ML20052F486
| ML20052F486 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/29/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Commonwealth Edison Co |
| Shared Package | |
| ML17194A668 | List: |
| References | |
| DPR-25-A-063 NUDOCS 8205130071 | |
| Download: ML20052F486 (52) | |
Text
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UNITED STATES 8'
NUCLEAR REGULATORY COMMISSION X
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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION UNIT *3 AMENDMENT TO FACILITY OPERATING LICENSE
- t Amendment No, 63 License No. DPR-25 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for, amendment by Commonwealth Edison Company (the licensee) dated January 11, 1982, as supplemented by letters dated January 21, February 23 and March 22 and 29,1982, and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set fcrth 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 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 Cc miss. ion's regulations and all appitcable l
requirements have been satisfied.
2.
Accordingly, the license is amended by changes to. paragraphs 3.B and 3.L.2, 3 and 5 of' Facility Operating License No. DPR-25 to '
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read as follows:
B.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 63, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
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5 3.L.2 The Minimum critical Power Ratio (MCPR)SafetyLimitwill be increased 0.03 (TS 1.1.A and 3.3.5C) 3.L.3 The MCPR Limiting Condition for Operation (LCO) and Figure 3.5.2 will be increased 0.03 (TS 3.5.K) 3.L.5 The APRM Scram and Rod Block Setpoints and the RBM Setpoints shall be reduced by 3.5% to read as follows:
T.S. 2.1.A.1 S<(.58 WD + 58.5)
T.S. 2.1.A.l*
Sl(.58 WD + 58.5) FRP/MFLPD T.S. 2.1.B S<(. 58 WD + 46.5)
T.S. 2.1.B*
S<(.58 WD + 46.5) FRP/MFLPD T.S. 3.2.C.
(TAFLE 3.2.2):
APRM Upscale <(.58 WD + 46.5) FRP/MFLPD RBM Upscale 5(.65 WD + 41.5) 3.
This license amendment is effe'ctive as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION M
Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing Attachraent:
Changes to the Technical Specifications Date of Issuance: April 29,1982 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 63 FACILITY OPERATING LICENSE NO. DPR-25 DOCKET NO. 50-249 Revise the Appendix "A" Technical Specifications as follows:
Remove Replace 1
1 2
2 5
5 6
6 6A 7
7 10 10 11 11 11A 13 13 14 14 15 15 15A 16 16 18 18 19 19 20 20 21 21 22 22 22A 34 34 36 36 36A 42 42 42A 42A 46 46 58 58 61A 61A 62A 62A 62B 628 63 63 64 64 65 65 78 78 81B 81B 818-1 818-1 81C-1 81 C-1 81C-2 81C-2 81C-4 81C-4 81D 81D l
4 A
4 i
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S 4
2-i 82 82 85A 85A 658 853 858-1 86A 86A 109 109 110 110 I
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1.0 DJJ ~.TIONS
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The succeeding frequently used terna are ex-placitly defined so that a uniform interpretation C.
Crit ical Power it.it io (CPR) - The critic.el
{-
of tha specifications may be achieved.
Iwwer rat io is the ratio of that assembly A*
power which causes sinne point in the assembly to experience transition boiling (Deleted) to the assembly power at the reactor i
- l condition of int erest as calculated by, application of the Xil,3 correlation.
(Reference XN-Ni'-S l ? )
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D.
Ilat Standby - Hot standby means operation with
! i the reactor critical, system pressure less than 600 psig, and the main steam isolation valves closed.
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i E.
Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.
lj F.
Instrument Calibration - An instrument cali-i bration means the adjustment of an instrument j
signal output so that it corresponds, within ac-ceptable range, and accuracy, to a known value(s) of the parameter which the inst rument monitors. Calibration shall encompass the
't entire instrument including actuation.larm,
- l or trip. Response time is not part of the routine instrument calibration, but will be I
checked once per cycle.
I G.
Instrinnent Functional Test - An instrurant functional test means the injection of a simu-lated signal into the instrissent primary sensor 3.
Alteration of the Reactor core - The act of to verify the proper instrument response j '.
alarm, and/or initiating action.
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'I-moving any component in the region above the core support plata; below the upper grid and within the ehroud. Normal control rod move- -
H.
Instrument Check - An instrument check i s sinalitative determination of acceptable oper-ment with the control rod drive hydraulic ability by observation of instrument behavior system is not defined as a core alteration.
l during operation. This determination shall is clude, where possible, comparison of the
- j instrument with other independent inst mments j
measuring the same variable.
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Amendment No.12, 63 g
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for opeestle.
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fyerable - a syetas. swa.sy ne.. te en s. es.p
.t, e, gogee gi.6;.;.g ci Jiilasi fir opidelon'_s[ltiil. ne
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al long_re.Jeale..
- .a ll ** eres el*3 e 'd'a 8 4 8e (*r*4'te er perteeneng 4to pWitr the se. M al d beiellen M. W irit In Hele detengsten ohen
'I pgelen e ecceftalete levels of systte peelsen-be else samespt ten tenet el3 emeteet.we y attenue,et go.sg.
- j i esc e pec e s s a ry t o a s s us e 6.f e S t.se t engo e nJ oge-
.e tlese, centrale, noei.e4 eed e=trs=*.cy eteetricot p. wor 4
eseglee of the fecllity. bleen Gliese cea.lilleas moun ces, seeling er east w.eter tut.rlcet te.se er elleer ese met. Ilse plaset case lie sapetsIcel safely enJ eues t seey e palemeest tenet as e reystre.t toe ts.e eyetes,
- ul**r e t e s to ute. e.unt= nent er emele s to per $ se i t.s l.
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.6-senet s tessan tens can les saf ely cent sellcJ.
synctIna( ) are else cepal.le of perfspelng Lanetr related support A.ctlen(el El=letas afety'syste= Srlling.(lam ). n e i
s I M " !.aA - o "s H as
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- e rM r
u.itt.g Tefany sysice set'sl'gs Ire ietsings e P.
n.c t i on s s "'a"t e r e.r. i t a.' e.e.or.""'"8 "' '""'
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- l. e..ent at tow i.l.ich int e let e the atit e.ut ic yeetece tte acelen et a level sesels that the safety
- i Itale s will not 8.e acces.itel. % e seglen 4
presaelag ycle,.laterwel between the end has een the safety limit enJ these swellegs o' ene se cling eistogo anal the end of the espresene s mergin with steenal operet ten tylog nest sealise.peemt refuelleg outage.
f below these sett amas. % e margin has base i
ll eeena.Innsi J se that with proper eyeestion of the 3,
primary rentelr.nsne lategrity - Primary leessionsniet tee the safety lletts will never be 7entilnTefng gestegrtty cani~ tliet she Jrywell
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sece4JcJ.
and pressiere seappresslee chaea.or are intact
'. l K. Pract.lon of Limiting Power Density (FLPD) enJ ell of the f ollowing con.Iltleas are satisfl Jl l
- i For finel fabricated by GE, t,he f rac tion g,
as
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,3 co.gege,e.t fieletgo.,el,e,,,
t, of limit.ing power density la the ratio lines co.a.,tle, le the reacte, coele.. eye.
of the Linear llent Generation 1(ate (LilGR) tem er contalement which are not re.pelred ij existing at a given locat.Lon to the to be open Juring occlJemt condittens are l!
design IJIGit for titut blindle type. f'LPD slescJ.
does not, apply to ENC fuel.
2.
At least one Joer in each elrlock is closeJ
' "'I %**I*E' gegt e f.y s s ta r n.c t l en Test. A log ic s y s -
l.,
m test of all relays iemliusci ssetia'l 't'es t seasT' eie es.cral.le er hectlweted in th laolete.
e i cJ Jew c I
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- l en s F**g'I*d-e, operehte per Jesign latent.
Wl'e's Possi.
ble, actlpe well go to casaplelloa. l.e.
peeses wil be started sad valves openeJ.
s n)E E 4'M M d ii'.' $ a" M J *
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le"ise_in" ~ m t-i-l' " i"
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t e el..
.sL nnaoog f.esi
,,,,,,,,,,,,,,,,,e,.,,,,,,,,see,,,,,,,_
j sessel.ly la t!ss core, g,
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11. se.J.. The react ee so.le is that whleh is
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- gun g. cal resguised to generate
~estiklished by stie e2Ju-selector-switch.
en.1 transent te e e rly systen a single trip j
signal relateJ to the plant parameter 3
j sunnitercJ by glast lastreament channel.
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i' Amendmeent ko. SN 63 i
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I 2.1 LIMITING SAFETY SYSTDi SE'ITING j
If 1.1 SAFETY LIMIT f
2.h.
FUI.1. CLAUD1'*G 15TrCillTY jyrt. C!.t.D31::C litTI:CRITY.
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~ p1 l ea'a t I i ty,
_Applienbility l
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A The 1.iuiting Safcty System Settings 1he Safccy Litr.its established to apply to trip settings of the instru-l
- 1 p r e..c ccc t h.: fuel cladding integrity ments and devices which 1rc provided variatnics uhich' l
ap,,1y to these to prevent the fuel cladding integ-c.onitor the f uel theimal behavior.
rity Saf ety Limits f ro:a being ex-cceded.
'i Objective _
,Ohj e c t iv e_
The objective of the Limiting Safe-The objective of the Safety Limits ty System Scttings is to define the i
is to establish limics'below which.
1cyc1 of the process variables at lt the Intenrity of the fuct cladding which auto..atic protective action is pre;ctved.-
is initiated to. prevent the fuel clad-s ding integrity Safety 1.imits from being cacceded.
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l Specifications
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^* ""
A.
i'cactasr Pressure >800 psig and Core
,[ lee > 107. of Rated.
The liiniting safety syste:a trip sectings shall be as specified i
The existence of a minimum critical below:
l power ratio (MCPR) less than 1.06 for GE 8x8R fuel, or less than b.05 l
for ENC or GE 8x8 fuel, shall constitute I
violation of the MCPR fuel cladding f
integrity safety limit.
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,unendment No. M, 63 3
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2.1 LIMITING S/JETY SETE. S TTIXG 1 :
i 1.1 SAftTY LIMIT I
1 APEM Flux Scran Trip Setting (Ren Mode)
When the reactor made switch is in the i
run position, the APN1 flux scram setting f )
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. shall bei i
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S6 58vg + 62-i
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with a maximum set point of 120% for core flow equal to 98, x 10' lb/hr and greater, wheres
- j S = setting in per cent, of rated power V " Por cent of drive flow required to produce i'i D
a rated core flow of 98 M1b/hr.
In the event of operation of any fuel assembly fabricated by GE with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
r--
t Wheres S 4 (.58WD + 62) L PD
,Il FRP = fraction of rated thermal power (2527) MWt)
HFLPD = maximum fraction of limiting power density for GE fuel i
The ratio of FRP/MFLPD shall be esat l
equal to 1.0 unless the actual operating value is less than 1.0, in which case the ectual operating l i, value wilL be used.
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Amendnu?nt No. E/, 63 f'
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i DR_I'- 2 5 1.1 SAFETY LlH1T 2.1 LIMITING SAFETY SYSTEM SETTING This ad,)ustment may also be performed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, which accomplisnes the same degree of protection as reducing the trip setting by FRP/MFLPD,.
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2.
APRM Flux Scram Trip Setting (Refuel or Startup and llot Standby Mode) ii When the reactor mode switch is in the refuel startup/ hot standby position, j
the APRM scram shall be set at less than
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or equal to 15% of rated neutron flux.
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6A Amendment No. 63 j -
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,q DPR-29 i
l l1.1 *SAFM Y LI; LIT 2.1 LIMITL'E SA?nY SYSTEM SETTDiG l
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- 3. Cor's Thermal Fower l'.f utt (Reactor _
.: s 3.
IRH Flux Scr: m T--in Settit e j 'g l
' Pressure g 800 psig) f q'
The IRtt fins scram setting shall be Wen the reactor pressure is < 800 set at less than or esiumi to 120/125 et Psig or core flow is less than 10%
full scale.
of rated, the core thermal power
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shall not exceed 25 percent of rated thereal power.
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3.
APRM Rod Block Setting C. Tc:ect Translent The APRK rod block settinC shall be!
1.
ne neutron flux shell not exceed the scran
- i setting established in Specification 2.1.A f or longer than 1.5 secunds as indicated by sf
.58WD + 50 the process computer.
. ne, definitions usel above for the AFFJf,
- j
- 2. 'llhen the process computer is out of service.
seven trip a;> ply.
this safety limit shall b e as s snoe d to b e In the event of operation of any fuel assembly
'4 i
esceeded if the neutron flux exceeds the coram fabricated by GK with a maximum fraction setting established by Specification 2.1.A limitiug power density (MFLPD) greater than the fraction of rated power (FRP), t.he setting and a control rod scram does not occur, shall be modified as follows:
S 4 (.58WD + 50) f3F.PD R
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D.
Reactor if ator Level (Shutdown Condition 1
,7 trip apply.
Wenever the reactor is in the shutdcun condition 4
with trradisted fuel in the reactor vessel, the
'l'h e ratto of FRP to MFl.PD shall be set equal to 1.0 unless the actual operating value water level shall not be less then that corres-is less than 1.0.
In which case the actual PonJing to 12 taches above the top of the active Operating value will be used.
'the adjust-fuet*when It is seated in the core.
ment may also he performed by increasing I t he Al'RM gain by the inverse ratio, MFI.PD/
- Top of ar.tive fuel is defined to be 360 inche s above vessel zero (see FRP, which accomplishes the same degree of D'""2 3 2)-
protection as reducing the t. rip setting by 7
' F9.P/lFLPD.
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Amendment No. N. 63 a
1.1 Safety I.imit liases t
FUEL CIADDING INTEGRITY The fuel cladding integrity limit is Sa fet.y I.imi t is defined with margin to the l
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set such that no calculated fuel dam-conditions which would produce onset of trans-ages would occur as a result of an ition boiling. (MCPR of 1.0.)
These conditions abnormal operational transient.
Ile-represent a significant departure from the cause fuel damage is not directly condition intended by design for planned j
observable, a step-back approach is operation.
The MCPR fuel cladding integrity l
used to establish a Safety Limit such Safety Limit assures that during hormal that the minimum critical power ratio operation and during anticipated operational l
(HCPR) is no less than the MCPR fuel occurrences, at least 99.9% of the fuel rods I
l cladding integrity safety limit.
In the core do not experience transition i l HCPR) the MCPR fuel cladding integrit y boiling.
See reference XN-NF-524.
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safety limit represents a conservative margin relative to the conditions A.
Reactor Pressure > 800 psig and Core required to maintain fuel cladding Flow > 10% of Rated l
l integrity by assuring that the fuel
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[ does not experience transit. ion boiling.
Onset of transition boiling results in a decrease in heat transfer from the clad The fuel cladding is one of the physical '
and the possibility of clad failure.
- However, and, therefore, elevated clad temperat ure 7
barriers which separate radioactive materials from the environs.
The the existence of critical power, or boiling integrity of this cladding barrier transition, is not a directly observable is related to its relative freedom parameter in an o;)erating reactor.
Therefore, from perforations or cracking.
Although the margin to boiling transition is calculated i
some corrosions or use related cracking from plant operating parameters such au core may occur during the life of the claddinge power, core flow, feedwater temperature, and j
fission product migration from this core power distribution.
The margin for each 1
source is incrementally cumulative fuel assembly.is characterized by the critical and continuously measurable.
Fuel power ratto (CPR) which is the ratio of the cladding perforations, however, can bundle power which would produce onset of i
result from thermal stresses which boiling divided by the actual bundle power.
occur from reactor operation signi-The minimum value of this ratio for any ficantly above design conditions and bundle in the core is the minimum crit ical I
,l the protmetion system safety settings.
power ratto (MCPR).
It is assumed that l
While fission product migration from the plant operation is controlled to the i
cladding perforation is just as measurable nominal protect ive setpoints via the l
l1 as that from use related cracking, the instrumented variables.
(Figure 2.1-3).
l thermally caused cladding perforat-lon li signals a threshold, beyond which still The MCPR Fuel Cladding Integrity Safety greater thermal stresses may cause gross 1.imit assures sufficient conservatism in the rather than -incremental cladding det er-operating HCPR limit that in the event of iuration.
Therefore, the fuel cladding an anticipated operational occurrence from the limiting condition for operation,4.t a
Amendment Nn #, 63 in fj~
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Safet y Limit Bases
)
1.1.A Heactor pressure > 800 psig and Core Flow > 10% of Rated.
(cont'd) least 99.9% of the fuel rods in the Integrity Safety Limit there would be no trans-l core would be expected to avoid it ion boiling in the core.
If boiling transition boiling transition.
The margin between were to occur, however, there is censon to C-calculated boiling transition (MCPR=1.00) believe that the integrity of the fuel would
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and the MCPR Fuel Cladding Integrity not necessarily be compromised.
Significant
- 1 Safety Limit is based on a detailed test data accumulated by the U. S. Nuclear N'
statistical procedure which considern Hegulatory Commission and private organizations the uncertainties in monitoring the indicate that the use of a boiling transition core operating state.
One specific limitation to protect against cladding failure uncertainty included in the safety is a very conservative approach much of limit is the uncertainty inherent the data indicates that LWR fuel can survive p
in the XN-3 critical power correlation, for an extended period in an environment of f
Refer to XN-NF-524 for the methodology transition boiling.
used in determining the MCPR Fuel Cladding Integrity Safety Limit.
If the reactor pressure should ever exceed the limit of applicability of the XN-3 i!
The XN-3 critical power correlation critical power correlation as defined in la based on a significant body of XN-NF-512, it would be assumed that the practical test data, providing a HCPR Fuel Cladding Integrity Safety Limit high degree of assurance that the had been violated.
This applicability pressure critical power as evaluated by the limit is higher than the pressure safety limit i
correlation is within a small per-specified in Specification L.2.
For fuel
,1 centage of the actual critical power fabricated by Cencral Electric Company, being estimated.
The assumed operation is further constrained to a m.iximum j
reactor conditions used in defining linear heat generation rare (LilGR) of 13.4 kW/ft the safety limit introduce conse.v-by Specification 3.5.J.
This constraint is atism into the limit because established to provide adequate safety margin
- r boundingly high radial power peaking to 1% plastic strain for abnormal operational factors and boundingly flat local transients initiated from high power conditions.
I peaking distributions are used to Specification 2.1.A.1 provides for equivalent il estimato the number of rods in safety margin for transients initiated from hoiling tiran s i t ion.
Still further lower power conditions by adjustina the APRM conse'rvatism is induced by the flow-biased scram by the ratio of FRP/MFLPD.
i tendency of the XN-3 correlation Specification 3.5.J establishes the maximum j
to overpredict the number of rods value of LilGR which cannot be exceeded during in Lolling transition.
These steady power operation for GE fuel types.
l conservatisms and the inherent accuracy of the XN-3 correlation For fuel fabricated by Exxon Nuclear Comnany, provide a reasonable degree of
( F.NC) fuel design criteria have been established I
assurance that during sustained to provide protection against fuel cent erline 11 operation at the HCPR Fuel Cladding melting and cladding strain, ENC has. performed Amen <iment No. 4, 63 0 f
s
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i Safety Limit liases l
1.1. A.
Reactor Pressure) 800 psig and Core Fuel > 10% of Ratett.
(cont'd) fuel design analysis which demonstrate that centerline melting is not predicted to occur during transient overpower conditions throughout the life of the fuel, Protection of the MCPR and MAPLilGR limits and operation within the power distribution assumptions of the fuel i
design analysis will provide adequate protection against centerline melt and ensures compliance with ENC's clad overstrain criteria for steady state and t ransient operation.
Since ENC's I
design criteria are more conservative than the 1% plastic strain limitation on CK fuel, the I.llGR limitation and APRM scram adjustment for CE fuel established i
in specifications 3.5.J and 2.1.A.1 respectively are unnecessary for the protection of ENC fuel.
The procedural controls of specification 3.1.It will i
ensure that operation of ENC fuel remains 4
within the power distribution assumptions I
of the fuel elesign analysis.
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!I Amendment No. 63 11a s
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- ij 2.1 L1mitinst Sarcty System Settintunes i
f 1.1 screty Limit Ibsca
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1.1.C Power Transient (cont'd)
FtJEL CLADDING INTEGitITY,
- 4 The.cbnormal operational transients
'Jhe computer provided han a ce iuence annuncia tion program *which applicable to operation of the Units will Indicate the sequence in which have been analyzed throughout the scrams occur such en neutron flux, spectrum of planned opereting cua-pressure, etc.
This program also dItions up to the rated therm 11 power 2nstectes,. hen tha scram setpoint 13 cond it ion of P'}**7 l~.Wt.
In additicn, cleared.
'Ihis will provide information 2527 bi'.It is the licensed maximus, steady on hot. long a scra:n condition ex tats state power level of the units. This cod thu, prov1dc nome measure of the maximum steady-stste power level will pII-IIf-7 9 - 71cver knowingly be exceeded. Seu reference energy adited durtnc a trenntent.
- Thus, cc: puter Information normally will be avellcble for analyzing scrams; h u s. -
ever, if the coriputer Informition abould ie not he available for any scram analys13, i ;
Speelfication 1.1.C.2 will be relled on to deterntnc if a carcty limit has been violated.
j.
During periods when the reactor is shut down. consideration must also be given to water level retutrenents due to the crfect of deccy heat.
1r reactor water level should drop below the top of the cctive ruel uurler, this tine, the cb:11t.y to cool the core la reduced.
Tals reduction in core coo.11ns cap-c*s titty could lend to elevated cladding te.perat.uren and clad perforation.
The core will he coolrd nurric1 cot.ly t.o pre-venFMiad rnelting nhould the water level be' reduced to tso-thirda the ccre helr.ht.
T.strblishr.ent of the safety limit at. 12
)
1:sches chove the tn;> of the rucl'provbics i
cale tuat.c eccreto.
"Ihla level w111 he con-Linuouily monitore!i.henever the recir-culet.1oa pu.'pa are not operatins.
t l.* Top of active fuel is defined t6 he 13 360 inches above vessel zero (see II.uae s 3. 2).
Amendment No, y, 63 4
,2 1
't i
Limiting Safety System Setting llases 2.1 FUEL CLADDING INTEGRITY (cont'd) 3 If
' ' *"F' 'h' 'h"*I ****'9"ences er Conservatism is incorporated into the the trans! cats, the HCFR's stated les darnIragdeonwratiosi transient analyses which define the 3.5.K as the Ilmitin9 can'titia o
HCPR operating limits.
Variables which bound those which are conservatively assumed inherently possess little or no to exist prior to initiation of the transients.
l uncertainty or whose uncertaint.y has A.
Neutron Flux Trip Sett1nsts
+*
'f little or no effect on the outcome I
of the limiting transient are selected APnN Flux Serem Trip Settinst_(Run Made) 1-at bounding values.
Variables which possess significant uncertainty that The everage power rance monitoring i'
may have undesirable effects on thermal (APHM) system, which ja calibrated margins are addressed statistically.
using heat belance data taken ducirq r
Statistical methods used in the steady-state conditions, reads in i
transient analyses are described in Be-percent or rated'ather nal power.nbers pro r:.1c the baaie i
XN-NF-81-22.
The MCPR operating cause fleston ch limits are established such that the 1 iput s tSnals, the I.PIGl syste:a responds occurrence of the limiting transient dtrectly to overage neutron riux.
l will not result in the violation Dur:n3 transients, the instantaneous l
of the MCPR Fuel Cladding Integrity rate or heat transfer from the fuel
~
Safety Limit in at least 95% of the (reactor ther.nal pouer) is less than random statistical combinations the instantoncous neutron flux due to of uncertainties.
In general, the t.he time constant of the fuel. There-variables with the greatest statistical fore, during abnormal operational significance to the consequences of transients, the thermal power'or the anticipated operational occurrences are fuct will be less than that indlcated the reactivity feedback associated with by the neutron riux at the scram setting.
the formation and removal of coolant Analyces demonstrate that, with a 120 voids and the timing of the control percent scram trip settine:. none or the j
abnormal operational transtentsian:1yzed l
rod scram.
elolete the fuel Sorcty I.letic and there 1s a r.ubs tantial mare.In trors riial J. rose.
q Therefore, t.he une or flott r.'rcreetc ed "g
. Steady-state operation without,.rorced scram trlp provides ever cdelition-1 r.r5*.A.
d.
arecirculation will not be permitted, r-e
excep', durins startup testing. The enalysis to support operation at various power and rios relctionships i
Ili has considere'd operation with either l
one or two recirculation puseps.
t The bases for individual trip eettings are discussed in the followin5 para-14 graphs.
'g Amendment No. H, 63 l
.a t
i r
t reducing the trip setting by FRP/
MFl.PD by raising the initial
- 1.1. A. Neutron Flux Trip Settings Al'RH reading closer to the trip setting such that a scram would be
- 1. APRM Flux Scram. Trip Setting, received at the same point in a (Run Mode) cont d) transient as if the trip setting g
An increase in the APRM scram trip
~
setting would decrease tiie margin 5.
Ah.OlutSetsmTri setting present before the fuel cladding Ih fuel _or Stort as lot 5tehdby Mode)
'l integrity Safety 1.imit is reacl The APRM scram trip setting wasled.
.i gg, determined by an analysis of margin'..
the reestor le sk low pressure, the APE 4 required to provide a reasonable teras setting of 15 percent of rated power range for maneuvering during operation, provides adequate thern:21 margin between the 8
Reducing this operating margin would the setpoint and the anrety liatt, 25 p er-increase the frequency of spurious etsit or rated. The margin is ade.4uete to scrams which h.4ve an adversa effeet
,,,,axnsda te antic ipated noneuvers ascociated on reactor safety because of the With pos.er plant startup. Effects or in-4tcustn't presuure at zero or low void con-
- l resulting thermal stresses.
- Thus,
- ent are t.itioe, cold water fron sources the APRM scram trip setting was e e e t trble dyr!.qt =tartup is not much colder selected because it provides adequate han that already in t.he system, tempera-margin for the fuel cladding integrity ture coefficients?are small, and son.
Safety 1.imit yet allows operating trol rod patterns are constrained to i
margin that reduces the nossibility be uniform by operatins procedures of unnecessary scrams.
back:d up by the rod worth minir.ttrer.s The scram trip setting must be Of all possible so.:rces of reactivity adjusted to ensure that t he LilGR input, unifor;n control rod withdrawal I
13 transient peak fo. G.is, fuel the most probable cause of signift..
l l
cant power rise. Eccause the flux is not increased for any combin-atton of Maximum Fraction of distribution ascoc:eted :lth uniform rod with trewala tio:s not involve high 1.imiting Power Density (MFl.PD) and reactor core thermal power, local peaks, and because several rods The-sepam setting is adjusteil in nust be nr.oved to chance power by a accordance with the formula in a lgnificant percent.3;;c of rated power, the rate of power r;se is very slow.
speci ficat ion 2.1. A. I. when t he Cencrally, the heat flux is in near MFl.PD is greater than the fraction equilibrium with t!e fission rate.
In of rated power (FRP).
an assumed untror:a rod withdrawal ap-The ad.}ustment may also be proach to the scre= level, the rate of accomplisheal by increasing po er rl:re is no =:re than 5 percent the APRM gain by the reciprocal or reted power per minute, and the of FRP/MFl.PD.
This provides the AP.Vi system would be more thsn adequate 15 same degree of protection as Amendment M
- 63
0 'l l
2 2.1.A.
Neut ron Flux Trip Set ting Motle )( Re fue l 2.
R RH Flux Scram Trip Settiny (cont (1) or Start anal llot S t analby
+
j to assure a scram before the power could exceed tin safety limit. The 15 percent A?.*1M scram re?.2 ::13 active un-til the = ode a*2!teh is pIrced in the RU;; position. 'Ih:2 criten occurs whert reactor pressure *s Creater than 850 psig.
I 3.
IIH4 Flux Sc'ren Tr!9 Setting The IAM system consists of 8 chambers, 4 in each of the reactor protection system loc!o chanc.els. The I.'IM is a 5-decede instrument which covers the ranse or power lev:1 between that covered by the Sir; and the APR;h The
$ decades are broken dcan into 10 ranges, i
,i each being one-half of a deccde in aire.
)i 15a
-I Mendment flo. 63 ij'
,f
.q f
i 4
2.1.A.
Keutron Fluz Trip Setting
'1 The 1Ri4 serem trip settin5 of 120 I
B1 visions is active in each cance.or Reactor power level may be v6fted by the IP.4.
For exos.ple, if the instru-moving control rods or by varyint the reoirculation flow rate. The APRM
- l neat were on rance 1, the scram cetting would be a 120 divisions for that r.anges system provides a control rod block 'to if the instrument were on range prevent gross rod withdrawal at coastant recirculction flow
- 13kewlae,
- 5. the scran would be 120 divis tons onThus, as the Illit is ranged
,,g,g,p7,gc,g,,210st crossly exceed-that range.
up to occo:sodeste the increase in power ing the MCPR fuel cladding inteqrity level, the screm trip setting is also satety limit. This rod block trip. setting, which is auto-I ranged up.
ratically vcried with recirculation The rost significant,Dources of reac*
loop flow rate, prevents-an increase in the reactor pruer Icvel to c::er s-t1vity change during the power increase stve values due ts control ros with-In cre duc to contest rod withdrawal.
order to ensure that the Illi4 provided d raw:41. The flow vcriable' trip actting' a
l edernte protection against the single provides substa:ittel margin from fuel l'
rod t41tisdrawc1 error, a range of rod danage, assu:.jng a stecdy-stato cpera-This g3on gg.he r.r19 sett:nr., over ti.e wIthdrcual accidents wris en:aly:cd.
analysts included starting the cccident entire recircolation flow range. Thg et vsrlous power levels. The most se-margin to the Safety L1: nit incre:)sco es were ecse involres an in$ttal condition the flow decreancs f:r the specified in which the reactor is just cubcriticcl trip se'tting versus flow relattenship; i'l rnd the I.T4 cystem is r.ot yet on sc.:le
- therefore the *.4 erst case l'.CP!! which I'
ceuid occur dur ste:dy-state opera.
Additiosial cor:servatism uns taken in this
. tion is e t 1004,ine; or roter1 thererl potter ent,1ysis oy assumitig that the Illit channel beccuse of the APRM rod block trip closest to the withdrawn rod 1s byptssed.
a ctt in,.
4he actual power diatt it:utton l
The s.esulcs of thta a:ielysis shou that the in the core is catnblished by c; ecified l
' reactor la scrar. :cd and peck power 11 Ited contral red sequcnces and is monitored i
maintaining to o:ee percent of rated porer, tnus centinuously by the in-core 1J'M. Syste:2.
ECPR above the PCPR fuel cladding integrity
- IU
- IUU I
salety limit. Based on the above the APAM 'co ! ulgck t:1p setting ta e'd-carilys*s, the Ill:3 provides protection s';ainst
.luated downward or APRM gain increased locc1 control rc,d withdre a1 crrors and con-ir the mnximum fraction of 11:altind tie o os ithdran t of control rods f ra ce.:*>*"sce c.r.: proviuua but:kup protectioni for the AP!'-1 power de:nsity for G.E. fuel exceeda the fraction of rated power, t b sa pre-f serving the APRM rod block safety f
Amendment No. p, 63 margin.
16 I.
1 4:
q Reactor Ceolant Low Pressure Initiates Main Steam _
C.
j Turb1po Step Velve Seren - The turbine stop valve Isolation valve closure The low pressure isolstica i
3,
~c}osure ncran trip caticipates the pressure, at 850 psig was provided to give protection against r.;utron flux at:4 h;at flux incresco that could fast reactor depressurization end.the resulting result fro rr.pid c3csu:s of the turbino stop rapid cooldown of the vessel. Advantage was taken 4
With a scran trio cctting of 10 of the scram f eature uhtch occurs wisen the main valv:3 percent of valvo cbsuru fron full open, the steam line isolation valves are closed,te provsde resultant ir. crease sn surface heat flux la f or reactor shutdown so that operation at acessures linited such that I CPR rernins above Ithe' MCpA er t una thou specified in tfie therns1 *hyderulic fuel cladding integrity safety limit, even safety 11: nit does not occur, although operation during the worst case transient that assu==s at a pressure loi.cr Lt.an 850 psig wooid not necessarily t-the turbine bypass is cloaed.
constitute an unsafe condition.
i;
'he H.11n stears Line Isolation valve closure scret M.
1ow pressure isolation of the main steam lines at 850 psis was provided to give protection against Ceneratar Imad Relecticei scram - The genera.
rapid reactor depressuiitation and the.resulting tor loed rejection scram is provided to rapid cooldown of the vessel. Advantage was,taken i;
i F.
~
i enticipate the rapid increase in pressure of the scrom feature which occurs when the main
.cnd neutros flux resulting f rom.
steam Itne isolation valves are closed, to provide fdt closure of the turbine control valves for reactor shutduun so 'that high power operation I
due to a load rejection and subsequent low reactor pressure does not occur, thus providing at for the fuel cladding integrity safety fallu:e of the byyttss: 1.o., it prevents protectio j
Operation of,the reactor at pressores lower a
CTr2 fre a tecoialns less'than the MCpR fuel 1t=1t.
than 550 pstg ueguires that th,e rescror.mo.le switch cladding integrity safety limit for this transient.
For the load rejection without be in; the startsp, position yhcre trotecctort of the
,j+
s bypass transient from 100% power, the peak fuel ciudding integrity safety Jiutt is provided by the rIRn hir,li neutron flux scr.m.11 bus, the coat,Instion lj
,J heat flux fand therefore t.lfGIO increases on it6e,Iow pressure isointion.and isolation l
gin f..
the order of 15% which provides wide mar $rline of seala steem p' valve closure scrass assures thrj evaflabilf ty 'el 4
lhel cent
[
l to the value *:orresponding to
.I neutron flux screa proteccion over the entire r:elt,1ng anti 1% clatbling ntrain.
j rangeofapoIled.silityofdhetfueijcladding integrity t
- In addition,'d.c isciatten valve j
l' d
safety limit.
closure scram anticipates the pressure and'llux l
transtents which occur daring norr at or inadvertent
/
{/
1 solation valve closure.' bith thd. screns set at
^
<l 10% valve closure,tliera is. no appreciable increase 6
t in neutrun fl'ux.
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4 1
r Amedment No. 47, 63
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g' If 2.2 LIMITlHP. S AFETY 'SYSTilt St'TTING l.2 SAITTY LIMIT
~
2.2 NEACTOR cnotAHT SYSTD4 REACTOR COOLANT SYSTEH_
1.2 gplicahitity:
gplienbility:
Applies to trip settings of the instruments and Appites to limits on resctor coolant system dovt ca wlitch are provided to prev.nt the reactor synten s.if ety ll=1ts f roes I,elng esceeded.
i pressure.
% leetIve_:
-(mjective To define the level of the process variables at To. establish a limit below which the. integrity which automatic protective action is initiated to the reactor coolsat system is not thseatened prevent time safety limits f rom being exceeded.
olidue to an overpsessure condition, SpecIIIcatlon:
Spest fIcat ton _t Reactor Coolant High pressure Scraa shall be,
A.
1he reactor coolant system pressure shall not eto60 psig.
exceed 13'19 pssg at any time uhen irradiated fuel l
is present in the reactor vessel.
primary System Safety Valve Nominal Settings 3
f
\\
stialt be as follows:
l l
I valve at 1)is pstg*
I 2 valves at 1240 psig 2 valves at 1250 pelg 2 valves at 1260 pais II 2 valves at 1260 pets 4
The allmeable seapoint error for each valve shall be 412.
e
,l
- Target Rock combination safety / relief valve g
i 19 Amendment No. A7, 63 t
I i
is a
l I
SeMet than 26,700 psi at an internal pressure 3.2 ne ve.ct.e e tant erstem sateerter se.. s.per-of 1250 psig:
this is a factor of 1.5 tant barra.e se th. prevents.n.c
- tr.13 4 re-below the yield strength of 40,100 psi te....t it..t.= pr.4.ct..
It te..unatet thee th*
at 575*F.
At that pressure 1imit of sneesetty et thi. ty te. Se protected by
.t. hts.ht.g 1375 PsiR' the Reneral membrane stress a pren.re it.it t. We eb erved t.c.11 eppettes will only be 29,400 psi, still safely l -
ceaatta.no end whenever there 1. tre.41sted to t e.
below t.he yield strength.
the react.r ve.ses.
%e ations N s M stm as b da to ne preuer. orety 11 1 et 1325 pets.==n r.4 yield strength are comparable for the j
I nr it.. unet etene.p c. prese re saatenter se primary system piping and provide a equivalent to 1375 pets oc the I
.t elev.at.a et tsne similar margin of protection at the r u ctor c.. tent erste. ne 1375 pata..
s.
established safety pressure limit.
j dertud tree the seeigi pree.oren er the react.c pressure was.el, c enant system piptng and t t The normal opnadng pusswe d de st. c denser. ne weepne tv. desar,= preneure.
reactor coolant system is 1000psig.
are taso ests et 575'r,117s pets et 160*r..no 1254 For the turbine trip or lo w of l
psis et 575*F. He presswee sarety 15.it weg ch.ee electrical load transients, the turbine
.. the 1.wer et the pres.ure transiente per.ttted l'
hy e.. epptteable dents,i c.4es: Amics.tser a
trip scram or generator load relection together with the turbine bypass prum.ce ve.ut c.de, secaten att ter the pressure
- scram, suset. 4 a...st c..eenser..a t sas sit.: c.4*
system, limit the pressure to approximately
{
rcr the react.r c tant erstes persas, ne astz 1100 siE (2).
In addition, pressure E
.!!er ar.4 Fressure Vesset C.ee pernite preza see relief valves have been provided to e
tr.
tente v7 t tar.pr entaa pren,re (lict reduce the probability of the safety x 1750 - 1175 pets), and the us>.si c,4e pernet.
valves, which discharged to t he drywell, prenure trea sent.., t. :ot var th-de.:
seenure (120: x 1175 - 14to psic). ne sarety operating in the event that the turbine J
Limit pressur et 1373 pets 1. rererenced to the bypass should fail.
lowest elevation of the reactor vessel.
Finally, the safety valves are sized l-The design pressure for the recire. suction t o keep the reactor vessel peak pressure [
line piping (1175psig) was chosen relative below 1375 psig with no credit taken to the reactor vessel design pressure.
for.the relief valves during the j.
Demonstrating compliance of the peak postulated full closure of all MSIV,s vessel pressure with the ASMis overpressure without direct (valve position switch) protection limit (1375psig) assures scram.
Credit is taken for the neutron compliance of t he suction piping with flux scram, however, t
the LISAsi limit (1410psig). I? valuation metrhodology used to assure that this The indirect flux scram and safety valve unrety limit pressure is not exceeded actuation provide adequate margin for any reload is documented in Heference below the peak allowable vessel pressure l
X ti-tlF 71.
The design basis for the of 1375 psig.
reactor pressure vessel makes evident the is continuously monitored o
i cactor Iiressure substantial margin of protect ion against-in t he control room during operation on failure at the safety pressure limit of a 1500 psi full scale pressure recorder.
1375 paig.
The vessel has been designed 1
for a general membrane stress no greater T4) SAR, Sect ion fl 7.2 -
also:
"Dresden 3 Second Reload License N Submittal," 9-14-73 also:
"Dresden Station Special Report tio. 29 Supplement B."
t W ndment No. f2<., f>3
D.-..
.j.
^
seses:
2.2 In temptIonce with Sectien III er ahe Asnt code, the serety velvee swet be set to open et me htsher then lon.c desse. ere.. ore..-a tiey e t liste the reacter prenewee to me more then 1801 et deelen path the neutron flux. cran en4 eefety presswee.
volve octuetCan are teeguired to prevent evertiree-esteleing the reacter preenwre weeget and ti.ve esceeding the preseere selety limit. The pseeeere scram is available as a backup protection to the di rec t. valve po:Il tlein trip : c ramn i
anti.the high flux ceram.
l If the high flux serase were to fall, a high presouro scram would occur at 1060 pelg. ' Analyses are per formed l
l nu described in rer erence Xfi-tiV-?) *(1 for each reload to assure that the pressus e narety limit in not exceeded.
4 r
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N 21 i
i Asu ndment No. A2. 63 t
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4.1
~
SUP.VEILIANCE Equtu!!!NT,
'j,1 LifettistC cot 401T10sl rot OttilATIOW 4.1 ffACT03 Fr.0Tr.CTIOM SYSTui j
, 3.{
pr.Act0R PPDTECTtos STStr;t 7polic.-b 111tv t
. g ottenbillert Applies to the survet11ance of the instrumen-f Applies to the lastressentation ed tatien and associated devices which intatate associated devices'tahtch initiate a reactor scram.
teacter seres.
Ob te et 1 en_:
gjecttvet To specify the type and frequency of
..I fe asswre the opershflity of the surveillance to be applied to the protecties i
l instruesntstfoe.
t teatter protection system.
_Spectitentient
,9v.elfteattent Insertzentation systees shell be f,
A.ftesetpoints,ntalousawneroftrip functionally tested and call' orated as' A.
erstems, miJ ntalmo ni..her of instrw-Andicated in Tables 4.1.1 and 4.1.2 f,
- e. tat chmenels that : net be operable respect twely".
tar each pe:Ittee of the reactor mode l
s-ettch sh..11 he ce ntwen in Teble 3.1. l.
Daily during reactor power operat. ion above 3.
from the 25% rated thermal power, t.he core power j
The system. response times distribution shall be checked for:
l opening of the sensor contact up toand including the opening of the trip i
exceed 50 Maximum fraction of limiting power actuator contacts shall not 1.
for fuel fabricated by GE density milliseconds.
and compared with the fraction the maximum (NFl.PD)
If durine oneration, limiting power density of rated power (FRP).
B.
fract-iron of'
^~
~
for f uel f abricated by Gli exceeds the For compliance with assumptions of the l
fraction of rated power when operating 2.
Fuel Design Analysis of overpower j..
3 above 25% rated thermal pouer, either:
conditions for fuel fabricated by ENC.
The APRH scram and rod bock settinh i
i l
shall be reduced to the values a.
given by the equations in Speci-fications 2.1. A. I and 2.1.11.
I t'
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A-mdment No. 42, 63 l-22 h
11 '
t li at t
--mm w_
~
1 I
3.1 I.IMITING CONDITIONS FOR OPERATION 4.1 SURVEILLANCE REQUIREME IT Specifications (cont'd) b.
The power distribution shall be changed such thac the maximum fraction of limiting power density no longer exceeds the fraction j
of rated power.
For fuel fabricated by ENC, operation of the core shall be limited to ensure the e
power distribution is consistent with that
!)
assumed in the Fuel Design Analysis for over-3 1
power conditions.
l s
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6 Amendraent No. 63 6
f
.kg_
d, i
a half scram and rod block condition. Thus, FortheAPRMsystradriftofokactronic j
if the calibration was perforesed during opera-apparatus is not the only consideration in de.
. )'
tion, flux shaping would not be possible.
cernining a calibration frequency. Change la power distribution and loss of chamber sensi-civitydicta Based on experience at other generating stations, drift of instruments, such as those 4
in the Flow Biasing Nework, is not significant Calibration on this frequency assures plant l'
and therefore, to avoid spurious scrams, a operation at or below thermal limits.
calibration frequency of each refueling outage l8 is established.
A comparison of Tables 4.1.1 and 4.~1.2 indicates that six instrument channels hhva 4i Croup (C) devices are active only during a not been included in the latter Table. These given portion of the operational cycle. For are: Hode Switch in Shutdown, Hanual Scram',
J-example, the IRH is active during startup and HigT Water Level in Scram Discharge Volume inactive during full-power operation. Thus, Float Switches, Hain Steam Line Isolation Valve the only test that la meaningful is the one Closurel Generator 1.oad Rejection, and Turbine l
'lj performed just prior to shutdown or startup; Stop Valve Closure. All of the devices or i.e., the tests that are performed just prior sensors associated with these scram functions Hi to use of the instrument.
are simple on-off switches and, hence, calibra-tion is not applicable; i.e., the switch is
'It Calibration frequency of the. Instrument either on or off. Further, these switches are i'
channel is divided into two groups. These mounted solidly to the device and have a very are as follows:
low probability of, moving; e.g., the switches in the scram discharge volume tank. Based on i
1.
Passive type indicating devices that can the above, no calibration is required for these be compared with like units on a con-six instrument channels.
t i
~
l,
tinuous basis.
B.
The HFLPD for fuel fabricated by CE ahall l
l 1
2.
Vacuum tube or semiconJuctor devices he checked once per day to determine if l
- t' sensitivity.
This may normally be done by checking the LPRM readings TIP traces, or process computer calculations.
Only
- j Experience with passive type instruments in a sm 11 number of cont rol rods are moved 1
Commonwealth Edison generating stations and thus um peaking factors are not
,i substations indicates that the specified cali-expected to change significantly and thus i
brations are adequate. For those devices which i
s a Q uate.
i
- lf employ amplifiera, etc., drift specifications l
call for drift to be less than 0.19/ month; i.e.e For fuel fabricated by ENC, the power an the period of a month, a drift of.19 would distribution will be checked once per l-occur and thus provade for adequate margin.
day to ensure consistency with the power distribution assumptions of the fuel design l
analysis for overpower conditions.
During i
l periods of operation beyond these power distribution assumptions, the APRM gains 1
Anendnent *10.
63 or scram settings may be adjusted to ensure consistency with the fuel design l
I c ri t e r.e for overpower conditions.
i '
34 9'
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3.2 IJMITING CONDITION FOR OPERATION 4.2 SilHVEII.I.ANCE HEQtlHIEM ENT C.
Control Hod Block Actuation
~
1.
The limiting conditions of operallon for the instnamentation that inillates control rod block are given in Table 3.2.3.
I 2.
The minimum number of operable i;
instrument chanm la sgwelfied in Table 3.2.3 for the Itod lilock Moni'.or i
I may be reduced by one in one of the trip systems for maintenance aml/or testing, provided that this condition i
skieu not last longer than 24 Imurn in any 30-day perloat. In aildi Lion, one channel may be bypassed I
above 30% rated power without
.i a time restriction provided that a limiting control rod pattern does not exist and the 1
remaining RBH channel is operable.
D. Steam Jet-Air Ejector Off Gas System 1.,Except as specified in 3.2.D.2.
below, both steam-jet air ejector off-gas system radiation mon-itors shall be operabic during l
reactor power operatlon.
The tri settings for the monitors i
sha 1 he set at a value not to exceed the erpiivalent of the stack release limit specified in Specification 3.8.
The i
time delay setting for closure of the steam jet-air ejector a
~'
isolation valves shall not 4
exceed 15 minutes.
I
-l i
36
.i j_
Amendment No. 63 4-es s
1 i
3.2 LIMITING CONDITION FOR OPERATION 4.2 SURVEILLANCE REQUIREMENT i
i 2,.
From and after the date that one of the two st eam-jet air ejector of f-gas system radi -
t ation monitors is made or found to be inoperable, continued i
reactor power operation is l
permissible during the next seven days provided the inoper-i able monitor is tripped in the upscale position.
t i
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i I
i 4
i Amendment No. 63 36a i
$~
90
j i
e
- INSTRUMENTATION THAT INITIATES ROD BLOCK Table 3.2.3 i
Hinimum No. of Operable Inst.
Channels Per s
Trip System (1)
Instrument Trip Level Setting l
1 APRH upscale (flow hias) (7)
M-0+
M LPD 1
APRH upscale (refuel and Startup/ Hot Standby, mode) 312/125' full' scale f,
2 APRH.downscale (7)
>3/125 full scale i
1 Rod block' monitor upscale (flow bias) (7) f(.65W + 45)
(2) l 1
Rod bloc.k monitor downscale (7)
>5/125 full scale
'3 IRH downscale (3)
>5/125 full' scale t
3 IRH upscals
$108/125 full scale 3
IRH detector not fully inserted in the core _
2 (5)
SRH detector not in startup position (4) 5 2 (5) (6)
SRH upscale 310 counte/sec.
q J
Scram discharge volume water level - high
<20 inches hbove the botton' of the instrument volume p
li, L,
1 h
s e
i
].
Pnandmer,t No.
-A, 63 J
42 1
(,
'i - l TABLE 2.3 (Cont..)
I t
NOTES:
1.
For the Startup/ Hot Standby and Run positions of the Reactor Hode Selec. tor Switch, there shall be two operable
.i i or tripped trip systems for each function, except the SRH rod blocks IRH upscale, -IRH downscale and IRH detector not fully inserted in the core need not be operable in.the "Run" position and APRH downscale, APRH upscale (flow blas), and RBH downscale need not be operable in the Startup/ Hot St sadby mode. The RBH upscale need not be operable at less than 30% rated thermal power. One channel may be bypassed above 30%. rated thermal power provided' that a limiting control rod pattarn does not exist.. For systems with more than one channel i
per trip system, if the first column cannot be met for both trip systems, the systems shall be tripped. For, I
the Scram Discharge Volume water-level high rod block, there is one instrument per bank.
t
~.
2.
WD Percent of drive flow required to produce a rated core flow of 98 Hlb /m. MFLPO= highest value of FLPD for G.E. fuel.
3.
IR.H downscale may be bypassed when it is on its lowest range.
4.
This function may be bypassed when'the count rate is >100 cys.
l S..One of'the four SRH inputs may be bypassed.
i 6.
This SRH function may be bypassed in the higher IRH ranges when the IRH upscale rod block is operable.
i e
~
7 Hot. required while performing low power physics tests at atmospheric pressure during or af ter refueling at 4
']
power levels not to exceed 5 HW(t).
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, haendment No.%, 63 42a
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l Table 3.2.1 whicts senses alte conditions for dich l
Basest is laHon 13 nqu ed. Such inMrunnentanen must be 3.2 in addluon to'rescior protection instrismentatlose avaltalste whenc ser prim.ary contalmnent Integrity which ini Isles a reactor screme protective Instras-is mu eil.
o @c le to isolate Hee primary menistion has been provided which inillates sellen enntainmerA so that the gulatellace of 10 CFR 100 bre to mitigate lies consespecaces of accidents whicle are not escecded disting en accident.
j tpeyomt the operators al>llity to control, or terml.
sales operator errora before tiecy restalt la serioens
.nic Indrumenth dich leilWlu @m gh This set of !%iecifications prowlites inplation is connected in a ehaal hiss arr ::ltement.
coneegimences.
ine Ismitlei:; conditions of eincration for tiec primary
.Heus. Ihc dimh $e n b % he M %@
systeni Isolation functioso, inillation of Ilic epicr-cation 3.1 la appilcable here.
gency core cooling systene, control rol talocic asul st.seulhy gas treatencnt systems. 'llie oh)ce:tives of vm. s. ew we sa. e eu meesoma. ee.use.eDe
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tlic specificalloas are (6) to assiere the ellectiscness
- n. e en. s.us se.ne su,.e was ens se e.es a e.
s.
sum...
n.us u.e..e.siw sa.=e., em em. e.se 1
of the proicctive lustrissnentation winen restiaired Iey r=*
- e's "= m*"..*.d' ** " '"..* *** *" "se"e "an m.. s so.
8 em use s.
,t i
. 6 en. e.,.a preserving its capaiselley to tolerate a single 1.sileire m.u.ese e.e e a m..
was.. ens s sen s.se
.n en
. See of any component of bucle systems even eherins:lerl=
s=m.. swow sm.. m uu ens m.ss%
..e eu s, i mn puum use li ods ulwn poslinses of sesch s} stents are out of ses tice
... e e..n. med s. em. s cn...e, for maintenance, asul (ll) to prescrline the trip met-sm.. u s, s.su.u. e s..w..a se, s e s p.e
.s.m.e i
s..e.e s
..s... n.e em - u s, em en sms.e s p,. se.e.e.
l y
lings requircel to assuee adeepeate perforniance.
- * *** ' "* ' ' 88 - '** * " "u s.,.e see s m.. ee When soccessary, osee channel niay tse maeli. Inoper-
.u..e u.". Isu" s..m..
6 e.,.e me s eats e e me-.wwe i
able for isticI intervals to cosuluct respelrc.i functional
..s...s.
. ean. em.
s.*.ssa n. es e m.e., men.es es b
eu esmeas.e m.u m. em one -s amte en. nies.,se tests smi callieratioses.
eene.s
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u.
vs. s s-. mew sms s u -e.es= s me e. use.a n=.
Some of the scitings oss the instruenentalian that
}*,',**," j,l'" ', {*,*l* "","*M"**,!*,"l,*" *,'*s'as n'.*s.'s"e "~
f' 6
lestilates or controls core and contalesment cooling l
hose lolerances exidicitly stated where the liigh and sum.. 6.n in,e.,se n u.. tese.
i This Iow valacs are hoth critical and niay 1 sve a substan-trip initi.stes closure of Group t primary containenent
[,,
llal ellect on safety, it shmaid Ise smotcel that ti c set.
isolat'on valves, lief. Section 7.7.2.2 Salt, enet also points of other instrunnectation, where only Ihc higi sclavales the ECC sulesyst enes, starts the emergency t
or low esul e,f the setting has a direct e caring oss slicscl generator seal trips the recircostation pumps.
s l
safety, are clausen at a level away fruen Ific nurnest
'thlu trip selling level was chcsen tu lse higle enstgh q
o;pating range to prevesit inadvceicnt act::allese of in prevent spi:rlous oparation tsat low encie;;h to lat-4lse ufety rystt ne involved and exposure la alesrmal llate ECCH er,wrutton as.d I rlmary system Isolello I i
Initaallons.
so that no mciting of the t sci cl::ehleng ulle occer and i
so tisat past accialent contint canlie accomplistical Isolation valses are installed in those lines Ihat and the I;uniclin-s of 10 CI'll 100 will nos he violated.
l penetrate the prionary containment snel si;ust be
}
isolateil slurisiel a loss of coolant acchient so liest the l'or the enenplete circuse:'erantial Isreal. of a 2s-1pcli recie rial.ilism line ate.1 mitti I?w ta ip setting given i
l raill:.llon dosc tintith are not caccealed sluring an clustc. UCS In'llaston anel primary syt'em Italation ac:liles t condelion. Actuatien of thcsc valves !s t
are Inillated in time to mes-t the alsore criteria, i
initi:steillay preelectise leibtriminentation shown in 6
u i
Amendment No. #, 63 O
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- 3. 3 LlHITING OWUITION FOR OPl! RATION 4.3 MHtVlilLt.ANCli iti;Qllllll35.NTS l
C.
Scram insertion Times C.
Scram Insertion Times 1.
D e averaFe scram insertion time, based 1.
After each refueling outage and prior to power on the de-energization of the scram pilot operation witti reactor pressure above 800 psig, valve solenoids as time zero, of all oper-all cont rol rods shall be sid>Joct to scram-time able control rods in the reactor power t est s from the failly withdrawn posit ton. He operation condition shall be no greater than:
scram times shall be measured without reliance on the control rod drive pumps.
% Inserted From Avg. Scram Insertion Fully withdrawn Times (sec) 2.
At 16 week intervals, 50% of the control rod drives shall be tested as in 4.3.C.1 so that 5
0.375 every 32 wecks all of the control rods shall 20 0.000 have been testcJ. Whenever 50% of the control l,
50 2.00 rod drives have been scram tested, en evalua-i l
90 3.50 tion shall be mado to provide reasonable assurance that proper control rod drive l
We uverage of the scram insertion times performance is being maintained.
i e
for the three fastest control rods of all l
grot.ps of four control rods in a two by two 3.1'o110 wing completion.or each set or scram '
l j
array shall be no greater than:
Len t,ltig as desc rilied above, the rienults will 1,e compared against the avt. rage 9
% inserted From Avg. Scram Insertion neram speed dint,ribut,ii.n used in tie i
j rully Withdrawn Times (sec) trann.ient analysis to vert ry the appli-catellity or the current MCPit Opernting -
i 5
0.398 1.imit.
Herer to Specificat, ion 3.S.K.
j 20 0.954 50 2.120 90 3.800 2.
De maximum scram insertion time for 90%
j insertion of any operable control rod shall not gacced 7.00 seconds.
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f, Amendment No.J7, 63 58 2j t
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DPR-25 indicative of a generic contrut ro.1 drive-provides a positive check as only uncoupled problem and the reactor will be shutdown.
drives may reach this position. Heutron in-e 5
Also,.if damage within tiie control, rod strumentation response to rod reovement pro-drive mechanisia and, in particular, cracks
'vides a verification that the rod is followin's L
in drive internal housings, cannot be ruled its drive. Absence of sucia response to drive
- I out, then a generic problein affecting a movement would provide cause for suspectling a b.
pusiber of drives cannot be ruled out.
Cir-rod to be uncoupled and stuck. Restricting re-
.. ld.cusforential cracks resulting from stress
' coupling verifications to power levels above -
assisted intergranular corrosion have 20% provides assurance that a rod drop during s
l occurred in the collet housing of drives at a recoupling verification would not result in a
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several BWR's.
This type of cracking could rod drop accident.
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occur in a nu,mber of drives and if the cracks propagated until severance of the 2.
The contirol rod housing support restricts the l
collet housing occurred, scrais could be pre-outward movement of a control rod to less than p
vented in the affected rods. Limiting the 3 inches in the extremely remote event.of a period of operation with a potentially housing failure. The amount of reactivity which severed collet housing and requiring in-could be added by this small amount'of rod witle-
{i crease'd surveillance after detecting one stuck drawal, shich is less than a normal singic with-rod will assure that the reactor will not be drawai increment, will not contribute to any op,erated with a large number of rods with damage to the primary coolant system. The desiga j
s failed collet housings.
basis is given in Section 6.6.1 of the SAP., and e
tlie design evaluation is given in Section 6.6.3.
3,.'The operability of the scram discharge This support is not required if the reactor e
volume vont and drain valves assurer the coolant system is at atmospherde pressure since h
j a
l proper venting and draining of the volume.
there would then be no driving force to rapidly
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'This ensures that water accuisulation does eject a drive housing.' Additionally, the support not occur which would cause an early termina-is not required if all control rods are fielly in -
Lion of control rod movement during a full sorted and if an adequate sleutdown margin with one These specifications provide for control rod withdrawn has been demonet, rated sinca lg core scram.
the periodic verification that the valves are the reactor would remain subcritical even in the 2
j e
open and for testing of these valves under event of complete ejection of the strongest con-f reactor scram conditions during each Re-trol rod.
~
- i fueling Outage.
f,
- B'.
3.
Control rod witledrawal and insertion sequences I,
Control Rod Withdrawal are established to assure th'at*the maximum In-sequence individual control rod or control rod' f,
I.
Control rod dropout accidents as discussed in Ref-sequences which are withdrawn could not be worth j
erence XN-NF-80-19. Vol.1, can lead to significant core enough to cause the rod drop accident design lisilt of I
280 cal /gm to be exceeded if they were to drop out of
! 3 doisa ge. If coupling inteirity is maintained, the possibility of a rod' dropout accident is the core in the manner defined for the aod Drop Acciderit.k I
eliminated. The overtravet position feature 3hese seguences are developed prior.to initial.
i operation i
i.
.i. !
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Annnlinant No. %, 63
,,_,61 a,
Bases (cont'd)
Parametric Control Rod Drop Accident analyses 6
j have shown that for wide ranges of key reactor parameters (which envelope the operating ranges of these variables), the fuel enthalpy rise during a postulated control rod drop accident j
remains considerably lower than the 280 cal /gm limit.
For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient effective delayed neutron j
fraction and maximum four-bundle local peaking factor are compared with the results of the I!
parametric analyses to determine the peak fuel rod enthalpy rise.
This value is then compared against the Technical Specification limit i
of 280 cal /gm to demonstrate compliance for 1
each operating cycle.
If cycle specific values j
of the above parameters are outside the range l
assumed in the parametric analyses, an extension of the analysis or a cycle specific analysis may be required.
Conservatism present in the analysis, results of the parametric studies, I
and a detailed description of the methodology for periorming the Control Hud Drop Accident analysis are provided in reference XN-NF-80-19, Volume 1 (Supplements 1 and 2).
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1 t-Amen &ient No.47, 63 l
62a
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pa ne s;. (con'J) a,
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The Rod Wotth llinimizer providen autumatte asupervision to annure that out of nequence control rods will not bc withdrawn or innerted; i
1.e., it limits operator deviations from planned j
f j
withdrawal sequences. Ref. Section'7.9 SAR. i j
itservesasabackuptoprocedural.controlo(
control rod worth.
In the event that the Rod l Worth Hint: sizer is out of service, when required, I
a licensed operator or other qualified l
technica'l eMildyeE 4 Air nonually Iuiti11 the i
iq control' rod pat'Leku'honformaance functions of the Rod Worth Hin'in1Ecf.' 1,n this case, procedural.
i l
control' 'is ' c'xc rci se'd' by"ver i f y Ing - al 1. cont rol
, j' rod positions' af tdrthe withdrawal of each group, prior to procheding 'to the next ll :
group.' Allowing.4bbstitution of a second
( ):
l 3 ent operator 'or' engineer in case.
j (i
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- Indep e:
d jg of RWH inoperbbil'ity. recognizes' the capability l
to adequately nonitor proper. rod sequencing in
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an alternate" manner without unduly.rcsfrict *
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lingplantoperations. Above 201 power, there is,
- i i
no requirement that the RICI be,operabic sipce the control rod drop accident with out-of-sequence rods will result in a.pcak fuel I
energy contont of less thar 28Q cal /gs. To j
assure high Rtal availability, the kl.7: is requeied to be operating during a startup for the withdrawal of a significant number control rods for any startup af ter June 1,1974.
i 4.
The Source Rango Honitor (liRM) system performs no automatic safety sys.tess function; i.e.,
it l
~j has no scram function.
It does provide the 1
0 i
I I
i 62h j
Amendment No.
E, 63 l.
)h f
J 4 l t
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L C
C with a visual indication of neutron C.
Scraai insertion Times operator level. This is needed for knowledgeable and ef ficient reactor startup at low neutron level.
The performance of the control rod insert. ion The consequences of reactivity accidents are system is analyzed to verify the system's functions of the initial ncutron flux. The ability to bring the reaetor subcritical at j
requircr-ent of at Ic.st 3 counts per second.
a rate fast enough to prevent violation of the MCPR Fuel Cladding Integrity Safety 1.imit essur.s that any traasient, should it occur begins at or above the initial value of 10 I and thereby avoid fuel damage.
The analyses j
i demonstrate that if the reactor is operated
[
i, of rated pcuer used in the analyses of transients within the limitations set in Specification
[
~
fret cnid cenditionn.' One operable SR:1 channel
.!.5.k. the negative reactivity insertion vonld be adequate to monitor the cpproach to rates associated with the observed scram 1
criticality uste:; hononeneous patterns of performance (as adjusted for statistical scattered control rod withdrawal. A ninimum variation in the observed data) result in of teu oper.:ble SMI's are provided as an added protection of the MCPR safety limit.
conservat i:.m.
In the analytical treatment of most tr.aunients 5.
The Rud filock Monitor (aral) is designed to auto-290 milliseconds are allowed between a neutron' l
matiently prevent fuel dasage in the event of sensor reaching the scram point and the erroneous rod withdrawal f rom locations of high start of motion of the control rods.
This 1,
power density during high power level operation.
is adequate and conservative when compared j
be to the typically observed time delay of T: o channels are provided and one of these may/or a
bypest:cd froi the cons. ole for iraintenance and about 210 millisecons.
Approximately 90 testi C.
Tripping of one of the channels will block milliseconds after neutron flux reaches erroneous rod withdrawat soon enough to prevent fuel the trip point, the pilot scram value solenoid dasanc. This systen backs up the operator who with-de-energizes and 120 milliseconds later the p
draws rods according to a written sequence. The control rod motion is estimated to actually speci.ffed restrictions with one channel out of henin.
Ilowever, 200 milliseconds rather than orwire conservatively assure that fu:1 damage 120 milliseconds is conservatively assumed for this time interval in the transient analy-0 will r:ot o: cur due to rod withdrawal errors when ses, and is also included in the allowable I
this condition c::1.sts.
Anandnents 17/l8 and 19/20 scram insertion tims specified in Speci-1 present the rer.ults of an evaluation of a rod block.
fication 3.3.C.
In the statistical treatment r.oaitor fa!!ure. Ther:e amendecnts show 'that during f the limiting transients, a statistical j
I reacter operation with certain limiting control distribution of total scram delay is uscJ i
rod aatterci..the withdrawal of a designated single rather than the bounding value described co. trol rnr'PRefd re: nit in one or nore fuel rods above.
.o s m i. 3 in.. sm. acen r ca.aan,
i.in..ir..iny ii.n.
o.. i., n. o r..o.
]
pa t t e n:n, at ts Jeann ni titat testing of the RB21 The performance of the individual control rod systen prior to witlntr::ual of such rods to assure drives is monitored to assure that scram its c;crability ullt assure that improper with-performance is not degraded.
Fifty percent dra;.al deem not acrer. It is the ree.ponsibillLY of the control rod drives in the reactor are of thi ::cclear Eng;l;:ccr to identify these lintting tested every sixteen weeks to verify adequate patt3rns a-l the denitnated rods either when the -
nerformance.
Observed nlant data were used patterns are initially established or as they to determine the averaPe scram nerformance O l
I devel due to the occurrence of inoperable control used in the transient anslyses, rods n other than limiting paLLerns.
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l-hnendment No. M, 63
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Scram Insertion Times (cont'd) and the results of each set of control rod scram tests during the current cycle are i
compared against earlier results to verify I,
that the performance of the control rod insertion system has not changed signifi-i cantly.
If an inuividual test or group of tests should be determined to fall outside
~
of the statistical population defining the scram performance charact eristics j
used in the transient analyses, a re-
! i determination of thermal margin require-ments is undertaken (as required by Specification 3.5.K) unless it can be shown that the number of individual drives falling outside the statistical population defining the nominal performance is less than the allowable number of inoper-able control rod drives.
If the number of statistically aberrant drives falls within this limitation, operation will be allowed to continue without rede-termination of thermal margin require-ments provided the identified aberrant drives are fully inserted into the core i
and deenergized in the manner of an inoperable rod drive.
3 i
The scram times for all control rods are measured at the time of each refueling i
outage.
Experience with the plant has shown that control drive insertion times vary little through the operating cycles hence no reassessment of thermal margin rqquirements is expected under normal cdnattions.
The history of drive performance accumulated to date i
indicates that the 90% insertion times of new and overhauled drives approximate a normal ii l
distribution about the mean ubich tends to become 64
~
skewed toward longer scram times as operating
'e time is accumulated.
The probability of a drise I
not exceeding the mean 90% insertion time by 0.75 se<ond is arenter than 0.999 for a normal distribution, kendment No. J7, 63 l-I J
-,, i
i
'i ii l
tivity varies as fuel depletes and as any burnable poison in supplementary control r
is burned. The magnitudo of this excess l
reactivity may be inferred f rom the critical ro.1 configuration. As fuel burnup progresses.
i anomalous behavior in the excess reactivity -
U may be detected by comparison of the crit leal rod pattern selected base states to the predicted rod Inventory at that state. Power operating base conditions provide the most
'j senaltive anil directly Interpretable data
,i relative to core reactivity. Fu rthermore.
i using power operating base conditions per-mits frequent reactivity comparisons.
Itequiring a reactivity comparison at the specified frequency assures that a compart-son will be made before the core reactivity change exceeds I'l AK, Deviations in core
}'
D.
Control llod Accumulators reactivity greater than 1% AK are not ex-pected and require thorough evaluation.
The basis for this specification uns not des-One tercent reactivity limit la conaldered i
crlhed in the SAll and, therefore, is presented safe etnce an Insertion of the reactivity into In its entirely. Ilequiring no more than one the core would not lead to transients exceed-i inoperable accumulator in any nine-rod square Ing design conditions of the reactor system.
array is based on a series of XY PDQ-I quarter enre calculations of a cold, clean core.
'I he worst rase in a nine-rod u tlhdrawal sequence G. Economic Generation Control System resulted in a kelf < l.0 -- other repeating rmt Operation of the facility with the Economic sequences with more rods uithdraun resulted Generation Gontrol System with automatic In keff > l.0.
At reactor pressures in excesh flow control is limited to the range of 65 '
jI of 300 psig, even those control rmis with in-and with reactor power above 20% the reactor 100% of rated core flow.
In this flow range t
operable accumulators will he able to meet re.
quired scram insertion times stue to the action can safely tolerate a rate of change of load i
of reactor pressure. In aihlition, they may le of 8 MW(e)/sec. (Heforence FSAR Amendment 9'-
l normall,y. inserted ualAg the control-rod-drive Unit 2, 10-Unit 3).
Limits within the Econo-l hyth sysicin. Procedural control will assur that control rods with inoperahf a accu-mic Generation Control System and Reactor Flow ji mulators will he spaced in a one-in-nine array Control System preclude rates of change I
rather than grouped together.
greater than approximately 4 MWe/sec.
E.
Heactivity Anomallen When the Economic Generation Control System l is-in operation, this fact will be indicated t
During each fuel cycle excess operating reac-on the main control room console.
The results l
of initial testing will be provided to the AEC i
at the onset of routine operation with the. -
l l
Economic Generation Control System.
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(flevised wit.h Chani'.es 27 and la lasued 3/29/71 )
1 45 l
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Amendment No. 63 p
8
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d.5 SURVE!!.t.ANCE REQ 1}IREHENT 3.5 1.1H171NG CONDITION FOR OPERATrott D.
Surveillance of the Autometts Pressure 1
9.
Automatic Pressure Relief subsystems Ro11ef Subsystem shall be performed as j
follows:
l 1.
Except as specified in 3.5.D.2 and 3 below, 1.
During each operating cycle the foltoutng I
the Automatic Pressure Relief Subsystem shall be performeJ:
~
shall be operable whenever the reactor A.
A simulated automatic initiation pressure is greater than 90 psig and irradiated which opens att pilot valvss, and fuel is in the reactor vess'e1 1
b.
With the reactor et pressure each i
2, From and after the,date that one of the relief valve shall be manus 11y opened.
i five relief valves 36ft.the automatic pressure I
metter valve opentna shall be verified bT rellef subsystem is cado or found to be a compensating turbine bypass valve er inoperablo when the reactor is pressurized control watve closure.
abo $e 90 psig with irradiated fust in the A logic system functional test shall be redctor v ssol, reactor operation is pornissible c.
I only durinC the succeeding seven days a:nless performed each refueling outage.
repairs are made and provided that during such -
,i slao the lirC1 Subsystem is operable.
- 22. When it is determined that one re11ef valve of the automatic pressure relief subsystem 3,
From and efter the date that more than one is inoperable, the llPCI shall be demonstrated of five.rclief valves of the outcmetic to be operable lamediately and weekly thereaften j
pressure retter subsysten naile or found j
Ii to be inoperable when the reactor is i
pressumi:ed above 93 psig with irr:diated fuel in the reactv - vessel,' reactor operation
!s permissible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless repairs are made and provided that durin2 such time the HPC1 Subsystem 3.
When it is determined that more than one is opor:ble.
l relief valve of the automatic pressure relief subsysten is inoperable, the HFCI subsystem shall be demonstrated to be operable lassediatog I
i f
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78 Amendment No. //1,f p, 63 I
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l 4.5
$URVEILIANCE REQUIREMENT LIMITING CONDITION FOR OPERATICet 3.5 I, Average Planar UlGit Av.orpqe Planne Linone treat ceneration I.
the During steady state power operation, Itato (APtJiciQ Average Planar 1inear llent Gene ration Itate
( APLl!Gil) or all the roda in any fuel a:.:;em-The APIJIGR for each type of fuel as a,
fuel bly, an a function o'r average planar exponu2 e function of average pinnar exposure f6r G.E.
bundle exposure for Exxon Ibel shall for G.E. fuel and average buiulle exposure tent average for Exxon fuel at be dete wined daily during reactor operation at
';l cny axial location, shall not exceed the 2 25% rated thermal power.
I anxiraum average planar ulcn shown in rigure 3.5-1.
If at any timo during cperation it is determined by normal sur-voillcnce that the limiting value for i
APulcn is being exceedo 1 action shall be initiated within 15 minutos to restore i
cparation to within the prescribed limite.
I' I
If the APulGR in not rot.urned to within the prescribed limite within two (2) hours, the reactor shall be brought to the cold shutdown condition within 36 Surveillance and corresponding i hours.
action shall continue until reactor opera-tion lo within the prescribed limite.
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rmenbent n>. E, 63 sis l
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Dl'R-2 5
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5 3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIh24ENT i
.t l'
3.5.J LOCAL LilGR J.
Linear lleat Generation Rate (l.IIGR)
During steady state power operation, the The LilGR shall be checked daily during l
linear heat generation rate (1.llGR ) of reactor operation at 2 25% r.ated thermall any rod in any fuel assembly fabricated by power.
j CE at any axial location shall not exceed
{,
the design value of 13.4 kw/ft.
I If at any time during operation, it is determined by normal surveillance that the l
limiting value for LilGR for G.E.
fuel is i
being exceeded, action shall be initiated within 15 minutes to restore operatton to within the prescribed limits.
If the
,l.
LilGR is not returned to within the
);
prescribed limits within two (2) hours,
,{
i the reactor shall be brought to the Cold Shutdown condition within 36 hoitrs.
Sur-veillance and corresponding action shall continue until reactor operation is within the prescrileed limits l
s s
I s
h e
e Amendment No. g,63 818-1 2
.4 s
.~
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0 10;000 20',0C0__
30,000 4;002 Bundle Average Dgosure ( Z /MT)
I Figure 3.5-1 Maxi =us Average Planar I.inear l
(Sheet 1 of 5)
Heat Generation Rate (MAP *IDR) vs. Bundle Average Expcsure i
NOTE: Figure 3.5-1 (Sheet 1 of 5) for Fuel Type XN802.69-5 is only valid to 10,000 ftfDPfT pending NRC approval of R0nEX2 code.
If RODEX2.'is not approved before 10,000 ft4DriT is achieved, additional analyses must be submitted and approved by the staff to veri fy the continued acceptability of the proposed MPL%R limits.
i i
I i
Amendment No.M, 63
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10000 20000 30000 E
PLANAR AVIRAGI IXPOSI.*RI (Kr.".3/T) t I
I 1
i t
hism Average flanar Linear 28 cat Generation Rate (MAP 23M l
'94. 'Flanar Aterste EKpasure Mgure 3.5 1 Uheet 4 of.5) l I
5 Amendment No. M, 63 5
\\
t,.
wk..1-W
j DPR-25 l
3.5 LIMITING CONDITIONS FOR OPERATION 4.5 SURVEILLANCE REQUIREMENTS l
K.
Minimum Critical Power Ratio (MCPR)
K.
Minimum Critical Power Ratio (MCPRJ During steady state operation at rated core MCPR shall be determined daily during a L
flow, HCPR shall be greater than or equal to -
reactor power operation at c.251 rated a
thermal power and following any change in 1.34 f3r GE 8 x 8R fuel p wer level or distribution that would g
1, 1.33 for ENC and GE 8 x 8 fuel cause operation with a limiting control rod pattern as described in the bases for for core flows other than rated, these nominal Specification 3.3.B.S.
l values of **CPP. shall be increased by a factor of.
i K, where K is as shown in Figure 3.5-2.
l f
g l
1 1
y i
i
't
]
i If at any time during steady state power operation, it is determined that the limiting d
value for MCPR is being exceeded, action sball be initiated within 15 minutes to restore operation to within the prescribed limits.
If the steady state MCPR is not returned to within the prescribed limit s
- j within two (2) hours, the reactor shali 1
be broultht,Io the Cold Shutdown condition J
wLthin 36 1murs.
Surveillance and corresponding l
action shall continue until reactor operation l
il is within the prescribed limits.
1 In the event the average 901 scram insertion time p
detennined by Spec. 3.3.C for all operable control i
rods exceeds 2.58 seconds, the '1CPR limit shall be j
increased by the amount equal to [0.0544T - 0.14]
where T equals the average 90% scram insertion time for the most recent half-core or full core surveillance Amendment [ 63 81D i.
4 data from Spec. 4.3.C.
l-J
- e
- 4 li n
d 0
O O
~
3.5 Limitin<,. conditions _ f_orAr_ation n_a_ncri desetoped in this refer-
~
A Core Soray and LPCI Hode _ of the fillR ence the repair period is foaml e,i s,e Ics, th3, Int Intenal. Tlils asnines that the
~-
Vystem - This specification assurc's i
y core spray and t.PCI sulerstems constitute a j
that adequate emergency cooling 1 out of 3 system. her.cre"r. the combined et.
capability is available.
rect or the two pystems to limit csec,,grc enad
, temperatures mu5t alto lic confldered. The Based on the loss of coolant analyses tcat infertal rpecificJ In Frecification 4.3 was i!
included in References (1) and (2) in I mer:ths. Therefore. an allmta' sic arti.11r jl 4.
g,'g, g g'"'y,"['[,,'{1c conjidcr-a accordance with 10CFR50.46 and Appen-
,i dix K, core cooling systems provide anal this specification is with!n this pertoel. '
sufficient cooling to the core to For mulilple fai!;srcy, a eh.?r:cr lanrral is i
dissipate the energy associated with sivcified and t<> Impleve t!.c apeurance that the loss of coolant accident, to limit the remaining systems will function. a daily' test is caHed fur.
Ut' mud it is rece;:atred j'
the calculated peak clad temperature thal the information given la reference 3 pro.
to less than 2200oF, to assure that vides a quanHtative c:etem,I to ept nite 3 l..
ahic repair Ilmes. the Iacl. of operating data to.
l; core geometry remains intact, to limit
'g the core wide clad metal-water reaction support the analytical approach presente enm-to less than IA and to limit the cal picte acceplance of this r. c:had at this time.
culated local metal-water reaction Therefore, the :Imesi a:3ted la the,pceirse llems were estabilphed tiith due regard lo I
i to less than 17U juilcment.
,l The alloweble repair times are es--
Should one core spray pub.ysicm become in-tablished so that the. average risk rate operable, the remaining core ep.sy and the for repair would be no greater than entire LPCI system are arallable should the i
the basic risk rate. The method and (2) NCDO-20566, General Electric l
concept are described in Reference Analytical Model for Loss-Compan{antAnalysisinAccordance (3). 'Using the results of-Coo with 10CFR50 Appendix K.
!j (1) " Loss of Coolant Accident Analyses Report (3) APED "Cuidelines for Determining i
for Dresden Units 2, 3 and Quad-Cities Safe Test Interynis and Repair l
Units I, 2 Nuclear Power Stationn,"
Times for Engineered Safcquards" =
Nt:DO-2 414 6 A, Itevisioni, Apr11 1979.
April 1969 1.H. Jacobs and 7
P.W. Marriott.
l (1 ) XH-NP lil '/'i "Dre aden Unit 3 LOCA Model I'
1 l?
Ifnina the ENC EXEM Evaluation Model knendment lio. y, 63 MAPLilGit llecults l
l 82
[l MO
,i
' ?
s
}
1 L
G'..
O-G
>\\
L1 s* tine Cerdition for Ornration Dases (cqt'd),
,3,3 I
I, p.ve-ere Plc. car Ifr.n This n;ccification accurts thst the pak
- j
. cla$ ding tenpernture fo11ouln6 a rectthted
(
docign basto lo n o,(-coole.nt accident 18111
' ~
not execcd the 2200 T limit opacific1 in i
IfCFR$0 App:r. dix K'censtccrire the po:tulated l'
ofrects of fuel pallot dencification.
The peak cladttng temperature follouing a postuletcJ loss-of-ccolant cecJdent 1*
prlicrily a funct1r.n of the avan.r:s II'CR of all ths rcla in a feel cece-itely et eny
),
l c.xini location er.d to o.aly d.*ptaient cocar.d-l arily,on the rol to red y:::or distritution willism a feel os::enbly. Sinco expcte:1 local varisticas in pos::r dit.trilution within a fusi as:.oa.bly affect the esiculated f:sk olad tonpratu:o by Icss than 120 7 rahtive to C
the pek tenprature for a 17t cal fusi design.
i ths llait on the asorar,e ple:t r IJil;R la suffic1 cat to escurs that calculated temp.
cruturas are belcu the 1ccFR50, APFondix K The :naximum average planar 1JIGlia for G.E. 9 r
5 -
i 11'51L *
,l ruel plotted in Pig. 3 5 1 at
[i hk.her alinnures renult in a l
The maximum average planar LilGRs shown calculated resk clad temprature of 1 :s l
in Figure 3 51 are based un calculations it-sn 226. littsever tho su: time:n aware.go
' employing the modelo denerthed in pinnar 12iCRs are cho:en on Plc. 3.51 es
- lleference (1) and in vercrence Xti-flP fil-75 Ile.its bscauso confor:ance calculaticas havo c
41
- . Power oppration with Al%IIGita at or below nct been perforr.:d to justify opsratica at I
. tho:v> y'hwn in Pig. 3 5.1 asuurra that, IllClio In exccos of those shoue.
. the phak cladding temperature following a postulated lons-of-coolant, accident will J.
Loc't1 fir.R
'(f*
not exceed the 22OO"P 11 mil,.
Li This specification asouses that the -
I{
niximun linear heat Generation rata in tiny fuel rod rubricated by U.E.
is l
l i
(1) *I.oss of Coolant Accident Analyses Report for lens 1.han the dealgn linear 5
presden Units 2, 3 and Quad-cities Units 1, 2 g$A t
Huclear Power Stations," NEDO-24146A, Revision I, i
Apr11, 1979.
i i
Amendment 40. #
63
-He
]
l
.5 Limiting Condition for operation Hases (cont'd)
I heat generation rate even if fuel pellet. den-the/ cycle-specific fuel loading, exposure and '
sification is postulated.
fuel type.
The current cycle's reload licen-sins analyses identifles the limit ing transient 4
For fuel fabricated by ENC,' p'rotect'lon o f the for'that cycle, o
/
MCPR and MAPLllGR limits and'operatihn within j'
E the power distribution assumptions of
,x the Fuel Design Analysis providas adequhte As described to speci fication 4h3.C.3 and the associated Bases, obseryed plant data were protection against cladding strain limics, /
used to determine the arcrage scram perfor-hence the LHGR limitation for GE fuel is'
+
unnecessary for the Trotection of ENC mance'p;ahd in the transient analyses for fuel.
\\ deter:afhing the MCPR Operating Limit.
If
'k the current ' cycle scram time performance i
The~_Minimym Critical Power Batio (PR s)pecified falls >outside of the distribution assumed g,
K.
MCPR steauy-state vaaues sur iiC 1
s
\\
in 4he analyses, an adjustment of the MCPR ;
- i in the Specification were determined using the THERMEX thermal limits methodology;
,l
, limit may be required to maintain margin to-r j.
described in XN-NF-80-19, Volume A.
The e
the MCPR Safety Limit during transients.
Com-
- l, l
safety limit ' implicit in the Operat.ing pliance with the assumed distribution end i/
limits is establLshed so'that duringf adjustment of the MCPR Operating Limit will sustained operation at,the MCPR 'aaf;ety;
,be performed as directed by the nuclear fuel l
ve:nder in< raccordance with station procedures.
limit, at least 99.9% of 'the fuel hods in-(.
I the core are expectedrto avoid %ril'ing transition.
The Limfting%Transienj 6 CPR' For core Mows lest than rated, the,.NCPR
- 'l implicit in the operating limits 04.s cA -
Operatina 1imit established in t'he'e' aveci'f t, '
icntion is adjusted 60 provide ~prbt ction of '
culated such that the occurrence of'.hc y
I limiting transiedt from..the operating limit
-l the MCPR Sa'fety Limit $n the event o f e.2n flow increase'te H
will not result in violation of the ttel'Re uncont rol)ed; recirculatiLis!
,'*s:'
Lte )hysical limit of pump Ilow.r This' pro-safety limit.in.at leanc-951 of _t he random t
tvetion is provided for menual and autcraatic statistical combinations of uncertainties.
s f.lhMeontrol"by chohsingirhe NCPR operating s
f s
jj 'irai t as'the value frsp figure 3.5-2~ Sheet 1 l
Transient events of each type anticipated j
the rated cor4 flow yalve, whichever la during operation of a hWR/3 were evaluated are yb "I
to determine which is most restrictive in
.greitter. _ For. Automat te Fl ow Control, in t
l II i
terms af' thermal margin requirements.
The v m adJitien yto'proiectini thk. MCPR E;afety Limit i
j genenuint load rejectios/ turbine trip without.
Lurin6the flow'run-up event. prote'ction is.
i *, :V bypass is typically the limitiilg event.
Q, provigted ir,ainst violating toe, rated flow 9 f
J MCPRroparatisp,Limitidurinp; an nutomatic-flow The thermal margin offects of the event are y
evaluated with the-THERMIX NethLdology and yincreate' to 'frared, core flo.s.
This protection l
?
i appropriate MCPR liraits consistint with in provided by the reduced flow MCPR limita'
~
( M3wn in Fiv,ure -3. 5-2 ' Sheet 2 where the curv'e i
the XN-3 critical power correlallon'are t
I
~, Almit' ponding 'to t'he c'urrent frated' fihrW1CPR
['
determined.
Several factors in f lue'es:e corres G 6seds(linect interpolationfbe'twecn i
4' which transient results in the largest reduction in critical power ratioysuch as,
<;che MCPR-limit liitids depicted is permissible).
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Amendment No! #, 63 sr id
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4.5 1.imiting condition for Operation Itases (cont'd) a 1lj Therefore, for Automatic Flow Control, the MCPR e~
Operating Limit is chosen as the value from Figure 3.5-2 Sheet 1, Sheet 2 or t.he. rated l
t flow value, whichever is great est.
It should be noted that. if the rated flow MCPR I.imit must be increased due to degradation of control rod scram times during the current cycle, the new value of the rated flow MCPR limit is i
applied when using Figure 3.5-2 Sheet 2.
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Amendment No. 63 858-1 i
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by hr*111once.fiequiree-nte Ilsees (cont'd)
K.
Minizias Critical Power Ratie (MCPR)
I.
Averere Flens? LlICR.
. -r l
At core theM1 power levele less than er.'
Atcorethermalpowerlevelslessthenegroh' equel to 25 por cent, operatin6 plant to 25 per cent, the reacter will.ge oprating; '
-l
=t mintoir: recirculotten.Finap speed p.id the..
exportence and thermal hydraulic enelysee moderator vold content will be very small, (
y, 1rdicste that the resulting eversge plansr 3
LitC.1 le below the soximua averace plasur LNCR alldesignatedcontrolgodpatternstJI.sp
, $6j.
- Poyedatthispoint,operatinfylaat(4perpt.
l by a considerable rargin; therefore, evaluation and thermal hydraulic aglydis indicites that, or the averece pionar LilCR below this power, I
'"I level is not nossosary. The delly require-1 rbo rg n Ni tb s" ow rent for cloculating overece planer LilCR content, any inadvertent core flow I crosse
- i above 23 p1r cent rated therent power is would only place operation in a more con.
4, ll suffiggen$teln power Cletribution ehtfpe,
servative mode relative to HCPA' are s1pv wisen re have not been elCnitb cent powet(or egatrol ro4 chances.
1he daily requirement for calculating j.
8 MCPR above 25 percent rated thermal power is sufficient since power distribuqitM l
shifts are very slow when there have not been l
J, I.ocal L'ICRr significant power er contrut rod changes.
g ite IECR t'o r 0. E. fuel chall I
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bs check *8 daily during reactor operation et In addition, the R correction applied to Crester then or eeuel to 25 p:r cent power to the Ixo provides pirgin for flow'increses l
d2tercins if fuel burnup or control rtrl movenent from low flows.
r-l3 Fas caused changre in power distribution.
A lisatting IJICR value is precluded by a considerable margin when employing a per
.I mLasible conttol rod pattern below 25% rated W 6al power.
i 86A Amendment No. #, 63 7
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3.7 L'IHITING CONDITION FOR OPERATIDH 1
4.7-SUILVEILLANCE REQUIREHENTS j.
l' 1
. Primary containanent lategrity shell be maintained 2.
The primary containment integrity shat! be demon-
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!j ; '
at all times winen the reactor la critical or when strated by conducting Primary Containment Leak
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5 jj ;l *I the reactor water temperature le above 212*F and Teste and shall be determined in conformance with l
fuel le in time reactor vessel except while per-the criteria specified in Appendia J of 10 CFR 50 forming low power phyelce teste at atmospheric using the methods end. references therein, f.
pressure at power levele not to exceed 5 Hw(t).
jt,j I
Three Type A teste (overall Integrated Con-a.
a.
Primary containment leakage rates are defined tainment Leakage Rate) ehall be conducted at 8
l:
from:
i approximately equal intervale during each 10 l
4 year plant in service Inspection interval at (1) The calculated peak containment internal either F or Ft with the last being done ll pressure, P, la equal,to 48 peig.
during the 10 year in-service inspection shut-
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' The containment vessel reduced test -
down.
(2)
I, pressure, P s le estual to 25 pelg, b.
If any' periodic Type A test falle to meet t
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either 75 percent of L or 73 percent of L.
~
li (3) The naminum allowable leakage rate at a the test schedule for subsequent Type A teste j
e t
j pressure of F L., is equal to 1.6 shall be reviewed and approved by the Cosmiission.
percentbywefg,htofthecontainment l
jI.'
air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 48 pels.
If two consecutive Type A tests fall to meet c.
j either 75 percent of L. or 75 percent of L.
j (4) The maximum allowable test leakage rate at a t
a Type A test shall be performed at each shut-pressure of P Lt.is less than or equal to down for refueling or approniaistely every 18 t
l.a (Ltm/ Lam).
If Ltm/ Lam is greater thqn 0.7, months until two consecutive Type A teste l!
Lg is (specified as equal to) La IP /Pa)).
meet the c.bove requirements, at which time t
j the normal test echedule may be resumed.
(5) The total measured leakage rates at
<i pressures of P, and rg are L.,,and 1.g.,
d.
he accuracy of each Type A test shall be lI respectively.
yerified by a supplemental test which:
l When primary containment integrity le re-(1) Confirms the accuracy of the test by 1i quired, primary containment leakage rates verifying that the difference between shall be limited to the supplemental data and the Type A terst-data is within 25 percent of La
~
(t) An overall integrated leakage rate for.
25 percent o f L.-
nr 4 j Type A teste oft t
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'I (2) lies a duration suf ficient to establish l
ji (a)
La. less than or equal to 75 per-accurately the change in leakage rate i
i.
cent of I...
between the Type A test and the supple-(b)
L't. less than or equel to'75 per-cent of L.
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', t A;;;endment No. -6b, 63
- 109 l
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3.7 LlH1TlHC CONDITION FOR OPERATIOil 4.7 5URVglLLANCE REQUlitEllENTS I
(2) (a) A combined leakage rate of less Elia n J(3) Requires the quantity of gas injected
- l' l
or equal to 60 percent of L for all lato the containment or bled from the li testable penetrations and Isolation containment during time supples ental test f;i valves subject to Type 8 and C teste to be equivalent to at least-9 scfm.
I
%ll f
encept for main. steam leolation i
valves.
- e. " Type II and C teste olist! be conducted at r.
l{
e (b) A teskoge rate of less tlian or equal at Intervals no greater than 24 anonths except
.i to 3.75 percent of La for any one for teste involvings l
f air lock when pressurlzed to 10 polg, j
(1) Hein steaci !!ne isolation valves vinich I'
(c) 11.5 SCF per hour for any main steam shall be tested at a pressute of 25 pelg
'+
isolation valve at a test prcosure of each operating cycle, 25 peig.
i (2) Bolted double gasketed seals which shall I
be tested at a pressure of 48 pels when-ever tina seat le closed af ter being opened and escle operating cycle.
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(3) A!r locks which shall be tested at 10 pels each operating cycle.
f.
Continuous Leak Rate Honttor
,)
(1) Mien the primary containment le Inerted, 1
1 the contaltunent shall be continuously i
j-
' monitored for groes leakage by review of the inerting system miske-up require-mie n t s.
i (2) This monitoring sys*.e stay be taken out of service for the purpose of snaintenance or testing but shall be returned to service as soon as practical.
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3 The Interior surf aces of Llie drywell shall be
, i:
I visually inspected escle operating cycle foy.__
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evidence of deterioratiosi.
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kendmert No. -60<, 63 110
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