ML19309G479
| ML19309G479 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/16/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309G480 | List: |
| References | |
| NUDOCS 8005070031 | |
| Download: ML19309G479 (53) | |
Text
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,,g UNITED STATES NUCLEAR REGULATORY COMMISSION 2
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j WASHINGTON, D. C. 20555 f
COMMONWEALTH EDISON COMPANY DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 License No. DPR-25 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by the Comonwealth Edison Company
~~
(the licensee) dated December 10, 1979, as supplemented on February 6 and March 24, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 3.B and 3.E of Facility License No. DPR-25 are hereby amended to read as follows:
< l 3.B Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 42, are hereby incorporated in the
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license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.E Restrictions Operation in the coastdown mode is permitted to 40% power.
Should off-normal feedwater heating be necessary for extended periods during coastdown (ies., greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) the Licensee shall perform a safety evaluation to determine if the MCPR Operating Limit and calculated peak pressure for the worst case abnormal operating transient remain bounding for the new condition.
~ ~'
This license amendment is effective as of the date of its issuance.
3.
FOR THE NUCLEAR REGULATORY COMMISSION
'$I.,-- cy Thomas A. Ippolipo, hief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: April 16, 1980 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 42 FACILITY OPERATING LICENSE NO. DPR-25 DOCKET NO. 50-249 Replace the following pages of the Appendix "A" Technical Speciffcations with the enclosed pages.
2 81B 4
818-1 5
81C-1 6
81C-2 7
82 8
85A 9
85B 10 86A 11 90 12 125 13 14 15 16 Add pages 81C-3, 18 81C-4 and 19-81C-5 20 21 22 24 29 34 38 40 42 42A 46 47 48 49 57 57A 62 62A 62B 63 71 i
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I.
Limiting ~ Conditions for Operation (I.CO)
'the N.
Mode - The reai: tor node is that which is limitiiig c~oiMitlons for operation specify the established by the mode-selector-switch, minimum acceptable levels of system perform-
.mce necessary to as iure safe startup and op-O.
Operable - A system or componont shall be crat ion of the facil (ty.
When these conditions ionsidered operable when it is capable of are met, the plant c :n be operated safely and performing its intended function in its re-abnormal sit et.iwas m be safely controlled.
quired manner.
J.
Limiting
), tem Setting (LSSS)
"Ihe P.
Operating - Operating means that a system limiting y / stem settings are settings on or component is performing its intended instrume at...m which. initiate the automatic functions in its required manner.
protective action at a Icvel such that the safety limits will not be exceeded. 'the region Q.
Operating Cycle -Interval between the end between the safety limit and these settings of one refueling outage and the end of the represents margin with normal operation lying next subsequent refueling outage, below these settings. The margin has been established so that with proper operation of the R.
Primary Containment Integrity - Primary instrumentation the safety limits will never be containment integrity means that the drywell cxceeded.
and pressure supprossion chamber are intact and all of the following conditions are satisfied:
K.
Praction of Limiting Power Density (FLPD)
The fraction of limiting power density is 1.
All manual contalnment isolation valves on the ratio of the Linear IIcat Generation lines connecting to the reactor coolant sys.
Itate ( LilGit) existing at a given location tem or containment which are not required to the design LIIGP. for that bundle type, to be open during accident -onditions are closed.
- 2.. At least one door in cach airlock is closed and scaled, Log'ic System Function Test - A logic sys-h.
tiim Tuiictiona1 test mcans a test of all relays
~
3.
All automatic containment isolation valves and contacts of a logic circuit from sensor are operable or deactivated in the isolated to activated device to insure all components are operable per design intent. Where possi-position.
ble, action will go to completion, i.e., pumps 4
All blind flanges and manways are closed.
will be started and valves opened.
S.
Protective Instrum*yt_ation Definitions H.
Ilinimum Critical Power I atio (MCPit)
'the 1.
Instnunent Charanel - An instrument chan-minimum in-core critical power ratio nel means an arrangement of a sensor and corresponding to the most limiting fuel auxiliary equipment required to generate assenibly in the core.
and transmit to, a trip system a single trip signa 1 related to the_ plant parameter ~
2 Amendment No. 42 manifored'by-tinateinstrument channel.
Z.
Secondary Containment Integrity - Secondary 8 11 Simulated Auto'matic Actuation - Simulated
- containment integrity means that the reactor autematic actuation means applying a simu-building is intact and the following conditions late 1 signal to the sensor to actuate the are met:
circuit in question.
1.
At least ong door in cach access opening CC.
Surveillance Interval - Each surveillance is closed.
requirement shall be perforned within the specified surveillance interval with:
2 The standby gas treatment system is
- operable, n.
A maximum allowa'ble extension not to exceed 25% of the surveillance interval.'
3.
All automatic ventilation system' isolation valves arc operabic or are secured in the b.
A total max'imum conbined interval time isolated position.
for any 3 consecutive intervals not to exceed 3.25 times the specified AA.
Sh'utdown
'Ihe reactor.is in a shutdown con-surveillance interval, dition when the reactor mode switch is in the shutdown mode position and no core alternations DD. - Fraction of Rated Power (PRP) -
are being performed. When the mode switch is The fraction of rated power is the placed in the shutdown position a reactor ratio of core thermal power to rated scram is initiated, power to the control rod thermal power of 2527 Mwth.
drives is removed, and the reactor prot ec-tion system trip systems are de-energized.
EE.
Transition Boil'ing - Transition boiling means 1.
Ilot Shutdown means conditions as abo've the boiling regime between nucleate and film
- wi th reactor coolant temperature greater boiling. Transition boiling is the regime than 212*F.
in which both nucleate and film boiling occur intermittently with neither type 2.
Cold Shutdown means conditions'as above being completely stable.
with reactor coolant temperature-equal to or less than 212*F.
FF.
Maximum Fraction of Limiting Power Density (MFLPD) - The maximum fraction of limiting power density is the highest value existing in the core of the Praction of Limiting Power Density (FLPD).
Amendment No. 42 4
3 J
2.1 LIMITING SAFETY SYSTEM SETTING 1.1 SAFETY LIMIT 2.3. FUEL CLADDI!:G ISTr.Cn1TY INTECitITY 1.1 [Itri. C!.An;st::C Ap_ plica 5tlity
_ Applicability _
The Limiting Safety System Settings 1hc Safety Limits established to apply to trip settings of the instru-preserve the fuel cladding integrity ments and devices which src provided apply to these variables which' to prevent the fuel cladding integ-conitor the fuel thermal behavior.
rity Safety Limits from being ex-cccded.
objective Objective The objective of the Limiting Saf e-The objective of the Safety Limits ty System Settings is to define the Is to establish limits'bclow which.
level of the process variables at the integrity of the fuel cladding which automatic protective action is preserved.-
is initiated to. prevent the fuel clad-ding integrity Safety Limits from being exceeded.
Specificat' tons Specifications A.
Pcactor Pressure >800 psig and Core Fi ct.- > 107,of Rated.
The limiting safety system Erip settings shall be as specified The existence o a cinimum critical below:
~
j poteer ratio (l:CPR) less than 1.07 shall constitute violation uf the MCPR fuel cladding integrity safety limit.
Amendment No.
42 5
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2.1 LIMITING SAFErY SY5TSM SEITING 1.1 SAFETT LIMIT
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1 ATF;4 F1.:x Scre-. Tri.) Settine (P.un Mcdo)
When the reactor code switch is in the run por.iticia, the APPat flux scram setting shall ues
-.6SJ + SS,-
a I,
S <
D with a es. int:m set point of 120% for core 6
flow equal to 98 x 10 lb/hr end greater, whores S - setting in por cent of rated power V;f-per cent of drive flow required to produco a rated core flow of CS Hib/hr.
In the event of operation with a mantran fraction of limiting :*ower density (F'JLPC) grcater tran the f raction of rated power (fm.'. the setting shall be modified as follows:
- 1 (.6W + 55) [ h ]
g l
Were:
fitP = fra:tica of rated thermal power (252'1 14Wt) g PJ'JD = maximes fraction of limitleg power density where the limititig M
j power density for each L.un:lle is the design linear be.-t i
g:r :iatiosi s ate for that t,ui.dle, kO) f' lhe ratio of IW'f tfD shall be set equal to 1.0 unless the actual 1.0. In which case the operatir.3 wafee is less than 9
actual operating value will be used.
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M Ei50 11r '
- z. ir.,no.,sm, w ceu-. 7,rc.3 e Strritis - s* Fet S tar *tv 1:q Mien t? e, reactor an.lc switch is in 8:ic refuel or s tart:n,*/tt : se c.t.ay pogg.
t ion. Lt.c Ar.*Jt s c r.sa. s?ia l t 1,e s e g at acalcisor et.at i. 15:.i ica.ieuse flus.
Amendment No. 42 6
i
1 1.1 "SAFnY LI:11T 2.1 ' LII'.ITING SA.7,TY SYSTEM SETING B. Core Thernal'Foner f.imit (Reactor 3.
I.M Flux Scran Trin Settine:
Pressure.1 000 psit:)
Tiic trat flis. scrata setting shall bc When the reactor pressure is < 600 set at less than or equal to 120/125'or Psig or core flou is less than 10
{ull scale, of rated, the core thermal power shall not exceed 25 perceitt of rated
- thermal po :ce.
3 AFaM Red Bicek Setting i T C. Power Transient b
The APRM rod block sotting shall bc" Q
1.
'Ihc neutron flux shc11 not exceed the scrs=
setting esteb11shed in Specification 2.1.A l
for Icocer than 1.5 seconc:, e indicated by l
S <
65f + A3 9
the process cosputcr.
C The definitions used above 'for'the AP?,M ocram W
- 2. Ynen the process co ? uter is out of.crvice, trip spply.
j this safety liett shall be assumed to Sc In the event or operatico with. cantzum tr:etton 11:stias power denciar (e....>)
creceded if the neutron flux exceeds the sorsa sreat ry an ste traction or rated power (>72), ti.e. ttias statt to sossrtes setting established by Specification 2.1. A 7M and a control rod scram docs not occur.
j s3(.65v,et.3)(
b 7te darinitices u:ed stove ror the A!?J: scras trip a; ply.
D.
Reactor Vater Level Shutdown con,3ition}
The ratio,og r,otP to M,itJn shall,be sc t, equal,,to 1. G unle,,ss the,,,act ual o,,perating
,,3, i,,,,
,,,3,
, 3,,,3,
,g
,ggi,
,,,a, Micnever the reactor is in the shecdcun condition with irradiated fuct in the reactor vessel, the
. water Ic ent shall not bc Icys than that corres-pes. ling to 12 inches above the top of the active I
it:el* vhen it is scated in the core.
- Top of active fuel is definied to be 7
360.. inches--above vessel zero - (see bases 3.2).
Amendment No. 42 i
)
l 2.1 LIMITING SM'ETY SYSTEM SETTIllG 1.1 SAFETY LIMIT Reactor 1crr water level scraa setting shall C.
fuel
- be 1 14 4= above, the top of the active at normal operating conditions.
Reactor lov untg'r Icvel ECCS initiation D.
shall be 84a d ) above the top of the O
active fuel
- at norcal operating conditions.
Turbine stop valve scram shall bc < 10%
E.
valve closure from full open.
F.
Generator Load Itejection Scram shall initiate upon ac:uation of the fast closure solenold valves which trip the turbine control valves.
G.
Slain Steaintinc Isolation Valve C!osure Scram shall be $ 10!. valve closure from full open.
C M
Main Steamline Pressure initiation of main H.
steamline isolation valvo closure shall bc EEE2) 1850 psig..
E-a ed Turbine Control Valve Fast Closure Scram on I.
loss of control oil pressure shall be set b
at greater than or equal to 900 psig.
8 y
- Top of active fuel is defined to be.360 inches above vessel zero (see. Bases 3.2).
8 Amendment flo. 42
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A
1.1 Safety Limit Bases FUEL CLADDI iG INTEGRITY threshold, beyond which still The fuel cladd!ng integrity limit,is Greater thermal stresses may set such that no calculated fuel dam -
cause gross rather than incre-age would occur,as a. result of an mental cladding deterioration.
abnormal opera tional transient.
Be-Therefore, the fuel cladding cause fuel damage is not directly Safety Limit is defined with cbservable, a step-back approach margin to the conditions which is used to establish a Safety Limit would produce onset of transition such that the minimum critical oower ratio,(i'.CPR) is no less than th'e MCPR fuel boiling, (HCPR of 1.0).
These l
conditions represent a significant cladding integrity safety limit.
MCPR pthe departure from the condition in-NCPn tuel cladding integrity safety limit tended by design for planned represents a conservative margin relative to the cosyditions required to maintain fuel cladding operation. Therefore. the MCPH fuel g;g;;;; p,p,UM..LIZ"'. f.%M," *,',;;"y,y,tja
,..., n. i. i... n....... o.,
i incesrity-
- t on s.u.
** "UE ' I." ' " "' ' '" * ** '" ' ""' 8 '
d""" *"* 8 a The fuel cladding is one of the N
physica1 baer1ers uhich separate L A.
Reactor Pressure >800 psig and rad ioactive raterials from the Core Flow > 10% of Rated.
environs.
The integrity of this
.M cladding barrier is related to its Onset of transition boiling results O
relative freedom from perforations in a decrease in heat transfer from g
or cracking.
Although some cor-the clad and, therefore, elevated reaion or use related cracking may clad temperature and the possibility M
occur during the life cf the of clad failure.
however, the c' adding, fission product m1Eration existence of critical power, or O
f ro:r. this source is incrementally boiling transition, is not a direc.tly M
cumulative and continueusly observabic parameter in an operating measurable.
Fuel cladd in5 per-reactor.
Therefore, the margin to
$379 forations, houcver, can result from bo111:1g transition is calculated ther al stresses which cccur from fron plant operating parameters such M
reactor operation significantly as core power, core flovt, feedwater D
aio/c design conditions and the pro-ternperature, and core power d istri-tect'on system safety settings.
bution.
.The margin for cach fuel b
%hile riusicn product migration from assembly is characteri ed by the c1Mding perforation 'is just as critical pcwer ratio (CPRI which is neasurable as that from use related the ratio of the bundle power which c r a c k : r:r., the therm 911y caused would produce oncet of transition cla M h:g peri' orations signal a 10 Amendment No. 42
.4arcty timit C.wes Ifowever, if boiling transition were to occur, clad perforation'would not 1.1.A Reactor Pressure W800 psig and be expected.
claddsns temperatures Core Flow > 10% of Ra ted.
(cont'd) would increase to approximately 1100 F which is below the perforation boiling divided by the actual bundle power.
t:mperature of the claading material.
The rainimum value of this ratio for This has been verified by tests in any benvile in the core in the minimum the General Electric Test Reactor critical power rat to (14CPR ).
It is (CETR) where s imilar fuel opera ted assuned that the plant operation is above the critical heat flux for a controlled to the nominal protective s1Cnificant period of time (30 setpoints via the ins trumented vari-minutes) without clad perforation.
ables.
(Fir,ure 2.1-)).
If reactor pressure should ever 1f00 psia during normal power I
The 14CPR fuel cladding integrity safety limit has exceed 1
sufficient conservatism to assure that opera tion (the limit of applicability in the event of an abnormal operational of the boiling transition correlation) transient ini tia ted from a norraal it would be assumed that the fuel O
caerating condition more than 99 9%
cladding integrity safety Limit has D
o'r the fuel rods in the core are ex-been violated.
g pected to avoid boiling transition.
'lhe margin between 14CPR of 1.0 (onset i
of transition bolling) and the safety In addition to the boiling transition lirait 11 r. I t' is derived fro:n a detailed O!CPR) operation is constrained to a maximum P. 8) stet!stical analysis considering all IJIGR - 17.5. kw/ft for 7 x 7 fuel and 13.4 kw/fr f r ali exo fuel types. Thia e n traint is st.ablinhed by o f
- o-l*..
unCe r ta j !!t les in monitorills
" - '7 specification 3.5.J to provide adequate safety margin to in C/
the core operatiac state includins viastic strain for annormai overating tronoienes initiated uncertainty in the bollin3. transition from high power conditions. Specification 2.1.A.1 provides FM Colrelations ac e e. g. Heference (1)*
I r equivalent safety margin for tranatents initiated from lower power conditions by adjusting the APRM flow blaced i scram by the ratio of FRP/MPI,PD..
Specification 3.5.J
{
,Because the Dolling transition Cor-established the IJIGR saax which cannot be exceeded under steady reletten is based on a large quantity Power operation.
of full scale data there is a very h'gh confidence that operatJon of a (1) aceneric Reload ruel Appilcation, Hens-24011-r-A*
' N'
- I nwi=ber at th reload fun analye
[,
I.
- t. i
! CI R f( eL addir g integrity p
u.s t e ty 1imit would not produce boiling-l transition..
11 Amendment No. 42 J
Safety Linit Bases (cont'd)
B.
Core Thermal Power Limit (Reactor Pressure < E00 psia)
At pressures below 800 psia, the which will not allow the reactor to core elevation pressure drop (O be operated above the safety limit during n. rmal operation or during o
power, O flou) is greater. than 4 56 psi.
At low pouers and flows this other plant operatin5 situetions which pressure dif ferential is maintained have been analyzed 'in detail.
In addition,. control red scrams are such in the bypass region of the core.
that for normal operating transients Since the pressure drop in the bypass the neutron flux tr7nsient is terni-region is essentially all elevation nated oefore a significant increacc head, the core pressure drop at low in surfoce heat flux occurs.
pouces and flows will always be greater than 4.56 pai.
Anal ses show that Control rod scram times are checked as with a flow of 28x10 lbs/hr. bundle required by Specification 4.3.C.
Exce e'd-flou, bundle pressure drop. is nearly independent of bundle power and has
(~qgf a value of 3 5 psi.
Thus, the bundle inn a neucron flux scram setting and a failure of the control rods to reduce flo. with a 4.56 pst drfv!ng head
())3 w111 be greater than 28x103 lbs/hr.
flux to less than the scrum setting within 1.5 seconds does not nececqarily, b5 3)
Pull scale ATLAS tes t da ta taken at E3531 pressures from 14.7 psia to 800 psia inply that fuel is da;aaged;
- however, I::d ica th tha t the fuel assembly fer this specification a safety limit.
critical powc r a t this flew is approxi-violation will be assumed any tine a.
((]3
..ctely 3.35 bt.
At 25% of rated neutron flux serem setting is exceeded thermal po,,er, the peak powered bun-for longer thfn 1.5 seconds.
die would have to be operating at
'~~
hbbk" 3.8% times the averate powered bundle If the scram occurs such that the neu-tron flux duell time above the limit-in order to achieve this bundle power.
cgg thermal power limit of ing safety system sett*:A is less than Thus, a core
$UEE3 25,9 for reactor pressure's.belos 800 1.7 seconds, the safety ilmit will r.ot be exceeded for normal turbine or gen-b-
psia is conservative.
era tor trips, which are the mos t se cre normal operatias transients expected; C.
Power Trans! cot Toese analyses shew that e icn f the bypass system fails to operate, the During transient operation the heat flux (thermal power-to-wa ter) would la6 be-dee lga lim.it of :.';Pa =. the MCPR fuel hind the neutron flux due to the inherent cladding integrity safety limit is not transfer time constant of the fuel exceeded.
- Thus, nsa ora j n heat
,.hich is 8-9 seconds.
Also, the limiting second limit provides additionai margin.
care v synt.m scram settings are at values fmendment No. 42 12 1
- 2.1 Limitincq Safety System Setting Bases 1.[
Safety L1rait Bases FUEL CLADDING INTEGRITY 101.C Power Transient (cont'd)
The. abnormal operational transients The computer provided has a celuence annuncia tion program *which applicable to operation of the units will indicate the acquence in which have been analyzed throug;hout the scrams occur such as neutron flux, spectrum of planned operating con-pressure, etc.
This program also ditions up to the rated thermal power indicctes when the scram setpoint is cond ition of 2S?"T MWt.
In addition, cleared.
This will provide information 2S07 Iult is the licensed maximum steady-on how long a scram condition exists state power level of the units.
This maximum steady-state power level will and thus prov3de some measure of the energy added during a transient.
- Thus, never knowingly be exceeded.
computer information normally will be available for analyzing scrams; how-Conservatism is incorporated in the if the computer 1ryformation should transient analyses in estimating the
- ever, not be ava ilable for any scram analysis, controlling factors, such as void
.ipec i fica t ion 1.1.C.2 w ill be relled on reactivity coefficient, control rod to deterraine if a safety limit has been scram worth, scram delay tirre, peaking factors, and ar.lal power chapcs.
These violated.
factors are' selected conservatively Dur1ng periods when the reactor is shut with respect to their effect on the k__3 down, consideration must also be given applicable transient results as deter-g to wa ter level requirements due to the mined by the current analysis model, effect of decay heat.
If reactor water Conservatism incorporated into the auvel should drop below the top of the transient analyses is documented in g
active "ucl ouring this time, the ite fe r e nce 1.
Transient analyses are ability to cool the core is reduced.
initiated at the conditions given in this This reduction in core cooling cap-reference.
.:9111ty could lead to elevated cladding The temperatures and clad perforation.
core will be cooled sufficiently to pre-vent clad melting should the water level M
(f-- '
be reduced to two-thirds the ccre height.
' Establishment of the safety limit at 12 l
1:sches above the top of the fuel *provides adequate margin.
This level will be con-tlnuou'ily monitored whenever the recir-colation pumps are not operating.
l j
^ Top of active fuel in defined to be 13 360 inches above vessel zero (see bases 3.2).
Amendment No. 42 s
E e l.
. hiting Safety System Setting Bases S.tcady-state operation without, force Fuel Cladeling Integrity (c.ont ' d )
reoirculation will not be permitted' except during startup testing.
analysis to support operation atThe The absolute value of the void reac-various power and flow rela tionsh1 is tivity coeffic!cnt used in the analysis has is conservatively e'stimated to be about considere'd operation with eith 25% greater than the no ninal maximura one or two recirculation pumps-va lue expected to occur during the core The bases for individual trip cetti lifetime.
The scram worth used has are discussed been derated to be equivalent to appro-graphs.
in the following para-xi:::ately 80% of the total scram worth of of the control rods.
The scram delay b[theanalysesareconservativelyset For analyses of the thermal consequences of t% and rate of rod insert 1on allowed the transients, the, MCPR's stated in paragraph 3.5.K as the l iny t t i ng connittoi o operation "4 ':a ^' to the longest aciny and slowest insertion ra te acceptable by Tec,hnica1 bound those which are conservatively assumed Specifications.
Tne effect of., cram to exist prior to initiation of the transients.
worth, scram delay tirac and rod in-A.
Neutron Flux Trip Sett$ngs~
sertion ra te, all conserva tively applied, are of greatest significance 1.
APHM Flux Scram Trip Settin((Run Mode)
In the early portion of the negative reactivity insertion.
The rapid in-The average power range monitoring sertion of negative reactivity is (APHM) system, which is calibrated l accured by the time requirements for using heat balance data taken during 5% and 20?.'re 60% inserted, s teady-s ta te cond itions, reads in insertion.
By the time the rods e approxi-mately four dollars of negative reac-percent of rated, thermal power.
Be-cause flesion chambers prov::!c the basic Q
tivity have been Inserted which input signals, the i.PHM syste:a responds strongly turns the transient, and d irectly to average acutron flux.
accomplishe3 the desired effect.
- The, M
ti:.es for 50% and 90% insertion are During trans ients, the instantaneous rate ~ of heat transfer from the fuel given to assure proper o:npletion of (reactor ther:nal power) is Icss than h
the er.pected performance in the the instantaneous neutron flux due to carl:er portion of the transient, the time constant of the fuel.
There-c=:=a and to establish the ultimate fully fore, during abnormal operational shutdown steady-stnte condition.
transients, the thermal power'or the c===3 fuel will be less than that indicated M
This choice of us1ng conservative values by the neutron flux at the scram setting.
l g of controlling parameters and initiatin$
Analyses demonstrate that with a 120 t'
transicots a t the design power IcVel, percent scram trip settinr,. none of the i
produces more pess imis tic ansucra than abnormal opera tional transients: analy cd t.o u l.1 result by using expected values of violate the fuel Safety Limic and there cor. trol narrmeters and analyzing a t higher is a subs tantial marr.in from fuel d,n.'!ge.
l no,.er ] chela.
Therefore, t.hc use of flow references:
scra:;. trip provides even ad<11 tion:1 nrrg'n. 34 i
Amendment No. 42 i
l
-..utron* Flux Trip Se ttings
. ) A.
s/
APRM Flux Scram Trip) Setting 1.
(P.un Mode)
(cont'd ture coefficients ure small, and con-in the APRM scram trip trol rod patterns are constrained to An increase
~
setting 5.ould decrease the mar 61n pre-be' uniform by operating procedures cent before the fuel claddin5 integrity backed up by the rod worth minimizer.6 Strety Limit is reached.
The APRI4 Of all possible sources of reactivity
= cram trip sett1ng was determined by an analysis of cargins required to pro-Input, uniform control rod withdrawal is the most probable cause of signifi--
y vide a reasonab'le range for maneuvering cant power rise.
Eccause the flux during oparation.
Reducing this oper-distribution associated i:sth uniform D
ating ::argin would increase the fre-rcd withdrawals dc?s not involve high
(;uency of spur:ous scrams which have an local peaks, and because several rods M
adverse effect on reactor safety because' must be moved to change power by a of the rceulting thermal stresses.
- Thus, g
the I.PRM scram trip setting was selected significant percentage of rated power, the rate of power rise is very slo.1.
M because it provides ade.guate r:3rgin for Generally, the heat flux is in near the fuel clndding integrity Safety Limit c(;uilibrium with the fission rate.
In Db yet allows operating r.argin that reducca an assur.cd uniform rod withdrawal ap-the possibility of unnecessary scrams, proach to the scram level, the rate of power rise is no mere than 5 percent 2B The scram trip setting must be ad, Justed of rated power per minute, and the
=z to ensure that the L!!3R transient ' peak 7
increased for any co:nbination of.
APRH syste:. would be more than adequate
's not to assure a OCram b0 fore the power
..se. re act ion of Lan t ting ro es pens 6 t y arrt ron en s m e.ctos Could eXCCCd tlDsafety limit.
The 15 core 2I.
Pb Is d N
het en t e
e p e r g. n t f,p 3?q O C r a.~ TcI2Ana aCt'iVe un-44 erection of mated ro-ee (ters.
til the mode 3;iteb is placed in the RU" position.
This nvitch occurs when reactor pressure is greater than 850 k
poig.
s 2.
PP:4 Flux Scram Trip Setting (itafuel or Start & !!ot Stahdby Mode) 3 IRM Flux Screm Tr19 Setting For operation in the startup mode while The IRM system consists of 8 chambers, the APRI4 the reactor is at low pressure, scram setting of 15 percent of rated power 4 in each of the reactor protection g
provides adequate thermal margin between the system logic channels.
The Iris is a the setpoint and tne safety limit, 25 pir-5-decade instrument which covers the cent of ra ted.
The margia is adequate to range of power level between that eccommodate ant:cipated maneuvers ascociated covered by the Sif-? and the APR:h The with pot.cr plant startup.
Effects of in-5 decades are bro %en ecwn into lo ranges,-
creasing presnure a t zero or lou void con-each being one-half of a decade in size.
tent are minor, cold watep from sources ev >iler,le d f rine, startup is not much colder 15 2
.i.
tF:t air mi:t in th syatem, ti rear i-1
2.1.A.
1:eutron Flux Trip -Setting *
- 3. IF.:4 Flux Scram Trip Setting (cont ' d')
2.1.B APRM Ro.1 Block Trip Setting The Inti scram trip setting of 120
'ivisions is active in each range.of d
Reactor power level may be varied by the IRFi.
For excmole, if the instru-movin5 control rods or by varying
- nent were on range' 1, the scram setting the recirculation flow rate.
The APRM would be a 120 divisions for that range:
i system provides a control rod block,to l!kewise, i f t.be instrument were on range pre vent gross rod withdrawal 5, the scram would be 120 d ivis ions on at constant recirculation flow th?t range.
'I hus, as the 1Rii is ranged rate to protect ac.* tinct grossly exceed-up to accomodnte the increase in power ing the MCPR fuel, cladding integrity IcVel, the scram trip setting is also safety limit.
This rod ran.,,d up.
block trip. setting, which is auto-matica11y varied with recirculation The most significant' sources of reac-1 p fl w rate, prevents an increase tivity change during'the power increase in he reactor power level to c::ces-are due to control rod withdrawal.
In 8IV" "UIUUs due 60 control rod with-I order to ensure that the 3Hi4 provided d rawa l.
'l,he f, low varihblo trip setting adequate protection against the single pr vides substa:1tial margin from fuel Q
rod uithdras.al error, a range of rod asage, a su::Ing a r.tead:1-stste cpera-withdrcual accidents was en?lyzed.
This
- he r. y en = ig.,
ov3r the a n:2 3 ys s included starting the accident ent ire rec irca.e t.f od f,t :u r: r.ge.
Th.;
at various now:. r Icvels.
The most se-b i$svolves an Initial cond ition
[.i q.n t th: S Tety L1: nit increases es vere case
- hU ' Uw decreaccS f0r 1,he spec;f'e.1 M
in t.hich the reactor is just cuberitical 1 rlp seto.ng vem:s flow reintien;33p; C==:3 e nti the IP..*I systun is not vet on sc.:le.
encrercre the ucrst case MCPR which g
Addit!onal conserva tism wns taken in this ceuid occur during stece y-state escra_
g entalysis ny assuul::g that the liti4 channel t:en is at 10C% or rated thermel jicwer c:: =.3 cloaest to the w*thdrawn rod M bypOssed.
becouse of the APDi rod block tria The results of t.his analysis shot. that the settin3 The actual power distribution 3
reactor is serar:..:ed and peak power limi ted in the core is established by specified E*
to o:se percent of rated pover, tous maintaining control red sequcnces and !s monitored j
EPR above the !!CPR fuel cladding integrity cor.tinuot: sly by the in-core LPR;*. sy te=.
.,s with the APRM scran t. rip setting, j sa fe ty licit.
Isased on the above tiac APh:i Por! block t"1p settins is ad-analys:3, the 11131 provides protection against q
loccl coatrol rod wit,hdrawal errors and con-justed downward if the iraxi:nu:
fraction t h.o in withdratal or control ruda In se.:e'mce of limiting power density evceeds the a m. prov icea ocu;<up protec c io: for the APM:4 f r ac t gn, o f r a t ed powe r, thus preservina
..:e A s.... red te loc k s 2 fe ty ma r:;in.
16 Amendment No. 42
Turbino Stop Vnivo Scran - Tho turbino otop valvo C.
_ Reactor Coolant Low Pressure Initiates Main Steam closure ucran trip caticipates tho vressuro, Isolation Valve Closure - The low pressure isolation
- r. cut.ron flux ar.4 h at flux incresso that could at 850 psig was provided to give protection against result frc., rr.pid c1ccuro of the turbine stop fast reactor depressurization and the resulting valvss. i!ith a scran trio cctting of 10 rapid cooldown of the vessel. Advantage was taken percent of valve closure fron full open, the of the scram feature which occurs when the main resultant increaso in surface heat flux is steam line isolation valves are closed.to provide lir. Mad such that I CPR re: rains above t the' MCPR for reactor shutdown so that operation at pressures
- E Y# """
fuel cladding imitegrity safety limit, even safety limit does not occur, although operation during the worst case transient that assumes at a pressure los.cr than 850 psig would not necessarily the turbine bypass is closed.
constitute an unsafe condition.
H.
Main Steam Line Isolation Valve Closure Scram - The low pressure isolation of the main steam lines at F.
Generatot Load Re_hetion Scram - The genera-850 psig was provided to give protect a n against tor loed rejection scram is provided to rapid reactor depressurization and the.resulting cnticipate the rapid increase in pressure rapid cooldown of the vessel. Advantage was taken of the scram feature which occurs when the main and neutron flux resulting f rom.
steam line isolation valves are closed, to provide fast closure of the turbino control valves for reactor shutdown so'that high power operation due to a loc _4 rejection and subsequent at low reactor pressure does not occur, thus providing fe.11ure of the byyans; i.e.,
it prevents Z5.'! frcr; becoming less' than the MCPR fuel Protection for the fuel cladding integrity safety c l.uld i ng integrity safety limit for this
. limit. Operation of the reactor at pressures lower transient.
For the load rejection without than 850 psig requires that the reacror mode switch byp.u;s transient from 100% power, the peak be in the startup. position where protection of the heal. flux (and t.he re fore LilGH) increases on fuel cladding integrity safety limit is provided by the onder of 15% which provides wide margin the Ilm high neut ron flux scram.
Thus, the combination i to the value corresponding to 1% plastic strain of main steam line low pressure isolation and isolation lof the cladding.
valve closure scram assures the availabijity of neutron flux scram protection over the entire 8 U range of applicability of the fuel cladding integrity b
safety limit.
In addition, the isolation valve closure scram anticipates the pressure and flux Q
transients which occur during normal or inaIvercent isolation valve closure. With the scrams ser at g
10% valve closure,there is no increase in neutron flux.
c===
&=E) c===>
18 i
g Amendment No.
42 y
i g===>
j
1.2 SAFETY LittlT 2.2 LitiITING SAFETY'SYSTFli SETTING l '. 2 REACTOR C00LAtlT SYSTI21 2.2 REACTOR COOLANT SYSTEM Applicability:
gplicability:
Applies to limits on reactor coolant system Applies to trip settings of the instruments and devicca which are provii.ed to prevent the reactor pressure.
system safety limits from being exceeded.
Objective:
Objective:
To establish a limit below which the integrity To define the level of the process variables at of the reactor coolant system is not threatened which automatic protective action is initiated to due to an overpressure condition.
prevent the safety limits from being exceeded.
Specification:
Specification:
% e reactor coolant system pressure shall not A.
Reactor Coolant High Pressure Scram shall be exceed 1325 psig at any time when irradiated fuel
$1060 psig.
is present in the reactor vessel.
B.
Primary System Safety Valve Nominal Settings shall he as follows:
1 valve at lll5 psig*
2 valves at 1240 psig 2 valves at 1250 psig 2 valves at 1260 psig 2 valves at 1260 psig The allowable setpoint error for each valve shall be +1%.
- Target Rock combination safety / relief valve 19 Amendment No. 42 s
J
Rases:
2.2 In compliance with Section III of the ASME Code the cafety valves sunt be set to open at no higher then 10)I of design pressure, and they avat 11 mitt the reactor pressure to no more then 110% of denign prensure. Both the neutron fluxecrc:a and safety valve actuation are required to prevent overtirce-surtr.ing the reactor preenure vcosel and thus exceeding the pressure orifety 11rntt.
"Ihe preneure scram in available as a backup protection I
to the high flux scr.un.
If t.he high flux scrara were to fail, a high pressure cram would occur at 1060 psig. ' Analyses are performed as described in the Generic Reload Puel l
Application, NEDE-240ll-P-A (Approved revision number at time reload analyses are performed) for each reload to assure that the pressure safety limit is not exceeded.
. i I
l l
l l
[
l 21 Amendment No.
p J
Basest ne relationships of stress levels to yield strength 1.2 De reactor coolant system integrity fe an impor-are comparable for the isolation condenser and tant barrier in the prevention of untontrolled re-primary system piping and provide a similar cLst-lease of fission products. It is essential that the gin of Protection at the established safety pressure integrity of this system be protected by establishing limit.
a pressure limit to ';e observed for all operating cenditions and whenever there is irradiated fuel in The normal operating pressure of the reactor coolant the reactor vessel.
system is 1000 psig. For the turbine trip or loss of electrical load transients, the turbine trip screm or he pressurst s.ifety limit of 1325 psig as o.casured generator load rejection scram, together with the by the vessel steam space pressure indicator is turbine bypass system, limit the pressure to approxi-e<;uivalent to 1375 psig at the lowest elevation of the cately 1100 psig (4).
In addition, pressen e relief reactor coolant system. The 1375 psig value is valves have been provided to reduce the probability derived from the design pressures of the reactor pT pressure vessel, coolant. systers piping and isole-of the safety valves,which discharged to tion condenser. Le respective desir,n pressures the drywell, operating in the event th. t Q
are 1250 psig at 575*F,1175 psig at 560*F, and 1250 -
the turbine bypass should fail.
psig at 575*F.
The pressure safety limit u.'s chosen g
as the lower of ti.e pressute transients pera.iitted i
Finally, the salety vaives are sizec to ateep by the applicable design codes: AS fE Coller and the reactor coolant system pressure below 1375 rsig pressure Vessel Coda, Sectien III for the pressure with no credit taken for the relief valves during the
.essel and
- isolation condenser and 1*SASI S31.1 Code
%j fcr the reactor coolant system piping. Re ASrrg postulated full closuro of all 14SIV's without direct (valve position switch) scram. Credit T, oiler and Pressure Vessel Code permits preunure Q
transients u;> to 10; over desir.n press are (11C%
is t.: ken for the neutron flux scrats, however*
x 1250 - 1175 psig), and the USASI Code pernit.s The indirect flux scram and safety valve pg pressure transients up to 20Z over tha dest.n actuation provide adequate margin below the g
pressure (120: X 1I 75 - 1410 psi:;). Re Safety peak allowable vessel pressure of 1375 psig.
I.1=it pressure of 1375 psig is referenced to the
,I, lovest clevatica of the primary coolant syutem.
l g
b g
Evaluation.?.ctbodology used to assure that this safety limit. pressure is not exceeded Reactor pressure is continuously monitored in the r,
f or any reload is docuinented in Reference 1.
control room duriry operation on a 1500 pui full 2
Oc de sign basis for the reactor pressure vessel
= des evident the substantial cargin of protection scale pressure recorder.-
aro ar.al-?t failure at the safety pressure limit of 1375 psit.
U.e v rsel has been designed for a general (4)
SAR, Section 11.2.2. -
ecch a-strers no greater than 26,700 psi at an I
also:
"Dresden 3 Second Reload License latr m i pressure of 1250 psig; this is a factor of Suba.i ttal," 9/14/7 3 1.~
-M the yield ntrength of 40.100 psi at 575'F.
At ti-a pressure Ifnit of 1315 psir, the general also:
"Dresden Station Special Report crt::7ne stress will only be 29,400 psi, still to. 29 Supplement 15..
20 n.
. e gy belca the yield strength.
i J
l a
4.1 SURVEILLANCE REQUIREtENT 3.1 LIMITING CONDITION FOR OPERATION i
(
4.1 _REACTO t PROTECTION SYSTEli it. /,CTOR p?OTECTION SYSTFll i
' 3.1 Ap ol i c.".b il it y :
l f.pplicabilitft
.Appites to the instrumentation and Applies to the surveillance of the instrumen-l tati n and associated devices which initiate
.t.sociated devices'which initiate a reactor scram.
~
reactor scram.
Objeetive:
CSJcctive:
To assure the operability of the To specify the type and frequency of surveillance to be applied to the protection reactor p rotection systca.
ins trurnen t at ion.
Specification:
5accification:
A.
Inc.trumentetion systems shall be A. The f.etpoints, ntnics: nurber of trip functionally tested and calibrated as D) sys te s, e.td iinit tra nu.Scr of ins t ru-indicated in Tables 4.1.1 and 4.1.2, M
r int char ncIs t!ict *: tot be cperabic
~
for each position of the reactor code respectively.
O reltch ch..11 be es r.1ven in Tebic y
3 1.1.
The recponse ti:nes of the B.
Daily during reactor power oporation, the coro power distribution shall be 1..d 1 t id ur.1 ft:. c tions shall not ex-I c1cci:c:1 for maximum fraction of ceed 0.10 seco.d.
limiting power density (MPLPD) and compared with the fraction of Rated Power FM (PRP) when operating above 25% rated g
si a;r ang opes at ion, the mas saua fi action of i t=lting thermal power.
[I
. a power des.s t ay enceeds the f r action of s ated power when atmve 2M s ated thermal puweg, eithers cpesatary a.
- as AP:?. screa aied sc.3 tJ sch settingc shall t,e s ed :ed to the values q s ven t,y the equations an 5.iecitacations 2.1.A.1 ar.J 2.1.B.
b.
Tr.e power d nat t it,ution shall be ch.nged such Endt the m.anawa Isaction of llanting power Jenesty e.o longer escoeJe the fraction of sated power.
Amendment No. 42 t
22
D G
O TA llLE 3.1.1 (cont)
No:cs:
1.
There shall be two operable et tripped trip systems for eact; function.
Pcsm.ssible to b}p.rss, witu control rod block, for scactor protection system reset in refuct and shutdown positions of the reactor mode switch 3.
Pernmsible to bypass when reactos picssure is< c00 peig.
4.
rcamissib!c so bypas. when first stJge turbine pressurc is less than that which corresponds to 457, rated steam flow.
5.
last's are bypassco. hen APRM's are onsc.ile and the scactor mode switch is in the tuti position.
The deagn pcts:uts closure of any one valve withoi.t a scram being Initiated.
7.
hten the reactor is sabcritical ar.d the reactor water temperature is less th.n 2120f'. only the following trip funcsions need to be operable:
a.
.Yode $ witch in Shutdown b.
?.fis:ual Scsam c.
llegS flus IRM d.
Scrara Dischange Volume liigh Level 5.
Not acquircd to be epc:abic when primary containment integrity is not acqu;ted.
L Not acquired while pesfoaming low power physics tests at Samospheric pressure during or aftes refueling at power levels not to exceed 5 MW(s).
' 10
- .
- ay be bypassed when nccessa,r) during purging for containment ine: ting or deinciting.
11 Not acq:.ited to be opciabic when the reactor pscssusc vciset head is not bolted to the sessel.
17 The APiM do.nscale trep functiori is automatically bypassed when the reactor mode switch is in the refuct and startup/ hot standby positions.
13.
The AP.4M downscale a:ip func::on is au;omaticall) b} pas:cd when the IRM instrumentat:o1is operabic and not high.
If the first column c..sunut tw arst for one of the trip sptems. that trip system shall be tripped.
If the f::s: column cannot bc.n.:t for both trip 6).nems, the approp:6 ic accions listed bc!ow shall be taken:
A.
Ina: ate.inication of operas:e :ods.ind con:picte in. cation of all opesabic sods within fous hours.
S.
- 4couce power Icsci to I:iM r cic and place er. ode switch in it.c it.ise ep/liot St ndby position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
g C.
Kedace tu:bsix le..d r.d.. e naam ste.in:Isn. i.olatiots *.ab es C..in i hc.are.
Q An Ah4M will bs cens:A red inopes.sble si thcsc are le.. tia n i' l.19M inputs per icscl or t$ cec arc icss tl.an W',of the normal corrpler icnt of layM's :o Jn Al5M.
.I l in:c:a oa the v. iter level ta.t:umentation is 7 504" above vessel zero (see 11ases 3.2).
Tr:ps i:pon actuation of the tas: cleiute so!cno7w hich trips the turbine control salves.
M GEED WeJ M
24 b
Amendment No.
42 O
i e
J
O O
O.
?..
DPR-25 The control red sirive scram system is designed so stop valve closure scram and causes a scram that ali of the water which is discharged from the before the stop valves are closed and thus the re-reactor by a scram can be accommodated in the sutting transient is less severe. Scram occurs at discharge piping. A part of this pipmg is an in-strument volume so-tube in the piping; which accom-23" lig vacuum, stop valse closure occurs at mudates in excess of 50 gallons of water and is the 20" lig vacuum and bypass closure at 7"lig sacuum.
!w psint in the piping. No credit was tal.en for
- h:- solume in the design of the discharge piping as etreerns the amount of water which must be accom-liigh radiation levels in the main steamline tunnel modated during a scram. During normal operation above that due to the normal nitrogen and oxygen L5e thscharge volume is empty; however, should it radioactivity is an indication of leaking Nel. A fill with water. the water discharged to the piping scram is iminated wheneser such radsation level fr om the reactor could not be accomm. dated which exceedsthreetimes normal background. The pur-would result in slow scram times or partial or no pose of this scram as to ' reduce the source of such control rod insertion. To preclude this occurrence, radiation to the extent necessary to prevent cxces-level switches have been provided in tic instrument sive turbine contamination. Discharge of excessive volu.ne which alarm and scram the rea, stor when amounts of radioactivity to the site environs is pre-sented by the air ejector off-gas monitors which the voiume of water reaches 50 gallons. As indi-cated above, there is sufficient volume in the pipin@
c ause an isolation of the mam condenser off-gas to accommodate the scram without impairment of line provided the limit specifs'ed in Specifica-the scram times or amount of insertion of the control tion 3. ti is exceeded, rads. His function shuts the reactor down while sufficient volume remains to accommodate the dis-The main steamline isolation valve closure scram.
is set to scram when the is tattort valves are 109 charged water and precludes the situation in which.
closed from full open. His scram anticipates the a scram would be rc< paired but not be able to per-
'Q Iarm its function adequately, pressure and flux transient, which would occur when the valves close. Ily scramming at this set-t..ss of condenser vacuum occurs when the con-g ting the resultant transient is insignificant.
h cr.uer can no Ions;er handle the heat input. loss g
<,1 condenser vacuum initiates a closure of the tur-bme stop valves and turtsine bypass valves which A reactor mode switch is provided which actuates or bypasses the various. scram functions a;'propriate eliminates the heat soput ta the condenser. Clos are h
of the tmhine stop amt bypass valves causes a pres-tion 7. 7.1. 2 SAll to the particular plant operating status. Ilef. Sec-sure transient, neutron flux rise, and an increase g- : ::::s in surface heat flux. To prevent the clad safety
%e m:.nual scram function is active in all modes.
g limit f rom being exceedes! if this occurs, a reactor thus providmg for a manual means of rapidly insert-8 cram occurs on turbine stop valve closure. 'the
,_-----s t ar!-%: stop valve closure scram function alore is ing control rods during all modes of reactor operation.
ad quate to prevent the elmi 8afety limit frem being M
- c. reeded in the event of a turisene trip transient
%e litit system provides protection against exces-f with hjpass closure.
The corden.ser to v vacuum scram is a lect ~up to the sive g=mer levels and short reactor periods in the stan t-up anti in actmediate power r:seges. Ilef.
29 Amendment No. 42 1
J ew
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l TADLE :!.2.'1 INSTRUMENTATION TilAT INITIATES PRIMARY CONTAMIMENT ISOLATION FUNCTIONS
{
Minimum No. of Operable Inst.
Chanacts per Trip System (1)
Instruments Trip Level Setting Action (3) 2 Reactor Iow Water
>144=above top of active fuel
- A 2
Reactor Low I,ow Water
,1'84"tbove top of aritve fuel' A
2 Illgh drywell pressure 52 psig rated (4), (5)
A 2 (2) liigh Flow Main Steam line
- 120'( of rated steam flow D
2 of 4 in each Iligh Temperature Main Steam Line
~
y of 4 sets Tunnel (200*F 11 C
2 liigh Radiation Main Steam Line s 3 times norrmi rated power back-Q Tunnel (6)
- round D
~
M 2
Low Pressure Main Steamline
'e850 psig B
liigh Flow Isolation Condenser Line D
1 Steamline Side 5 20 psi diff. on steamline side C
D 1
Condensate Return Side E 32 water diff. on condensate
[
{gp)
- eturn side C
w g
g 2
Iligh Flow IIPCI Steam I,ine 2150" water D
k g
4 Ifigh Temperature llPCI Steam 1,Inc Area 5200*F D
g O'
bs r\\)
Notes:
l.
Y.henever peinusy containenent entegelty as seguired, there stials be two ope.etale or tsipped Leip systerns for each function, except foe low piessuse main stearnline which only need be available in the ItUN position.
2.
Pcr each steamline.
N 3.
,Sxtiois: If the first column canr:nt he met for onc of the trip systeins, th.it trip systeni shall be tripped.
38 (see Bases 3.hro for all al>ove veer.el z of active fu is defined as 360"
' Tod>e r levcin usede 1in the I,0CA analyses
).
w.
Ancadsent flo.
O O
o n-Tall 1.E 3.2.2 INSTitlJMENTATION TIIAT INITIATES Oil CONTHOIS Tile COHE AND CONTAINMENT COOLING SYSTEMS Min. No. of Operable Inst. Channels lair Trip System (1)
Trip Fu'nction Trip level Setting Hemarks
=
84"( *0,[) above top of
- 1. In conjunction with low reactor 2
Heactor low low Water Icvel
,cggg c g,3cg a pressure initiates core spray arul LPCI.
~. In conjunction with high dry-well l
pressure 12'O sec. time delay,
~
and low pressure care cooling interlock initiates auto blowdown.
- 3. Initiates IIPCI and SUGTS.
- 4. Initiates starting of dicscl generators.
- 1. Initiates core spray, LPCI, 2
Iligh Drywell Pressure 5 2 psig IIPCI, and SilG'13.
(2), (3)
- 2. In conjunction with low low water level,120 sec. time delay, and low pressure core cooling inter-lock initiates auto blowdown.
- 3. Initiates starting of diesel generators.
1 Heactor Iow Pressure 300 psigsps350 pelg
- 1. Permissive for opening core spray aiul LPCI admission valves.
- 2. In conjunction with low low reactor water levelini-aics core spray and LPCI.
k Containraait Spray Prevents inadvertent operation R
Interlock of containment spray during g
1(4) 2/3 Core licight 22/3 core height accident conditions.
- s 2(4)
Containment liigh 0.G psigs ptl.5 pelg Pressure 1
Timer Auto Blowdown 5120 seconds In conjunction with low-low o
reactor water level, high A
, dry-well pressure, and low pressure core cooling inter-l l.wk_ initiates auto blowdown.
for all 40 al(>ove bessel z3ro fuel.is hefined as 360"e LOCA analyses l
- Top of activa see ases 3. ).
wal.er levels used in t s
Amendment No.
INSTRalSNTATION TImT INITIATES ROD BLOCK
~~l".inirs:. No. O f Cp rnbl' Inst.-
Channels Per Trio Level Settin'q Inst rur..cnt Trio Systen(l)_
AP?.M upscale (flow bias) (7) 4 65Wg43)'
@g (a
1 a
l APM upscale (refuel end Startup/Ilot
$12/125 full scale
~
Standby code) 2 A?rJ1 downscale (7) 2.3/125 full scal ~e
- Red block monitor uptcale (flow bias) (7) {65w+i2[
(2) 1 1
~
Rod block monitor downscale (7)
> 5/125 1011 scale N
3 IPJ1 downscale (3)
,15/125 full scale g
1108/125 full. scale Q
IM upsc'alc-3 M,..
Ird detectdic noti fully inserted in 3
the core 55? 2 )
su detector not in ctartup position (4) 2(5) c ::s d.105 ccunts/sec M
2 (5_) (6)
SPlt upscale P
Anendment No. 42 42 2
3
O O
O.
TA3L': 3.2.3 (cont)
Note:;-
1.
For the Startup/ Hot Standby and Run positions of the Ecactor Mode Selector Switch, there-:shall bc two cperabic or tripped trip systens for cach function, c:: cept the SPJ1 rod bloci:s, IPJi upscale, IRIi downocals ar.d IPJ4 detector not fully inserted in the core need not be operable in the " nun position and APiUI downscale, AFR.4 g
~ Standby monte.
and RDtl dotinscale nec' not be operabic in the Startup/Ilot 1
upsenlo (flow bias),
d
' Die ItD11e upner.le need not be operable at lens than 30% rated thermal power.
One chnonel meiy be bypassed nbove 30% rated thermal Ivver provided that.
n limiting cont.m1 md pattern doces not exist. For syoters with more than I
one channel per trip system.
If the first column cannot oc men ror noen trip sy steras, une systems shall be tripped.
2.
U nercent nf drive rios required t.o produce a rated core flow or g-5 9b a90/m.
IPJ4 downscale may be bypassed when it is on its lowest rancie.
3.
4.
This function may be bypassed whe'n tha count rate is 2100 cps.
5.
One of the four SPJ4 inputs may be bypassed.
This SIO1 function may be bypassed in 'che higher IRM ranges when the IRM upscale red M
6.
D block is operable.
M Not recuired while performing low powar physics tests at atmospheric pressure during 7.
or af ter refueling at power levels n6t to exceed 5 m!(t}.
M
=
@B
====>
2ssa f,k Amendment No.
42 t
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The instrumen-Temperature monitoring Instrumentation is tation also covera the full ra T) e or spectrum of provided in the main :,tcamitne tsumel to'dctcct breal:s and meets the above criteria.
leaks in this area. Taips are provided on thin in-i strumentation and when exceeded cause closure of Group 1 Isolation valves. Its setting of 200*F is 11.c high drywell pressure Instrumer.tation is a back.
Iow enough to detect leaks of the ordet of 5 to 10 up to the water level instrutnentation and in achiilion gym; thus, it is capable ol covering the entire to initiating ECCS It causes isolation of Group 2 Isc-spectrmn of breaks. For large breaks, it is a lation valves. For the breaks discussed aboec, this back-up to high steam flow instrumentation dis-Ins rumentation will initiate ECCS operation at about cussed above, and for small breaks with the result-the sarne time as the low law water level lastrumen-Ont small release of radioactivity, gdves isolation tattoa; thus the results given above are applicable before the guidelines of 10 CFil 100 are exceeded.
here - ho.
Group 21 solation valves include the dryaell vent, purge, and sump ise.iahen valves.
Illgh radiation momfors in the main steamline liigh drywell pressure activates only these valves tunnel have been provided to detect gross fuel failure.
because high drywell pressure coubl occur as the This instrumentation causes closure of Group 1 rc:. alt of non-safety related cau:,cs such as not valves, the only valves re<luired to close for this purging the drywell air during :.tartup. Total sys- -
J accident.'With the established setting of 3 times tem isolation is not denirable for these conditions normal background, and main steamline isolation an:t <mly the valves in Group 2 are a ertuired to valve closure, fission preduct release is limited so close. The low low water level instrumentation that 10 CFit 100 guidelices are not exceeded for this initiates protection for the full spectrum of loss of accide nt. Itef. Section 14.2.1. 7 sal!. The per-coolant accidents and causes a trip of Group I pri-fortuance of the process radiation monitoring system mary system isolation val.es.
relative to detecting fuel lenhage shall be evaluated during the first five years of operation. The conclu-Venturis are provided in the main steamlines as a sicus of this evaluation will be reported to the means of measuring steam flow and also limiting Atomic Energy Commission.
No)- tl:o loss et mass inventory from th: vessel during
.a :.teamline break accident. In addition to moni-Pressure instrumentation is provided which trips toring :,tcam flow, instrumentation is provided when main steamline pressure drops below 850 psig.
b wiuch can.es a trip of Group 1 isolation valves.
A trip of this lustriunentation results in closure of M3 The primary function of the in:,trumentation is to Group 1 isolation valves. In the "Itefuel** and detect a break in the snain steamline, thus only "Startup/llot Standby" mode this trip function is by-Q Grcep i valves are closed. For the worst case passed. This function is provided primarily to pro-.
e protection against a pressure regulator vid acch'ent. main steamline break outside the drywell, ilu*. trip setting of 120T of rated steam flmv in con-malfunction which wouhl cause the control aml/or Q
b:mh.n v.ith the !!cw limiters anil main steamline bypass valves to open. With the trip set at S50 psig vatte clo:.ure. limit the mass inventory loss such inventory loss is limited r,o that fuel is not uncovered
~
pg th.it locl is m,1 uncoverett, hiel tem 3.cratures re-and peak clad temperatures at e much less than m.ein less than 150n F and release..f radii.activdy 1500'I'; thus, there are im fit.sion products available to the environs is well below to Crit 100 guidelines.
for re:...;. r.tl.c. than those in the s cartor ats r.
h it A 8. ciii.ns 14. 2. :i. ti.uut i 1. 2. :t... sait.
Ite!. s cii.,n i t.2. : : alt.
17 Amendment No. 42 t. _..........
-. - ~
%.?
nQs' Two sensors on the is.ulation condenser supply.and siay be reduced by one for a slsort period of time to,
return lines are provided to detect the failure of allow for seatntenance testinst. nf calthratissn.
isol tlon condensur line auJ actuate *.oiation action.
T:ais time period is only -3% of the operating time The sensor 6 on the supply and return sides are 2n a month and does not significantly increase the arranged in a 1 out of 2 logic and, to meet the risk of preventing an inadvertent control rod with-single failure criteria, all sensors aid instrumen-drawal.
tation are required to be operable. The trip settings of 20 ps.lg and 32" of water and valve closure time The APRH rod block function is flow biased and are such as to prevent uncovering the core or ex-prevents a signific, ant reduction in MCPR expecially cceding site lisetts. The sensors will actuate due.
during operation at reduced flow. The ARPH provides to high flow in either direction.
l Rross come pruteccion; i.e.. limit. the gross withdrawal Tine IIPCI high flow and tes perature instrumentation are provided to detect a break in the IIPCI piping.
Tripping of this instrumentation results in actuation of IIPCI isolation valves.
i.e., Group 4 valves.
In the refuel and startup/ hot standby modes, Tripping logic for this function is the same as that the APRM rod block function is set at 12% of for the isolation condenser and thus all sensors I
rated power.
This control :'od block provides are required to be operable to steet the single fail-the same type of protection in the Refuel and Startup/
ure criteria. The trip settings of 200*F and 300Z llot Standby snode as the APRH flow biased rod block of design flow and valve closure time are such that does in t he run stode; i.e.,
core uncovery to preventel and fission product Q
release is vf*
- 1inits, prevents control rod withdrawal before a scram is reached.
"R.e ins t -.....;..ma u s ch initiates ECCS action is bO) arranged in a dual bus systesa. As for other vital The RBM rod block function provides local protection instruaientation arranged in this fashion the Speci-of the core, i.e.,
the prevention of transition fication preserves the effectiveness of the system boiling in a local region of the core, for a single even during periods when maintenance or testing rod withdrawat error frors a limiting control rod h,
is being performed.
Pattern. The trip point is flow biased. The worse case single control rod withdrawal error The control rod block functions are provided to.
l
, is analyzed
' h prevent toe e x h resoad to assure that w 6 th t e..
speci f ic t r ip se t t ings, exccasive control rod withdrawal so that so.6 with.sa...t as stucked t,. rose the r.cra reaches the acra t 1 HCPit oces not go below the iiCPlt fuel clad-
' # "* """Y **'**Y l'"
C:C' b aing integrity safety limit.
The trip s.now ta p.ncent p
- r. the worst ca.. withdrawat or a sangt.
cor contrat saa witnous roa bioch action want not violate the acen p logic.
this function t o, 1 out o ri; e.g.,
t.et ct.aeing ina.gsity s.t.tr statt. m s, the man sea block any trip on one of the six nPitM's, 8 IRM's, or
- f. action 1. not :.quis.4 t,. tow ths. po.: levet.
m b
4 Sitai's will result in a rod block.
The
(
.ni n isaus:-. instrument channel requirements e
ssure nu2Ticient instrumentation to assure the single failure criteria are net.
The sc i n i.nura instrument channel requirements for the itiff-l 48 Amer.dmerit No. 42 O
-N t
ne I!i'l ro.1 block function provides local settings given in the specif.tcation are adcua/te to-De ocollag assure the above criteria are met. Ret. Sec'. ion
rell c greca core protection.
is occh that trip settin:; in Icss G. 2. G. 3 SAR.
'llte specification preserves Qc 4 r range-/:nt a fr.ctor of 10 chove the indicated Icvel.
efL.ctiveness of the sy tem d* ring periods of main-the 2 t.n:170!o of :tc vorct ccee accident results csung, or cnhoratian, and airo mini-
~
C in red bicc's cction before l'CPR approaches mizes tne risk of inao,vertent o;:cration; i.e.
ontv
~
l t.he f.1CPit fucl cladding or.c instrument channel out of service.
integrity 3afety limit.
A deenscale indication on an APRM or IR!." is an 1: ;tcation the instru:nent has failed or the instru-Two air ejector off-gas monitors are provided and ment'is not sentitive enough. la c! hor case the when their tr.o point is reached, cause an isol: tion instrum.:nt will net respr.d to chan ;cs in control Of the air c.icciar off-gas line. Iso!ation is initiated red motion an i thus control roti motion is prevented.
when both instruments.rcach their hi;h trip point Ti:e downscale trips are ret at 5/125 of full scaic.
or one hes an upscale trip and the other a doc.n-sc.'!c trip. %cre is a fif:cen minute delay before '
W U-E'S IU I*U O valve is closed.
- The rod block which occurs when the IRM Bis oclay a.s accounicd for by t.ne 30-minute actectors are not fully inserted in the holdup time of the off-gas before it is released to core for the refuel and startup/ hot the stack.
standby position of the mode switch has ocen provided to assure that these Both instruments are required for trio but the detectors are in the core during reactor 3nstruments are so designed that any instru:nent utartup.
This, t.nerefore, assures that failure gives a daen.ccate trip. He trio settings of the instruments arc set so that the i$stantanc-instruments arc in' proper position ous stack reicase rate limit given in Specification these to provide protection during reactor
- 1. R is rot execeded.
The IIDl's primarily provide startup.
Y' protection ngninst local rer.ctivity cour radiation :nonitors are provided v.hich g
cffccts in the source and interr:cdiate initiata Isolation of the reactor building and operation of the standby gas treatment system.
neutron rangc.
The m"uitors are located in the reactor bul! din?
~
M ventilanon duct and on the refuelir; I!oor. The ror cIfective c:ncrgency core cooling for smallpipe trip lo;ic is a 1 out of 2 for cach set and each S
b cti:s, the liPCI system must function since reac-set can Initiate a trip independent of the'other tor pressure decs not decrease rapidly enough to set., Any upscale trip will cause the desired EE!$
- llow cither core spray or LPCI to operate in time, action.
Trip settirgs of 11 mr/hr for the C":=2 The cutematic pressure relief function is provided monitors in the ventilation dact are based upon
$8 as a back-up to the liPCI in the event the ilPCI does initiating normal ventilation isolation and sten {by' The arrangemen' of the tripping con-1;as treatmeat system operation to limit :he dose not oper ste.
M tacts is such as to provide this function when rec-Ti:c trip essary nnd minimin'e spurious operation.
fi 49
. Amendment No. 42 1
I
~
O O
G 3.3 1.IMITIlid Colli,ITIO:I FOR OPERATIO!I 4.3 SURVEILL,ANCE REQUIREMENTS 3.
(a) Control rod withdrawal sequences shall be 3.
(a) To consider the rod worth minimizer established so that anximum reactivity that operable, the following steps acust be could be added by dropout of any ' increment performed:
j of any one control blade would be stah that j
the rod drop accident design limit (1) The control rod withdrawal sequence of 280 cal /gm is not exceeded.
for the rod worth minimizer computer shall be verified as correct.
(b) Uhenever the reactor is in the utartup or (ii) ne rod worth minimizer computer -
inode below 20% rcted therini power, on-line diagnosite test chall be run the Hod Worth Hlulmizer shall be operable.
successfully completed.
A second operator or qualified technical person may be used as a substitute for an (iii) Proper annunciation of the select inoperable Rod Uorth Minimizer which falls error of at least one out-of-sequence after withdrawal of at least 12 control rods control rod in cach fully inserted to the fully withdrawn position. The Rod group chall be verified.
Uorth liinimizer may also be bypassed for low power physics testing to demonstrate the (iv) '1he rod block function of the rod shutdown rurgin requirements of specifications worth minimizer shall be verifica 3.3. A.1 if a nucicar engineer is present and by attempting to withdraw an out-veriften the step-by-step rod movecients of of-sequence control rod beyond the the tent procedure.
block point.
8 (b) If the rod worth minizer is inoperable while the reactor is in the startup or g@
I run mode below 20'a rated thercal power and a second independent operator or engineer is being used, he shall verify that all rod positions are correct prior to commencing withdrawal of each rod group.
57 mE i
Amendment No.
42 i
t
f LIMITING CONDITIONS IDR OPERATION 4.3 SURVEILLANCE REQUIREMENTS 4.
Control red shall not be withdrawn for 4.
Prior to control rod withdrawal for stattup' startup or refueling unless at least two or during refueling verify that at least two source range channels have on observed source range channels have been observed count rate equal to or greater than three couat rate of at least three counts per counts per second.
second.
5.
During operating uith limiting control rod 5.
When a. limiting control rod pattern exists, patterns, as determined by the nuclear an instrument functional test of the RAM engineer, either:
shall be perforacd prior to withdrawal of the designated rod (s) and daily thereafter.
a.
Both RD:t channels shall be operable; or b.
Control rod withdrawal shall be blocked; or c.
The operatin6 powcr level chall bo 11aited co tho the F.CFil will unnin above the MCPR fuel cladding integrity safety limit assuming a single error that results in complete withdrawal of any single operable control rod.
.N L3 C3 2e)
GEES
- ass E83 M
TED IJ
' Amendment No.
42 57A i
)
~
O
.o o
o, i
sraall amount.of rod withdrawal, which is less inJiotive of a generic cont.ol 'rnd drive I
than a non, il single withdrawal inercsent, will probics and th: se.ictor will be shutdown.
l r.at ccatritu:= to any da::ge to the priv.ary A so if damage within the control rca de tyc coolant syste:n. The design basis is given in rechanista and in p. articular, cracks an drave Ecction 6.6.1 of the SA:1. and the design evalua.
lat ernal housings, estnot be ruled out, then a tion is given in Section 6.6.3.
This support gencric prchica affecting a nu:bcr of dr:ves iz not re';uired if the reactor coolant systes cannot be ruled out.
Circumferential cracas is at atmosphcric pressure since there would resulting frca stress assistod intcrt:r' anular then be no drivir.g force to rspidly eject a corresica have cccurred in the collet hcusing drive housing. Additionally, the support is of drives at several INits. 1his type of -
not required if all control rods are fully en cracking cculd occur in a number of drives inserted and if an adequate shutdcwn rargin cr.d if t;.c cracks prcpagated until severance with cne control rod withdrawn has been demon-of the collet hcusing occurred, scram eculd strated since the reactor would renain suberitical'
~
be prevented in the affected rods. Limiting even in the event of complete ejection of the the perice of operation with a potentially strongest control rod.
severed collet housing and requiring increased sure:illar.cc after detecting or.o stuck 3.
Control rod withdrawal and insertion scquences-are red will :ssure that the reacter will not established to assure that the naximum insequence La cperated with a 1srgq nue.ber of rods with individual control rod or control rod stg:ents f:.ile.1 collet ht.u:ings..
which..rc withdtc-n could r.ot bc isrth cncunh to the rod drop accident design limit of cause d.
Centrcl ro.! n'ithdr:ws!
I 280 cal / gam to be exceeded 1.
Cor. tral red <Ircpout accidents ns discussed if they were.to d:ap ca: of the core in the r:.anner Jefined for the R:3 Drcp Accident.N in Reference 6 c.:r. Icad to significant coro -
11;csc sequences arc devcicped prior to laitial c u-.ge.
If coupling integrity is c:.intcined.
cperatica of the unit following a.:.y rc. fueling cutage ti c possibility cf a red drepeut accident is and the ra.M rc :ent that an optrater ' allo.s the<c cl ir.ir.2 t e d. Tne overtr vel positica f;:turc p vides : pczitive ch::k as caly un:cupies or a second qualified station employee.
q sequences is backed up by the.cperation of the I;*21.
Oo af t.ees :y :::2ch this position. Neutron These sequences are developed to limit b
ir.::::: a.:ation r:r; ensa to rod r.: Cent reactivity worths of control rods and y;cv. des a verifi:cticn th.t the rod 1. fol-i M
lowing its drive. Absence of such response together with the integral rod veloci:y lit.itens and the c:tton of the co a rol to drice cove: cat would provide cause for rad dr ive systen, liritt retential reactivity su:,pecting a rod to be uncoupled and stuck.
insc: tion 31.ch that the results of a control rod Restrict 1rg reccupling verifications to pover drop accident teill nor exec:d a nazir:::: fuel carrSy, Icvels atove 20: provides assurance that a content of 2S3 cal /gc..
The peak feel enthstpy cf D
rad diop during.: :ccoupling verification h
2E0 c.sl/gr. is below the er.crgy content at which wou d ro: result in a rod drop accident.
C:::: 3 rcpid fuel dispersal cnd p ;csry systera dassgc have been feu. d to occur based on experi= cats 1 data as 2.
The con:rel rrd heutin?. stT.;.::t rcruicts is discussed in Reference 1.
t;u sutward n:,t a nt ct a ccAttel tce to led th7n 3 ink s in the cL;tren:17 rm:tc 1hc en:1ysis Of the centrol rcd drop accident :::s eve r.: cf a Lcasing tailure. lac :Cr..nt'ei l
criginally presented in 5::ticr.s 7.D.3 14.2.1.2 reacti,ity M.ich could ha ad2:2 L/ t.is j
sud 14.2.1.4 of the Safety Analysis P.cPer:. I:;.rcve-n;.nts is: sn:lytical capability have !!:-cd s n:sc 62
. Amendment No. 42 refiae.d 't.nslys'is of the '::nt s1 red 2 ra **ci '
es id Bases (cont'd)
The following parameters and worst-case bounding assumptions have been utill:ed ness techniquegre described 11) a in the reload snalysis to determine com-Pcal(I In dd on, banked pliance with the 280 cal /gm peak fuel i
t ments.
position withdrawal sequence" described enthalpy.
Method and basis for the rod drop accident analyses are documented in in naference (4) has be'esi developed.
Reference 6.
Each core reload will be to further reduce increatental rod analyzed to show conformance to the worths.
limiting parameters.
By using the analytics1 models
,, ( An inter-assembly. local power peaking described in those reports coupled IN factor with conservative or worst-cace input paraccters, it har been determined b.
The delayed neutron fraction chosen k
')
l that for paver levels less than 2f/h for.the bounding reactivity curve.
of rated power, the specified limit (typically 1.3% 4K)
Q on insequence control rod or control c.
A beginning-of-life Doppler reactivity feed-g rod r.ctment worths will limit the peak back.
fuel enthalpy to less than 280 cal /gm.
O' l Al,ove 20% power even single operator d,
scram tirras slower than the technical errors cannot result in out-of-r.cquence Specification rod scram insertion rate a M
control rod worths which are sufficient
-(Section 3.3. C.1) l to acach a peak fuci enthalpy of 230 c.
The maximum possibic rod drop velocity cal /g a should a po:.tulated control rod (3.11 ft./sec.)
dsop accident occur.
7M f.
The design accident and scram reactivity shape function.
gb g.
H e mininum coderator temperature to reach
{
(1)Paone, C.J., Stirn, R.C. and Wooley,
'*" ' I*
2 J. A., " Rod Drop Accident Analysic for 3
Large Eo111ng k*ater P.cactorG (4) C.J. Paone, " Banked Position Withdrawal 1;Eco-10527, liareh 1972.
Sequence" Licensing Topical Report (2)Stirn, P..C.,
Paone, C.J., and Young, NEco-2123, January 1977 g
n.!!., Rod Drop Accident Analysis for 1.at ge Intn's", Supplcment 1 - ELDO-To include the paer spike effect caused by gaps 10527, July 1972 between fuel pellets.
- Stirn, R.C., Paone, C.J., and Ihun (6) Gesic:~1c Reload Fuel Application e
J.l!.. "tod Drop Accidcat Anclyeis for IEDE-2 4 0ll-P-A
- Large In:R's Addendun 1:o. 2, Exposed Cores", se.,plecent 2-NEDO 10527,
- Apo oved revision number at time 62a Jenuary I W.
reIchd fuel analyses are perfor.T.cd.
J
parte rt (con'd)
It in recognized that these bounda are
- fhe Rod k' orth litntmizer providen aut omatic connervative with reupect to expected nuperviolon to annure that out of nequence operating conditions. If any one of the control rods will not be withdrawn or inserted; above conditions is not satisfied, a more i.e., it limits operator deviations from plan:ied detailed calculation will be done to show withdrawal sequences.
Ref. Section 7.9 SAR.
j co:npliance with the 280 cal /gm design limit.
It serves as a backup to procedural control of control rod worth.
In the event that the med In most cases the worth of in sequence Worth Mini:ntzer is out of service, when requirs.A rods or rod segments a licensed operator or other qualified technical employee can manually fulilll the control rod pattern conformance functions of the in conjunction with the actual Rod Worth Mininizer.
In this case, procedural-o values of the other important accident analysis control is exercised by verifying all control parameters described above would most likely rod positions af ter the withdrawal of each I
result in a peak f uel enthalpy substantially 1cos group, prior to proceeding to the next than the 280 cal /gm design limit.
group. Allowing substitution of a second independent operator or engineer in case.
of RWti inoperability recognizes the capability to adequately nonitor proper rod sequencing in an alternate manner without unduly resCrict
- I ing plant operations. Above 201, power, there is Should a control drop accident result in a peak no requirement that the RW:1 be operable sipce fuel energy content of 280 cal /gm less than the control rod drop accident with out-of-660 (7 x 7) fuct rods are conservatively ocquence rods will result in a peak fuel cutinated to perforate. This would result in an energy content of less than 280 cal /gs. To offsite done well below the guideline value of assure high Rtal availability, the M.7: is D
For 8 x 8 fuel, less than 850 rods requeled to be operating during a startup Q
are conservatively estimated to perforate with for the withdrawal of a significant number g
nearly the sai.ie consequences as for the 7 x 7 control rods for any startup after June 1, 1974.
fuel case because of the' rod power dff ferences.
4.
The Source Rango !!onitor ('JRM) systen performs no automatic safety system function; i.e.,
it M
has no scram function.
It does provide the r ----
=bb cm Amendment No.
42 62b J
7.
O>
' I. 3 '
C.
Scram Insertion Tic >cs
$ith a visual indication of neatron i
Icvel. His is needed for knowledgeabic and operator The control rod eyote'a.is-nalyzed to bring ths efficient reactor startup at low neutron Icvel.
react r suberitical at a rate fast enough to Ec consce;tences of reactivity accidents are Prevent fuel da:cige; i.e.,
to prevent the h*CFit
.~cactions of the initial neutron flux. The requircr:ent of at 1 cast 3 counts per second f rom becoming 1 css th.an. t.the MCPR f.uel cladding Integrity safety limt Analysis of should it occur assures that any transient, or above the initial value of 10-g the limiting power transient shows that benias at the negative reactivity rates resulting of rated pcuer used in the analyses of transients f rom the scrata with the aver. age response f rc:: cold conditions. One operabic SRM channel of 11 the drives as niven in the above
~
uonid be adequate to monitor the approach to specification, provide the requir'ed pro-criticality using hocanencous patterns of tection, and MCPR remains greater than the scattered control rod ulthdrawal. A miniwum MCPR fuel cladding integrity safety limit.
of L.co operable Si2M's are provided as an added conservat i:,:n.
So ne nod illock Monitor (RSM) is designed to auto-rutien11y prevent fuel da aage in the event of crreneocs rod withdrawal fro = locations of high
- cus tiensity during high power icvel operation.
bc I
Tuo che: ncis are provided and one cf these may/or bype ;::cd f ro a the console for caintenance and
(
3 esting. Tripping of one of the channels vill block g
err ;ncous rol withdra tal soca enough to prevent fuel d:=::nc. %is sys te:2 backs up the cperator who with-7;,e rd nicat= c ount of reactivity to be dr:.:s rods according; to a written sequence. The inser;ed duting a scrati is controlled by creci.fied restrictions with ene channel out of pe r.:itting. no =;,rc than 10:: of the operabic g
terrice conservatively assure that fuel damage rods t'o.have lcng scra 2 tims.
In the M
g will not occur due to rod withdraual errors uhen analytical t rastrent of t he t r. nsients, 370 g
this conditica exists. h.:,:nincats 17/18 and 19/20
=tilise.: cuds are all. ved between a neutrc,n g
present the results of an evaluation of a rod block' sensor reachin;; the scram point and the r*
r c.aitor f ailure. These amendments show tilat durinS start of otio -of the centrol rods. This y
reacter operation with certain limiting control is adequate and cdiservative War. co: pared eb red ratteren, the withdrawal of a designated single to the typicelly observed tien delay of (f-3 g
cc-tret rod ceuld re-it in one or nare fuel reds about 270 =illineconds. Approxi;.:tcIy 70
.: ? s.7: s t e :.s t:..in t ho xt'i k f.e : c;adziv)
. rte,..;y :..fety 1& oat.
Ou s t av,.:6e of hu;n C,..,. incConds a f t e r neut ron fli:x reaches the j
p.:: t e rn :, :t is, juan ::t tact testing of the RBM 1
- ,yst,.r: prior to withdrcual of such rods to assure its oper.:bility will assure that inproper with-drar.il does not occur. It is the responsibility of,t:4e ::ccicar Engineer to identify these liniting
- .
- tt erns a. f the designated reds either when the l
- at erns are initially established 'or as they g
de celop due to the occurrence of inoperable control l
?
O O
O
^
Components of the system are checked periodically as described nbove and make n' functional test of pages:
the entire cystem on a frequency of Icss than once A.
The design objective of tht standby liquid during each operating cycle unnecessary. A test control system is to provide the capabvy ar of one installed explosive charge is made at least brinning the reactor from full power colt l once during cach operating cycle to assure that the f
xenon-free shutdown assumtag that s.,
of LM charges have not deteriorated, the actuation circuit withdrawn control rods can be Inserted. To is functioning properly, the valve functions properly, meet this objective, the lhp:id coatrol r,yr. lcm and no flow blockagcu exist. The replacement charh.c is designed to inject a quantity of boron which will be selected from a, batch for which there has been I
produces a concentration of no Iess than a successful tout fitinn. Recor.mendations of Llic. :
i 600 ppm of lxsron vendor shall be folicwed in maintaining a five-yeay in the reactor core in less than 100 minutes, life of the explosive charges. A cont timal check,0f 600 ppm boron concentration in the reactor
. the firing circuit continuity is provided by pilot, core is required to bring the reactor from full lights in the control room.
power to a 3"cak or more s'ubcritica1 condition considerin9 The relief valves in the standby liquid control the hot to col 1 reactivity swing, :wnon stem protect the system piping and positive poisoning and an athhtional ma'rgin (2r,'4) for displacement pumps which are nominally possib!c imr.erfect mixing of the chemical destf;6cd.for 1r:00 psig protection from over-solution in the reactor water. A mminuun pressure. The pressure relief valves discharge quantity of 3171 gallons of solution having a back to the standby liquid'contiol solution tank.
13.17 sodium pentaborate concentration is required to meet this shutdown requirement.
D.
Only one of the two standby liquid control T
pumi,lr.,, cli cuits is needed for proper ope.a-(c-g
.The time requirement (100 minutes) for inscr-linn of the system. If one pumning circuit is tion of the boron solution was selected to over-found to be l'noperable, there is no immediate a
ride the rate of rcactivity inseition due to threat to shutihmn capability, and reattor oper-M tooldov.n of the reactor following the Acnon ation may continue while repairs are being poisen paak. For a required pumping rate of made. Assurance that the remaining system 30 gallons per minute, the maximum starage will perform its intended function and that the volume of the boron solution is est:iblitheil reliability of the system is good is obtained by as 1,059 gallons (IT,3 gallons are conia;ned demonstrating operation of the pump in the Iselow the ;u:mp suction and, therefore, cannot operable circuit at least once daily.
be inserted).
C.
- fhe solution saturation temperature of 137 N
Caron concentration, solution tem;)crature, and sodlum pentaborate, by weight, is 59'F. To telume are checked on a frequency to assure a guard against boron precipitation, the solution -
high reliability of cperation of the system inchuling that in the pump suction piping is shoubt it ever be requireil. Ihperience with kept at 1:ast l'."F above: the saturation teeper-pump operability indicates that monthly testinC eture by a tank heater and by heat tracing in is adequate to detect il failures have occurred.
the puni T. suction piping. The 10*F margin is Amendment No.
42 t
71
~
l l
DPR-25 l
4.5 SURVEILIANCE REQUIREMENT 3,5 LIMITIliG CONDITION FOR OPERATION i
I.
Average Planar Linonr IIcat Generation Averaqo Planar UtGR Rate (APuiGR)_
During steady state power oparation, the The APUIGR for each type of thol as a g
Average Planar Linear Heat & nci4 tion Rate (APLilGB) function of average pinnar exposure shall y
of all the rods in any fuel assembly, as at be determined daily during reactor opera-a function of average pinnar exposure, tion at,1 25% rated thermal power.
any axial location, shall not exceed the mnximum average planar UlGR shWn in l
Firparo 3.5-1.
If at any time during oporntion it is determinod by normn1 sur-voillanco that the limiting value for APUlGR is being exceedod, action shall be initintr;d within 15 minutos to restore operation to within the proscribed limits.
If the APutGR in not returned to within the proscribed limite within two (2) hourn, tho reactor chall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillanco and corresponding action nhall continue until reactor opera--
Linn in uithin tha prescribed limita.
T h- )
d 3 e)
W5EU E
T DE579 T
~
M 81s Amendment No. 42 ll 3
h DPR-25 p,_..
3.5 T.I!!ITING COIIDITION IVR OPERATIOli 4.5 StJRVEILLMICE REQt1IREMd!IT
^
J.
T.ocal LitGR J.
Linear 11ent Generation Rato (L11GR)
During steady state power operation, the The LIIGn as a function of core height nhall 1incar heat generation rate (LIIGR) of be checked daily during reactor operation i
uny rod in any fuel assembly, at any axial at >-- 257. rated thermal power.
locatiog shall not' exceed the maxtmum nilowable LitGR as calculated by the following equation:
1 L11GR LilGR fsp I
^
[TF L '
t I
d 1-max \\
max u
LIIGR Design LitGR = 17.5 W /ft. 7x7 d
n fuel 3
1.1. l kw/tt toa al1 8x8 l' ue l t y ner, I
\\
r.ax = naximum power spiking pennity =
jjAP/P i
L
)
0.036 for 7x7 fuel and 0.0 for Ox0 fuel
.LT *= Total core length = 12 ft.
Axial position abovo bottom of.
N L
=
coro LD If nt npy timo during operation, it in g~
dutoriained by normal survoillanco thnt the D
limiting vnluo for LitGR in being exconded, nction shall bo initiated within 15 minuton C
to rest, tore operation to within the preacribed limitu.
I f the 1.IIGR in not roturned to with-g i.s the proncribed limi.tn within two (2) hourn thn ro.tetor nhall bo brought to the Cold Shut-rg down condition within 36 hourn.
.Surveillanco and corronponding action ohnll continue un-y t.1L reactor operation in within the pre-01n-1 ner3 bed Ibnito.
y
=
Amendment No.
42
}
- =%
~~
DM-25
==
- n-g
- =:.. -
~m 14.5 N.-
Dresden Unit 3 -
's 14.0 i
\\
Fuel Tyee i
7D212 No GAD i
\\
- s. ',
u.s 13.5 s
\\
b 1
\\
w g
s er<-
w m-
="
13.0
-i
== %
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.=
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- i i..
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~
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- a. ac wz eo 12.5 5*
liii;:;i:!l:!!!:
<w r
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ou lii:i:::
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12.0
,._. :.: : :::--l:. -
.:.::.:.l.::.:
11.5 11.0 1
i 1
i l
l l
10.5 '
0 10.000 20.000 30.000 1
PLANAR AVCRACE EXPOSURE (.%'D/T) ncN: 3.5-1 MA:1.v. AVERACE PLANAR LINEAR fSheet 1 of 5 h HEAT GENERATION RATE (P.APLHGR)~
vs. PtAnAR AVERAct,(IPOSURC Amendment No. 42
N I
U CD DPR-25 6-4 s.s
.T.
su'
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<m
. i ww
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........1
== N ac s 1)*0
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0 100C0 20C00 3C000.
4C000 Planar Aver Inosupe (G'D/SM "icure 3.~,-1 PAXrm AMAm MEAR LMN1 HER' N.6 (chact 5 of 5) 9 ATE mm.Jna) vs. FLWAP AV"W2 ".*W t
Amendment No. 42 l
O O
O
't. S I.inyi t:i ng cond I t:f ons for cdra_ tion nanes cnce the repair period is found to lie less than des etoped in this refer-A.
Core Sorjn',_jind T.PCI Mo,de _of _t help,r!R-1/2 the test Intercal. This assumes that the y,ystem - This specification assuren core spray and I.PCI subrystems conrtitute a that adequate emergency cooling 1 out ol 3 system, hov.crcr. the combined ef-capability is available.
feet of the two systems to limit esec8sive clad
, tesoperatures must also lic considered. The ICt interval specified in F gl liased on the loss of ecolant analyses i cification 8. i uns
'l m n...s.
.heref rc. an al m.:.nt tr;mr included in References g1) and (2) in accordance with 10CFRSO.46 and Appen-lod did mWh & Iw s wh ing single failures should he less than t...iavs dix K, core cooling systems provide and this specifica: oa is within this period. ~
sufficient cooling ?o the core to For snultiple failares a oh. r:cr inwrval is dissipate the energy associated with specified and tr> imprere th( assuranec : hat the loss of coolant accident, to limit the remaining systems u di fenetion a daile~
the calculated peak clad temperature that the information ;;iven in reference 3 pro-to less than 2200 F, to assure that vides a quantitative cwthod t.i esi: iue allow-core geometry remains intact, to 1imit ahic repair times, the lack of operating data to the core wide clad metal-water reaction support the analytical approach presents com-to less than 11 and to limit the cal--
plete acceptance of this tre: hod at this time.
culat.ed local metal-water react ion
- lherefore. the :imc8 stated la the specific to less than 17L judgment.
The allowable repair times are es-Should one core spray subsystem become in-tablished so that the. average risk rate operable. the remaining core sp ay arr.1 the for repair would be no greater than entire LPCI system are available should the the basic risk rate.
The method and.
concept are described in Reference (2) NEDO-20SG6, General Electric (3)-
tising the results Compan Analytical Model for Loss-of-Coo ant Analysis in Accordance with 10CFRSO Appendix K.
(1) "I.oss of coolant Accident Analyses itepo r t (3) APED " Guidelines for Determining for Di eselen IJn i ta. 2, 3 and Quad-Cities Safe Test Intervals and llepair i
lin i t :.
1,
? Nuclear power S t a t. ions,"
Times for Engineered Safeguards" u t:lx )
.! 4 1 % A, Itev is ion i, April 1979.
April 1969, I.M. Jacobs and P.W. Marriott.
02 Amendment No.
42 m.
su.a. -
~
^ Q-O O
'I.5 linttine contition for op3 ration Iinnen (cont'd)
I, 1.verre Pl..rar Id.n s
Thin specification annurta that the palc cin'iding tenperature follouing a pcotulated dor.ir.n baalo loco-of-coolant accidont uill not excccd tho 2200 F limit opacific1 in ICC?nSO Appendix K'cencicering tho postulated ofrccta of fuel pallot dcucification.
Tho 1.cak cladQng tonporaturo follouing a g
postulated loss-of-ccolant ccc3 dent is priscrily a functicn of tho avan.ca ID;il b
of all tha rcdc in a fuol ccccably et any N
c.xtal location ur.d in o.;1y dop:ndent ::ocond-nrily on the ro.1 to red p.nor dictribution uithin a feel om:cnbly. Sinco expaci.cd local variaticna Jn poner dit,trl'aution uithin n g
f ual acno.7.bly affect the calculated pak olnd tonporaturo by less than i20 7 rolat.1vo to C
the pack tenparaturo for n typical fuel design, M
ths limit on the nvorano plen'r IJR;R in cufficient to accure that calculatcd terap-craturco nro bolcu tho ICCFRSO, Aprondix K y
The linit.
rnix.ust nverano pintur IJ!G1:0 plotted in The r ximum avorago plenar IllCRJ choun in Fic. 3.5.1 at hicher c::posurec recult in a g
Figura 3.5.1 aro haced on calculatico; c ploy.
calculated peak clad temperaturo of 1 r3 g
g-Ing tho ;no.lcla Occerihcd in Hoforanco (1).
than 2200 F.
!!cucver the ruxinum aver p g
Pcuer op ration eith !!!CBa ot. or t. Ic:: theco planar I!!CRo aro cho:in on Fic. 3.5.1 as choun in Fig. 3.5.1 encoroa that the r'eak lii: tta becauso conforr.ac.co calcuintica, have r*
c1rddin. to:qoraturn follouing a postuinted nct been perforr.:d to ju-tify oparatica at 3csa-c,f-ccalant. accident vill not excced the 1110113 in excess of thooc choun.
220007 linit. Thonc values reprocent 11oita for op: ration to ensuto ccnforcanco uith J.
I.nen1 I1!GR iccFn50 crd Appendix K only if thcy :u o moro
.1 citing then other denign pirar.ctorn.
Thio opecific~ntion nsnures that tho 1
tcximun lincar heat genera
- lon rata in "1.onn of Coolant Accident. Analynes Report for any red 10 loan tinn tho dc31ga 1.incar
('(I)
Dresden Units 2, 3 and Quad-Cities Units 1, 2
!!uclear Power Stations," tJEDO-24144A, Hevision 1, MA j
1 h
Apr i 1, 1979.
)
s' ],
(.,4 s
/.
Condition for Operation lisses (Cont'd)-
The most limiting transients with roope,.
to
.5 1.ioitint MCPn are generally:
beat generation rato even if foel pellet a) nod withdrawal error densiftcation is postulated.
11.e power in b) ejection or turbine trip without
- 5'
- 5
- a e
r nc 2
c)
Loss of feedwater heater Itnearly increasing variation in asia,1-gaps between core 1ottom and top, and trans}innts em. l [ actors influence which of these assumes with 951 confidence, that no.
results in the largest reduction more than one fuel rod exceeds the design n critical power ratio such as the specific IUCR due to power spiking.
fuci loading, exposure, and fuel type.
Tho current cycles reload licennin9 analyses l
specifies the limit ing transient for a given exposure increment for each fuel type.
The v:n lue:s specifled as the Limitir.g Condition of Minimm Critical rower Ratio (HCPR)
Operation are connorve.tively chonen raost restrictive over tl.o entire cVc e gN" K.
(
each fuel type.
The steady state values for HCPR specified an this Specif tcation were selected to p ovide cargin to accostso.lat e t ransients and uncertainites in monitoring the core
[""\\C) operating state us well as uncertainties Q
la tt e critical power correlation itself.
These values also assure that operation For core flow rates less than rated.
Q will be such that the initial. condition the steady state 6 '* is increased by the M
formia given in the Specification. This assened for the IOCA analysis. plug 2% un-is satisfied. For any of the special assure that the MCPR will be caiutained ce r t.a i nlyof transients or disturbance cause.1 trY-greater than that specified in Specift.
Q a
set catson 1.1.A even in tha event that the ringle operator error or single eqisipmen.
ralfe% tion. at is required that design ro'.or-generator set speed corcroller analyses initin114ed at this steady stato cat:ses the scoop tube positioner for the less than fluid coupler to novo to the manie.u:n operating linit yield a HCPR of not that specified in Specification 1.1. A at asiy speed position.
7d tine doctng the transient asstning instau:sent For trip settings given in Specification 2.1.
4 analysis of the thest.at consenguences of these the v.ilue ut nCFR stated in t his specif icItion fo, (2), Generic Reload Fuel Application'"
I salttal
"*1"' "I NEDIC-2 4 0 l l-P-A
- t r. ins sent s.
}
ti.e t init iin conait ion of occiat non touads t he ncra assumed to esist psius t o t he anassation of the transients.
Tt.as inasial conIttson which is used in the t ransient. analyses, I
the MCPH fuct claddang intrysitt safety want preclude violation or an i.alculating the scquised limat.
Assu. net ions,and methoJs va.cdeach veload cycle.se documented in og I.
(
revision Ottmber
..L time reload I'
steady state HCru limit tog tact easing conservatism while poterence 2.
The results apply witsi IUOl aftalySeS die performed, opesating with MCPMs greater than specified.
6
. Amendment No.
42 853 i
j
ep
~
h.S Surveillence Requirec--nto Panes (cont'd)
I.
Averare Plansr LHCR.
K.
Minimum Critical Power Ratio (MCPR)
At core thercol power levela less than o--
At core thermal power levels less than or equal equal to 25 p:r cent, operating plant to 25 per cent, the reactor will be operating experience and thermal hydraulic analyses at mininu= recirculation pu.ap speed and the indicate that the resulting overage plansr moderator void content will be very small. For LtGn is below the c.aximum overage planar LHGR all designated control rod patterns which may be, by o~ considerabic marg n; therefore, evaluation empt yed at this point, operating plant experience' i
of the overage planar LilGR below this power.
and thermal hydraulic analysis indicates that the level is not necessary. The daily require-resulting MCPR value is in excess of requirements i
rent for 'cloculot.ing overage pioner LHCR by a considerable margin. With this low void content, any inadvertent core flow increase above 25 per cent rated thermal power is would only place operation in a more con-cufficient, since pcuer distribution shif te servative mode relative to NCPR.
are slow when there hn te not been signifi-cant. power or control rod changes.
lhe daily requirement for calculating
~
MCPR above 25 percent rated thermal power is sufficient since power distribution J.
Local I.9CR shifts are very slow when there have not been T1.e Ll'GR es a function of core height shall be check 2d daily during reactor operation at in addition. the K correction applied to g
crec ter than or caual to 25 p:r cent power to the 1.C0 provides margin for flow increase 9
determine if fuel burnup or control rM covenent from low flows.
O bas caused chenCes in power distributicn.
g A limiting LilGR value is precluded by a considerable margin when euploying a per minuible control rod patter n below 25'a rated thermal power.
M..
m Amendment No. 42 86 A
J
A.tt1ITIrlG CollDITIori FOR OPERATIO:t G G-3.6 4.6 SURVEILIAtICE REQUIRE?tEITF i
- 4. n orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition within 24' hours.
.2. The primary containment sump campling
- 2. The primary containment sump sampling system and an air sampling system shall and air sampling nyatem operability be operable during power operation.
If will be observed dally as In rt of
)
either a sump water sample or a contain-4.6.D.2.
ment air sample cannot be obtained for any reanon, reactor operation is permissible M
only during the succeeding seven days unless period.
b the system is made operable during this M
u b,
.W C. Safety and Relief Valves E. Safety and Relief Valves b
- 1. During reactor power operating conditions A minimum of 1/2 of all safety valves and whenever the reactor coolant pressure shall be bench checked or rep.sced with a bench checked valve each refuci 'ng is greater than 90 psig and temperature outages.
The popping point of the greater than 320 f. all eight of the safety valves shall be set a.s follows:
safety valves shall be operable.
The lumber f Valves Set Point ( Ps iq),
solenoid activated pressure valves shall g
1 1115*
be operable as required by Specification-2 1240 3.5.D.
{
}?]
2 IMO The allowable set point error for each valve is 11%
All relief valves shall be checked for set pressure cach re fueling outage.
The set pressures shall ber 2.
If Speci fication 3.6.U.1 is not met, an llumber of Valves Set Point (Psid orderly chutdown sha1.1 be initinted and l
1115*
~
the reactor coolant pressure and temper-
{
~~ j3
^
ature shall be 790 psig and 2320*f within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I 2
- Target Rock combination safety /rcitef valve Amendment No.
42 90 J
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