ML20149B268
| ML20149B268 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/27/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20149B274 | List: |
| References | |
| NUDOCS 8103130118 | |
| Download: ML20149B268 (38) | |
Text
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UNITED STATES
,yo e
g, NUCLEAR REGULATORY COMMISSION y
.C WASHINGTON, D. C. 20555 Oi
,i s,*.v TENNESSEE VA'LLEY AUTHORI Q DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 68 License No. DPR-33 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendments by Tennessee Valley Authority (the licensee), dated fiarch 1,1979, as supplemented by letter dated August 7,1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2 C(2) of Facility License No. DPR-33 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 68, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
slosiso16
2-3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:~ February 27, 1981 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 68 FACILITY OPERATING LICENSE NO. DPR-33 DOCKET N0. 50-259 Revise Appendix A as follows:
1.
Remove the following pages and replace with identically numbered pages:
39 40 41 48 1
124 240/NT 2.
The underlines pages are those being changed; marginal lines on these pages indicate the area being revised. Overleaf pages are provided for convenience.
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McTES ma Tact.E !.. A. A Initially the minimum f requency f or the indicated tests shall be once.
1.
per month.
A description of the three groups is included in the Basec of this 2.
specification.
Tunctional tests are not required when the systems are not required to 3.
If tests are be operable or are operating (i.e., already tripped).
missed, they shall be performed prior,to returning the systems to an operable statua.
This instrumentation is expapted f rom the instrument channel test 4.
instrument channel functional test will consist of This definition, injecting'a' simulated electrical signal into the measurement channels.
The water level in the reactor vessel will be perturbed and the corres-5.
ponding level indicator chan5es vill be monitered. This perturbation test will be performed every sench af ter completion of the senthly functional test program.
6.
The functional test of the flow bias network is performed in accordance with Table 4.2.C.
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TaaLE 4...a REacTom re0 Fact 10m steTIM (ecmAM4 Instasmetart cALIsaATION 4
mINEMUN c1LIBBATION FREQU8tCIES FOR RE. ACTOR P90fSCTION 1RSTRUMENT CM&MMEES I
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t lastrument m a==e1 Group (1)
Calibration Minimun Frequency (2) 4 East sigen Flus C
Comparison to APRM on Control-Isote (4) i led seksteoeste (tij l
Aften sight Flam i-omtpet signale a
seat antamos once every 7 dare Flow Stae signal 5
Ca11tarate Flow sine Signal (?)
Once/ operating cycle l
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Right poector Pressere Standard Pressure source Brery 3 Isaattne i
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Sigen Drywe11 Freesare Standard Proseure source Every 3 seonthe i
Beector saw teater Imvel Pressure Standard Every 3 Isonthe j
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semia steam s. lee neolatism valve cloeure 1
I sama steam Esse algen meetatiam a
steadare current source (3) svery 3 seonthe 1
Tsartsime First stage Proeswe permiselve Stam4ard Pressure source Every 6 seanthe i
ttarbine control valve - Emee of oil pressere &
standard Pressure source cacetoperettag cycle smendae sea, valve closure
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i Amendment No. 68 I
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i NOTF.g FOR TASt.E 4.1.R I
A description of three groups is included in the bases of this 1.
specification.
Calibrations are not required when the systees are not required to If calibrations are missed, they shall l
2.
be operable or are tripped.
be performed prior to returnin2 the system to an operable status.
I Cali-The current source provides an instrument channel alignment.
3.
braties using a radiation source shall be made each refueling outage.
4.
Hasimus frequency required is once per week.
Physical inspection and actuation of these position switches will be 3.
l performed once per operating cycle..
6.
Ce controlled shutdowns overlap between the IBM's and AFRM's will l
be verified.
f The Flow Bias signol Calibration will consist of calibrating the 7.
sensors, flow converters and siCusi of f set netvorks during esch The instrumentation is an anslog type with redun.
operating cycle.
dant flow signals that can be compared. The flow comparator trip and upscale will be functionally tested according to Table 4.2.C to Refer to ensure the proper operating during the operating cycle.
4.1 geses for further explanation of calibration frequency.
8.
A complete tip system traverse calibrates the LPRM signals to the e
The individual LPRM meter readinrs will be process computer.
adjusted as a minimum at the beginning of each operating evele before reaching 100% power.
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4 Amendment No. 68 4
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%i1El The reactor prutection systee autosatically initiates a rasctor seras to:
1.
Preserve tt e integrity of the fuel cladding.
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Preserve the integrity of the reactor coolant systes.
3.
Ministse the energy which must be absorbed following a loss of coolant i
accident, and, prevents criticality.
This specifiestion provides the limiting conditions 'for eperation necessary t o preserve the ability of the system to tolerate singis f ailures and still perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one. channel any be made inoperable for brief intervala to conduct required functional est's and calibrations.
The reactor protection system is made up of two independent trip systema (refer to,5ection 7.2, TSAR). There are usually four channela provided to ronitor each critical parameter, with two channels in each trip systes.
The outputs of the channels in a trip systes are cozhined in a logic such snat either channel trip will trip that trip system. The simultaneous tripping of both. trip systems will produce a reactor scras.
This system meets the intent of IEEE - 279 for Nuclear Power Plant Protec-
. tion Systems. The systes has a reliability greater 'than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 systes.
With the exception of the Average Power Rante Monitor (ATEM) channels, the Intermediate Range Monitor (ILM) ehannels, the Main Steam Isolation valve closure and the Turoine Stop Valve closure, each trip systes logic has one f
ins.t ruse 9 C c hannel. When the minimus condition for operstion on the nu=ber et operable instrucent channels per untripped protection trip systes is set or'if it esanot be set and the effected protection trip system is placed in a tripped condition, the ef fectiveneus of the protection systes is preserved; i.e., the systan can Colstate a single f ailure and still perform its intecded function of scrammins the reactor. Three A?1M instrdment channels are pro-vided for each protection trip systas.
Each protection trip systes has one acre AFRM than is necessary to meet the minimum nunbar required per channel. This allows the bypassing of one AFKM I
per protection trip systes for maintenance, testing or calibration. Addi-tional IRM chanecle have also been provided to allcw f or hypassin8 of one such channel. The bases f or the actsa setting for the !1M, AFEM, high rese-ter pressere. reactor low water level M51V closure, turbine control valve fast closure, turbine stop valve closure and loss of condenser vacuum are Jiscussed in Specification 2.1 and 2.2.
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d.1 8A%F.S The frequency of c.ilibration of the APtM Finw Mla ains Network han been establis.hed am cath refue11mg outaxe. There are neverai lui.trustents which nust be calibented and it will take several haitra to perform the calibration of the entice network. While the calibration is being per-formed, a aero fler signal will be. sent to half of the APRM's resultLng d
in a half scram and rod block condition. Thus, if the calibration were Based performed during operation, fluz shaping would not be pnosible.
other generating stations, detf t of inscrcsents, such on experience at and therefore, as those in the 7109 Biasing Network, is not significant to avoid spurious scrame, a calibration frequency of each refueling out-age is estabitshed.
Croup (C) devices are active only during a given portion of *the opera-tional cycle. For example, the IRM is active during startup and inactive during full-power operation.
Thus, the only test that is meaningful is the one perf ormed just prLor to shutdown or startup; i.e.,
the tests that are perf orned just prior to use of the instrument.
Calibratinn f requency of the instrument chanAe1 14 divided ince two groups. These are as followe Passive type indicating devices that can be compared with like 1.
units on a continuous basis.
Vacuum tube or scciconductor devices and deccctors that drif t or 2.
lose sensitivity.
Experience with passiv type instrucents in generating stations and sub-For stations indicates that the specified estibraticns are adequate.
r.
those, devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4%/ month; i.e., in the period of a month a For de if t of. 4% vould occur sed thus providing f or adequate margia.
not the only considera-the APRH system drift of c1cceronic apparatus tu I
tion in determining a calibration frequency. Change in power distribu-tion and lose of chenber sensitivity dictate a calibration every seven Calibration on this f requency assures plant operacion at or below days.
thormal limits.
I two instru=ent A comparison of Tables 4.1. A sad 6.1.3 indicates tnst channels have not been included in the latter table. These are: mode switch in shutdown and manual scrsm. All of the devices or sensors associated with these scras functions are simple on-of f rwitches and, hence, eslibration during operation is not applicable, i.e., the switch 1's either on or of f.
. '.1 The ratio of Core llaximun Fraction of Limiting Power Density (MFLPD) to
?
Traction of Rated Power (FRP) shall be checked out once per day to determine if the APRM scram requires adjustment. This vill normally be done by checking l
Only a small number of control rods are moved daily i
the APEM readin6s, 47 e
Amendment No. 68
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during steady-state. operation and thus the retto is not expected to
" t change significantly.-
The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. The APRM systen, which uses the LPRM readings to detect a change in thermal oover, vill be calibrated every seven days using a heat balance to compensate for this chsnge in sensitivity. The RBM system uses the LPRM reading to detect a localized change in thermal power. It applies a correction factor based on the APRM output signal to determine the percent therral never and therefore any change in LPRM sensitivity is compensated for by the APRM calibration. The technical specification limits of C M PD, CPP.,
MAPLUGR and R ratio are determined by the use of the process computer or other backup methods. These methods use LPR" readings and TIP data to determine the power distribution.
Compensation in the process computer for channes in LPRM sensitivity will be made by performing a full core Tip traverse to update the computer calculated LPRM correction fr.ctors every 1000 effective full power hours.
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As a minimum the individual LPRM meter readings will be adjusted at the beginning of each operating cycle prior to reaching 100 percent oower.
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StjRVE711.ANCE REQUQDN 1/iv.tTENG CONDRTTOH5 FOR OPERATROM
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4.3.1 Control Rods
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.Coetrol Reds _
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- b. During the shutdown procedure-b" no rod movement is permitted a.
The capabill:7 of the RSCS to pro-
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between the testing performed perly fulfill its function shall be above 20% power and the rein-verified by the following tests:
statement of the RSCS re-3*9"**** ?'881'" ~ 3*1*'E
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straints at or above 20%
and atta=p: to wi:hdraw a rod in the powe r.
Alignment of rod renaising sequences. Move one rod groups shall be accomplished in a sequen:e and select the remain-prior to performing the tests.
ing sequences and atta=pt to move a rod is each. Repeat for all sequences.
Whenever the reactor is c.
in the startup or run modes Croup notch portion - For each of the below 20% rated power the six comparator circuits go th:ough Rod Worth Minimizer shall be test initiate; comparator inhibit; operab1's or a second licensed verify; reset. On seventh atta=pt test is allowed to con:inue ustil operator shall verify that the operator at the reactor completion is indicated by console is following the illuminstien of test co= place light.
control rod progras.
b.
The capabili:7 of the Red A second licensed operator VCh Minini:er (R'4M) shall may not be used in leiu of
testing in the startup or run modes below 20 percent 1*
The cc rectness of the of rated thermal power.
control red withdraval saquence input to the
'.R'JM cc:puter shall be verif ied before reactor startup or shutdou.
2.
The RW. computar en line d*
If Specifications 3.3.3.3.a diagnos:ic test shell be through.c cannot be met the successfully pet.'ormed.
reactor shall not be started, or if the reactor is in the 3.
Prior to startup, proper run or startup modes at less annunciatien of the selec-than 20: rated power, it tion error of at least one shall be brought to a shut-out-of-sequence contrel rod down condicios imediata1.
shall be verified.
7 4
Prior to startup. the rod block fune:1on of the F/N.
shall be verified by :oving an out-of-sequence control rod.
5.
Prior to obtaining 20% ra:ed power during rod ir.sertion at shutdevn, verify ths 1acching of the proper red 3**"? **$ ?#87*r anzuncia:1:n 3:33 M3 after inser: errors.
Amendment 35 as m
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y 'TtHC CnNalTIONS FOA orEAAftoW SURVF.!Lt.ANCE REQUIRD @rr3 3.3.5' Centrol Rede 4.3.8 Centrol Rede 4.
Centsel roda shall not be the presense
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rw withdrawn for startup er of e sessed eposeemr g
refuelias ualeas at leaag to m ify tb failsstos of tuo eserce range channels tb eerrest M psegram M have as observed count rate.
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ogast to or greater than
'= PFlorr to sentrol red withdrowel three counts per ascend.
for eCartup er during refuelir.g.
5.
.Durins.eperation with verify that at leses rue searce 4
I' range channels have se observesL limiting centrol red pat.
count rata of at least three j!
, tarns, as determined by the I
designated qualified permes.
counts per second.
mel,either j..
a.
Both RaN shamaels shall S.
When a limiting control rod l'
be operablas pattern exists, an instruseet functional test of the RAM er shall be perfe:wed prior to b.
Concrel rod withdrawal withdraval of the designated shall be blocked.
rod (s) and at least, ease per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
C.
Scram Insertion Times 1.After each refueling outage all operable rods shall be scram time tested from the fully withdrawn position with the nuclear system i
pressure above son pois This dy, testing shall be completed prior to C. Scram Insercien Times exceeding 40% power.
Below 20%
,4, peuer', only. rods in those sequences 1.
The average scram insertion g,gg7vithdrNnintN) region (A
34.or 3 and B which
,,1,2 and A time, based on the deenergi-sation ef the scram pilot valve from 100% rod density to 50% red estenoids as time sare. of all density shall be scram time tested.
operable centrol rods in the The sequence restraints imposed rpon reesser power operation condi-the control rods in the 100-5C tien shall be no greater them:
percent rod density groups to t's
!amertad From Avg. Scram Inser.
preset power level may be removed Fully Withdrawn tien Times (see) by use of the individual bypass switches associated with those 5
0.375 control rods which are fully or 20 0.90 partially withdrawn and are not 50 2.0 within the 100-50 percent rod density 00 3.500 groups.
In order to bypaso a rod, the actual rod axial position must be
'l known; and the rod must be in the correct in-sequence position.
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68 124 Q
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Sevendary containment inte-g, Arity shall be maintained in lance shall be performed as the reacter sone at all times Ndicated beln :
encept as specified in 3.7.C.2.
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3.7.C Secondary containmenc 4.7.c Second.ir, containmen't
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Secondary containment caps-bility to maintain 1/i+ inch a veter vacuum under cales wint
(< $ uph) conditions with a system inleakage rate of e
a not more than 12,000 efs.
{4 shall be dem nstrated at each refueling outase priur to refuelina. ~
2.
If reactor zone secondary con-2.
After a secondary containment tainment integrity canaec be violation is determined the maintained the following con-standby gas treatment system V
ditions shall be mett will be operated inanediately aft'er the affected gones are isolated from the remainder of s.
The reactor shall be made the secondary containment to subcritical and specifica-confirm its ability to esin-tion 3.3.A shall be met.
tain the remainder of the secondary containment at 1/4-b.
The reactor shall be cooled inch of water negative pres'sure down below 212*F and the under.cals wind conditions, reactor coolant systen vested.
ca Fuel movement shall not 4
be permitted in the reac-
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Primary containment integrity maintained.
3.
Secondary containment integrity.
shall be maintained in the re-feeling tone. except as spect.
!!ed in 3.7.C.4 241 V
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Amendment No. 68 3
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UNITED STATES NUCLEAR REGULATORY COMMISSION y
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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 64 License No. DPR-52 i
1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendments by Tennessee Valley Authority (the licensee) deud March 1,1979, as supplemented by letter dated August 7,1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license anendment and paragraph 2.C(2) of Facility License No. DPR-52 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 64, are hereby incorporated in the l
license. The licensee shall operate the facility in accordance with the Technical Specifications.
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- 3. 'This license amendment is effective as of the date of its issuance.
1 FOR THE NUCLEAR REGULATORY COMMISSION A-
@ h- [oZP) c Thomas A./Ippolito, Chief Operating Reactors Branch #2 Division of Licensing j
Attachment:
Changes to the Technical Specifications Date of Issuance: February 27, 1981 c
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ATTACHMENT TO LICENSE AMENDMENT NO. 64 FACILITY GPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260
. Revise Appendix A as follows:.
1.
Remove the following pages and replace with identically numbered pages:
39/40 41 4f 48 1
Tf4 239/240 241/NE 2.
The underlined pages are those being changed; marginal lines on these pages indicate the area being revised. Overleaf pages are provided for convenience.
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McTES rne TAGt.E 4 1. A Initially the minimus frequency for the indicated" tests shall be once 1.
per month.
W.52:
2.
A description of the three groups is included in the Bases of this specification.
3.
Functional tests are not required when the systema are not required to be operable or are operating (i.e., already tripped). If tests are missed, they shall be performed prior to returning the systems to an operable status.
4 This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels.
5.
The water level in the reactor vessel vill be perturbed and the corren-pending level inJ!cator changes wLil be monitored. This perturbation test will be perfor:ed every month after completion of the monthly functional test program.
6.
The functional test of the flow bias network in performed in accordsoce with Table 4.2.C e
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TABLE 4.1.5 REACTOR PROT 9CTION SYSTEtt (SCSAMl INSTRthatmT C&12BAATIO08 utstaeust CmLIsaattest rasgossctas rom atacrom reoracTrom Instsosetur Canteessa Instrument (Snannet ergep (1)
Calibration stinimum Frequency (2)
C Comparison to ArtM on control-Note (4) l 3888 Eigth Fles Red ene* h a (6) 4 i
Armut sigen Fles
(
outpet signate a
seat annanos ence esory 7 dare Flow B&se signal S
Calibrate Flow aims signal (?)
caceteperating cycle
{
j 1.fmet signal B
Tir system Traverse (8) spery teet gifactive Full Power Boere Rigen Beactor Pressure standard Pressure source arery 3 saanthe f,
Nigen Brywe11 Pressere staneerd Pressure source Svery 3 stanthe teactor Een unter Lesel Pressure standard Deery 3 Isouthe
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migh seter smel in scram stecherve velene a
ante (si inste (se Testsame heer Eme samem standere vecean source avery 3 samathe i
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f o
e mim steam Line Isolation valse Cleomre a
pote (s) uste (sl maa n steen Line B>h mediation a
standard Corrent source (33 Svery 3 seonthe l
Tartname First stage Pressure permissies Stan4ard Pressere soereo Beery 6 annatte 1
4 serbt.e contret valse - smee or mal pressere a steaansa pressure sourse encomperstans crete i
Termane stav vatee closure a
inste (si note (se i
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M rf.s Ton TAst.E 4.1.n_
A description of three groupe is included in the bases of this 1.
specification.
calibrations are not required when the systees are not required to be operabis or are tripped. If calibrations are missed, they shall 2.
be performed prior to returning the system to an operable statua.
Cali-The current source provides sa instrument channel alignment.
3.
bretion using a radiation source shall be made each refueling outage.
4.
Maximum f requency required is once per week.
Physical inspection and actuation of these position switches will be 5.
performed once per operating cycle.
On controlled shutdowns, overlap between the IRM's and APRM's will 6.
be verified.
The Flow Bias signal Calibration will consist of calibrating the sensors, flow converters. and siCual offset networks during each 7.
The instrumentation is an analog type with redun-operating cycle.
The flow comparator trip dant flow signals that can be coespared.
and upscale will be functionally tested according to Table 4.2.C to Refer to ensure the proper operating during the operating cycle.
J.1Basesforfurtherexplanationofcalibrationfrequency.
A complete tip system traverse calibrates the LPRF signals to the j
8.
The individual LPRM' meter readines will be process computer.
adjusted as a minimum at the beginning of each operatiftg evele before reaching 100% power.
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3.1 nAsts The reactor pt.' et t ti u sy.et em autos. tically initiates N;;;;;;..
a reactor scram to:
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- 1.
Preserve the integrity of the fuel cladding.
- . Preserve the integrity of the reactor coolant system.
3.
Minimize the energy which must be absorbed following a loss of coolant accident, and prevents criticality.
Thie specification pr.$vides the limiting conditions for operation necessary in preacrve the ability of the ny*c*m to tolerate eingic failures and still channels perform it s intended f unction even duting periods when instrument asy be out of, service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional 1ests and calibrattune.
The' reactor protection system is made up of two independent trip systems (refer to Section 7.0, PSAR). There are usually four channels provided to eenitor each crit!:41 parameter, with two channels in each trip system.
The out;uts of the channels in a trip system are combined in a logic such that either chann:1 trip will trip that trip system. The cimultaneous tripping of both trip syste=s wil'. produce a reactor scram.
This system meets the intent of IEEE - 279 for Nuclear Power Plant Protec-tion Systems. The mystem has a reliability greater than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 systes.
With the exception of the Average Power Kange Monitor (APRM) channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valve
~
cloncre and the Turbine Step Valve closure, each trip system logic has one h
b in.orume,c channel.
- 4 hen ene minimun rendition for operation en t e num er everable instrument channels per antripped protection trip system is met etor if.It csanot be ret dnJ the ef f ected protection trip system is placed in a t ripped ennd Lt hm. the c!!ectivenena of the protection system is preserved; i.e.. the =vnte:a can t,let ate a sing!c failure and still peiform its intended function of scramm r.4 the reactar. Thrse APRM instrument channels are pro-vided fer each prats:ti): trip systen.
Each protection trip system has one more APRM than is nccessary to meet the minimum number required per channel. This allows tha bypassing of one A7KM per protection trip mystem for =aintenance, testing er calibration. Addi-tional ILM channele nave also been providad to allow for bypassing of one such channel. Tne bases f or the sc:an setting for the IRM, APRM, high resc-cor pres. ore, inactor luw water lovel, 'tSIV closure, turbine control valve fast closure, turbine stop valve closure and less of condenser vacuum are discussed in Specification 2.1 and 2.0.
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l 4.1 SA5F.S The frequency of c.elibratinn of the APRM Flow sin etnu 'fetwork han been There are neveral Instrumenta established an each refueling outage.
A which must be cattbrated and tt will taL* several hoern to perform the While the calibration la being per-4 calibration of the entire netuork.
formed, a aero flow signal will be sent to half of the APRM's resulting j
\\d in a half scram and rod block condition. Thus, if the calibration were performed during operation. fluu shaping would not be possible. Based J
l on experience at other generating' stations, drift of instruments, such as those in the riov Biasing Network, la not significant and therefore, to avoid spurious scrams, a calibration frequency of each refueling age is established.
Croup (C) devices are active only durir.; a given portion of *the opera-For example, the 1RM is active duttn; Startup and inactive tional cycle.
during full-power operation. Thus, the only test that is meaningful is the one perf ormed just prior to shutdown or startup; i.e.,
the tests that are perf orned just prior to use of the instrunent.
divided into two Calibratinn f requency of the instrument chanhel 19 These are as fnilows:
groups.
Passive type indicating devices that can be compared with like 1.
units on a continuous basta.
Vacuum tubs or scciconductor devices and detectors that drif t or 2.
lose sensitivity.
Expretense esth passivs type instrucen's in generating stations and sub-For stations it.dicates that the specified calibrattens are adequate.
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those devier= which employ amplifiers, etc., drif t specifications call for drift to be less than o.4%/ month; i.e., in the period of a month a For hif t of.4 *wos1d occur sed thus providing for adequate margia.
not the only considera-the APRM system drif t of electronic apparatus in tion in determining a calibre. tion frequency. Changu in power distribo-tion and loss of chamber sensitivity dictate a calibration every seven Calibration on this f requency assures plant operation at or belee.
days.
thermal limits.
two instrument A comparison of Tables 4.1.A and 6.1.3 indicates that These are: mode channels have not been included in the latter table.
All of the devices or sensors switch in shutdovn and nanual scram.
asseciated with these scras functions are simple on-of f switches and, hence, calibration during operation is not applicable, i.e., the switch is either on or of f.
The ratio of Core Maximum Fraction of Limiting Power Density (MFLPD) to Fraction of Rated Power (FRP) shall be checked out once per day to determine I
if the APRM scram requires adjustment. This vill normally be done by checking 4
the APEN readings. Only a small number of control rods are moved daily 2>.
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Amendment No. 64
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The sensitivity of LPM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. The APRM systen, which uses the LPRM readinr.s to detect a change in thermal power, will be calibrated every seven days using a heat balance to compensate for this change in sensitivity. The RBM system uses the LPRM reading to detect a localised change in thermal power. It applies a correction factor based on the APRM output signal to determine the percent therral never and therefore any change in LPRM sensitivity is comoensated for by the APRM calibration. The technical specification limits of CFPD, CPP.,
MAPLHCR and R ratio are determined by the use of the process computer or other backup methods. These methods use LPP.M readings and TIP data to determine the power distribution.
Compensation in the process computer for changes in LPRM sensitivity will be made by performing a full core Tip traverse to update the computer calculated LPRM correction fr.ctors every 1000 offective full power. hours.
As a minimum the individual LPRM meter readings will be adjusted at
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beginning of each operating cycle prior to reaching l')0 percent power.
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Amendment No. 64
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LMtT T HC COMU}.,.u,,!'? ' " ~ ~ ~'- ~ ~' 4.3.3 concrei Rada 3,g g trei and. . hI
- b. During the shutdown procedure The capabill:y of the F.SCS to pro-no rod movement is permtted perly f ulfill its f unction shall be a.
between the testing performed verified by the follovir.g tests: above 20% power and the rein-st st ement of the RSCS re-S*9uence portion - Select a sequesco straints at or above 209, and attemp: to withdr:w a rod in the Ali gnment o'f rod *** hs 889"'" '
- 88' 'M power.
groups shall be accon:pli shed in a sequen:s and select the re=412-prior to performing the test.. int, sequen:cs and atte=pt to c:cve s a rod in es:h. Repast for all sequen:es. Whenever the reactor is Group notch por: ton - Tor each of the c. in the startup or run modes si.x co= pars::: circuits go th:ough below 20; rated power the t es t ini:ia:e; ce:psrator 1r.hibi:: Rod Vorth Ministser shall be verify; reset. On seventh att:=:p: operable or a second licensad is alleved to continue until test operator shall verify that comple:icn is indica:ed by the opers:or at the reactor 111 W v.:b: of test ce=ple:n light. console is f olleving the control rod progra.m. b. The capab111:7 of the Rod A second licensed operator V'f th M13'mi:er (. M) shs11 G may not be used in leiu of " r*11rd h :h: f:11.wi.o S the RUM during scram time checks: testing in the startup or run mo6es below 20 percent 1. The corre:: ness of the of rated thermal power. con:rol rod vichdrswel I sequer.:e input to tha ?'aM c =puter shall be verified before rea::or s:ar:c; er shg:d: :. 2. The m c::pu ar cc line If Specifics: ions 3.3.5.3.s diag.os:1: es: shall 'ce d* r.hrough.c cannot be met the successfully performed, reactor shall not be started, or if the reactor is in the 3. Frier te star;up, proper ~ ane. uncia:1:n of the selec-run or startup modes at lass tion er:cr of a: least ose than 20" rated power, it our-of-seque-:e contrel ros t shall be brought to a shut-shall be verified. down condition ir:=ediately, i 4. Prior to startup, the rod block fun:: ion of the RW. shall 'ca verified by :oving ~ an out-of-sequen:e con::e1 rod. 5. Prior to obtai=ir.3 20: rated power durks rod inser: ion at sha:down. verif y ths la:ching of the proper rod group asi pr:per at unciatici. E"? g3 after inser: a.rors. i 35 Amendment No., y w w+-
[:. y 'TINC CnNutTLONS FOA 0FtAAT10M 5URVEIDANCE REQUIAgMgyrs 4 3.5.$' Cesarel,,Ag1J, 4.3.3 Centrol Rede 8 4. Centset rede shall not be h '*981' b P'****** I withdramm for startup er of a seemed 11esseed opesseer ( sofeeling unless at tenac te verify the feinsets et two source range channata th terrest red psegree shalt have as observed count rate. be verified. egnet se er grasser than 4 Frier to esmerol red withdrawet a <bres eenmaa per second. for startup er during refueling. i verify that at least two source
- 5..Durins, operation with
. nage channels have as observed r limiting control rod pac. cowoc rate of at leses three .tarus, as determined by the counts per second. danignated qualified persos as1,either: S. ilham a limiting centrol red s. Sesh asM channata shall pattern exists, an instrument b. rables fonctional test of the R3M shall be perforwed prior to withdraeat of the designated b. Centrol rod withdrawal rod (s) and at least, ease per s h11 he blocked. 24 hours thereafter. C. Scram Insertion Times 1.After each refugling outage all h operable rods shall-be scram timeV tested from the fully withdrawn. Position with the nuclear system pressure above 800 paig This testing shall be completed prior to C. M an Inserrien Times exceeding 40% power. Below 20% power', only. rods in those sequences r4 1. The everage seren insertion g,ggyfithdrNaint$k) region .o r 5 and 5 which (A' ,,1,2 and A 3 time, h==d en the deenergi-sation of the scram pilot valve from 100% rod density to 50% rod solenoids as time sere, of all density shall be scram time testod.
- g. ",
The sequence restraints imposed upon g. g e e the entrol reds in th.e 100-50 tien shall be no greater them: percent rod density grcups to the
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preset power level may be removed Fully Utahdrame tien Times (see) by use of the individual bypass switches associated with those 5 0.375 control rods which are fully or 30 0.90 partially withdrawn and are not 50 2.0 within the 100-50 percent rod density 90 3.500 groups. In order to bypass a rod, the actual rod axial position must be I known; and the rod must be in the entract in-sequence position. 4": f~~ 9 124 k 2 Amendment No. 64 9 eptaquead eq= eye we e Iy.-, -,-,e m e- - -ee-,-wm -..-r--ww- -w=,,-w.,e.--.-n,... www..,3--w.,,,.,~,n we,..... .,w-..sw*em--me+.- w w - erw ww.w-enw,.eewe-.-we.w-w,w-+.,=r-e.-e=-- .w.v=..,,*me-<a+,*
Ur. '. t I i.IMITING CONDITICHS FOR OPERATION SURVIILLANCE RIQUIRZ.WNTS ff=- e.7 CONTAINMENT sys?rvs
- 1. 7 Co$rtA DMINT sysT y g c.
When one train of the standby gas treatment system tecomes inoperable the other two trains sna11 te demonstrated to be operatie within 2 hours and daily thereafter. 4 If these conditione cannot be met, the reactor shall be placed in a condition for which the sta ndry oas treatment syste-is not required.
==. 239 A.T.en::~ent U0. 44 AME.. e e-.e. w-a w.= '-n-w w -Me -t 4
Unit 2 t.!MITfxc enNntTfn*as rna net: TAT 10W SURVE11.1. ANCF REQtitw.CitMTS t 1.7.C Set **ddrJ..Cesc a t nment 4,y,c Secondary Concetament 1. Secondary containment inte. g, g,,,,g,7,,,,,g,,,,,,,,,gg, stity shall be w intained in g , g,gt 9 p,g g,, the reacter zone.sc all times wg,,g,4 wgg escept as specifled in 3.7.C.2. t i t 7.., a ir +, ,,i v l 1 l L' eps y. ?.o
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.7.....- j l ~ 3.7.C Secondary Containment 4.7.C Secen.iar, containment { j 4
- a. Secondary containment capa-bility to maintain 1/4 inch a vecer vacuum under calm wint
( < 5 mph) conditions with a system inisakage rate of not more than 12,000 cfr.. shall be demonstrated at each refueling oucase prior to refuelina. 1 2. If reactor zone ascondary con-2. After a secondary containment i tainment integrity cannot be violation la determined the maintained the following con-standby gas trea ment system V j dicions shall be mets will be operated leasediately i aft'er the affected genes are isolated from the remainder of i a. The reactor shall be made the secondary containment to suberitical and Specifica-confirm its ability to main-tion 3.3.A shall be mee. tain the remainder of the i secondary containment at 1/4-b. The reactor shall be cooled inch of water negative pressure down below 212'T and the under cala vind conditions. p reactor coolanc eystee vented. c. Feel movement shall not be permitted in the reac-ter sene, d. Primary containment integrity asintained. 3. Secondary containment integrity shall be maintaineJ in the re-fuellas sene. except as spect-fled in 3.7.c.f.. I 241 d v Amendment No. 64 .k 3
i o TINC CONDIT!nNS FON nP LRAT ION $URVEILLANCh XICU!r..<h2.st! l [j;;,.'.R] 7. Secondary containment 4.7.C Secondary Conestnesnt 4 If refueling aone secondary containacnt cannot be maintained the following conditiona shall be mets s. Handlinr. of spent fuel and all operations over spent fuel paole and open reat-tor wells containing fuel shall be prohibited. b. The standby gas creaguent system suction to the re-fueling kone will.be blocked except for a con-trolled leakage aren si:ed to assure the achieving of a vacuum of at least 1/4-inch of water and not over 3 inches of water in all three renctor zones. P rima ry Conrainment'laolation Valves. D. Prinary Contai.nent Isol.itfen Vs*vai 3. Durinr, reactor power operatien, 1. The primary containe.cn: isols-all 1selation valves listed in tien valves surve111r.i.ce shall Tobic 3.7.A and all reactor be performed as follows; coolant systen instrunent line flow check velves shall be a. At least once per operating operable except as specifiad cycle the operable isols-in 3.7.D.2. tion valves that are power operated and auto-matica1*.y initiated shall 1c tested far simulated autoe. etic initiation and closure times. l r b. /,t least once per, quarter:
- 41) All nore.sily cpen power operated isolation valvas (except for ths main sesam line pover-operated isolation valves) shall be fully closed and re=pened.,
242 G ~e w p-1r w <eva--eg .-tw-.mi-p t -y-w-ne-n yr vw-- mm v --+eg gw-g- -iy-- w-gy-y -rr ,wr-w-
/p ua<,q(0,, UNITED STATES y g NUCLEAR REGULATORY COMMISSION E WASH WGTON, D. C. 20559 %....f TENNESSEE VALLEY AUTHORITY D0dKETNO.50-296 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 3 t AMENDMENT TO FACILITY OPERATING LICENSE t Amendment No. 40 License No. DPR-68 1. The Nuclear Regulatory Commission (the Cornission) has found that: A. The application for amendments by Tennessee Valley Authority (the licensee) dated March 1, 1979, as supplemented by letter dated August 7,1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without adangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Concission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility 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. 40, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
4 2-3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION _ /- / _ 4) Thomas lito, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: February 27, 1981 e o I -...-s..,.+--w-+ my.y,.-,. -r .-w.y,r,y ,ww.y.,,-a.,,,-*m---* e-+my--e+--- ,--gy,--rw47 emw~,-- w y- .e-w (, - -, - 9,e-', e v.
ATTACHMENT TO LICENSE AMENDMENT N0. 40 l FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows: l t 1. Remove the following pages and replace with identically numbered pages: 39 40 46 47 128 251 j 252 2. Marginal lines on the above pages indicate revised area. e
a 4. 1 Thats 4.1.3 j R2ac10R 790TECT10m sYsTEN (SC3 TAM) INSTRONENT CALIBRATION MIu1NUBt ekt.IBRATION FREQUENCIES FOR REACTOR PRotaCTION INSTRUMEttr CHAleNEIA toetriment Emannel Group (1) calibration minimum Frequency (2) 2 j IBM Rigen Flus e ccespartoon to APRM on Control-leote (4) led Samutdoene (6) ] APstet Eigen Flam output signals a seat melance once every 7 daye riow sias signal a calibrate riow sine signal (7) once/ operating cycle I LPets signal a TIP system Traverse (8) Every 1000 Effective Full Pouer Noere i Eigtn Reactor Freesere A standard Pressure source Brery 3 sconthe l Righ Drywell Freesere standard Pressure source avery 3 seanthe Beector Iow tanter Imvel Prwesure Standard Every 3 aoomthe EigskEpster Level in Scram Disct g e Toleos A stote (5) Inste (s) i Turbine condenser Low Tecuum A standard Vaeema source avery 3 seenthe (s) '
- =,te (se seat. et..a Lt e rootation volve closure a
note aman steen Line migh Radiation 5 Standard cerrent Source (3) Every 3 seosthe m i.e,1 - e eure -...e ... _t-rd _ e.re r.e m t.e rabine contset valve - Imee of mit pressere a standard reessere source asceroperatine crete j verbi e sea, valve closure a note (s) Imote (se 4 3 i 5 J 4 Amendment No. 40 1 i = - - 41 ,m- ~ w
o HOTES TOR TABLE 4.1.R e 1. A description of three groups is included in the bases of this specification. 2. Calibrations are not required when the systems are not required to be operable or are tripped. If calibrations are missed, they shall be performed prior to returning the system to an operable status. 3. The current sourte provides an instrument channel alignment. cali-bration using a radiation source shall be made each refueling outage. 4. Haximus frequency required is once per week. 5. Physical inspection and actuation of these positten switches will be performed once per operating cycle. 6. On controlled shutdowns, overlap between the IRM's and APRH's will be verified. 7. The T1ov Bias Signal Calibration vill consist of calibrating the sensors, flou converters. and signal of fset networks during each operating cycle. The instrumentation is an analog type with redun-dont flow signals that can be compared. The flow comparator trip and upscale will be functionally tested according to Table 4.2..C to ensure the proper operating during the operating cycle. Refer to 4.1 Bases for further explanation of calibration f requency. 8. A complete tip system traverse calibrates the LPPF signals to the process computer. The individual LPRM meter readines vill be adjusted as a minimum at the beginning of each operating evele before reaching 100% power. 40 Amendment No. 40 ~ -r
The f requency of calibration of the APM Flow Biasing Network has been established as each refueling outage. There are several instruments which must be calibrated and it will take several hours to perform the calibration of the entire network. While the calibration is being performed, a zero flow signal will be sent to half of the APM's resulting in a half scram and rod block condition. Thus, if the calj bration were performed during operation, flux shaping would not be possible. Based on experience at other generating stations, drift of instruments, such as those in the Flow Biasing Network, is not significant and therefore, to avoid spurious scrams, a calibration frequency of each refueling outage is established. Group (C) devices are active only during a given portion of the operational cycle. For example, the IM is active during startup and inactive during full-power operation. Thus, the only test that is meaningful is the one performed just prior to shutdown or startup; i.e., the tests that are performed just prior to use of the instrument. calibration f requency of the instrument channel is divided into two groups. These are as follows: 1. Passive type indicating devices that can be compared with 'like units on a continuous basis. p 2. Vacuum tube or semiconductor devices and detectors that drift T or lose sensitivity. h* Experience with passive type instruments in generating stations and substations indicates that the,specified calibrations are adequate. For those devices which employ amplifiers, etc., drif t specifications call for drift to be less than 0.45/ month; i.e., in the period of a month a drift of.45 would occur and thus providing for adequate margin. For the APM system drift of electronic apparatus is not the only consideration in determining a calibration frequency. Change in power distribution and loss of chamber sensitivity dictate a calibration every seven days. calibration on this frequency assures plant operation at or below thermal limits. A comparison of Table 4.1. A and 4.1.B indicates that two instrument channels have not been included in the latter table. These are: mode switch in shutdown and manual scram. All of the devices or sensors associated with these scram functions are simple on-off switches and, hence, calibration during operation is not applicable, i.e., the switch is either on or off. The ratio of Core Maximum Fraction of Limiting Power Density (CMFLPD) to Praction of Rated Power (FRP) shall be checked out once per day to determine if the APFM scram requires adjustment. "'his vill normally be done by checking the APRM readings. Only a small number of control rods are moved daily during steady-state operation and thus the ratio is not expected to change significantly. 46 k endment No. 40
l A The sensitivity of LFRM detectors decreases with exposure to neutron ^ ( flux at a slow and approximately constant rate. The APRM systen, which D uses the LPRM readings to detect a change in thernal power, vill be calibrated every seven days using a heat balance to compensate For this change in sensitivity. The RBM system uses the LPRM reading to detect a localized change in thermal power. It applies a correction factor based on the APRM output signal to determine the percent therral power and therefore any change in LPRM sensitivity is compensated for by the APRM calibration. The te'chnical specification limits of CMFLPD, CPR, MAPLilGR and R ratio are determined by the use of the process conputer or other backup methods. These methods use LPR" readings and TIP data to determine the power distribution. Compensation in the process computer for changes in LPRM sensitivity will be made by performing a full core Tip traverse to update the computer calculated LPRM correction fcetors every 1000 effective full power hours. As a minimum the individual LPRM meter readings will be adjusted at the beginning of each operating cycle prior to reaching 100 percent power. s m D ,^ I 1 I \\ O j m 47 _s j Amendment No. 40
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS m[ 1.3 REACTIVITY CONTROL 4.3 REACTIVITY COtffROL j C. Scram Insertion Tithes 1. The average scram 1. After each refueling outage all insertion time, based operable rods shall be scrom time on the deenergization tested from the fully withdrawn of the scram pilot position with the nuclear system valve solenoids as pressure above 800 psig time zero, of all This operable control rods in the reactor power testing shall be completed prior operation condition to exceeding h0% pcver. Belov 204, power, only rods in those shall be no greater thant sequences (A12 and A h or B12 3 and B3L) which were fully with- ) % Inserted From Avg. Scram Inser-drawn in the region from 100% i Fully Withdrawn tion Times (sect rod density to 50% rod density shall be scram time tested. The 5 0.375 sequence restraints imposed 20 0.90 upon the control rods in the 50 2.0 100-50 percent rod density groups 90 3.5 to the preset power level may 2. The average of the be removed by use of the indi-scram insertion times vidual bypass switches associeted for the three fastest with those control rods which operable control rods are fully of partially withdrawn of all groups of four and are rot within the 100-50 control rods in a percent rod density groups. In two-by-two array order to bypass a rod, the shall be no greater actual rod axial position must thant be known; and the rod must be in 5 Inserted From Avg. Scram Inser-the correct in-sequence position. Fully Withdrawn tion Times (sect 2. At 16 week intervals,10% of the 5 0.398 operable control tod drives O shall be scram timed above 90 3.800 800 psig. Whenever such scram time measurements are made, an 3. The maximum scram evaluation shall be made to insertion time for provide reasonable assursnee 905 insertion of any that proper control rod drive operable control rod performance is being shall not exceed 7.00 maintained. seconds. (~ ,p 128 Amendment No. 40 _ _. ~ .._ ~_ -, _. _
~ ! j Unit 3 g,IMITINC CONDITIC:ts FOR OPERATICt! SURVEILI.ANCE Rh0UIRE.".ENTS i 3.7 CONTA f ttM DIT SYsT!Ms 4.7 CONTATtfMENT SYSTINS C. Secondary containment c. secondary containmant, i 1 1 i 1. secondary centainment 1. secondary containment integrity shall be surveillance shall be maintained in the performed as reactor zone at all indicated below: times except as specified in 3.7.C.2. l m 251 Amendment No. 40 1 i i )
l LIMITING CONDIT10tl5 FOR OPERATICII SURVEILLANCE MQUIREMDfTS 3.1 $9NTATNMENT SYSTEMS
- 8. 7 CONTAIMMENT SYSTEMS i
a. secondary containment capability to maintain 1/4 inch of water vacuum under calm wind (<5 mph) conditions with a system inleakage rate of not more than ,12,000 cia, 3 shall te descnstrated at each refueling outage prior to e reiuoling. 2. If reactor zone 2. After a seccndary secondary containment integrity cannot he containment viciation maintained the is determined the following conditions standby qas treatment shall be met: system will he operated lemediately a. The reactor after the af f ected shall be made sones are isolated suberitical and from the remainder of specification the secondary 3.3. A shall be containment to met. confirm its ability to maintain the b. The reactor remainder of the shall be cooled secondary containment down below 212*F at 1/4-inch of water and the reactor negative Pressure coolant system under calm wind vented. conditions. c. Fuel rovement shall not be permitted in the reactor zone. d. Frimary containment integrity maintained. 2$2 L Amendment No. 40 . - -. - -. -.. -. -.....}}