in Shutdown X
X X
X 1.A X
X X
X 1.A.
1 Manual Scram M
O IRM (16) 3 High Flux 5 120/125 Indica ted on scale X(22)
X (22)
X (5) 1.A
+
X X
(5) 1.A 3
Inoperative APpH (16)(2%)(25)
X t.A er f.3 w
2 High Flus (Flzed Trip) $ 120 percent 2
High Flus (Flow Blased) See Spec. 2.1.A.1 I
1.A or 1.8 2
High Flux i 15 percent rated power It2l) 1(17)
(15) 1.4 or 1.s 2
Inoperative (13)
I(21) 1(17)
I
- 1. A or 1.5 -
2 Dcunseale 3,3 indicated on scale (11)
(11)
I(12)
- 1. A or 1.8 High Reactor Pressure s 1055 psig X(10)
X X
1.A 2
8 2
liigh Drywell Pressure (14)
$ 2.5 psig X (8)
X(8)
X 1.A 2
Reactor Low Water Level (14) 2 538" above vessel zero X
X X
l.A 2
1.iqh Water Level in scran Discharge Tank 5 50 Gallons X
X t 2)
X X
1.A
.ain Steam Line Isola-w 1
tion Valve closure 5 10% Valve Closure X (3) (6)
X (3) (6)
X (6) 1.A or 1.C 2
Turbine cont. Valve Fast Closure or 2 550 osta X (4) 1.A or 1.D 4
Tur!.ine Trip e
1 t
4 u.
J 24.
The Average Power Range Monitor scram function is varied (ref.
Figure 2.1-1) as a function of recirculation loop f1w (W).
The trip settinC of this function must be maintained in accordance with 2.1.A.
25.
The APRM flow biased neutron flux signal is fed through a time censtant circuit of approximately 6 seconds. This time constant may be lowered or equivalently removed (no tims delay) without affecting the operability of the flow biased neutron flux trip channels. The APRM fixed high neutron flux signal does not incorporate the time constant but responds directly to instantaneous neutron flux.
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Amendment No. 70 35a
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TABLE 4.1.A REACTOR PROTECTION SYSTEN (SCRAH) INSTRUMENTATION FUNCTIONAL TESTS
)
g MINEMUM FUNCTIoHAL TEST FREQUENCIES TCA SAFETY INST R. AND CONFSOL CIACUITS CD
- s r*
croup. (2) nanctional Test
, Minimum Frequency (3) 2:
,o Mode switch in Shutdown A
Place Mode Switch in shutdown Each Refueling outage l
Manual'swram A
Trip Channel and Alarm Every 3 Months IDH U
High Flux C
Trip Channel and Alarm g4)
On'.e Per Week During Refueling an.1 sefore Each startup j
Inoperative C
Trip Channel and Alarm (4)
One.e Per Week During Refueling and sefore Each startup APRM High Flux (151 scram)
C Trip output Relays (4)
Defore Each Startup and Weekly When Required to be Operable liigh Flux (F10W Blased)
B Trip Output Relays (4)
Once/ Week Righ rius (Fixed Trip) 8 Trip ostput pelays (4)
Once/ueek i
g Inoperative 8
Trip Output Relays (4)
Once/ Week Downscale 8
Trip Output Relays (4)
Once/ Week riov Bias a
(6)
(6)
High Reactor Pressure A
Trip Channel and Alarm once/Honth (1)
High Drywell Pressure A
Trip Channel and Altra once/Honth (1)
Reactor Low Water Level (5)
A Trip Channel and Alarm once/ Month (1)
High Water 14 vel in Scram Discharge Tank A
Trip channel and Alarm Once/Honth 3
Turbine Condenser Low Vacuum A
Trip channe'. and Alarm Once/ Month (1) l O
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LIMITING CONDITION 3 TOR OPERATION SURVEILLANCE REQUIREMENTS s
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3.5 CORE AND CONTAINMENT 4.5 CORE AND CONTAINMERT COOLING CoCL!':0 SYSTEM)
_ SYSTEMS I.
Averace Planar Linear I.
Maximum Averace Planar Heat Generation mate Linear Heat Generation Rate (HA PLHG R)
During steady state power operation, the Maximum' The KAPLHGR for each type Average Planar lleat of f ucl as a f unctior of Ganeration Rate (KA PLHCR) average planar exposure tor each type of fuel as a shall be determined daily f unction of averaae plance during reactor operation exposure shall not exceed at 2 25% rated thermal
- the limiting value shown Power.
!in Tatlet 3.5.I-l throu ;h
~5. 5.1-7.
If at any time during operation, it is Jetermined by normal surveillance that the
-~
limiting value. for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the
~
prescribed limits.
If the APLNGR is not Jeturned to within the prescribed limits within two (2) hours, the reactor shall be trought to the Cold
{
Shutdown condition within 36 nours.
surveillance ar.C corresponding action shall continu,e until reactor operation is within the prescribed -
,1: mi t s.
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6 165
- k. enc lent No. k 70 r
yt 3.5 BASES testing to ensure that the lines are filled.
The visual checks no will avoid starting the core spray or RRR system
.with a discharge line not filled.
In addition to the visual
('
observation and to ensure a filled discharce line other than prior to testing, a pressure suppression chamber head tank is located approxima tely 20 feet above the discharge line highpoint to supply makeup water for these systems.
The conden sate head tank located approximately 100 feet above the discha rge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service.
System discharoe pressure indicators are used to determine the water level above the discharge line t.tqh point.
The indic.itors 'will re flect - approxima tely 30 potq f or a wa t e r level at the nich point and 4 5 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge linen are filled.
Wh*n in their normal standby condition, the suction for the HPCI and RCIC pumps are aligned to the condensa te storage tan \\, which is physically at a higher elevation than the HPCIS and RCICS piping.
This ascures that the HFCI and RCIC discha rge piping remains filled.
Further acr.urance is provided by observing water flow.from these systems high points monthly.
4.
t.aximum Averace Planar Linear Heat G en er a t i on Hate (MAPLHCR1 This specification assures that the peak cladding temperature foll owing the postulated deuign basis loca-of-conlant accident will not exceed the limit specified in the 10 CTR 50, Appendix K.
The peak cladding temperature following a postulated loss-of.-
coolant accident is primarily a.! unction of the averaqe heat generation rate of all tne rods of a fuel assembly at any axial location and is only dependent seconda rily on the rod to rod power distribution within an assembly.
Since expected local variations in power distribution within a fuel assembly af f ect the calculated peak clad temperature by less than +
20*r relative to the peak temperature for a t)pical fuel design, the limit on the average linear hea t generation rate is sutficient to assure that calculated t enpera tures are i
within the 10 CFR 50 Appendix K limit.
The limiting value tor MA PLHGR is sho.n in Tables 3 5. }-1 t h re n mh 7.
Th.
s apporting these limiting values is presented in referenc'emI,e l ve.s.s J.
Linear Heat Generation Rate (LHGn1 This specification assures that the linear heat generation rate in any rod is less than the, design linear heat 176 i
Amendment No.
70 l
i g
n ge% emeA ep.
- /
a~
's d D.^ W reported within 30 davs.
It must be recoont ed that there as always an action which would return any of the pa rame te rs (KAPLHCR, LHCR, o r MC P R) to within prescribec l imats, namely power reduction. Under most circumtsnees, t..ts will not be the only alternative.
n.
References 1.
-Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3, NEDO-24194A and Addenda.
2.
"EWR Transient Analysis Model Utili::ing the RETRAN Program," TVA-TR81-01-A.
3.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.
- m -e 3
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179 f*
AmendmentNo.[,70 r
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TABLE 3.5.1-7 MAPLH3R VERSUS AVERAGE PLANAR EXPOSURE Plant: 3F-3 Fuel Type: BP8DRB284L Average Planar Exposure MAPLHGR (mwd /t)
(kW/ft) i
200 11.2 1,000 11.3 5,000 11.8 10,000' 12.0 15,000 12.0 20,000 11.9 25,000 11.3 30,000 10.8 35,000 10.1 40,000 9.4 45,000 8.8 u.e. a l'
Amendment NO.
, 70
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Figure 3.5.K-1 MCPR Limits for 8x8R, P8x8R and LTAs 182c Amendment No. 70
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