ML20126H357
| ML20126H357 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/31/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126H361 | List: |
| References | |
| NUDOCS 8104090646 | |
| Download: ML20126H357 (52) | |
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COMMONWEALTH EDISON COMPANYg l
D_0CKET NO. 50-237 i
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DRESDEN NUCLEAR POWER STATION UNIT NO. 2 AMENDMENT TO PROVISIONAL OPERATING L_ICENS[
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Amendment No. 58 License No, DPR-19 1.
The Nuclear Regulatory Commission (the Commission)' has found that:
A.
The application for amendment by the Commonwealth Edison l'
Company (the licensee) dated January 28, 1981, supplemented by your letter dated February 23, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules i
and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the apolication, the provisions of the Act, and the rules and regulations of I
the Commission; C.
There is reasonable assurance (i) that the' activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety I
of the public; and E.
The issuance of this amendnent 15 in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable I
recuirements have been satisfied.
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2.
Accordingly.. the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B and 3.F of Provisional Operatino License No.
DPR-19 are hereby amended to read as follows:
i 3.B Technical Specifications i
The Technical Specifications contained in Appendix A as revised through Amendment No. 58, are hereby j
incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
.3.F Restrictions i
Coeration in the coastdown mode is permitted to 405 power.
Should off-normal feedwater heating be necessary for extended periods during coastdown (i.e., 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 8
for the worst case abnormal operating transient remain bounding for the new condition.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION i
JMMb!
Dennis M. Cruten.1e1, Cglet Operating Reactors Branch *5 Division of Licensing
Attachment:
Changes to the Technical i
Specifications Date of Issuance: March 31, 1981 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 58 PROVISIONAL OPERATING LICENSE NO. DPR-19 p0CKET NO. 50-237 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
2 81 4
81B 5
818-1 6
81 C-1 7
81C-2 9 (Blank page) 810 10 82 11 85A 12 85B 13 86A 14 90 15 16 18 20 21 22 Add pages:
81 C-3 26 81C-4 and 29 81C-5 34 42 42A 45 46 47 48 49 57 57A 60 62A 63 64 71
DPR-19 I.
Limiting Conditions for Operation (LCO) - De N.
Mode - The reactor mode is that which is limiting conditions for operation specify the established by the mode-selector-switch.
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minimum acceptable levels of system perform-ance necessary to assure safe startup and op-O.
Operable - A systen or component shall be cration of the facility. When these conditions cor:sidered operabic when it is capable of are net, the plant can be operated safely and performing its intended function in its re-abnormal situations can be safely controlled.
quired manter.
J.
Limiting Safety System Setting (LSSS) - W e P.
Operating - Operating means that a system liniting safety system settings are settings on or comronent is performing its intended instrwaeatation which initiate the automatic functions in its required manner.
protective action at a level such that the safety limits will not be exceeded. The region Q.
Operating Cycle -Interval betweca the end between the safety limit and these settings of one refueling outure and the end of the next subsequent rofuoling outage.
represents margin with normal operation lying below these settings. The margin has been established so that with proper operation of the it.
Primary Containment lategrity - Primary instrumentation the safety limits will never be containment integrity neans that the drywell exceeded.
and pressure supprossion charaber are intact and all of the following conditions are satis ied:
r K.
Praction of Limiting Power Density (FLPD) -
The fraction of limiting power density is 1.
All manual containment isolation valves on the ratio of-the Linear lieat Generation lines connecting to the reactor coolant sys.
llate ( LilGR) existing at a given location tem or containment which are not required to the design LilGIt for that bundle type.
to be open during accident ronditions are closed.
- 2.. At 1 cast one door in each airlock is closed and sealed.
L.
Logic System Function Test - A logic sys-tem tunctional test means a test of all relays 3.
All automatic containment is olation' val ves 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.
g ble, action will go to completion, i.e., pumps
- 4. ~All blind flaneces and manways are closed.
will be started and valves opened.
S.
Protective Instrummtation Definitions M.
Minimum Critical Power Ratio (MCPR) - The 1.
Instrument Charanel - An instrument chan-niairaua in-core critical power rati nel means an arrangement of a sensor and corresponding to the most limiting fuel auxiliary equipment required to generate assembly in the core.
and transmit to a trip systen a single trip signal related to the plant parancter 2
Amendment No. 58 monitored by that instrument channel.
3
I DPR-19 t
flB. ' Simulated Automatic Actuation - Simulated l
2.
Secondary Containment Integrity - Secondary -
automatic actuation means applying a sinu-
- containment integrity means that the reactor.
1:ited signal to the sensor to actuate the building is intact and the following conditions circuit in question.
are met:
CC.
Surveillance _ Interval - Each surveillance 1.
At least ong door in each access opening requirement shall be performed within the l
is closed, specified surveillance interval with.
2 2.
The standby gas treatment system is A maximum allowa'ble extension not to _
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a.
I operable.
exceed 25% of the surveillance interval.
i All automatic ventilation system isolation 3.
b.
A total maximum conbined interval time valves are operabic or are secured in the for any 3 consecutive intervals not-to isolated position.
exceed 3.25 times the specified surveillance interval, I
AA.
Shutdown - The reactor.is in a shutdown con-1 dition when the reactor mode switch is in the shutdown mode position and no core alternations DD.- Fraction of Rated Power (FRP) -
I are being performed. When the node switch is The fraction of rated power is the ratio of core thermal power to rated i
placed in the shutdown position a reactor thermal power of 2527 Mwth.
m scram is initiated, power to the control rod drives is removed, and the reactor protec-l tion system trip systems are de-energized.
Transition Boiling - Transition boiling neans EE.
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I the txiiling regime between nuc1cate and film l
1.
Ilot Shutdown means conditions as above boiling. Transition boiling is the regine -
I with reactor coolant temperature greater in which both n cleate and filn boiling
- than 212*F.
occur intermittently with neither type
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i being completely stable.
j 2.
Cold Shutdown means conditions as above I
with reactor coolant temperature equal FF.
Maximum Fraction of Limiting Power,
to or less than 212*F.
Density (MPLPD) - The maximum fraction of limiting power censity is the highest value existing in the core of j
the Fraction of Limiting Power Density I
(FLPD).
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i Amendnent No. 58 5
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5 1
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I DPR-19 2.1 IJMITING SAltiY SYSTEM SBITING 1.1 S <r,iY LIMIT 2.5 Ftll'.L CI.ADDI:C INTEC!!ITY 1.1 } itri. C!. Anal::G I:iTECRITY
_ Applicability
[qplicaStlity The I,imiting Safety Syste.m Settings The Saf ety Limits cut 1 lished to apply to trip settings of the instru-preserve the fuel cladding integrity ments and devices which are provided-variables uhicli-the fuel cladding integ-apply to these to prevent cenitor the fuel thermal behavior.
rity Saf ety Limits f rom being ex-cceded.
Objective Cblective 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.
1cvel 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.
Specifications Specifications A.
ficutron Flux Trip Settines_
A.
Ee ctor Pressure >800 psig and Core Fleu > 107. of Rated.
The limiting safety system trip settings shall'bc as.specified The existence of a minimum critical below:
power ratio (t:CPR) less than 1.07 l
. shall constitute violation of the MCPR fuel cicdding integrity safety limit.
Miendment No. 58 5
DPR-19 1.1 SAFETY LIMIT 2.1 LIMITING St.FEIT SETE:t SETTD:G
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1.
APPJi Flux Scran Trin Settine (Run Moh)
When the reactor mode switch is in the run positiott, the APitt flux scram setting shall be n I
S <
.6SJ + 55 D
with a maximum set poliit 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 = per cent of drive flow required to produce D
a rated core flow of 98 Mlb/hr.
In the event of operation with a maximien fraction of Ilmiting power density (MFtPD) greater than the f raction of rated power (IRP), ttie setting shall be modified as follows:
5 $ (.65Wg 55) [ h D I Where:
FRP = fraction of rated thermal fiewer (2527.tiWL)
Itf LPD = maximum fractior: of limiting power density where the llattin0 power. density for each bundle is the design linear heat genceatli,n eate for that bundle.
The ratio of IRP/flFtrD shall be set equal to 1.0 unless the actual cpetaling value is less than 1.0, in which case the actu.il operating value will be used.
Thl-
'tjustment may also be performed by increasing the APRrt gain by the inverse ratio, P7LPD/FRP, C ich accomptithes the same degree of protection as reducing the trip setting by rRP/MFLPD.
= _ _
2 fr: Plux Se t i ?;1o s 'h*
(Fere?1 e-Stertt. v t-6 Met Stann, I *g Amend: rot No. 58 u.en t'in. reace.c ~4 witch is in t :ie t c f vel o r s t a r t :.* /h..: se,. thy yo=1 t ion. l'ac Al'."1 a c r a s's. Il M,e t
.s t le*s' then we et.st t a IM of ested newtes.i f l..s.
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1.1 'SAFR Y LIMIT 2i ' LIMITING SAFETY SYSTEM SEITING e
- 3. Core Thermal'Po*cr Limit (Reactor 3.
IRM Flux Scram Trin Settine; Presst:re (COO pu.t:)
Tne Itut flim scram setting shall be Vaen the reactor pressure is < 800 set at icss than or equal to 120/125' o f psic or core flots is less than 10%
full scale.
of. rated, the core thermal power shall not exceed 25 percent of rated j
thermal po:cr.
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Be AFRM Rtyi Blech Settin.g C. Pmser Trans!cnt i
The APRM rod block cettinC shall be:
1
)
1.
The neutron flux shall not c:teced the scram 2
setting established in Specification 2.1.A l
7 for lonner than 1.5 seconds as indicated by l
S <
.6SJ D the process con puter.
. The, definitions used above for the APF,.4 scru 2.
Tacn the process cocputer is out of service, trip-apply.
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j th15 safety 11Dlt shall be assuDed to bc In the event of operation with a sentmum fraction 11rities power density (W1.FJ) l C7.cceded if the acutron flux cxecedS the Soram greater than the traction of rated power (H'P), the setting shall be moJ1ftee as follows:
setting established by Specification 2.1. A 4
P end a control rod scran does not occur.
sa.(.65v,+43)7
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_R..enete k.*ater L-vel (Shutdown Condition]
value la less than 1.0 la which case the actual operating value will t>= used.
The rat to of FAP to MFLPD shall tie act equal to 1.0 unless the actual operating i
1 Thle adjustment may also be perf ormed by increasing the APRM gain by the Inverse 1
retto, M!LPD/tpr, which accomplishes the same degtee of psotection as redocing b.senewr the reactor is in the shutdtren condition the trip etting by rar/Mrtro.
with irradinted fuel in the renctor vessele the
.uater ic cr 1 shall not bc Icys than that corres-peini:a.. to 12 inches above the tep of the active f p *uhen it is seated in the core.
- Top of active fuel is defined to be Amendment No. 58 7
360 inches above vessel zero (see i
Bases 3.2).
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DPR-19 Q..I _
Safety Linit Bases Revised w/ Change 22 dated 8/29/75.
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F'JEL CDDDING INTEGRITY
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Tha fuel cladd:ng integrity limit,is thresho)<* beyond which still set such that no calculated fuel dam-greater thermal stresses may 1
'w'e would occur as a.resulm of an cause c>ross rather than incre-v nboormal opera tional transieno.
Be-mental cladd ing> deteriora tion.
l cause fuel damage in.n.ot directly Therefore the fue l c., add in,,,
i g
c m *nble a step-back approach Y
ubed to establish a Safety Limit margin to the conditions which t s st.ch that the minimum critical power would produce onsev of. transition i
k ratio (ECPR) is no less than the.MCPR fuel bo iling,. (MCPR _. oi
- 1. 0 ).
gle,,,
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i clodding integrity safety limit.
MCPR7 the conditions represent a significant I
IICPR fuel cladding integrity safety limit departure from the condition 2n-
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represents a conservative margin relative to tended by design for planned.
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the conditions required to maintain fuel cladding operation.
Therefore, the MCPR fuel
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y g g)7ggY' cladding integrity safety limit to estab!!shed such that no calculated fuet damage is espected to occur as a result
,j The iuc1 Cladding is one.of the abnormat operational transtent. Basis of the values derived of an-for thle sarety H.it for each ruet ype is docu.esaed !=
physical barriers which separate i
- '"'a" 8-i radloactive mater 10lo. Arom the A.
Reactor Pressure t 800 psig an,s,.
l.
environs.
The integrity of this Core. Flow > 10,. oi Ra te a.
j cladd*n$ barrier is related to its m
relativ freedom from perforations Onset of trancition boiling results or cracking.. 'Although some cor in a decrease in heco transfc.
rc 1
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realon or use related cracking raay the clad and, therefore, elevated occ ur durin:7, the life of the elad temperature and the poss'ib.111t,7 cladding f'ission product migra tion of elad failure, l' owe ve r, the i
from this source 13 incrementally existence of critical power, or t
cumulative and continuousl,,,
boiline transition, is not a dirce,wly I
observable parameter in an opera
,n, o
measurable.
Fuel t ladd ing per_ -
vi forations however can result from reactor.
Therefore, the marSin so the:nl s tresses w[lich occur from boiling transition is calcula ted reac tor opera tion significant1y from plant ~ opera ting parameters such abo te desien conditions and. the pro-as core power, core, go>t,. fcedwatc" tec tion c.ys tem safety se ttings.
temperature and core power-d1stri--
j th11e fission product migration from
- bution*
The margin for each fuel j
c1Scd ing perrora tion -is s,use as assembly i s c ha ra c te r.,.,.,ed by. %...
nea surab,ic as that.. rom use related critical power tatio-(CPR) which is 3
cracking, the-therxilly caused tha ratio of the bundle power.which cin.hn ng peri orations o mna a
. Amendment No. 58-would produce oncet of transition 10' 4
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DPR-19 4
safety timit Bases However, if boiling transition were to occur, clad perforation'weuld not 1.1.A Reactor Pressure > 800 psig and be expected.
Claddins temperatures Core Flow > 10% of Ra ted.
(cont'd) would increase to approximately y
O llOO F which is below the perforation 4
boilin:; d ivided by the actual bundle power.
temperature of the cleading material.
The minimum value of this ratio for This has been verified by tests in any bundle in t,he core is the minimum the General Electric Test Reactor c ri tical power ra t,lo (MCPR).
It is (GETit) where similar feel operated a s s u-,e d that the plant operation is above the critical heat flux for a
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controlled tu the nominal protective significant period o f t i r:e (30 i
setpoints via the instrumented vari-minutes) without clad perfera tion.
4 ables.
(Fir,ure 2.1-3).
If reactor pressure should ever i
l The t1C Pit fuel cladding integrity safety limit has exceed 1400 psia during normal power j
sufficient conaervatism to assure that opera tion (the limit of applicability i
in the event of an abnormal operational of the boiling transition correlation)
I t ra ns ient initiated from a normal it would be assumed that the fuel cperating condition tr. ore than 99 9%
cladding intec, ity safety Limit has of the fuel rods in the core are ex-been violated.
i pected to avoid boiling transition.
i
'1he margin between FiCPR of 1.0 (onset l
I of tra:.sition bolling) and the safety In addition to the boiling transition limit 1
1*mit is derived from a detailed 3:CPR) operation is constrained to a maxinum s ta tis t ical ana lya is considering all Il!CR - 17.5 ku/f t for 7 x 7 fuel and la.4 kw/ft i
tr 11 ex fuel types.
This nstraint is stablished by o r t t' - uncerta j nt ies in monitorin5 j
Specification 3.5.J to provide adequate safety margin to 14 the Cer0 ope ra t in'*, s ta te inc lud in5 plastic strain for abnormal operating transients initiated' t'1c e r ta in ty in the bolling transition from high power conditions. Specification 2.1.A.1 provides correlation.
See e.
- g. Hererence (1).
E r equivalent safety margin for transients initiated from lower power conditions by adjusting the APRM fit,w biased L scram by the ratio of TRP/Hrt.PD..
Specification 3.5.J
,l'eaause the boiling tratis ition Cor-established the LHGR max which cannot be exceeded under steady rela tien is based en c large quantity Power operation.
l of full scale data there 13 a very h
- gh co:1ridence the t opera t3on of a (1) "Gener ic Reload Fuel Application," NEDE-24011-P-A*
f uel asse:nbly at the condition of i
L:CPR = the MCPR fuel cladding integrity
, Approved revision number at time reload fuel analyses are performed.
safety limit would not produce boiling d transition.
yy Amendnent No. 58 E
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- 2.1 Limitin:r Safety System Settine Bases 1.1 Safety Linit Bases FUEL CLADDI!IG INTEGRITY lol.C Pover Transient (cont'd) '
The.cbnormal operational transients The computer provided has a applicable to cperation of. the units se.;uence annuncia tion program *uhich have been analyzed threushout the will ind ica te the ccquence in which spectrum of planned operating con-nerams occur such as neutron flur, d itiorec up to the rated thermal power pressure, etc.
This program also cond ition of 25'?7 F0 t.
In additica, indiertes <. hen tha scram setpoint is 2527 Mt is ' the licensed maximum s tendy-cleared.
'Ihis will provide informa tion state power level of the ur.its.
This
~
long a scram condition exists maximum steady-state power level will en now and thus provide some measure of the never knowingly be exceeded.
caergy added during a transtent.
- Thus, computer informat ion normally will be Conservatism is _ incorporeted in.the a va ilcb le for analyzing scrams; how-ever, if the computer informition should transient analyses in estimating the controllins factors,.such as void not be ava11able for any scram analysis, reactivity coefficient, centrol red Specification 1.1.C.2 uill be relied on scram worth, scram delay time, peaking to determine if a safety limit has been factors, and ax121 power shcpec.
These v;olated, factors are' selected conservatively with respect to their effect en the During periods when the reactor is shut applicable transient results as deter-d o..n, consideration must also be given mined by the current an.' lysis medel.
~
to w: ter level re:liairements due to the I
effect of decay hent.
If reactor Nater Conservatism incorporated into the
.tevel should drop below the top of 1.he transient analyses is documented in ac..ve.ucl ouring ohis u me, che Beference 1.
Transient analyses are ab;11ty to cool the core is reduced.
initiated at the conditions given in this-2his reduction in core cooJ.ing cap-reference, ability could lend to elevated cladding temperaturen and clad perforetlon.
The core will be cooled sufficiently to pre-i I
vent clad melting should the water level Le reduced to two-thirds the cere height.
Amendment No. 58 limit a t 12 l Establishment of the safetythe top of the fuel
- prov! des inches above e
ade.!uate margin.
This level will.be con-tinuou'ily monitored whenever the recir-culatica pumps are not operatins.
I'
,3 I
- Top of active fuel is defined to be
~
360 inches above vessel zero (see ~
Bases 3.2).
2*la Il ting Sarety System Setting Bases Steady-state opera tion without.foreca except durinG startup testin,g, recircu i
DPR-19 Fuel Cladding Integrity (cont'd) analysis to support operation at The cbaolute value of the void reac-7he i
has considered operation with evarious tivity coefficient.used in the analysis in cohserva tively es timated to be about 2:5 greater than the nominal maximui one or two recirculation pumps. ither value expected to occur during the core The bases for individual trip settin lifetime.
The scram 1. orth used has are discussed in the following para gs been derated to be equivalent to appro-graphs.
x ::.a tely 80;; o f the total scram worth of c
o
%- $;: control roda.
The scram delay d
ate of rod ' insertion allowed v e-For analyses of the thermal consequences of ty the analyses are conservativeyy w,.e g the transtents, the MCPR's stated in k> ara raph equal to the longes t dcIny and slowest 5.5.K as the limiting conditiot of operation Inse-tion rate acceptable by Technical bound d"ecificationa.
Tne effect of scrDm I
those which are conservatively assumed 5
-i to exist prior to initiation of the transients.
-th.
scram deley time and rod in-A.
ell conservatively Neutron Flux Trip Settings
-n-*'
k.a'lkonrata ed, or$'of greates t s ignificance in' the ecriy portion of the nega tive 1.
APR!1 Flux Scram Trip Setting
(
_ Run Mode) reactivity insertion.
The rapid in-
$N ured byn-clon of negative reactivity is The average power range monitoring the time requirements for
( APitM) system, which is calibrated
\\ c.; n i.d 2 0 1 insertion.
Dy the time using heat belance data taken during i
Ehe roda are 60% inserted, approxi-s teady-state concitions, 1
reads in 1
-.a tely four dollers of nega tive recc-percent of rated,thcrmal power.
Be-t:v atv have been inserted which cause flesien chambers prov'3e the basic etrcn' gly turns the trans ient, and input alunals, the iPItM syste:a responds acccmpliches the desired errect.
The.
directly to average neutron flux.
i During transients, the instantaneous times for 504' cod 02; 1"3erti n a re j
g: ren to assure proper congletion of ratc~ of heat transfer from the fuel (reactor thermal power) is the expected performance in the less than enr3*er portion or the transicut, the Instantaneous neutron flux due to i
and to establish the ultimate fully the time constant of the fuel.
1 There-nbutdo:.n s tecdy-a te tc cond ition.
fore, during abnormal eperational t ra ns ie n ts, the thermal power'or the i
This choice of ualng conacrvative valuc3 fuel will be less than that indicated er centrollinS parametera and initiatin6 by the neutron flux a t i
tranalcots at the design power lev Analyses demona tra te tha t with a 120the scrcm setting.
preduces mere pessimistle answers,el, 4-than percent scram trip settlng, i
none of the v.cul) reault by using expected values of abnormal opera tional trenaientat analyzed ne>.er 1ckeln.
analy:Ing at higher violate the fuel Safety I,le:ic and there control naremetera and is a nubs tantlal marr,in from ft;el damage.
(
There fore,
1 the une or flow referenced I
Amendment No. 58 cermu trlp providea -even additionel nerg'n. 34 3
i
,..me c
r 4
tron
- Flux ' Trip Settinc',s_
c.1.A.
DPR-19 1,
- PR:4 Flux -Scram Trip Setting (Rid I~. ode)
(cont'd) ture coefficientorarb small, and con-An increase in the APRM scram trip setting would decrease the margin pre-t'rol rod patterns are constrained to acnt before the fuel claddins integrity be' uniform by operatins procedures Safety Limit is reached.
The APRt4 backed up by the rod worth minimizer.5 scran trJp setting was determined by Of all possible sources of reactivity an analysis of rargins required to pro-input, uniform control rod withdrawal v de u reasonable range for maneuvering is the most probabic cause of signifi -
during opera tion.
Reduc ing this oper-cant power rise.
Because the flux at:ng margin would increase the fre-distribution associated,:ith uniform quency of spur *ous scrams which have an rod withdrauals does not. involve high adverse effect on re.'ictor safety because' local peaks, and because several rods of the resulting thermal stresses.
- Thus, must be moved to change power by a the f.PR;l scran trip setting was selected significant percentage of rated power, t'e ca us e it provjdes adequate margin for the rate of power rise is very slou.
the ft.el cladding integrity Safety Limit Generally, the heat flux 13 in ncar yet allows operating margin that reduces equilibrium with the fission. rate.
In an assumed uniform rod withdrawal ap-the possibility of unnecessary scrams.
proach to the scram level,,the rate of The scram trip setting must be adjusted power rise is no more than 5 percent 1
to ensure th-t the Ll!GR transient peak of rated power per minute, and the is not increased for any combination of APR!4 system would be more than adequate to assure a scram before the poner outrue Fix t ion of t im6 t in1 Powes t>ca-It y qMr trm and s cactor core t h. i s. 31 g=ame s. The ssl tem he t t i tig tS ad just eet in acce.gdance with the Cottld t'XCCed tlnSafety limit.
The 15 h
to. la in specteicition 2.t.A.I. when the titTLn is greater than the Fo xt ion of Rated ruwet (fpri. The adjustment way be accomplished by pgpgggg gp,M ggpag pgggggg ggggyg gg,
]
in.ne.istag the APPM qien by the tecipsoCJI Ut FPr/MFLPD. This pro
- Ides til the mode switch is placed in the the same degree of psottst6cn as seducir.q the is ip setting by Fpr/MFLPD g
e g
by r aising t he initial AFFM te. ding closer to the trip setting such that a h
sesam would be seteleed at the sama gioi nt in a taansient as if the trip reactor pressure is greater than 850 setting haa teen eae,d.
psig.
2.
}i& fuel or Start (c liot Stahdby Mode) 3 IRi4 Flux Se' rem Trip Setting For operation in the startup mode while the reactor is at low pressure, the APRi4 The IRM system consists of 8 chambers, scram setting of 15 percent of rated porter 4 in each of the reactor protection provides adequate thermal margin between the system logic channels.
The IR;~ is a the se'.po 'n t and the safety limit, 25 ptr-5-decade instrument which covers the cent of rated.
The margin is adequate to range of power level between that acce== eda te antic ipa ted maneuvers associa ted covered by the SRM and the APRMs The
.1 *.h ps..c r p la nt s ta rtup.
Effects of in-5 decades are broken det.n into 10. ranges, c rs.as ing preature a t zero or low void con-each being one-half of a dec.ade in aire.
teat are minor, cold v.ater from sources c. ille:.le duriv. startup is riot much colder Amendment No. 58 15
.........e..q, t ene ra-s.
DPR-19 4
2.1.A.
Neutron Flux Trip-Setting' 3, IRM Flux Screm Trip Setting (cont ' d')
2.1.3 i
The IRM screm trip setting of 120-i APRM Rod Block Trip Setting
' iv is ?.cn3 is active in each. range.of d
Reactor power level may be rcried by the.iRM.
Fer exampic, if the instru-nent were on range 1, the scram setuing moving control rods er by varying the recirculation flow rate.
The APRM v.ould be a 120 diviaions for tnat range:
i system provides a control red block to l iiCen isa Jr tho inatrument were on range 4
-e preven.
gross rod withdrawal u
5, the scram would be 120 divisions on that range.
Thus, as the IRM is ranged a t cons en, rec i rc ula t ion ?'. m.-
o a
d-i ep to accomod Le the increare in power ing the MCPR fuel cladding integrity level, the screm trip setoing is als
^
' safety limit.
This rod n
renged up.
block trip setting., which is aute-i matically varied with recirculation Tha most significant. sources of reac--
loop flow rate' prevents an
- c '-a a s e -
1 tivit" change durinc the power increase in the re' actor power level to axcas-3 e
cr dua to control rod withdrawal.
In 31ve values due o control roc w'*"-'i-4 u
.he IHn.. provided d rawa l.
orde-to ensure that T.he
- f. low varin.la. trio s **"'in5-u o -
cdequate protection agains,. the single provides subs,antial margin from fue-l u
rod uithdra unl crcor, a range of rod 4
i x!thdretal acc idents was en,1yned.
This gamage, assumJ ng a steady-state epera-
.hg r. rip gettgyg,- over the u
n au enclysis ineleded starting the accident ent ire rec arca le t)oa i nu r: re -
--s at various power 1cvels.
,,.he mon ce-i o
marg in to. the
.,.,.3 fety Limit increaser as vere ecce involves an initial cond ition the flo.s.. d e c re a s e's f"-
tha c ~'
'"*a.
in which the reactor is just subcritical m ng versus fl w rela" c - otionship;
- 9**
- ^
i a"~i tha IP.M system is not yet on scale.
therefore the ucrst case MCPR which i
Addit *onal conservatism was taken in this could occur durins stecdy-state opera-
.tien is et 100% of rated thern:rl pcwer cn.; lysis oy assuming that the IBM channel clouest to the withdrawn rod 13 bypassed.
because of the APRM rod block trip The results of th13 analysis show that the c e t t ins.'
The actual power dis tributien reactor 13 scrane. icd cnd peak power limited in the core is catablished by specified to one percent of rated power, thus maintaining control red sequences and is conitored j
n MCPR above the MCPR fuel cladding integrity continuounly by the in-core LPMM sys tem.
safety limit based on the above As with the APdM scram trip sett:ng, p
analys*s, the IHH provides protection against the APHM rod block trip setting 13 ad--
locc1' control rod w ithdrewal errors and con-Justed downward if the ma.zimu a.f raction i
.tinous withdrawal of control rods in ses:e?nce of limiting power density arcaada the fracti rated power, thos'oreservin3 and prov vies backup proteccion for the APHM.
Pthe Ai,on ofHH red block s afety reargin.
j i
16 Amendment No. 58
4 DPR-19 L Tr:Mm Ste, Yelve Scron - Tho turbino otop valve -
G.
Reactor Coolant Low Pressure Initiates Main Stcan 1
c'caun ceran trip enticipates the pressure, Isolation Valve Closure - The low pressure isolation n:ut:ca flux ard heat flux incresso that could at 850 psig was provided to give' protection against result frc, rnpid cIccuro of the turbine stop fast reactor depressurization and the resulting valves.
Uith a scrsn triu cctting of 10 rapid cooldown of the vessel.. Advantage was taken Ierc:nt of valve clonure fron full open, tho of the scran feature which occurs when the main re.:ltant increaso an uurface heat flux is steam line isolation valves are closed.to provide linited such that I;CPit rerains above the'MCPR for reactor shutdown so'that operation at pressures
'"P""
fuel cladding integrity safety limit, even safety limit does not occur, although operation Lduriny the worst case transient that assumes
} the totbine bypass is closed.
at a pressute lowr i.han 650 psig woord not necessarily, constitute un unsafe condition.
11.
Main Steam Line Isolation Valve Closure Scram
'The Iow pressure isolation of the main steam lines at l
r.
p nerator T.oed Re jection Scram - The genera-850 psig was provided to give protection against tor lend rejection scram is provided to rapid reactor depressurization and the.resulting caticipate the rapid increase in pressure rapid cooldown of the vessel. Advantage was taken cad.:cutron flux resulting f rom of the scram feature which occurs when the main fast. closure of the turbino control valves steam line isolation valves are closed, to provide for reactor shutdoun so 'that high power operation -
i due to a lead rejection and subsequent at low reactor pressure does not occur, thus providing j
retlure of the bypans; 1.e.,
it prevents
'"'D frcn becoming less' than the MCPR fuel protection fot" the fuel cladding integrity safety limit. Operation of the reactor at pressures lower cladding integrity safety limit for this
.than 850 psig requires that the reacror mode switch transient.
For the load rejection without l(, bypass transient from 100% power, the peak be in the startup. position where protection of the h neat flux (and therefore LIIGR) increases on fuel cladding integrity safety limit is provided by the order of 15% which provides wide margin the IRM high neutron flux scram. Thus, the cochination
! to the value corresponding to 1% plastic strain of main steem line low pressure isolation and isolation -
e o t' the cladding.
valve closure scram assures the availability of neutron flux _ scram protection over the entire-range of applicability of the fuel cladding integrity safety limit.
In addition, the isolation valve closure scram anticipates the pressure and flux-transients which occur during normal or inadvertent isolation valve closure.
With the scrcms set at-107. valve closure,there is tio increase in neutron flux.
Amendment No. 58 i
i 18
DPR 9 l
Bases _:
The relation = hips of atre== 1.v.1, ta yt.14 ottenxth 1.2 De reactor coolant system integrity Jn nn impor-tent barrier in the prevention of un' controlled re.
are conparnhie for the tacIntien ec.t e, or e-d lease of fission products. It is encential that the.
primary systen piping end prevt.!c n n i 3111e e.ir-j integrity of this system be protected by establishing gin of protection at the estab119hcd gn!cty recature 4
l ic'i t -
a pressure 11 nit to be oboerved for all operating cenditions and whenever there is irradiated fuel in The normal operating pressure of the renetor ecolant the reactor vensel.
system is 1000 pain. yor the turbine trip or ic-g of 4
electrient load trar.sients, the turbin. trip serrei or The pressure safety limit of 1325 psig no measured generator load rejectien nerne, tegether with tSe by the vessel nteam space pressure indicator is equivalent to 1375 psig at the lowest elevation of the turbine bypass synten, 11ott the pre.sure to.rpro.1-
'estely 1100 psig (4). In addition, pre sure relief reactor ccolant systen. The 1375 psin value is i valves have been provided to redu e the pro'anbility 4
crived f rom the design pressures of the reactor pressure vessel, coolant. system piping and inola-of the safety valves which discharged to tion condenser. The respective design pressures the drywell operating in the event that 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 uns chosen as the lover of the pressure traitsients permitted Finally, the natety valves are sizec to 4eep i
I by the applicabic design codes: ASlE Boller and the reactor coolant system pressure belo.s 1375 p<te, Pressure Vessel Code, Sectien III for the pres sure with no credit taken for the relief valves d*:t.ing th3 vessel and' isolation condenser and l'SASI 331.1 Code fer the reacter coolant system piping. The t.S!3 postulated full closure of al' MSIV's withcut d irect (valve position switch) scram. Credit geiler and pressute Vesici Code perotts pretinure j
transtents up to 10% over design pressure (110%
is taken for the neutron flux scram, however 7.1250 - I'l?5 psig), and the USASI Code pernits the indirect flux scram and safety "alve pressure transients up to 20I over the desi.n actuation provide adequate margin below the pressure (120% X 1175 = 1410 psig). The Safety lpeak allowable vessel pressure of 1375 psig.
Licit pressure of 1375 psig is referenced to the Iovest elevation of the primary coolant nyutem.
Evaluation methodology used to assure that this safety limit prensure is not exceeded Reactor pressure is continuously monitored in the 2
j for any period is documented in Reference 1.
contro'l room during operation on a 1500 pai full ne de sinn bar.is for the reactor pressure vessel scale pressure recorder.-
=akes evident the substantial c argin of prutection a ;; i l - = t failure at the safety pressure limit of 1375 psig.
The versel has been designed for a general (4)
SAR, Section 11.2.2. -
merb ma strer.s no greater than 26,700 pst at an inteen I prev;ure of 1250 psig; this is a factor of 1. 5 be l.r.e tbc y1cid ntrength of 40,1no psi at 575*F.
At tl, prennure linit of 1375 psig, the genern1 eet:t-ace a tre,s vill s.nly be 29,400 poi, still safely below the yield strengtlg.
20.
Amendment No. 58
4 4
DPR-19 f.
I
'Reses:
2.2 In ec rpliance with Section III of the ASME Code. the carcty volvco cunt be cet to open at no higher then 2
1033 of design preocure, and they nunt limit the re.sctor preanure to no core then 110% of denign prennute. Both the neutron f hxacra., and anfety valve actuatlon are re guired to p. event overpree-purir.ing the reactor preenure vennel and thus exceeding the pre.sure nnfety limit. The prennure j
l1 scram is available as a backup protection to the high flux scram.
If the high flux scram were to fail, a high-pressure scram would occur at 1060 poig. Analyses are performed as deceribed in t'he Generic Reload Fuel i
Application, NEDE-240ll-P-A (Appro0ed revision number at time reload analyses are performed) for each reload to assure that the pressure safety limit is not exceeded.
i.
r i
Amendnent No. 58 21-O w -
,e r.
-ew-e,,,___e w
)
u
.._ c 3.1 LIMITING COSDITION EVR' OPERATION 4.1 SURVEILLANCE REQUIRE!ENT,
t 3.1 REACTOP. pFDTECTION SYSTE?t 4.1 _ REACTOR PROTECTION SYSTEM Apolicability:
g pitcability_:
. Applies to the instrumentation and Applies to the surveillance of the instru::en-
.nsociated devices'which initiate a tatha -and associated devices which initiate i
reactor scram.
react or scram.
i Gjbj ect ive:
Objective:
To assure the operability of the To specify the type and frequency of reactor protection systca.
surveillance to be applied to the protection ins trurr.e n t at ion.
Specifications Specifientton:
A.
The setpoints, minimum number of A.
Instrutrentetion systeins shall be trip systems, and minimum number functionally tested a:3d caliGrated as of instrument channels that must indicaEed in Tables 4.1.1 and 4.1.2, be operable for each position of respectively.
the reactor mode switch shall be 1
as given in Table 3.1.1.
The Da ly during nactor power oporation, I*
i system response times from the opening of the sensor contaet up the core power distrib : tion shall bo I
checi:cd for maximum fraction of to and including the opening of k
the trip actuator contacts shall
_ limiting power _ density (MFLPD) and not ex ceed 50 millise conds.
compared with the fraction of Rated Power (FRP) when operating above 25% rated f n.
11 dosis** operation. the cianimum fraction of limiting 1
thermal power, power 4 nsiay caces-dr. the Ieaction of rated power when i
i
)
operatinq -ituve 2% s at ed ther mal gewer. c ither s a.
Tbc ArrM scram and rod block settingt shall be l
verbs. ed to the v31u s given t>y the equations in o
5:recifications 2.1.A.1 and 2.1.B.
This may be acce ptished by increasing the APRM gain as described therein.
b.
The power distribution shall be chaeged such that time maximum fraction of limiting power density no longer exceeds the f raction of rated power.
22
}-
Amendment No. 58
f DPR-19
. I TABLC 4.1.1 (cont)
Notes:
1.
Initially once per month until exposure hours (M as defined on Figure 4.1.1) is 2.0x105; thereafter, according to Figure 4'.l.1 with an interval not less than one month nor more than three months.
The compilation of instrument failure rate data may include data obtained from other Boiling Water Reactors for which the same.
i design instrument operates in an environment similar to that of Dresden Unit ;t.
a 2.
An instrument check shall be performed on low reactor water level once per day and'
~
on high steamline radiation once per shift.
3.
A description of the three groups is included in the bases of this Specification.
4.
Functional tests are not reouired when the systems are not required t,o be operable or are tripped.
If tests are missed, they shall be. performed prior to returning the systems to an operable status.
5.
This inntrumentation is exempted from the Instrument Functional Test Definition
( 1. F ).
This instrument Functional Test will consist of injecting a simulated electrical signal into the measurement channels.
6.
If reactor start-ups occur more frequently than once per'wcek,the functional test need not be performedt i.e.,
the maximum functional. test frequency shall be once'per week.
Amendment No. 58 26
i 4
DPR-19 i
The c..ntrol rod drive scram system is designed so stop valve closure scram and'eauses a scram t!ht ali of the w ater w hich is discharged from the before the stop valves are closed and thus the re-reacter bv a scram can be accommodated in the sutting transient is less severe. Scram occurs at discharge piping. A part of this piping is an in-23" lig vacuum, stop valt e cl6sure pecurs at strument volume ste-tube in the piping) which accom-20" lig vacuum and bypass closure at 7" lig modates in c.scess of 50 gallons of water and is the s acuum.
'm.
griint in the piping. No credit was tahen for i
fm-udume in the design of the discharge piping as Iligh radiation levels in the main steamline tunnel et ccerns the amount of water which must be accom-above' that due to the normal nitrogen and oxygen mot!ated therin;t a scram. During normal operation radioactivitr is an indication of leaking Scl. A j
LV d:, charge volume is empty ;.however, sh<sutd it scram is initiated wheneser such radiation level fill v.ith water, the water discharged to the piping exceedsthreetimes normal background. The pur-f rom the reactor coubt not be accommodated which pose of this scram is to reduce the source of such v.o:44 result in slow scram times or partial or no radiation to the extent necessary to prevent exces-control rod insertion. To preclude this occurrence, site turbine contamination. Discharge of excessise j
level switches have been pr ovided in the instrument amounts of radioactivity to the site environs is pre-i vol nue which alarm and scram the reactor when vented by the air ejector off-gas monitors which
{
tho veduim nf water reaches 50 gallons. As indi-cause an isolation of the main condenser off-gas j
cated abere, there is sufficient volume in the pipin@
line provided the limit specified in Specifica-t s accommodate the scram without impairment of tion 3. 8 is exceeded.
the scram times or amount of insertion of the control j
rads.
lhis incetion shuts the reactor dewn while The main steamline isolation valve closure scram sulficient vobime remains to accommodate the dis-is set to scram when the isolatiim valves are 105 ch irged water and preciwles the situation in which.
closed from full open. This scram anticipates the a scram would be rerpaired but not be able to per-pressure and Ilux transient, which would occur
~
farm its function aderpiately, when the valves close. Ily scramming at this set-3
[ ting the resultant transient is insignificant.
1..n of condenser vacuum occurs when the con-
{
j' cmcr c.,n no Ir.nger handle the heat input. Inss of comdenser vacuum initates a closure of the tur-A reactor mode switch is provided which actuates tano stop valves and turbine bypass valses which or bypasses the various scram functions appropriate i
chminates the heat input to the condenser. Closure to the particular plant operating status, llef. Sec-ut the tuihine stop amt bypass valves causes a pres-tion 7. 7.1. 2 SAll.
sure transient, neutron flux rise and an increase a
in surf ace heat flux. To prevent the clad safety
- Ihe manual scram function is active in all modes, limit f rom being exceeded if this occurs, a reactor thus providing for a manual means of rapidly insert-scram occurs on turbine stop valve closure. 'the ing control rods during all modes of reactor turbine stop valve clostnac scram function alone is operation, adequate to prevent the clad safety limit from being execeded in the event of a turbine trip transient
- Ihe IllM system provides protection against exces-
^
with bypass clusure. Itef. Section 4.4.3 SAH. "ihe sive power levels and short reactor periods in the -
condenser hnv vacutan scram i:- a back-up to the stan t-up and in 8.ermediate power ranges. Itef.
29 Amendment No. 58
i s
DPR-19 a half scram and rod block condition. *lhu s, For. the APIDI system drif t of electroalc
}
if the calibration were performed during oper-apparatus is not the only consideration In de-l ation, flux shaping would not be possibic.
termining a calibration frequeccy. Change In
{
Ilased on experience at other generating power distribution and loss of chamber seast-t stations, drill of instruments, such as those tivity dictate a calibration every seven days.
i in the Flow liiasing Nework, is not significant Calibration on this frequency assures plant I
mal theiefm e, to avoid spurious scrams a operation at or below thermal limits.
calibratica frequency of each refueling outagt.
is estabhahed.
A comparison of Tables ll.1.1 and 4.1.2 indicates that six instrument ch:innels have not Group (C) devices are active only during a been inclut!cd in the latter Table. These are:
giten portion of the operational cycle. For
.T! ode Switch in Shutd.m n,.ila: mal Scram, liigh e.'.ampic, the 1101 is active during startup and WaNr I.evel in Scram Ifischarge Tank,.Tlain inae:ive during full-power operation. Thus, Steam I.ine Isolation valve Closure, Generator the only test that is meaningful is the one per-1.nad Itejection, and Turbine Stop Yalve fortnett just prior to shuldnwn or startup; i.e.,
Closure. All of the devices or sensors associ-the tests that are performed just prior to use ated with these scram functions are simple of the instrument.
on-off switches amt, hence, calibration is not applicable, i.e., the switch is either on or.
Calibration frequency of the instrument chan-olf. Further, these switches are mounted nel is divided into two groups..These are as solidly to the device anti have a very lo.v follows:
probability of moving, e.g. the switches in the scram dischar ge volume tanu. nased on 1.
l'assive type Indicating devices that can the abuse,'no calibration is required for these be compared with like units on a continu-six instrument channels.
oos basis.
3 The MFLPD shall be checked once 2.
Vacuinn tube or semiconductor devices per day to deternine if the APItil and de:ectors that drif t or lose acra:1 requires adjust *:'.cnt.
This may sensiti vity'.
raay nonully be done by checking the LP.W1 readl'nga, TIP tracca, or Experience with passive type instruments in p::ocess com.puter calculations.
Commonwealth Edison generating stations and Only a snall number of control substation; indicates that the specified calibra-rods are nored daily and thus the tions are adequate. For those devices which
';ca, ding 4 actors are not expected cmplov vinpliliers etc., d:,ift specificatloas call fo'r dritt to bu less than 0.4%/ month; i.e.,
I to chnngo alcnificantly and thus in the period cf :t month a drift of.4% would g
it cally check of the MFLPD is occur and ti. :3 providing for adequa*e n'argin.
udcquate.
Amendment No. 58 y
DPR-19 INSTRGIENTATION THAT INITIATES RCD BLOCK Table 3.2.3
!*.in L= u:. L*O. of Conrnb10 In s t.-
Trio Level Settin'q _
.q Channels Per Instrument
~
I'RP Trin S/ste,(1)__
APRM upscale (flow bies) (7)
~
~ _* g5g f y-2.
I2I I
MPPED 1
'1 APml upscale (refuel and Startup/ Hot
$12/125 full scale Standby no6o) 2 APRI1 downscale (7) 2 3/125 full scale Red block monitor uprenle (flow bins) (7) I 65W + 'i2{
}2) 1
^
Rod block monitor downscale (7) 2 5/125 fell scale 1
1 3
Int downscale (3)
,15/125 full scale Iai upsc' ale-f.100/125 full. scale 3
Ird detectdi not fully inserted in 3
the core SRM detector not in startup position (4')
7(5) 6.105 counts /sec 2 (5) (G)
Smi upscale Amendnent flo. 58 42'
l TA3LE 3.2.3 (cont)
DIH-19 d
1:otes:
Startup/ Hot Standby and Run positions of the Reactor Mode Sclector Switch, f
c:: cept the Fer the there-nhall be two cperable or tripped trip cystens for cach function, 1.
IPGi downscala ar.d I?J4 detector not fully inserted in 570; red blocks, IitM upscale, the core need not be operable in the " nun" position and.TPPJ1 de.:nscale, APPJi 1
and Rilti downscale necil not be operabic in the Startup/Ilot Standby mode.
upscale (flow bias),
l
'1he IEl' upscale need not be operable at Jess than 30% rated thermal power.
One chennel may be bypassed above 30% rated thermal power provided that.
limiting control rod pattern does not exist.
For systems with more than If the firsti a
I one chonnel per trip system.
trip sy rccms,- tne systems shall be tripped.
coluain cannot oc mcc ror nocn i
4 2.
U nercent or drive rios required t.o produce a rated core ricw or p
d 90 i:2b/hr.
4 1
IP.14 dounseale may be bypassed when it is on its lowest range.
3.
This function mcy be bypassed.whc'n the count rate is 2100' cps.
4.
5.
One of the four SPJ" inputs may be bypassed.
This SP21 function may be bypassed in the higher IRM ranges when.the IRM upscile red 6.
block ir. operable.
Not required while performing low power physics tests at atmospheric pressure.during d
7.
or af ter refueling at power levels n6t to exceed 5 MW(tl.
tj t
Amendnent No. 58 42h
-.,s,
,m
.v r.,
. -..... -,.~
,.,s,-
-.~,. -.-
'AT11J. 4.2.1 (cont) patest 3
1 Intitally ence pt month tintti espesme boon (M et defined on flgttre 4.1.1) le 7.0 a 10 : actenitet, eccording to Figure.4.1.1 with en tenervel set less then ette enonth not more than three momhv. The complistion of Imtmment failure save data mey include date obtained inom othet l
Itallitig Wete Itcotton fee whicle the same dettgn Imtmment sterates tre an environment timlist to that of Dresden Unita.
7 functional test callhistions and tretmment checks are not ecquired when these trutntments are tiot required to be operable et are tilpped, functional tests shall be peeformed belove each startup with a sequired krequency not to enceed once per week. Caltheettom thall be performed dming each startup or du Ing contmtled shutdowns with a seguired frequency not to enceed once pet weelr. trutetiment checks shall he performed at leset once ps ween. Insisument cheebe shall be perfoemed at feest once per dey dueing thme periods when the Irwitumeme see required to be
- opeselle, 3.
This lmemmentation Is encepted from the functional tett defintthm. The functiomt tent will comtit of Infecting a timelsted electitcel elgnet ine, the mesentement thannel. See tiote 4 Th*se lminiment thennels will be callbeste.1 nelng almulsted elect:1 cal signait once every three months. In s Alltlon. celthention?oclading the 4
seemon will be penfmmed <beits.Ir each seineting entage.
5.
A tetnimine of' tuo chnnnels is rerluired.
6.
From and af ter the date thnt one of these parnmeters (... either drywell-torus differential prennute or torus water Icvel Indtention] is reduced to one indiention, continued operation I
is not permissible heyond thirty days, unless such instrumentation is sooner innde operable.
In the event that all indientions of tliese parameters [...either drywell-torus dif f erential pressure or torus water icvel) in finnbled and nuch Indication cannot be restored in ntx (6).
i hourn, an orderly shutdown shall be initiated and the reactor ahnl1 be in a cold shutdown condition in twenty four hours.
T 4
i a
2 i-45 l
Amendment No. 58 1
6 a
ne
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1 DPR-19 Bases:
Table 3.2.1 which senses the conditions for which isolation I,s required. Such instrtunentation nust be 3.2 In addition to reactor protection instrumentation which initiates a reactor scram, protective instru-
"?ver primary containment integrity i
is required. 'Ihe objective is to isolate the primary mentation has been provided which initiates action ntainment so that the guidelines of 10 CFR 100 are to mitigate the consequences of accidents which are not exceeded during an accident.
beyond, the operators ability to control, or termi-d nates operator errors before they result in serious 4
consequences. This set of Specifications provides instrumentation wldch initiates primary systern s connecM,n a dual bus artWment n
i 2
the limiting conditions of operation for the primary us, seussion g m.n in l>ases for Specifi-system isolation function, initiation of the emer-cation 3.1 is applicabic here, gency core cool.mg system, control rod block and standby gas treatment systems. The objectives of
'"***"*"****"*'"""'"*'"'"'""***"'r**"
the specilications art- (i) to assitre the efIectiveness inthee on the level instrwnent (top of active fuel to dettned to of the proleelive itbtttintetitution Wh'en reyttircel hy be 360 inches above vessel serol and af ter alto-Ine for the full preScrying its capability to tolerate a Slugle Iailure P"""
P"'
"'" d rop *er o8 8 the steam drYtr th* lo'* 18 vel t r iP le at 504 inches above veesol pero, or 144 inches above the top of ol any enmponent of such systems evelt during peri-active suet, stetrofit one fuel has an actin fuel length 1.24 ods vihen portions of Sitch sySteins are out of Service Inche. io-,er then e.riier f..i de ir.n. iio o.er.
re.ent tri,
r*'"'****'"***'c^"""'r"'**"'^"r'"-
for tuaintenance, and (ii) to prescribe the triIi set-This trip initiates closure af Croup 2 and 3 primary conte'ent tillgs ret}ttired to assttre adeyttate performance.
1sotation valves but does not trip the recirculation pumps treter-WIlen necessary, one channcI may be made inoper-ena w t section 7.7.2).
For a utp utting of 504 Inchn Won vessel rero (144 inches above top of active fuell and a 60.necomt ahl0 for brief Illtervals to conditCt reqtitred functional es i e cios... t i.e. t he.e i.e s. i t,e oto,ed,e,o, e,cr,or.t.on o, j
tests and Calibrations.
the ctedding occurs even for the enamtmo breake the setting to therefore adequate.
Some of the settings on the instrumentation that The now to-rnetor t ="1 inst r=~atastaa la *** ** t r t P **** "*-
Illitiates or ControIs core atid cotilattifnent Coolitlb' fuel defined as 360 lett.ee above veneel sero. 59 inches t5 P4 have tolerances explicitly stated where the high and inch.s a..e th, toe of.ct ive r.-ti.
low values are both critical and may have a substan-TNs tial cifect on safety. It should be noted that the set-trip initiates closure of Group 1 primary contaltuant points of other instrumentation, where only the higli isolation valves, Ilef. Section 7. .2.2 SAH, anti also or low cud of the setting has a direct I earing on activates the ECC subsystems, starts the emerge? cy J
safety. are chosen at a level away from the normal diescl generator and trips the recirculation pumps.
=
operating range to prevent inadvertent actuation of This trip setting level was chosen to be high enough thr Fafety system involved and exposure to abnormal to prevrut spurious operation but low enough to ini-situations.
tlate CCCS operation and primary system isolation l
so that no molting of the fuel cimiding will occur and l
Isolation valves are installed in those lines that so that post accident cooling can be accomplished penetrate the primary containment and mttst be and thc* guidelines of 10 CFit 100 will not be violated.
isolated during a loss of coolant accident so that the For the complete circumferential break of a 2t!-inch radiation dose limits are not execciled during an recirculation litie aml with the trip setting given I
i accident condition. Actuation of these valves is chove, ECCS Initiation and primary syr'em isolation initiated by protective instrument:ition shown in are initiated in time to meet the above criteria.
l l
[
Amendment No. 58 46 i
~,.. _, - -
n
DPR-19 k
Temperature monitoring Instrumentation is The Instrumen-provided in the main stcan:line tennel to'dctect tatin also covers the full ram;e or spectrum of leaks In this area. Trips are provided on this in-
~
brea;:s and meets the above criteria.
strumentation and when exceeded cause closure of I
Group 1 Isolation valves. Its setting of 200*F ls low enough to detect leaks of the order of 5 to 10
'll e hiit drywell presrure Instrumentation is a, back.
gpm; thus, it is capable of covering the entire t p to t.he vater level instrutnentation atni in athlition spectrum of breaks. For large breaks, it is a to initia:ing ECCS It causes isolatica of Group 2 Iso-back-up to high steam flow instrumentation tils-lation valves. For the breaks dit eussed abo >c, this cusser! above, and for small breal;s with the result-Inritumentatioa will initiate ECCS operation at about the ra:re titre as the lovi low watcc level Instrumen
~ 1;ives isolation ant small release of ratlicactivity, taha; t' es the result, given abovt are applicable before the guidelines of 10 GFil 100 are exceeded, a
baro ~t ro.
Greep 2 Isolatica valv; s luclede tbc dry. cil vent, purge, an 1 sump Isciation valves.
Illgh radiation momtors in the main steamline iga drywell pressure activates only there valves tunnel have been provided to detect gross fuel failure
- t 'rmse h,gh <lrywell le essere cochi occur as the This instrumentation causes closure of Group 1 i
r c: att of non-safety related cau:;c:. noch as not valves, the only valves required to close for this I u : ia.: the drywell air duringytartrp. rotal sys-accident.' With the established setting of 3 times t -ra colatma is not ih rirable mr t_itse conditions normal background, and main steamline isolation cir! only the valves in (,roup 2 atlc required to Valve closure, fission product release is limited so thee. i he low it.w - ater level m*.trumentatien that 10 Cril 100 guidelines are not exceeded for this accident. Itef. See, tion 14.2.1.7 SAll. The er-im:iates prdection for the full r.lp.etrum of loss of co da: accidents and caure:. a trip nf Group I pri-f."
oIt1i ptmss rathau n m nitoring system ma r. s'n te:o isolation valve ~,-
relative to de:ccting fuel leahage shall be evaluated during the fit st five years of operation. The conclu-
,,,,,,ris are provided in the ma!n steamlines as a
- ns of this evaluation will be reported to the mer as o. n.casaring n;"am floa aint also limiting Atomic l-inergy Commission, 11 e loss oI mass inventory from tir versel during a
- teandire in e k acci&nt. In a
- hiitloa to moni-Pressure instrumentation is provided which trips teri.;;
- tea:n Iknv. in:-trumentatina is provideel when main steamline pressure drops below 350 psig.
which causes a trip of Gt oup 1 isolation valves.
A trip of this instrumentation results in closure cI TI-e primary function of the instrinerntatiim is to Group 1 isolation valves. In the "Itefuel" and detect a break in 1:.c main sica:nline, thus on!y "Startup/Ilot Standby" mode this trip function is by-Grrrp 1 valves are closed. For the worr.t case passed. This function is provhled primarily to pro-
- s t -ide t", main steamline break outside the alrywell, vide protection against a pressure regulator
- his trip setting of t.'iih of rated steam flo v in con-malfunction which wouhl cause the control and/or pmetion v.ith ti e flow limiters and inain steamline bypass valves to t pen. With the trip set at 350 psig valse cim.;ure, limit the mass inventory loss stich inventory loss in limited to that fuel is not uncovered that fuel is nut unemered. !act temi eratures re-aint peak clad temperatures are much less' than
- r. min less than 1500T aml release of radir. activity 1500'F: thus, there are no fission products available to the environs is well below 10 CI'It 100 guidelines.
for reb.3. r.tl.cr than those iii the t eactor water.
!! ! Sectis.n s 14. 2. ") tl and 14. 2. *t. I o Tali, llei Sect ie,n I l. 2. ;; Salt.
Amendment No. 58
?.co.,ennors on the isolation condenaer supply nnd may be reduce'd by one for n short period of tir-i to return lines are provided to detect the failure of allow for maintenance, teattne, nf calibrntion.
i.al.. tion condenser line and actuate i.viation action.
This ti:nc period is only -37. of the operatin. ti=c The sensors on the supply and return sides are in a month and does not significantly increase the e.rrany.ed in a 1 out of 2 logic and, to racet the risk of preventing on inadvertent control rod with-sinr,le f ailure criteria, all uennors and lustru:nen-drawal.
tation are required to be operable. The trip settings of 20 psin and 32" of water and valve closure t itre The APRM rod block function is flow biased and at e such as to prevent uncovering the core or ex-ptevents a significant reduction in MCPR expecially cteding site l isai t s. The sensors will actuate due.
during operation at reduced flow.
The APIN provides te high flow in either direction.
gross cose proteccion; i.e., Jimit. the gross dithdr&wal of control rods in the nor-nal withdrawal sequence.
The IpCi high flow and temperature instrumentation 1
are provided to detect a break in the lipCl piping.
Trippinp. of this inst ru:rentation results in actuation In the refuel and startup/ hot standb'f rtodes, of i!PCL isolation valves.
i.e., Group 4 valves.
Tr ip;iinn lonic f or this f unction is the sar-e as that the APRM red block function is set at 12a of 6,r the isolation condenser and thus all sensors
,d rated po.wer.
This control rod block provides required to be operable to treet thb single tail-the same type of protection in the Refuci and'Stcrtup/
are ure criteria. The trip settings of 200*F and 300%
llo t Standby mode as the ArnM flow biased rod block of design flow and valve closute time are such that does in the run mode; 1,c.,
core uncovery is preventel and fission product relcar.c in viehin t irsits.
prevents control rod withdrawal before a scra:s is reached.
W in:.t rvment stion which initiates ECCS action is arranned in a dual bus system.
As for other vital The RNI rod block function provider. local protection
- astrumentation arranged in this fashion the Speci~
of the core, i.e.,
the prevention of transition fication p res e rve s the effectiveness of the system boiling in a local region of the core, for a single even durirn period; when maintenance or testing rod withdrawal error from a limiting control rod D being perforned.
j pattern. The trip point is fica biased. The worst cane single control rod withdrawal error is anafy~ zed the spectrte trip cestiass'.
Tn.' cont r,1 rod block f unctionn arc.provided to for each reloaa to assure that with
- 'r+ vent exc e. live contro) tod ulthdrawal so that rod withdrawai is blocked bcInte the PCPR reaches the HCPR f asel I
- l
- d " "9 ' " * "' i t 7 ' " Y l'"'t-KPR does not Co below the iiCPit fuel clud-eing integrity safety limit.
The trip nelow 30 percent power, the.orst case witharawat at a str.ste
" "All ",t 'l **** t "CP"
- "*l
' d "ith **
d bl ** 'cti l o<
- ic sur this fuaction is 1 out of n; e.o.,
fuel cladding integrity safety limit.
.hus the RBM rod biv:a function is not requirce below this power level.
_i ny ttip on one of the six nPRM's, 8 IRM's, or j
4 SPai's will result in a rod block.
The
)
ininisaum instrument channel requirements i
ossute su2ficient instrumentation to assure the single f a ilure c r ite r ic a re ine t.
The min imura instrument channel requirements for the
.f. 0 R BI.1 Amendnent No. 58
DPR-19 Tc I!i'i rod block runction provides local nL' scitings given in the specificatica arc ade.ite to we ocalina assure l.ac acove criteria are Inc.
Ret. Ecc'.ica-s-
Il ca oso core protection.
is occh that trip setting is less G. 2. G. 3 SAR.
Tne specif..ica'.icn preserves 1.tc crranga.: at than a fcc:or o,10 above the infliceted level.
cifcetiveness of the system dur. "c acr. ds of m:fa tn io f,-17aio of :t c vorct c..ne accident results tenance. testin ;, or cah..orat. ton, and a.tso mini-
- n red bicek cction before I;CPR approachc3 rnizes.J.c ris,.s. of. av.tcrtent operation; f.c., only in ene instrument channel out of service.
the MCPit fuel cladding integrity safety limit.
A devinscale indication on an APRM or IRM is an l- ;!cetion the instru:nent has failed or the instru-Two air ejector off-gas monitors are provided and '
- h. In either case the n
- :nt is net Ednritive enon;!! to chan ;cs in control when the!r trip point is reached, cause an isolation I the air c,iccior eff-gas line. Isolation is initiated instramant veill r.ct res;m red motirm an i thus control ro i motion is prevent 5d.
when both instruments.rcach their hi';h trip point Ti:e dovenscale trips are ret at 5/123 of full scale.
or one has an upscale trip and the other a. dor.n-
lhere is a fincen minute delay beford.
scale trip.
the air ejector off-t;as isolation valve is c!csed.
4 This o. ia.,. is accounted for,.re t.ne ~,0-minute c
- The rod block which occurs when the IRM detectors are not fully inserted in the holdup tiEc of the cif-gas before it is retcased to the stack.
core for the refuel and startup/ hot standby position of the mode sveitch has been provided to assure that these Both instruments are required for trio but the in the core during reactor instrurnents are so designed that any instrument startup.
This, therefore, assures that failure gives a dovinscate trip. Tne' trip settings cetectors are of the instruments arc set so that the instantanc-these instru:nents arc in proper position aus stack reicase rate limit given in Specification to provide protection during reactor t r,13 rot execeded.
i' startup.
The 1RI-l's pri:narily provide ur radiation monitors are provided whic.
nrotection against local recctivity a
ro 5
cffects 2.n t.ne source and 3.nterraediate initiate isolation of the reactor building and I
operation of the standby gas treatment sys*cm.
neutron range.
The :non!! ors are located in the reacter bcilding j
ventilation duct and on the refueling ficor.
The-For clfective emergency core cooling for smallpipe trip logic is a 1 out of 2 for cach set and each 4
breaks. the liPCI system must function since reac-set can initiate a trip independent of the'other tor pressure uces not decrease rapidly enou;;h to a
set.. Any upscale trip will cause the desired allove either cere spray or LPCI to operate in time.
action. -Trip settin; s of 11 mr/hr for the The aute:natic pressure relief function is provided monitors in the ventilation duct are based upon as a back-u;) to the liPCI in the event the IIPCI does initiating normal ventilation islation nnd standby, The arrangement of the tripping non-
- as treatment systera operation to limit the da
- .e not oper ste.
tacts is r.uch as to proviile this function when nec-I The trio essary and minimia'e ::purious operation.
49 Amendment No. 58
DpR-19 3.3 LIMITING CONDITION FOR OPERATION' 4.3 SURVEILLANCE REQUIREIENTS 3.
(a)
Control rod withdrawal sequences shall be 3.
(n.) To consider the rod worth tiinicizer established so that anximum reacti.vity that operable, the following steps crust be could be added by dropout of any ' increment perforned:
of any ore control blade would be such that the rod droti accident design limit (i)
The control rod withdrawal sequence of 280 cal /<pn is not exceeded.
for the rod vorth minimizer cor.puter shall be verified as correct.
2 (1 ) 1?henever the reactor is in the startup or (ii) The rod worth minimizer computer 3
inade below 20% rated thereni pc,wer, n-line diagnosite test shall be run the Rod 1 orth rtintminer shall be operable.
successfully cortpleted.
A second olierator or quallfled technical person may be used as a substitute for an (iii) Proper annunciation of the select i
inoperable P.od 1:arth !!!nimizer which fails error of at 1 cast one out-of-sequence after withdraual of at Icast 12 control rods control rod in cach fully inserted 3
to the fully withdrawn position.
The Rod group shall be verified.
I? orth !!1nimizer may also be bypassed for low I
power physica testing to demonstrate the (iv) The rod block function of the rod chutdown margin requirements of specifications worth minimizer shall be verified-3.3. A.1 if a nuclear engineer is present and by attempting to withdraw an out-verifica the ntep-by-step rod movements of of-sequence control rod beyond the l.
the test procedure.
block point.
a (b)
If the rod worth minizer is inoperable while the reactor is in the startup or run code below 20% rated thercal power and a second independent operato'r or 1
engineer is being used, he shall verify that all rod positions are correct prior to commencing withdrawal of each rod group.
57 Amendnent No. 53
f DPR-19
~
i LIMITING CONDITIONS IDR OPERATION 4.'3 SURVEILLANCE REQUIRDINTS
.~
4.
Centrol rod 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 Ic.'.st tva source range channels have an observed source range channels have been observed count rate equal to or greater than three count rate of at least three counts per l
second.
counts per second.
5.
During operating uith limiting control rod 5.
When a. limiting control rod pattern exists, an instrument functional test of the R:UI pa tterns, as deteralned-by the nuclear shall be perforsted prior to withdrawal _of ce;; inter, either:
the designated rod (s) and daily thereafter..
Both RBM channels shall be operabic; or a.
b.
Centrol rod withdrawal shall be blocked; or c.
The opcInting power Icvel chall bo i
p lialted co that the I;CFR ulll l
terrtin above the MCPR fuel cladding
[
integrity safety limit assuming a single error that results in complete withdrawal of any single operable control rod.
i
+
1 Amendment No. 58 1
57A wi
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m
o J
J DPR-19 3.3 L1311 TING CONDITION FOlt OPEllATION 4.3 Sl!ItVEll.l.ANCE llEQUlitEllENT E.
Itcactivity Anomatics -
E.
Iteactivity Anomalies l
The reactivity equivalent of the difference During the startup test p ogram and startups betwcca the actual critical rod configuration following refueling outages, the critical rod configurations will lic compared to the expected and the expected contiguration during power configurations at selected operating conditions.
operation shall not exceed 1C a1C. If this hmit is execeded, the re.netor will be. shut-These comparisons will be used as base data down until the cause has been determined and for reactivity monitoring during subscriuent corrective actions have been taken if such pow er operation throughout the fuel cycle. At i
actmns are appropriate. In accordance with specific power operating conditions, the Specification G.6, the URC shall be notified criticai rod configuration will be compared l
of this reportable occurrence within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
to the configuration expected based upon appropriately corrected past data. This com-F.
If Specifications 3.3. A through D above are not parison will be made at least every equivalent met, an orderly shutdow n shall be initiated *and full pow er month.
4 i
the reactor shall be in the Cold Shutdown con-dition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Economic Generat.on Control System G.
Economic Generation Control System l
i G.
' Operation of' the unit with the Economic Weekly, the range set into the Economic Generation Control system with auto-Generation Control System shall be recorded.
matic flow control shall be permissible only in the range of 65-100-4 of rated core flow, with reactor power above 20%.
u 60 Amendment No. 58 J
d
~
DPR-l"'
small runount.of rod withdrawal, which is less inJic.stive of a generic cont.ol 'rnd drive' than a norr al single withdrawai incrc nent, will probles and th: reactor will be shutdowis.
l nat contribute to any da.:;c to the pri c.ary Aeso if da: cage within the control red doive coolant syste:n. The design basis is given in recMniso and in particular, crncks in drive Section 6.6.1 of the SA:1, and the desisn evalua.
i.,cern:1 housings, cannot be ruled out, then a tion is given in Section 6.6.3.
This support generic nroblem affecting a nu.Scr of dr:ves is not required if the reactor coolant syste.
c:ianot b'c ruled out.
Circumferential cracks e
is at at:r.ospheric pressurc ince there would resulting from stress assistod intergranular then be no driving force to rapidly eject a ccrresicci have cccur. red in the collet housing drive housing. Additionally, the support is of drives at several IKts. Ris typc of not required if all control rods are fully craciing cculd occur in a number of drives in>erted and if an adequate shutdcwn r:argin
- 3 if the crecis pr:pagated until severance with one control rod withdrawn has been deson-of the collet hcusing cecurrcd, scra:n could '
strated since the reactor uould renain subcritical be prevented in the effectcl rods. Li:niting even in the event of cumplete ejection of the the pericd of operation with a pe cntially strongest control rod.
e scured collet housing and requir' ng increased sur :illance a f ter detecting or.., stuck 3.
Control rod withdrawal and insertion sqaences-are red will assure that the reacter will not established to assure that the naximu:n insequence La cperated with a large nu.:.ber of rods with individual control rod or control rod stI ents f;.iicd collet hcucings.
1 s.hich nre withdrawn could not be uarth enough to cause the rod drop accident' design limit of i
E.
Centrcl Eol Withdrcwal i
280 cal / gram to be exceeded if they were to drap cat of the ccre 1.
Cor.t:ul red drcpout accidents as discussed in the r.c.nner defined for the Rc3 Dr:,p Accident.N in Pe fe rence 6 car. Icad to significant coro u cse sequences are develcped prior to initial damage.
if coupling integrity is unintained.
cperation of the unit following z.:.y refueling outage ti e possibility cf a rod drepeut acciJcat is '
and the req *iirc-mnr that an optrater folloa these c i t r.ir.a t e d. The overtrevel positica ft:ture
{
sequences is backed up by _the.cneration of the TCt.
p vides : pczitive check as c::1y un:cupied or a secon'd qualified station employee.
dN.es ::y te::ch thic position. Nautron These sequences are developed to limit inst:= ntation r:rrenso to rod r.cvv ent reactivity worths of control rods and prcvides a verifit:tien th:.t the ro1 is fol' L
together with the integral leving its drive. Absence of such response rod valecity lititers and the a:: tion of the control to drive movement would provide cause for rod daive system, liriit pctential reactivity suspecting a rod to be uncoupled and stuck.
inscrtion such that ti.c results of a control rod Rcstricting recoup 11ng verifications to pover drop accident will not' exceed a naxinum fuel energy, Icvels above 20% provides assurance that a I
contcet of 230 cal /go. The pesk fuel enthalpy cf rod drop during a recoup 11ng verification 250 cal /g:n is belew the er.crgy content at hich would not result in a rod drop at, dent rcpid fuci dispersal cnd pricary system damage have nore severe than analyzed.
been fcund to occur based on experinental data as 2.
The contrel red Ecusing sept.trt rct:ricts is discussed in Reference 1.
th: :utuard marc =>.nt cf a ccr.tt:1 tcd to les: thar. 3 inches in the extre:217 r v..to The snalysis of the centro! red drop accident :::s event cf a housing failure. Tne aracunt or criginally presented in 5::ticr.s 7.9.3,14.2.1.2 reactielty which could ba adl=3 b/ this j
and 14.2.1.4 of the S'afety Analysis P.cpert. I:; rove-j nents in analytics1 capability have allend a mere 62 refine.d ' analysis of the' Ontrol rod drcp accident.
. Amendnent No. 58
~
.. =. -...
w
I DPR-19
~
Bases (cont'd)
The following parameters and worst-case bounding assumptions have been utilized These techniquegre described in a in the reload analysis to determine com-(ISETN In )addIbon, banked Pliance with the 280 cal /gm peak fuel ments.
position withdrawal acquence described enthalpy.
Method and basis for the rod drop accident analyses are documented in
[
in neforence (4) has be'eri developed.
Reference 6.
Each core reload will be to further reduce incremental rod analyzed to show conformance to the worths.
limiting parameters.
i By using the analytical models
,, ' ( An f actor ( 5) inter-assembly. local power peaking described in those reports coupled with consetrative or vorst-cace input parametern, it has been determined b.
The delayed neutron fraction chosen that for pouct levels less than 2&;,.r for,the bounding reactivity curve.
of rated power, the specified limit (typically 1.3Y. 4K) on insequence control rod or control c.
A beginning-of-life Doppler reactivity feed-j rod sennent worths vill limit the yenk back.
fuel enthalpy to Jess than 200 cal /gm.
Above 20% power even ningic operator d.
Scram times slower than the technical crtors cannot result in out-of-sequence Specification rod scram insertion rate g control rod worths which are suf ficient
-(Section 3.3. C.1)
{
to reach a peak fuel enthalpy of 280 c.
The maximum possible rod drop velocity cal /cm should a postulated control rod (3.11 ft.fsec.)
drop accident occur.
f.
The design accident and scram reactivity shape function.
g.
The minimum moderator temperature,to reach Faona., C.J., Stirn, R.C. and Wooley, criticality.
J.A., " Rod Drop Accident Analysis for targe Boiling Water Reactors"*
(4)
C.J. Paone, " Banked Position Withdrawal HEDO-10527, March 1972.
Sequence" Licensing Topical Report (2)Stirn, R.C., Paone, C.J., and Young, NSDO-2123, January 1977.
R.!!., " Rod Drop Accident Antlysis for tat ue m R's",
supplement 1 - KEDO-(5)To include the power Spike effect caused by gapr 10527. July 1972 between fuel pellets.
- Stirn, R.C.,
- Paone, C.J., and Haun (6)Gennaric Reload Fuel Application e
J.M., " Rod Drop Accident Analysis for targe BWR's Addendum No. 2, Exposed.
NEDE-240ll-P-A, August 1978*
j cores", supplement 2-NEDO 10527
- Approved revision numl.er at time 62A January 1973.
- relohd fuel analyses are performed.
l Amendment No. 58
epciafor bith a visual' indication of neutron level. This is needed for knowle'dgeable and DPR-19 c:ticient reactor startup at low neutron level.
Tac consequences of reactivity accidents are rods in other than limiting patterns.
-ene.iuas of the init.7 il neutron flun.
The req ti rc: cr.t of at lecs t 3 cotats per second C,
Scra:n Insertion Ticies cut n s that any traanient, should it occur The control rod cyatch.:is-anely:cd to bring the 1.:.: r. i a:: af or above the initial value of 10 $
i er ra:cd pcuer used in the analyses of transients reactor subcritical at a rate fast enough to he celd conditionn.
One operable SiCI channel Preeent fuel-d azu ge; i.e., t o p reven t the ECTR' eculd be adequate to runitor the cpproach to
.fr a bece:ning less than the MCPR fuel cladding cri t ic;1ity using hc,r.en. ::.:ous pat terns o f.
Integ{1ty s.dety limit.
Analysis of
- a: tere-i con trol rod s:i t h ira tal.
A minimum the limiting power transient shows that ef iee operabic Sint's are provided as an added the negt.tive reactivity rates resulting
. c :n:.c rvat is::i.
fr m the scram with the average response of 11 the drives as givan in the above Specification, provide the're uir'ed a.
an.,, lock Mon., tor (RBM) is designed to cuto-Ero-tection, and MCPR remains greater thaa the i
- .atit.:117 prevent fuel da. age in the event of MCPR fuel cladding integrity safe ty limit
- rod wit,oraval acer. locations of high c erc:.:9m n
density during hir.h power level operation.
i
- r. r.
' a charnel: are prnviled and one ef these may be j
h pr ned f ro t the console for rainte:iance and'/or i
/
L.r t in g.
Tripping of one cf the ch'anncIs uill bloc 1t j
crr:. nous rod withdcreal seen erot:gh to prevent fucl d ra,c.
~iliis systea backs up the cperator who with-j i
Ua::s rods according to a written sequence. The j
i L~:ci.fidd restrictions vitji one channel out of I
- er.Ir conservatively ansure that fuel danage 1
rill net occur due to red vidiuraual errors uhen
~
I
..ir cer:Jirien enists.
.mn&: cuts 17/1S and 19/20 i
- rarcr t.a
- results of an ev.,luation o, a rod block, the minin:um a:9unt of reactivity to be m
inset.ted c,uling a scra:2 is controlled bY
. rx :titor i n iurc.
,Thene 2ren_
- ts shou.t.nat during
- r *cter operation with certain ltm1 tin" control Pc, ot t tin g. no ec,rc t han 1n.,
of th'c operable tods t'o.have.icng scram titnes.
In the
)
)
rc:. -atterr:, t c wi thdrz:va.i c.y a cesignateo. sinale ana1.. ca_i t reat ren t of the t ransients 290 u
D 2
1-i i
o.
-a l rn. coule,. re :ul t,in on or nare fuel. rods
~
\\
..t.
a us us ua:- u e n:2:. ra. : cia.: tim
_a 3;tse m e.2 are alleved between a neut,roa
{
tit i s.t e ritr s dety linit. 1:ur ir, use of s uch patterni, it is juon ::t tant testing of the RB.'l sensor rcaching thC s c ra*2 point and the priar to v1,, draual of such rods to assure s t.arr of motion -o.c t.ne centrol rods.
This
, e s t e_ r:
a rdequate and cd: scrvativc vhen cocparcd 15 tn o; t r u, lity will ass :re t,nat improper with-n o
, to -ac typically observad ti:n c.olov of
..:: n a.
- t. es not occur.
It is the renponsib111ty e.
about 210 milliseconds.
~
- J.the ::ccicar Engiacer to identi.fy these limiting ret a res ael the deninnated-re?s either when the
)
[
]
rn._cr,s cre initially estnblid.ed'or as they Amendment No. 58' l
develop due to the occurrence ef' inoperable control f
63
~. -
DPR-19 l
Approximately 90 millise conds a f ter neutron flux reaches the trip point, the pilot scram valve solenoid de-engerizes and 120 milliseconds later the control rod motion is estimated to actually begin.
However 200 milliseconds rather than 120 milliseconds is conservatively assumed for this time interval in the transient analyses, and is also included in the allowable scram insertion times specifled in Specification 3.3.C.
the scram times for all control rods will be lhe history of drive performance accumulated det ermined at the time of each refueling to date indicates that the 90*4 insertion times outage.
of new and overhauled drives approximate a normal distribution abcut the mean which Fi f ty percent of the control rods will be tends to become skewed toward longer scram checked every 16 weeks to verify the per-times as operating time is accumulated.
formance.
'the probability of a drive not exceeding the mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution.
The measurement of the scram performance of the drives surrounding a drive exceeding the expected range of Amendnent No. 58 5
+
4.-
k DPR-19 Components of the systen are checked periodically 2
as described above and make a' functional test of-nares:
t.hc entire system on a frequency of less than once -
A test during each operating cycle unnecessary.
A.
The design objective of the standby liquid of one installed cr. plosive charge is nade at least l
contrcl system is to provide the capabihty of once during each operating cycle to assure that.the~
f ijrin;;ing 1:ic reactor from full power to a'colti,*
charges have not deteriorated, the actuation circuit mron-free shutdown asstuning that none of the is functioning properly, the valve functions prcperly, 1
with>lrawn control rods can be Inserteti. T5 meet this objective, the liquiti control r.yntem isnd no flow blochanca exist. The replacement chart;c l
is desigued to inject a quantit of boren which will be.nelected from a. batch for which there has'been.
l proth res a concentralien of no less than a successful test Itrinn.
Recomncndations of the.:
j 4
600 ppm of bciron' vendor shall be fo11 cued in maintaining a five-yeap j-in the reac!cr core in less than 100 mi:mtes.
life of the explosive charges.. A continual check Af 600 Finn boron concentratitm in tbc reactor the firing circuit continuity is provided by pilot j
cere is requited to bring the reactor from full lights in the. control room.
mwer to a 3%ok or more s'ubcritical j
condition considering The reller valves in the standby liquid control j
the :uat to colil reactivity sa.ing, xenen system protect the system piping and positive
]
i poistening and an attditional ma,rgin (2Ts4) for displacement pumps which are nominally possi >!c imr.crfect mixing of the chem,ical destfinctLfor 1r:00 psig protection from over-pressure. The pressure relief valves discharge i
sclation in the reactor water. n nummmn l
quaatity of 317'l n.tllons of solutinn having a ljack to the standby liquid' control solution tank.
- 13..lk-so.lium pentaborate concentration is Only one of the two standby liquid 'ontrol j
requireil to meet this shuttlown requirement.
B.
c pumptr.g circuits is needed for proper ope..i-The time rc<tuirement (100 minutes) for inscr-tion of the system, if one punming circuit is l
tion of the boren solution was selected 1:3over-found to be Inoperable, there is no immediate l
ride the. rate of rcactiv!ty insertion chie to threat to shutdown capability, and reacter oper-co:hkw.n of the reactor following the xenon ntion may continue while repairs are being i
j poisen peak. For a required pumping rate of made. Assurance that the remaining system 33 -allcas per minute, the maximum starage will perform its intended function and that the t
volume of the boton solution is established reliability of the system is good is obtained by i
as 1,0T,Nr.allons (IT,3 gallons are containett dcmonstrating operation of the pump in the i
below thq pump suction anil, therciore, cannot operable circuit at least once daily.
be inserted).
C.
The solution saturation temperature of 13'1 Enron concentration, solution temperature, and sodlum pentaborate, by weight, is 59'F. To
~
velume are ci:ceked on a frequency to assure a guarti against Imron precipitation, the solution high reliability of operatloa of the syste:n including that in the pump suction piping is shoubt it ever be required. Experience with kept at least IT ob9ve the saturation tee:per-pump operability in licates th9 monthly testing ature by n tank heater and by heat tracing in is adeintate to detect if failures have occurred, the punk Isuction piping. The 10*F margin is y
Amendment No. 58
^
n.
DpR-19
- l 3.5 LkIt1 TING CONDITION FOR OPEPATION l
4.5 SURVELLLANCE REQUIREMENTS' and control rod drive maintenance performed provided that the spent fuel pool gates are open, the fuci pool water icvel is maintained above the low level alarm point, and the minicum total condensate storage reserve is maintained at 230,000 gallons, and provided that not more than one control rod drive housing is open at one time, the control rod drive housing is blanked following removal of the control rod drive, no work is being performed in the reactor vessel while the housing is open and a special flange is available which can be used to blank on open housing in the event of a leak.
5.
When irradiated fuel is in the reactor and the vcusel head is removed, work i
that has the potential for draining the vessel may be performed with ices than 112,000 ft3 of water in the suppression pool, provided that:
- 1) the total volume of water in the suppression pool, dryer separator above i
the shictd blocks,' refueling cavity, and the fuel storage pool above the i
bottom of the fuel pool gate is greater than 112,000 ft3; 2) the fuel storage pool gate is removed; 3) the low pressure coolant injection and core spray systems are operable; and 4) the automatic mode of the drywell sump pumps is disabled.
'H.
Maintenance of Filled Discharge Pipe H.
Maintenance of Filled Discharge Pipe Whenever core spray, LpCI, or HPCI ECCS The following surveillance requirements are required to be operable, the discharge shall be adhered to, to assure that the piping from the pump discharge of these
. discharge piping of the core spray, LPC1, systems to the last check valve shall be and HPCI are filled:
filled.
Amendnent No. 58 -
a-- a-one nn. 20 dated 6/9/76, 81.
~..
.'3 4.5. SURVEILLANCE REQUIREMZhr
- i. 5 LTi1ITING CONDITION FOR OPERATION I.
Mer7Qo Planar Linonr IIcat Generntion Avornqc Planar UIGR Rate ( APLIIGR)_
D.: ring cteady state power oporation, the The APU1GR for each type of 1\\sel as a Aserage Planar 1.inear ticat caner ation pate t r, Puton) function of averngo pinnar exposure shall of all the rods in cny fuel anne::bly, as t
at be determined dal.ly during reactor opera-o function of avorage planar exponure, any axial location, chall not exceed the tion at p_ 25?.' rated thermal powor.
anx 6ura averago planar LnGR nhown in rimire 3. 5.1.
If at any timo during crernt.ien it in determined by normal our-vc111r.nce that tho limiting value for A77 :IJR is being exceeded, action shall bo initinted within 15 minutoo to restoro operation to within the prescribed limito.
IE the APLliGR in not returned to within the proscribed limito uithin two (2) heuro, the reactor ohnll be brought to the Cold Shutdown condition within 3G houro.
Surveillanco and correnponding l
netion chall continuo until reactor opera-tica 10 ithin the prescribed limito.
Amendnent No. 58 OlB O
DFR-19 1.
3.5 LIMITI;'G CONDITION FOR OPERATION 4.5 SURVEILIANCS REQUIREMENT J
Linenr Hent Generation Rato (LEGR)
LCcal LHGR The LHGR as a function of coro height shall 4
During steady state power operation, the be checked daily during reactor operatica linear heat generation rate (L11GR) of nny rod in any fuel assembly,at any axial at > 257. rated thermal power.
locatiog shall not exceed the maximum nllowable LilGR as calculated by the folicuing equation:
fan f Lh (T/ max [\\l7f)-
LilGR LilGR d
d,1 I max u
LilGR Design LIIGR = 17.5 '10t/ft. 7x7 d
=
4 fuel I
r 13.4 kw/ft for all 8x8 fuel tvoes
=
f
\\ r.nx = Maximum power spiking penalty =
gj AP/P t
0.037 for 7x7 fuel and 0.0
(
j for 8x0 fuel LT = Total core length a 12 ft.
A::ial position abovo bottom Of.
l L
a CorO If nt n.ny time during operation, it la d<:termined by normal curvoillanco that the liraiting vnluo for L1:GR in being excended, nction ohnll bo initiated within 15 minutes to restore operation to within the prescribed lir.i t s.
I f the IJiGR in not roturned to with-
~
.in the prescribed limi.ts within two (2) houro the reactor shall bo brought to the Cold Shut down condition within 3G hourn.
Surveillance and corresponding action shall continue un-til reacter operation 10 within the pro-ceribcd lisalto.
[
gyg,y I
Amendment NO. 58 4
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istM cmA;m MTI WPtilM) vs
- p. /w An,u,.t,Dr 05tn.E hnendment No'. 58 p.,
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DPR-19 3.5 LIP.ITIt;G cow ITION roa OPr;tt. TION 4.5 SunvrILI.h5CC REQUIREME:GS K.
nini: un critic::1 Pctier ftetio (t:c?a)
K.
rtininen critieel Pc.er natio (n0P*t) 4 During steady state operation, V.CPR shall bo' NCPR thall b= deteratined daily during a greater than or equal to -
reactor power cperation at 2, 25% rated ther: sci power and Collo. ting any change in -
t! nit 2 pos= = lev =L or distribution that would 1.24 (7 x 7 fuel) cauce operation wi.th a li:nitir.g control.
rod pattern as de:cribed in the bases for I1.31 (8 x 8/8x8H/p8x8R fuel)
Specification 3.3.D.5.
et rated pe,cr and ficw.
ror core ficws other th:n rated, these nc=inal values of I;cPa shall be increased by a factor of Kr, where Er is as shown jn rigure 3.5-2.
1 I C at any tir.o durinJ steady state power o.r.c :c t ion, it is detetuined that the liraiting value for I;cPa is beir:9 cxeceded, netion shall be initinted,sithin 15 :rinutes to restore operation to uithin t.hc pre-1.cribed li:aits.
If the stcody state *'.CPR is not returr.cd to within tl:c. pec: cribed 3iniits within t'a'o
( 2 ) hours, the rcOctor
- bali Le brought to the Cold Shutdo sa con-d ition within 3G hours.
Surveillance end corresponding action shall continue until i
rocctor op ration is within the. prescribed limits.
81" Amendment No. 58 e
e M
e
- =i 4
m-
DIR-19 l_5 Liraiting Conditions for Operation Dases developed in th,s refer-i A.
Core Spray and LPCI Mode of the RHR ence, the repair period is found to be less than 1/2 the test Interval. This assumes that the system - This specification assures core spray and LPCI subsystems constitu:e a that adequate emergency cooling 1 out of 3 system, hov.ever, the combined ef-ca pability is available.
feet of the two systems to limit excessive clad
, temperatures mtist also l'c considered. The pased on the loss of coolant analyses test in1erval specified in Frecification 4.:. was 3 m - r.,. c, 3.
. hr re fo re, a n a s !.,n ;:.,:. trpa r t' included in References (1) and in accordance uith 10CFR50.46 and Appen-iod d % m ie w & M s m g ing single failures should be less than 4.~. days dix L core cooling systems provide and this specifica: ion is v.lih n this period.
sufficient cooling to the core to For multiple failures, a eh.sr:er imerra: is dissipate the energy associated uith specified and to impre.c he assuranec : hat the loss of coolant accident, to limit the remaining systems will f:nction, a <!aily' the calculated peal clad temperature test is called for.
-\\ lta.ws.n it.is reen ni: red that the information given in reference S
-r to less than 2200 F, to assure that vides a quantitative n etha I to estimate all5wo-core geometry remains intact, to limit able repair times, the lack of operating data to the core wide clad metal-water reaction support the analytical appreach pres ents com-to less than I and to limit the cal-prete acceptance of this n:e: hod at this time.
culated local metal-water react ion Therefore, the Unw. Mated in the specific to less than 171 items were established with due recard to judgment.
~
The alleuable repair times are es-Should one core spray subsystem become in-tcblished so that the average risk rate operable, the remaining core spray and the for repair would be no greater than entire LPCI system are available shcu!d the the basic risk rate.
The inethod and.
concept are described in Reference (2) NEDO-205G6, General Electric (3).
Using the results Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFRSO Appendix K.
(1) "1.oss of Coolant Accident Analyses Report (3) ApCD " Guidelines for Determining for Dresden Units 2, 3 and Quad-Cities Safe Test Intervals and Repair Units 1, 2 Nuclear Power St.ations,"
Tiines fcr Engineered Safeguards" a
NEDO-24146A, Revisionl, April 1979.
April 1969, I.M. Jacobs and P.W. Marriott.
02 Amendnent No. 58
~. _.. _. _ _. _ _ _ _
?.5 Li-.' tin-coraition for operation Ilanen (cont'd)
I.
1.verav Phrar IEEll '
Thio c.:ccification anneren that the ponk chilin:; temperature follouing a pcotehted
-coohnt nccidont util docir.n t2aio lonn-o{? lirdt opacific1 in not c>:c:cd tho 2200 1CCFT:SO Appendix K 'cena1Cering tho pontulated effecta of fuel pallot dcncification.
The p nk claddng tonporaturo folloulnh a p:stuhted loss-of-ccolcut. ecc) dent is prl:;;rily a functica of ths avornca IEGR of all the rcdn in a fuol ccccably ct cny cz.ic1 Iccation und la o.aly depen:Icnt cocond-r.rily cn the rol to rcd f:nor dictribution within n fuel ascenbly.
Sinco expaci.ed local variaticna in po::ar dit.tribution uithin n fnol acno.bly affect the calculated Tonk olad tonycr.~.turo by lean than i20 F rolative to C
the T: k tenparatum for a typical fu3l design, the limit on the avoraco plen'r 1)iCR 10 cufficient to assure that calculetcd tonp-cret rca cru bolcu tho iccFitSO, Appendix K The 11r.1L,
~
r.aximum nyerato planar D:GRG plotted in The r1xitun nvoraco plenar li!GRa choun in Flc. 3.5.1 at bicher c::pceure result in a Figura 3.5.1 cro bacca en calculatics; c:tploy-calculated peak chd temperaturo of 1.22 Ing tho noich dcccritcd $n Hoforanco (1).
than 2200 F.
Houcver the nixime.1 aver go Fcuer cparation eith I!!Cits at or t:lcu those planar UICRo aro choun.on Fic. 3.5.1 as choun in Fig. 3.5.1 cnnures th1t the pbak limits becauno confor.anco calcuhtions have cleddin: to:Toratum follouing n postulated nct been perforn:d to justify oparatica at Icsa-cf-ccolant accident vill not exceed tho 111G113 in oxecca of thoac choun.
2200CF 11rit. Thono valuos reprocent linits for cperatica to encuro ccnforcanco uith J.
I.oen1 IJ!GR ICCFRSO crd Appendix K caly if they aro more 11nitin;; then other design pararctors.
Thio opecific'ation annurcs ths.t the n2ximun linear heat genera *. ion rato in any rod io loos than tho decisa Lincur (1) " Loss of Coolant Accident Analyses Report for Dresden-Units 2, 3 and Quad-Cities Units 1, 2 tJuc ear er Stations," NEDO-24146A, Revision l' Amendment No. 58 85A
u
- l.ini t ing ~run ll t i on f or Oper et t an n u e s,,Kon,tf.1).
The most l im i t l eig trentsleints with ronpou
. ?.
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to P1Cl It aren even if f **e l llle' genernity heat genesnelidi sato
.le ns i f i e.it ion i s pos t isl a t e.l.
the power q
g gg g
i spit e eenalt y s u il i ues n :.... I i ii 8'l l
I h. l.. i.. n.... s a nd a ""'" 5
- f'.nad rejcetion or turbine trip without t
Y l'4 " '
linearly nac s bes ing variat iosi in animi c)
I.o a a of feedwater henter g.,ps t.c t u r e n ( " 8 e It t om and top, asul Several factors influence whicle of thersen a s suocs wi th 9% s onfident e, thit no -
nose ti.an one fuel md cced5
'h*
"S En t o gusw e r s p i k a,"g.
in critical power ratto sv.-h an the specific IntT due n
fuel l oail a ng, exposure, and fuct type.
The current cycles reload 11cenoin9 analyr.en slicci fic s,the Ilmiting transient for a given e x po-s u re a ncrerne nt for each fuel type.
The v a lin.
spec i fleal as t.he I.imit ing Conalit ton of K.
3_:i_St i.a_nn_C_'.I '.I ' 'l s' owe r R a t i o,M,9.
Ogn? ration are connorvetively cho icu to gynnny gn iso st re strictive over the entire cycle for
'Ibc steady St*te values f or ItCI'st s peci fic.)
cath fuel type.
in this Sp.tification Oere s e l et t e'I t' pre, vide n;ingin t o at t.inno.la t e t r atis t esit s a.id un(catatulles in anonstoring the core o;,c i a t ing
.t at e e s well n uen et t ninit i es la al c c rit ical power cur s elat ion itself.
n cse values also assure that opesatson for core flow rates less than rated, will be such that the initial. condition the steady state PICrit is increascil lay the I 4:e?:
nss.:,,cd fue the IOCA analysis. pluS M un-fa' asia t ven in the Speci fication. This i
1.s i nt y IS " ' ' I S I I *'t.
Fo* any o
'c SPet assure that the I!CI'R
- ell t be na nn:ained set of t a ans s ent s or tistensbant e c ause 1 by.
greater than that speelfled in Specift.
c'Eui.sen:
. i ng;t e ope s a t or error or sinzie l
cation 1 l A even in the event that the ralfenction, it i s s e'Eul s e.1 that.irstyn eso*or-tener a t or s et speed controller ana;yses initintlac4 at i t. i s s t e.e.ly stata the scoop tube positioner for the causes operating linit yield a tit 3Ts
<>I not less than fluid coupler to novo to the manir.we that spcca fled in Specificat ion 1.1. A at any speed position.
t s ne eher asig the t ransient a s sins i n g Instavncut I
given in 5.eci ficat ion 2.1.
Ter trip setting 3 3
an.nlysis of the thess21 consealueno:s of these f
- e. nsients. ihe,an.
on wrn c as e s sa ihl= ar-8 8 8"t iaa I",
(2),Gener Ic lteload 1'tiel Applical iols "
I t i.e 1 ::e L t h.e
- ini.la t e.we of ol'en at 8""
3..un ls t i.e anittal value el N13)U 2 401 1 -pa
- p:t g'at.ensom.-J t u e n t st I.s los I., t l.c e n i t s.at i.m..(
t t.c taan.sents.
Tl. i s inis i.nl tua. lit iews which i s ut.c I ih t h.= trae ient an.a t y a.e s.
the futp f ise l, l a*Li t "'I ine. age f ty sald y wiIb i nni. i.s e. l le v io l at non as
. m
.t
., an.: -eih.s=
.-iia.atota'ia
'he ' m '
- e
1 4
s e.. ty r.t s e e tu ru inmit f.s e e B.
- e. loa.s. ys le,se.l.a.
- -s* t e.l 8 85 e
s.
u.-
.co li. ari.l y -ii n n-i ca'= ~e nesn ut i m e,htle
,I3, g t U V g rs k On nillillie l-al { g, gygg. g g,. ), g, gg l -
.......,60.n....,.e...
t h
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Iori.ina 1 ynen,it e Pel formed.
Amendment No. 58 asa m
m.-
.r.
I !,
l DpR-19 Surveillence T'couiree-nte Isaeen (cont'd) i h.5 K.
Minimum Critical Power Ratio (MCPR)
I.
Averere Plansr LHCR -
At core therrol power levela leaa than or At core thermal power icvels less than or equal to 25 per cent, the reactor will be operatine, equal to 25 per cent, operating plant et ricicun recirculation pw.p speed and the c:ctnrience and therr.al hydraulic onalyses moderator void content will be very small. For
~
I in.11ca te that the resultirg overage planni.
all designated control rod patterns which may be, Ln0R is belev the coxirnua average plamr LEGR employed at this point, operating plant experience' by a considerable rargin; therefore, evcluation and thgrmal hydraulic analysis indicates that the of the overage plar.or LUCR below thic power.
resulting MCPR value is in excess of require:nents
~
1ev.1 is not necessary.
'Ihe da ily require-by a considerable cargin. With this low void rent for claculating average planer LUCR content, any inadvertent core flow increase above 25 per cent rated thermal power is would only place operation in a more con-cuf ficient since pcuer distribution shif ta servative mode relative to MCPR.
are slou ul.en there have not been siCnifi-esnt power or control rod changco, lhe daily requirement' for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution l
shifts are very slow when there have not been J.
Local L'ICR significant power or control rod changes.
'n:e LUCR es a function of core height shall be chect;ed da f]y during resetor opera tion at In addition, the K correction applied to g
Greeter than or eccal to 25 par cent power to the LCO provides margin for flow increase from low flows.
deter =ine if fuel burnup or control rod novement l
bas caused chcnces in power distribution.
A limiting LilGR value is precluded by a considerable margin when employing a per missible control rod pattern below 25% tated thermal power.
Amendnent No. 58 ESA
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1.I:111 II:C C0!!DITInti FOlt OPERATION 4.6 SURVEILLANCE REQUIREMENT an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l I
E.
Safety and Relief Valves E.
Safety and Pelief Valves 1.
During reactor power operating conditions A minimum of % of all safety valves shall be and whenever the reactor coolant pressure bench checked or replaced with a bench checked is greater than 90 psig and temperature valve each refueling outage. The popping point
~j greater than 320*F, all-nine of the of the safety valves shall be.-set as follows:
nafety valves shall be operable. The solenoid activated pressure valves shall Number of Valves Se't Point (psid be operable as required by Specification I
1115*-
3.5.D.
2
-1240 2.
If Specification 3.6.E.1 is not met, an 2
1250 orderly shutdown shall be initiated and 2
1260 the reactor coolant pressure and temp-2 1260 erature shall.be less than or equal to 90 psig and less than or equal to 320*F The allowable set point error for each valve within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
is +1%.
All relief valves shall be checked for set pres-sure each refueling outage. The set pressures shall be:
Humber of Valves Set Point (psid 1
1115*
2 Less than or equal to 1130 2
Less than'or-equal to 1135
~
- Target Rock combination safety / relief valve Anendment No ' 58 90-
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