ML19270G420
| ML19270G420 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/24/1979 |
| From: | Vollmer R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19270G421 | List: |
| References | |
| NUDOCS 7906070010 | |
| Download: ML19270G420 (35) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMis3lON 3 'v
$.) M i h WASHINGTON, D. C. 20555 E$[o/
COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESOEN NUCLEAR POWER STATION UNIT NO. 2 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 43 License No. DPR-19 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated January 15, 1979, as supplemented March 2, 1979, April 6, 1979, April 12, 1979 and April 20, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
7906070 0/O
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1
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraphs 3.B and 3.F of Provisional Operating License No. DPR-19 are hereby atended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 43, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
F.
Restrictions Reactor power level shall be limited to maintain pressure margin to the safety valve setpoints during the worst case pressurization transient.
The magnitude of the power limitation. if any, and the point in the cycle at which it shall be applied is specified in the Reload No. 4 licensing submittal for Dresden Unit 2 (NED0-24160). Subsequent operation in the coastdown mode to 70% rated power is permitted based on the Generic Reload Fuel Application (NEDE-24011). Plant operation in the coastdown mode from 70% to 40% rated power shall be limited to the operating plan described in NED0-24034 (Dresden 2 Reload No. 3) using full recirculation flow.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR EGULATORY COMMISSION
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au 9,aw R' chard H. Vollmer, Assistant Director for Systems & Projects Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 24,1979
i ATTACHMENT TO LICENSE AMENDMENT NO. 43 PROVISIONAL OPERATING LICENSE NO. DPR-19 DOCKET NO. 50-237 Revise Appendix A Technical Specifications by removing the following pages and by insertirg the enclosed pages.
The revised pages contain the captioned amendment number and marginal lines indicating the area of change.
REMOVE PAGES INSERT PAGES 5
5 6
6 7
7 8
8 9
9 10 10 11 11 12 12 16 16 18 18 19 19 24 24 38 38 40 40 42 42 46 46 48 48 49 49 57 57 57A 57A 62 62 62A 62A 62B 628 63 63 81C 81C 81C-1 81C-1 81C-2 81C-2 81D 81D 82 82 85A 85A 90 90 125 125
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1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM ',ETTH" 1.1 FUEL CLADDING INTEGRITY 2.1
_ FUEL CLADDING INTEGRITY Applicability Applicability The Limiting Safety System Settings The Safety Limits established to preserve the fuel cladding integrity apply to trip settings of the instruments and devices which are apply to those variables which monitor the fuel thermal behavior.
provided to prevent the fuel cladding integrity Safety Limits from being exceeded.
Objective Objective The objective of the Safety Limits The objective of the Limiting Safety is to establish limits below which System Settings is to define the level the integrity of the fuel cladding of the process variables at which is preserved, automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.-
Specifications Specifications A.
Reactor Pressure 7800 osig and Core A.
Neutron Flux Trip Settings Flow >10% of Rated The limiting safety syst..a trip The existence of a minimum critical settings shall be as specified l
power ratio (MCPR) less than 1.07 below:
shall constitute violation of the fuel cladding integrity safety limit.
5 f
Amendmenc No. %, 43
2.1 LIMITING SAFETY SYSTEM SETTING 1.1 SAFETY LIMIT 1.
APRM Flux Scram Trip Setting (Run Mode)
When the reactor mode switch is in the run position, the APRM flux scram setting shall be:
r S
f=.65W + 55 LTPF
_TPF l
with a maxium setpogntof120%forcore flow equal to 98 x 10 lb/hr and greater, where:
S - setting in percent of rated power W - percent of drive flow required to produce a rated core flow of 98 Mlb/hr.
T'!F - LTPF unless the combination of power and peak LBGR is above the curve in Figure 2.1-2 at which point the actual peaking factor value shall be used.
LTPF : 3.05 (7x7 fuel assemblies) 3.01 (8x8 fuel assemblies) 2.98 (8x8 R fuel assemblies)
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2.
APRM Flux Scram Trip Setting (Refuel or Startup and Hot Standby Mode)
When the reactor mode switch is in the refuel or startup/ hot standby position, the APRM scram shall be set at less than or equal to 15% of rated neutron flux.
6 knendment No. Ji, 43
emus = m -
2.1 LIMITIITG SAFETY SYSTEM SETTING 1.1 SAFETY LIMIT B.
Core Thermal Power Limit (Reactor 3.
IRM Flux Scraru Trip Setting Pressure 6 800 psig)
When the reactor pressure is f=800 psig The IRM flux scram setting shall be or core flow is less than 10% of rated, set at less than or equal to 120/125 the core thermal power shall not exceed of full scale.
25% of rated thermal power.
C.
Power Transient B.
APRM Rod Block Setting
- 1. The neutron flux shall not exceed the The APFM rod block setting shall be:
scram setting established in Specifica-tion 2.1.A for longer than 1.5 seconds r-n as indicated by the process computer.
S 4_.65W + 43 LTPF J
- 2. When the process computer is out of TPF'
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service, this safety limit shall be assumed to be exceeded if the neutron The definitions used above for the APRM flux exceeds the scram setting scram trip apply.
established by Specification 2.1.A and a control rod scram does not occur.
D.
Reactor Water Level (Shutdown Condition)
Whenever the reactor is in the shutdown conca'. ion with irradiated fuel in the reactor vessel, the water level shall not be less than that corresponding to 12 inches above the tcp of the active fuel when it is seated in the core.*
- Top of active fuel is defined to be 1
360" above vessel zero.
(see Bases 3.2) 7 Amendment No. 2T,a3
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OO 20 40 60 80 10 0 CORE THERMAL POWER (PERCENT OF RATED)
Figure 2.1-2 PEAK LHGR VERSUS CORE THERMAL POWER FOR A LIMITING TOTAL PEAKING FACTOR 9
Amendment No. g, 43
1.1 SAFETY LIMIT BASES FUEL CLADDING INTEGRITY _
boiling, (MCPR of 1.0).
These conditions represent a significant departure from the is set such fuel cladding integrity limit The that no calculated fuel damage would occur as condition intended by design for planned a result of an abnormal operational transient.
operation.
Because fuel damage is not directly observable, A.
Reactor Pressure 7 800 psig and Core a step-back approach is used to establish a such that the minimum critical power Flowt710% of Rated Safety Limit ratio (MCPR) is no less than 1.07.
MCPR > 1.07 represents a conservative margin relative to the Onset of transition boiling results in conditions required to maintain fuel cladding a decrease in heat transfer from the integrity.
clad a nd, therefore, elevated clad possibility of clad temperature and thethe existence of critical fuel cladding is one of the physical barriers failure.
- However, The which separate radioactive materials from the power, or boiling transition, is not a The integrity of this cladding barrier directly observable parameter in an environs.
freedom from perfor -
Therefore, the margin is related to its relative operating reactor.
Although some corrosion or use to boiling transition is calculated from tions or cracking.
related cracking may occur during the life of the plant operating parameters such as core fission product migration from this feedwater temperature, source is incrementally cumulative and continuously power, core flow,
- cladding, The margin and core power distribution.
Fuel cladding perforations, however, for each fuel assembly is characterized by measurable. from thermal stresses which occur from the critical power ration (CPR) which is can result reactor operation significantly above design the ratio of the bundle power which would conditions and the protection system safety produce onset of transition boiling diviced fission product migration from by the actual bundle power.
in the The minimum settings.
While cladding perforation is just as measurable as value of this ratio for any bundle that from use related cracking, the thermally core is the minimum critical power ratio caused cladding perforations signal a threshold, is assumed that the plant (MCPR).
It beyond which still greater thermal stresses may cperation is controlled to the nominal cause gross rather than incremental cladding protective setpoints via the instrumented deterioration.
Therefore, the fuel cladding variables.
(Figure 2.1-3).
is defined with margin to the Safety Limit conditions which would produce onset of transition 10 kaendment No. jH', 43
1.1 SAFETY LIMIT BASES If reactor pressure should ever exceed 1400 1.1.A Reactor Pressure 7800 psig and Core psig during normal power operation (the lbmit Flow p 10% of Rated (Cont'd) of applicability of the boiling transition co& relation), it would be assumed that the l
The Safety Limit (MCPR of 1.07) has sufficient fuel cladding integrity Safety Limit has conservatism to assure that in the event of been violated.
an abnormal operational transient initiated from a normal operating condition more than In addition to the boiling transition limit 99.9% of the fuel rods in the core are ex_
(MCPR), operation is constrained to a maximum pected to avoid boiling transition.
The LHGR - 17.5 kw/ft for 7 x 7 fuel and 13.4 kw/ft margin between MCPR of 1.0 (onset of for 8 x 8 and 8 x 8R fuel.
This constraint is transition boiling) and the safety limit, established by specifications 2.1.A.1 and 1
1.07, is derived from a detailed statistical 3.5.J.
Specification 2.1. A.1 established analysis considering all of the uncertainties limiting total peaking factors (LTPF) which in monitoring the core operating state in-constrain LHGR's to the maximum values at cluding uncertalnty in the boiling transition 100% power and established procedures for correlation.
See e.g. Reference (1) adjusting APRM scram settings which maintain equivalent safety margins when the total Because the boiling transition correlation peak factor (TPF) exceeds the LTPF.
is based on a large quantity of full scale Specification 3.5.J established the LHGR data, there is a very high confidence that max. which cannot be exceeded under steady operation of a fuel assembly at the con-power operation.
lditionofMCPR= 1.07 would not produce boiling transition.
However, if boiling transition were to occur, clad perforation would not be expected.
Cladding temperatures would increase to (1) NEDO-20694, " General Electric Boiling approximately 1100 F which is below the Water Reactor Reload No. 3 Licensing perfcration temperature of the cladding Submittal for Dresden Nuclear' Power material.
This has been verified by tests Station Unit 3."
in the General Electric Test Reactor (GETR) where similar fuel operated above the critical heat flux for a significant period 11 of time (30 minutes) without clad perforation.
Noendment No. y(, 43
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2.1.A.
Neutron Flux Trip Setting' 3.1Di Flux Screm Trip Setting (cont ' d')
2.1.3 APRM Rod Block Trip Setting The IHt4 screm trip cettin5 of 120 Reactor power level may be varied by d iv isi ons is active in each rance.of moving control rods or by verying the IRii.
F o r e r.c.T.p l e, if the instru-the recirculation flow rate.
The APRM ment were on range 1, the scram cetting cycte : provides a control red block to would be a 120 dia/icions for that range; prevent rod withdrawal beyond a given 12kewise, if the instrument were on range point at constant recirculation flow 5, the ceram would be 120 divic tonc on rate t.o protect a c. r i nc t the cond : tion-t.h a t range.
Thus, as the 1RM is ranged l
o f a.4CPit lecc than 1.07.
This rod up to accomodete the increace in power gg oleck trip cctting, which ic.'ute-level, the scram trip setting is also 4
- a t.ically varied w ith rec ircula tion ranged up.
'i f loop flow rote, prevents an 2ncrence W
in t h'! reactor power level to exccc-y The moct cie*nificant cources of re'-
-'M nive valucc due to control rod w i '.h -
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"w tivity change during the power increase drawal.
The flow vuriable trip cctting provides cubc tantial m:trgin from fuel cre duc'to control rod withdrawal.
In order to encure that the IRM provided
' ?j d r r.a ge, accum3ng a steady-ctate cpera-cdequatc protection against the cincle t io.1 at the n: io actt'in, ov;r the entire recirca:Utica fiv. rei.ge.
T:d.
rod withdrawal crcor, a range of' rod c;3
.mnrg!n to the Sa fety Litr.it increacer es w j thdru'.,tl accidents v:oc en >1yced.
This o
analycic incleded cterting the cccident L*/
the flow decrca.:ec for the specified et variouc power levels.
The noct sc-
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- t. rip cesting vercus flo. relatienchip; vere case involvec en initial cond ition
'M therefore the vorst cccc F.CPR which in bhich the reactor is,just cuberitical could occur durine; s teady-c ta te opera-b and the I.51 cyctum is not yet on scalc.
tion is at IOC% ot' rated thermal power
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because of the f. pdm rod block trip 6 d,s,,:.. o 0 0 4 cot:ccrva ticm was take-n in thic o;
mhe ac tral power d ia tribution c c at in';.
i c:. lyaic oy aucu:n)::q that the 11M channel in the core is catab1'ched by specified c ocect wo the w.whdrawn roo is byeccced-cont.ro,i red sequenccc and
,c ronitored i
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c on.c 4.nuous ly.ay. hc in-core LP:,...
cyctem.
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.n r e r. c t o r,c scran..:cd and peck power lim *t..rd 7, u. t.h 'Je li Pli.'1 c e ra n 1,r i p c e t t i n g,
o 0: c percent of re t.ed power, thuc m'in-t he I /.G" rod b loc k t'.'1 p c e t c i n. is ad-yc i r.., n g..C F R n o o v e 1. 07.
2ced on the above l
,; r'; t e.,- d o w n w o rt,.
2 t.ac tra x imuli tota 1 a.. ' 3. y c,. u, the,lii4 p ro v i<f oc prot **c tion aga inst p g,., g. g c.
factor exceeda the 11ntiti:ig
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r a a m.lo ' 4 0. L.au rct.a l o f. centrol roda In uetuence t.he T. N. rod ' b l oc. ac re ty ma r.'-in.
e.e proviuca oackup protection for che A l'l { M.
y Amendment No. )(, 43
S.
Turbino Stop Velvo Seren - The turbine stop valvo C.
Reactor Coolant Lov Pressure Initiates Main Stesta closure neran trip enticipates the pressure, Isolation Val.*e Closure - The low pressure isolation neutron fit c and heat fltty. incresso that could at 850 psig was provided to give protection against result frc., rr.pid cIccuro of the turbine stop fast reactor depressurization and the.resulting valv.c.
With a scran trio cctting of 10 rapid cooldown of the vessel. Advantage was taken of the scram f eature which occurs when the snatu g rcept of vnive clocure fron full open, tho steam line isolation valves are closed to provide resultant increaso in sttrface heat flux is l
lini:ca cuch that MCPR rc:tains aboyc 1.07 cven for reactor shutdowr so.that operation at pressures lover than those specified in the thermal hydraulic derl- : the.uorit care transient that escua.es the safety lintit does not occur, although operation turbine bypsnc is cles:d.
lower than 850 psig would not necessarily at a pressure constitute an unsafe condition.
H.
Main Steam Line Isolation Vnive Closure Scram - The low pressure isolation of the main steam lines at 850 psig was provided to give protection against F.
Generator I.ond Rejection Scram - The genera-rapid reactor depress irization and the resulting tor load rejection scram is provided to rapid cooldown of the vessel. Advantage was taken anticipate the rapid increase in pressure of the scram feature which occurs when the train and neutron flux resulting from secam line isolation valves are closed, to provide fast closure of the turbino control valves for reactor shutdown so that high power operation due to a load rejection and subsequent 3cw reactor pressure does not occur, thus providing at failure of the byynns; i.e., it prevents protection for the fuel cladding integrity safety l
E m frca beco dng less than ' 07 for this
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limit. Operation of the reactor at pressures lover trannient. For the load rejec ion fro:::
than 850 psig requires that the reactor mode switch 10% pouer, the LiCR increases to only be in tiie startup. position where protection of the 1C6.S.0 of its rated value uh.ch results feel cladding integrity safety limit is provided by in only a sntil decrecsc in f*OPit.
the 11C4 high neutron flux scram. Thus, the combination of main steam line low pressure isolation and isolation valve closure scrum assures the availability of gh neutron flux scram protection over the entire p
range of applicability of the fuel cladding integrity pQ)
~jD safety limit.
In addition, the isolation valve 7,,
closure scram anticipates the pressure and flux transients which occur during normal or inadvertent A
isolation valve closure. With the scr.'ms set at g;s}dg
@yi 10': valve c1cuore,there is no increase in neutron
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18 Amendment No. Jg, 43
1.2 SAFETY LDlIT 2.2 LilllTING SAFETY SYSTE.T1 SET 11SG
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1.2 REACTOR COOI. ANT SYSTE1!
2.2 REACTOlt COOLANT SYSTE.TI.
Applicability:
. Applicability:
Applies to limits on reactor coolant system Applies to trip settings of the instruments and pressure.
devices which are provided to prevent the reactor system safety limits from being execeded.
Objective:
Objective:
To establish a limit below which the Integrity of To define the level of the process variabics at the reactor coolant system is not threatened due which automatic protective action is initiated to to an overpressure condition.
prevent the safety limits from being execeded.
Specification:
Specification:
The reactor coolant system pressure shall not A.
Reactor Coolant !!igh Pressure Scram shall be execed 1325 psig at any time when irradiated fuel 510G0 psig.
Is present in the reactor vessel.
B.
Primary System Safety Valve Nominal Settings shall be as follows:
1 valve at 1115 psig*
g 2 valves at 1240 psig
[6 2 valves at 1250 psig
- (h; 2 valves at 1260 psig QS;,.
2 valves at 1260 psig l'f
",(
The allowable setpoint error for each 4_
valve shall be 11%.
p..
- Target Rock combination safety / relief Yf valve L
V<
by 19 43 AmendmentNo./,
TA11LE 3.1.1 (cont)
Noter 1
There shall be two operable or tripped trip systems for eact; fanction.
C.
Petm.sinb;c to b pe... witt cont:ol rod block, for reactor protection system reset in refuel and shutdown positions of the reactor mcCe switch.
3 3
Permisiib'c to by pai w hen react 0* pressure ti< s.C0 p.tg.
Permisub;e to bypan w hcn f.r.t a ge !;rb:nc pressec is less than th t w hich corresponds to 47. rated ste am flow.
.i.
IRMi arc b pas.cd *nen APkM's are cr..cale ar.d the reactor :-ode switch is in tr.e r n poststora, 5.
y The de..p. perni: clowre of any o'c val.c wirtcut a. cram being inttutec.
s.
ht.en tr.c sc c:o: i< su5e r:tical r..: h reactor = ater tempe rature is Ic s t un 2100r. only the following trip functrom need to be operable:
.t. Mocc $w itc!. tit Sh tta. O b.
meaal 5stam c.
Higt Flus IRY d.
Sc r m Di.:targe Ve!.mc P :g* Un el S.
Not re q..:sc to be ope r.b;c w her. primary containment integrit) H not rcquired.
So: r::q ;:sc w r.ic perictmq 10. pc.cr p5)ncs te:t..it atmo,prer.
pre.~ c dar.ng or after refueling at pow er levels net to exceed 5 Mk[t).
Ma y be o pa.cd w!,en i ecenary Curia, prctng for containment inerting er dc:nerting.
l '..
f 11 Not req..rsJ to be epc:.b!c w 'sen t'.c reactc preist.rc se.s:1 he o 16 r.ct bclted to the sessel.
I '..
The A FW co. i c.!c tr.p f ar..non 2. a u:cmat:cally by pmed w he n tne reac:ct me%c sw itch is in the refuel a nd startup/ bot 5:an foj pc.t: ens.
15.
The APkM cc...c.'c tr;p f ar.c: on :. aatonuticall b)pa.:--d w "c n :51RM instra re ttat son is operable and not high.
p
- 14. The APPS. 15% scra:- is bypassed in the run Inode.
['[_
If tl.c first co'.umn c nno; bc et !ct cx of the trip r,.ste:m that ? np sys;c n shall be tr pped.
If tk firs: cei mrt canac: ).
n: fcr Doth tr:p.;.:emi. I'.c a ppropa te actio7s !Lited below shall be taken:
'~
A.
Ir iU.tc i; s::sor, e' ope.:Jc im and ccTp;c e in.c:t.an of all opd4 die rodi w ithin foar hours.
S.
e4 c C a c pu. c : :ese; u IP.
a n. c as.d p;n c r.ece, u in tt c 4:.ro.p/ Hot 5t ndby potitaon w ithin e hours.
gr P'
C.
rec.c c :ura.. s
..s 2... 6
-.:n vea mle..:c.2t.c n. a;u...
'c ur..
A n A - : m i'. o s e... ws :s c :.ob r. ';.s :t t'.(re a re Ic, t n. LPe?.' try ts yr lesel or t'ere are les than ' % of the normal complement of dl 1.f9M'. to a n A FC'
-[
g 1 ad, on t'.c w.:ct les e; tr.c. c.s r* t cn is 2 5 04 "
above vessel 0 (See Bases 3.2).
t Tr.ps upon ac:.:;on of e.e f a.: clo.a;e sole noid w h.ch trips t* e t rb.r.c control s alses.
(4 U
'g.,.9 v
24 Change No. M klendment No. 43
~
TABLE:1.2.1 INSTRUMENTATION TilAT INITIATES PRIMARY CONTAINMENT ISOLATION FUNCTIONS Minimum No. of Operable Inst.
Channels per Trip level Setting Action (3)
Instruments Trip System (1)
A
>144"above top of active fuel
- 2 Reactor Iow Water
.) 84"above top of active fuel
- A 2
Reactor Low Low Water A
2 liigh drywell pressure
$2 psig rated (4), (5)
$1207 of rated steam flow B
2 (2) liigh Flow Main Steam line 2 of 4 in each Illgh Temperature Main Steam Line B
(200*F of 4 sets Tunnel 2
liigh Radiation Main Steam Line 5 3 times normal rated power back-B
- round T mnel (6)
B 2
tow Pressure Main Steamline
?850 psig liigh Flow Isolation Condenser Line C
520 psi diff, on steamline side 1
Steamline Side 5 32" water diff. on condensate 1
Condensate Return Side C
..eturn side D
2 liigh Flow IIPCI Steam Line 3150" water 4
liigh Temperature llPCI Steam Line D
s200*F Arca Whenever primary containment integrity is required, there shall be two opemble or tripped trip systems for each function, exce Notes:
1.
steamtene which only need be available in the RUN position.
2.
Per each stesmline.
If the first column cannot be met foe one of the trip systems.that uip system shaft be tripped.above vessel 3.
Action:
- Top of active fuel is defined as 360" in the LOCA Analysis (See Dases 3.2).
38 Char.ge No. 3d Amendment No. 43
T A B LE 3. 2. 2 INSTRUMENTATION THAT INITIATES OR CONTRO1E Tile CORE AND CONTAINMENT COOLING SYSTEMS Min. No. of Operable Inst. Channels per Trip System (1)
Trip Function Trip Level Setting Remarks
+4" 2
Reactor low Low Water g4a ( -0 ") above top of
- 1. In conjunction with low reactor active fuel
- pressure initiates core spray level and LPCI.
- 2. In conjunction with high dry-well pressure 120 sec. time delay, and low pressure core cooling interlock initiates auto blowdown.
- 3. Initiates IIPCI and SBGTS.
- 4. Initiates starting of diesel generators.
- 1. Initiates core spray, LPCI, 2
Ifigh Drywell Pressure s 2 psig flPCI, and SBGTS.
(2), (3)
- 2. In conjunction with low low water level,120 sec. time delay, and low pressure core cooling inter-lock initiates auto blowdown.
- 3. Initiates starting of diesel generators.
1 Reactor Low Pressure 300 psigsps350 psig
- 1. Permissive for opening core spray and LPCI admission valves.
- 2. In conjunction with low low reactor water levelini-ates core spray and LPCI.
Containmmt Spray Prevents inadvertent operation Interlock of containment spray during 1(4) 2/3 Core lleight
?2/3 core height accident conditions.
2(4)
Containment liigh 0.5 psigt pal.5 psig Pressure 1
Timer Auto Blowdown 5120 seconds la conjunction with low-low reactor water level, high dry-well pre saure, and low pressure core cooling inter-lock initiates auto blowdown.
- Top of active fuel is defined as 360" above vessel zero for all water 4o levels used in the LOCA analysis (See Bases 3.2).
Amendment No. 43
INSTRD1INTATION TEAT IEITIATES RCD ELOCK Table 3.2.3
\\
- .Ln t := ::0. o[
I Opsrnble Inst.
Chan.01s Per Trio level f atti_- 7 T r i -) 'i'/ s ten ( l)
Instrur.cnt i
A??J4 upscale (ficw bics) (7) 1
.,. 4 4 (p}
l4 1
A P.'J: upscale (refuel and 5tcrtup/Het l
$12/125 full ecalc Standby redo)
Ii l
'" -~
2 I,F.~O1 dcwnscale (7) 2 3/12I l
Rcd block renitor upreale (flow bics) (7)
C.y.-l+42)
(2) 1 Rod block ;r.onitor do cscale (7)
I 2 5/125 C011 secle I.
/125 full reale
>5 3
I?;I dcwnsecle (3) 3 IR:-1 upscale.
1103/125 full.seclg 3
IrGi detector noti fully inserted in the core 2(5)
SPM detector not in startup position (4')
2 (5) (6)
SPJi unseal 4 TOS counts /see
,:p gh,,,A,,'e.,
k :.w/
6-
"/
42 n)~0 n,,
u,n
'7
- 6 f
/,;>
9/f//
Amendment No. 43
Ihc Lstrumentation whlet hittat, primary system Bases:
isolation is connected in a dual bus arrangement.
Thus, the discussion given h the bases for Specifl-3.2 In addition to reactor protection instrumentation cation 3.1 is applicable here.
which initiates a reactor scram, protective instru-
- 1" ct*' *"" 1*1 1""mant o s n 1..a l
mentation has been provided which initiates action to mitigate the consequences of accidents which are to trip when r..etor veter te nt 1 504 * (t* on the beyond the operators ability to control, or termi-tusru nt). son.....t n ro, 504. ben vu t nates operator errors before they result in serious
..ro corr..pona. to 144-abon the top or setin
'"1
' ' 'YP** 7 "7 7 " in ena e maa rot.
conse quen ce s.
This set of Specifications provides How var.
inc. Z a.t typ. 3 x 8 Das ( 8 x 8 retrortt) the limiting conditions of operat. ion for the primary hn an.ctin eat 1-ngth which i 1.24 tonger than 7 = 7. 7 = n na e a so.
.na th. acts.1 wet.c system isolation function. initiation of the emer-1.nte e r m.cm and wcs instietton in the toca gency core cooling system, control rod block and an.ty t. r.m tn.4 nrhangea, th top or active to.1 Standby gas treatment Systems. The objectives of ror att o chate=t sp.e tetc t ton na tie.n tng
- ty t. purpo... 1 4.rta.4 36o adow. v..
1 the specifications are (i) to assure the effectiveness of the proteetise insiru-entation when required by
,'*'l;, $ *[,st, ['**i, n,s'"** * **'** *'l"'.**
,, t,, u i t,
preserving its upability tu tolerate a single failure
.uty.i...ue. the 1.nl r. r.r.ne.a to... 1
..ro, weach t....a in tw tzzo -2414c toca nstria. b..
of any component of such systems evea during peri-r This tr.=unna==cune.al.ip initiates closure of Group 2, and 3 primary con-ods when portions of such systems are out of service for main;enance, and Gi) to prescribe the trip set-ainment isolation valves but does not trip the recir-tings required to assure adequate performance.
culadon pumps.
For a 504 ; Mon 7.L2.2 MR.
Rq When necessary, one channel may be made inoper-trip setting of above vessel zero, I
able for brief intervals to conduct required functional and a GO second valve closure time the valves w,ll be i
tests and calibrations.
closed before perforatior. of the clad ocet:rs even for the maximum break and therefore the sening is Some of the settings on the instrumentation that adequate.
initiates or controls core and containment cooling have tolerances explicitly stated where the high and The low low reactor water level instrumentation is r; g,
- ow values are both critical and may have a sub? tan-set to trip when reactor water level is 444" above tial effect on safety. It should be noted that the set-vessel zero
(-59" on the instrument). This gr.
points of other instrumentation, where only the high trip initiates closure of Group 1 primary containment 7
E or low end of the setting has a direct bearing on is lation valves, Ref. Section 7.7.2.2 SAR, and also safety, are chosen at a level e.way from the normal activates the ECC subsystems, starts the emergency operating range to prevent inadvertent actuation of diesel generator and trips the recirculation pumps.
the safety system involved and expowre to abnormal This trip setting level was chosen to be high enough pp E
situations.
to prevent spurious operation but low enough to ini-M.
tiate ECCS operation and primary system isolation cT Isolation valves are installed in those lines that so that no melting of the fuel cladding will occur and b
penetrate the primary containment and must be so that post accident cooling can be accomplished
-3 isolated during a loss of coolant accident so that the and the guidelines of 10 CFR 100 will not be violated.
{d radiation dose limits are not exceeifed during an For the complete circumferential break of a 28-inch CT.-
accident condition. Actuation of these valves is recirculation line and with the trip setting given E
initiated by protective instrumentation shown in above, ECCS initiation and primary system isolation Table 3.2.1 which senses the conditions for which are initiated in time to meet the above criteria. Ref.
isolation is required. Such instrumentation must be available whenever primary containment integrity 46 is required. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an accident.
Amendment No. 43
may be reduced by one for athhort period of time to Two sensors on the ise J. tion condenser supply and return lines are provided to detect the failure of allev for naintenance, tening, or calibration.
This time period is only ~M of the operating time isolation condenser line and actuate isolation action.
in a conth end does not significantly increase the The sensors on the supply and return sides are rie.k of preventing an inadvertent control rod with-the arranged in a 1 out of 2 logic and, to meet drawal.
si:mle failure criteria, all sensors and instrumen-tation are required to be operable. The trip settings Ec APRM rod block function is flou biased' and of 20 psig and 32" of water and valve closure time are such as to prevent uncovering the core or ex-prevents a significant reduction in MCPR cct;ccially
~
,The sensors will actuat e duc during operation at reduced flow.
D e APIUl provides ccedtn; site Itatts.
to high flow in etther direction.
gross core protection; i.e.,
limits the gross core control rods in the normal withd rawal sequence. The Tbc llPCI high flow and temperature instrumentation trips are set so that 1;CI'R is maint:1aed greater than 1.07.
are provided to detect a break in the llPCI ptptng.
Tripping of this instrumentation results in actuation De APPM rod block function which is cet et.
of IIPCI isolation valves; i.e., Group 4 valves.
is the same as that 12% of rated power to functicoal in the re fuel Tripping logic for this function and Startup/ Rot Stendby twde.
This control fo r the isolation condenser and thus all sensors are required to be operable to meet the single fail-rod block provides the same type of protection Startup/Ilot Standby mode as in the P. cruel cod ure criteria. The trip settings of 200 F and 300*.
the tT; ficv biased red block does in ttte run of design flow and valve closure tine are such that code; i.c..
it prevents i'CP3 fron decreasine core uncovery is prevented and fission product below 1.07 durinc control rod wit.hdrcuals and -
relcase is eithin limits.
prevento cent rol rod withdtcval before u T2]
dq ocraa io reached.
inst rumentation which initiates liCCS action is ne As for other vital instut rientation arranged in this fashion the Speci-n c Pat red block function provides local M
arranged in a dual bus system.
cD protection of the core, i.e., the pre-ficat ion preserves the effectit eness of the system ventica et tranultion boiling in a local ru lon -
enn during periods when maintenance or test ing for a cin ;1c rcd uithdtcual errorM cf the core.
it he i tq: perfotwed.
frca a liciting ccctrol rod pattern.
T7.c Sg in f l e-: biased.
'Qc voret casesin%Ie:
t rip poict ccatrol red vithdraval error h n beca cnaly:cdM5)
The control rod block functions are provided to end the re:ulto abe-i that utth the specified prev :nt c.ves<:!ve cont rol rod withdrejisl so that trip nett inm red vithdrevnl is blocked M
1.07. Tne t rip logic f or l
before th: ;;CPJ waches 1.07,thus g
l FCI3 docs not c.ppree.ch c.S.,
eny trip en one 3170 41n<; edec,nste n2rgin*
b th i:, f enction is 1 out of n; ef the s ix APRM's, 8 1RS's, or 4 SRT s vil1 result l
Below 30 Icrcent po scr, the worst caso in a rod block.
De r:intn m inr.trueent channel instrumentation to
- u. g.,,;.1 of a nl.vlo control red resulto requ' rewats assure suf f' ci ent The l
In a 1*J'?."I crec.ter than 1.07 ulthout rod ae,ure the s ingle f ailure criterio are rnet.
r,Moum Instrument channel requi rer=ent: f or the RBM bicci: ccticn. Thus, bclera thi t-power Icvol i t I n not rc'lui red.
48 Amendment No. 74, 43
Tne III'! rod block function provides local c scitin s given in the specincation are adecutte to i,. 11 c: g oos core protection.
'Ib e ocali 0 assurc,6ac above criteria are Tr.ct. Ect. Sec'a.on is ouch tMt trip settia; 13 less G. 2. G. 3 SAR.
'Ihe spectitcat,on preserves t. c a
a r r ao +-tr.:o t t F :.c n f r.c t o r o f 10 cb o.e t e t w. i c e c d leve l.
cffecti'. cress of tye system d'irin- ~*rtods of main-o t- -lycio of tt c steret c_ce.,ccident res ul t s tenance. ! cst.trg. or enh.. oration. and also ra..ta;-
./.c ris,< o. tanvcertent operation; i.e. ' ontv
- n rod bicek cction t,0 fore I.,..R appro?.ches mires..
c.
.w l
- 1. 07.,
cr.e a:u tru~ent channel out of service.
A derc.nscale indication on an APIDI or IIU.I is an in '.ic ton the irstru:nent has failed or the instru-Two air ejector off- :s monitors are provided and J
n:. cat is not sen_ itive enoc ;;:. In either case the wnen tacir trip point is retened, cause an isoletion i: ' rum;nt vci11 net rc::po.d to chan';es in coa'.rol of the air cicetor eff-gas line. Isolation is initiated red mo.c..on nn th us con.rc.i r ou m e., on I.,. pre s e n.,.o.
wl:a bd, :n:,trummts renen t.ne t r,.u. ;; tr.ip pont Ti:e downscele ' rips arc ret at 5/12'> of full scale.
or en.; hr s en upscale trip and inc other a dov ::-
scale trip. 'lhere is a fificen minute del y 'oefore the nir ejector off-ttas iuotation valve is closed.
- The rod block which occurs wnen the
.M,.1 Tnin dela. is acceanteti for be the ':0-minute detectorc arc not fully inscrted in the ho!. g ti;.c cf the off-gas before it is released to the s:ach.
for the refuel and startup/ hot core ctandby pcciticn of the mcde ruitc'n has been orovided to cscure that.thece Ecth tr. trumen.s are required for trio but the cctcetors are in the core during rocctor instruments are so designed that any instru-:ent utartup.
This, therefore, ascurcs that fain rc gives a dovensente trip.
me trip settir's -
of the instruments are set so that the instarenc e,
in proper position ces strek re! case rate lirnit given in Specifica*.ionG, these instru;nents :T rc to provide orotection during reactor
. m. execcc,cc,.
g~
t
.n, u.
s t a r t.u p.
The IIOl's primarily -nrovide F
protection n-cinst local re r.c t iv i cv
.our rad...ia.;on rnonttors are provided w. ich r.
r cffcctc in toc cource ar.d in te r mea :. ate gn;,ja igg;;.ioa of the reactor buildir.; and S
opeiation of the standby gas treatment system. ScN neutron range.
muniters are located in the reactor bui!cingM Tne vc:.ilation duct and un the refueling floor.
'I hc Q For cifcctive emergency core cooling fer smallpipe tr:p lo;ic is a 1 out of 2 for each set and each I:6 b:cahs. the liPCI sys'.cm must funcuen since reac-set een in tinte a trip independent cf the other tor pressure dces not decrease rapidly enoui;n to set.
An; upscale trip will cause the desiz ed G
elitwc citbar con. sprry or LPCI to operate i:: time.
n cD on.
Trip settir;s of 11 mr/hr for ti:e CM
'ibe auica:atic pressure re!.cf function is prevn:cd I
nu nit": s in the ventilation duct are basco uran es a bac;:-up to the ;IPCI : the event -he IIPCI dec3 mrinal v' ntih: tion isolation nr.d r:.niby.
mt at of the tri;.: iOor-
- '!! Ui not ope r t'... Tha nr:
rec-r.?
. : cal eg c;cs. o;w:. utio.1 to hr. git t. e dm e to ; rowi; Jun fu :ttion -
- acts is s ch Ti
- e triI essary ;.tvl mini:aic.e.purians creratina.
49 Amendment No. M,43
3.3 LIMITING CONDITION FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS 3.
(a) Control rod withdrawal sequences shall be 3.
(a) To consider the rod worth minimizer established so that maximum reactivity that operable, the following steps must be could be added by dropout of any increment performed:
of any one control blade would not result in a peak fuel enthalpy in excess of (i)
The control ro.. withdrawal sequence 280 cal /cnn.
for the rod worth minimizer computer (b) Whenever the reactor is in the startup or shall be verified as correct.
l run mode below 20% rated thercul power, the Rod Worth Minimizer shall be operable.
(ii) The rod worth minimizer computer A second operator or qualified technical on-line diagnosite test shall be person may be used as a substitute for an successfully completed.
inoperable Rod Worth Minimizer which fails ef ter withdrawal of at least 12 control rods (iii) Proper annunciation of the select to the fully withdrawn position. The Rod error of at least one out-of-sequence Worth Minimizer may also be bypassed for low control rod in each fully inserted power physics testing to demonstrate the group shall be verified.
shutdown margin requirements of specifications 3.3.A.1 if a nuclear engineer is present and (iv) The rod block function of the rod verifies the step-by-step rod movements of worth minimizer shall be verified the test procedure.
by attempting to withdraw an out-of-sequence control rod beyond the block point.
(b)
If the rod worth minizer is inoperabic while the reactor is in the startup or run modo below 2 0% rated thermal power and a second independent operator or engineer is being used, he shall verify that all rod positions are correct prior to commencing withdrawal of each rod group.
Change No. )$
Amendment Nc. 43 M
3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIRD!ENTS 4.
Control rod shall not be withdrawn for 4.
Prior to control rod withdrawal for startup startup or refueling unicso at least two or during refueling verify that at least two source range channels have an cbserved source range channels have been observed count rate equal to or greater than three corat rate of at least three counts per secoad.
counts per second, 5.
During operating uith lielting cot. trol rod 5.
'a' hen a liniting control rod pattom exists, patterns, as detercticed by the nuclear an instrument functional test of the RB:t enginect, either:
shall be perfor::cd prior to withdrawal of the designated rod (s) and daily thercafter, a.
Lotn RDP. chanucls shall be operable; or b.
Control rod withdrawal shall be blocked; or c.
The operating power level chcIl bo I
lialted to tho the MCD will l
rcnnin above 1. 07 css =ning a cinclo error that resultc in complete withdrai: 1 of cny cingle operable centrol red.
6'G -(
h,,
v
/
57A Amendment No. Jef, 43
indiutive of a generie cox.vl r-
- d-i'ee N.
c l l cu:ount of r;d withdrawal, which is less problem and the reactor will be snut hwa.
thnn a normal single withdrawal inercient, will not contribute to any datage to the pricarv Aac. if dar. age w tthin the contro; tcd di.vc l
cect.snism and in particular, cands in drive i
ccolant systez. The design banis is give.1 i-(; O int e rnal housings, cannot 'oe rulec cut, then a Section 6.6.1 of the S.W. and the d:sigr. evalua-g a gene ric nrobicm af fectirg a ns.;cr of dri.cs tion is t;iven in Sect ion 6.6.3 L.is sapport Mi cannot be rulcd out.
Circumferential erse s is not required if the reactor coolant systo.
cW' resulting fro:n stress assisted irtergranular is at atrtos;5eric pressure since there would corrosion have occurred in the collet hcusing then be no driving force to rapidly eject a hsy of drives at several BNRs. W.is type of drive housing. Additionally, the support is
].9 cracking could occur in a r.u-Ser of driver, not required if all control ecds are fully g6 c<d and if the cracks prcpagated until severance inserted and if an adequate shutdcwa cargin 1
of the collet housing occurred, s:ran could wtth one control rod withdrawn has been deron-ci be prevented in the affected rods.
Liniting strated since the reactor would recain suberitical h a the period of operation with a potcatially even in the event of co r.plete ejection cf the y;;2 severed collet housing and rec,uiring increased strongest control rod.
surveillance after detecting one stuck rod will assure that the rez: tar will no:
3.
Control rod withdrawal and inse-tian s :y:rcas are be operated with a large nu:,ber of rods with established to assure that the tax.inun in cg::nce iailed collet hcucings.
individeal control rod or control red sLment.
s.5ich nre withdrawn could not be isrth crcur. to L.
Centrcl c.od Withdrawra result in a peak fuel enthalpy of 280 cal /gm l
i: they wer:: to dran ca: of the ca g 1.
Contral rod scpeut accidents a Jacu.2.d in the r unner d* fined fcr the 't:d Drop Accideat. M in the SAR car.1c cd t o signific?nt coro The?: sequences are develcped prior to l
Ou..ge.
If coupling integrity is t.in:..inel, opera tion of the unit followin2 :iy refueling outve t'cc pc.ssibility cf a rod drcrout ceciJcnt is and the requirenent that an opcrator follcw these c l i'- i na t ed. The overt ravel pos iticn fiature set,uances is backed up cy the operation c,f the W.'4.
provides a po:itive c' cck as c::1y un:capled These sequences are developed to limit reactivity worths of control rods and dri es riay reach this position.
.. trcn instrumentaticn responso to rod t cVe ent together with the integral prc eides a verificatien that the red i'- fci-red velcei:y i:.iters and the a::isn of th: co crol lowLng its drive. Absence of such response M d r i*.' sy s t e m ; lir:it p M entici reactivity to drive movew nt would provide cause for l
i:iscrtiren such t h : ti.e result s of s ce': :1 -od suspecting a rod to be uncoupled and stuck.
drop accident will not excced a mayintc: fuel ener g*
t Restricting recoupling veriffcations to power cont ant c f 280 cal /gca.
The peak fuel enthalpy of levels above 20% provides assurance that a l
200 ca!/g.r. u below the energy content, 425 cal /gm, at rod drop during a recoupling verification which rcpid feel dispersal cnd primary system da, age have i
In$fre "[everNhan'aIOb.Ne"d."Ci been feuad to occur based on experinental data as is discussed in Reference 1.
2.
The control rod housing support rcetrtets the outward movement of a cor. trol rod to l
The snclysis of the control rc,i drco accidenc was less than 3 inches in the extreeely ree:te originally prescnted in Sectict.s 7.'9.3. 14.2.1.2 event of a housing failure. The an unt of and 14.2.1.4 of the Safety Analysis P.cper.
I _.a.c Ve -
reactivity which could be added by this t
4
- ents in analytical capability have allowed a :acrc l
refined ' analysis of the control rod drop accident.
62 Amendment No. M, M, 43
Bases (cont'd)
ThesetechniquegIpedescribedina The following' parameters and worst-case topical (2) po rt and pvo supple-bounding assumptions have been utilized in the re (3) In addition, a banked reload analysis to determine ccxnpliance with the ments.
position withdiawal sequence described 280 cal /gm peak fuel enthalpy.
5.n Reference (4) has beeri developed Details of this analysis are contained in Reference 6. Each core reload vill be to further reduce incremental rod l
analyzed to show conformance to the li.iting para-meters.
By using the analytical models An f actor (5) inter-assembly. local power peaking 3,
described in those reports coupled with conservative or worst-cace input parameters, it has been determined b.
The delayed neutron fraction chosen that for Pover levels less than 204 for the bounding reactivity curve
- of rated power, the specified limit (typically 1.3% 4K) on insequence control red or control A beginning-of-life Doppler reactivity f eed-c.
rod sentent worths will limit the peak back.
fuel enthalpy to less than 280 cal /ge.
I Above 20% power even eingle operator d.
Scram times slower than the technical errors cannot result in out-of-sequence Specification rod scram insertion rate control rod worths which are sufficient (Section 3.3.2.1) to reach a peak fuel enthalpy of 280 e.
The maximum possible rod drop velocity cal /gm should a postulated control rod (3.11 ft./sec.)
drop accident occur.
f.
The design accident and scram reactivity shape function.
g.
The minimum moderator temperature to reach g' Paone, C.J.,
Stirn, R.C. and Wooley, criticality.
J.A., " Rod Drop Accident Analysis for tors",
(4)
C.J.
Paone, " Banked Position Withdrawal ED -10 7
l9 Sequence" Licensing Topical Report
- Stirn, R.C., Paone, C.J., and Young, NEDO-2123, January 1977.
R.H.,
" Rod Drop Accident Analysis for Lat B'
upplement 1 - NEDO-(5)To include the power spike effect caused by gaps between fuel pellets.
(3)Stirn, R.C., Paone, C.J., and Haun, (6) Generic Reload Fuel Application J.M., " Rod Drop Accident Analysis for NEDE-240ll-P-A, August 1978 Large BWR's Addendum No. 2, Exposed Cores", Supplement 2-NEDO 10527, 623 Change No. p January 1973.
Amendment No. 43
Bases (con'd)
The Rod Worth Minimizer provides automatic It is re.ct;nized that these bounds are supervision to assure that out of sequence conservative with respect to expected control rods will not be withdrawn or inserted;
~
operating conditions. If any one of the i.e., it limits operator deviations from planned Ref. Section 7.9 SAR.
above conditions is not satisfied, a more withdrawal sequences.
detailed calculation will be done to show It serves as a backup to procod ural control of compliance with the 280 cal /gm design limit.
control rod worth. In the event that the Rod Worth Minimizer is out of service, when required, r
in most cases the worth of insequence a licensed operator or other qualified rods or rod segments technical employee can manually fulfill the control rod pattern conformance functions of the In this case, procedural Rod Worth >11nimizer.
in conjunction with the actual control is exercised by verifying all control values of the other important accident analysis rod positions after the withdrawal of each likely parameters described above would most group, prior to proceeding to the next result in a peak fuel enthalpy substantially less Allowing substitution of a second group.
i than the 280 cal /gm design limit.
independent operator or engineer in case.
of RWM inoperability recognizes the capability to adequately monitor proper rod sequencing in an alternate manner without unduly restrict-j ing plant operations. Above 20% power, there is {
no requirement that the RRM be operable since Should a control drop accident result in a peak the control rod drop accident with out-of-fuel energy content of 280 cal /gm less than sequence rods will result in a peak fuel 660 (7 x 7) fuel rods are conservatively energy content of less than 280 cal /go.
To estimated to perforate. This would result in an assure high R'41 availability, the RVM is offsite dose well below the guideline value of requried to be operating during a startup 100FR 100.
For 8 x 8 fuel, less than 850 rods for the withdrawal of a significant number are conservatively estimated to perforate with control rods for any startup after June 1, 1974.
nearly the same consequences as for the 7 x 7 fuel case because of the rod power differences.
The Source Rango Monitor (SRM) system performs 4.
no auto =atic safety system function; i.e.,
it has no scram function.
It does provide the 62b Change No. jMI Amendment No. 43
C, Screm Insertion Times operator with a visual indicatien of neutron level-This is needed for knowledgeable and e f ficien' reactor startup at low neutron le.21.
Tho control red eyotem.is analyzed to bring the The ccanequences of reactivity cecidents are reactor suberitical at a rate fa.se enough to t
prevent fuel damage; i.e.; to prevent the ECPR functiuns of the initial neutron flux.
The g
f rom becontng 1 css th an 1.07,. Th e lici t in g rec,u i rcren t of at leest 3 ccents per second pcuer transient is that resul ting f roa a tur-assures that any raarient, should it occur the initial value of 1(, g bine stop valve closure with f ailure of the bee ns at or above i
turbine bypass system.
Analysis of this of rated neacr used in the cualyses of transients a
trancient shcrrs that the negative reactivity frer told conditions.
One operable SR:I channel rates resulting f rom the scram with the could be adequate to monitor the cpproach to averste response of all the drives as given in criticality us ing homo;;eneous pa t te rns o f g
g gg ca ered control rod aithdrawal. A inninum protection. Ond,...,TR re ca i n s
reater then a
m of t. o opetable SPM's arc provided es an added 1 07 Refcaence (1) shcus the centrol rod cm.ervatism.
cerre. reactivit.y use.1 in analyzing tha t ra as ic:. '.3, Reference (1) should not be
'Ih e Mod 'iloc,n Moni tor (RUM) 1s designed. to auto-cei:!csed ui ch the total coatrol rod vorth atic..lly prevent fuel damage in the event of 33t.k,
- a. listed fu come arendeents to the SAR erroneous rod withdra.al f rom locations of high nc I m.k vclue re p res ca t s the ar:oun t of pc.cer density dering high power Ievel operatien.
recccivity availab7e f o r withdra. cal in the Tuo ch mnels are provided and one of these cry be cold clean core, whereas the control rod byn:med f rom the con:. ole for raintenar.cc and/or vorih t thcM in Reference (1) Icytv-ten t in g.
Tripping of one of the cha:mcis vill block s en t the an.>unt of reactivity availnble for erruacous rod withdrawal soon enou J to prevent fucl i ns e r t i e.' b. cram) in the hot eperat ing cere.
da aue.
"Ihis syrte, backs up the operator vha with-c Q 7;,e j n n.n, amun c of rt ac t i vi ty t o be d r..
rods according to a written.cquence.
The
@(
inset ted du:ing a scran is coM rviled by specified restrictions with one chamel cut of
%i; perrittian no c ac than 1C:1 of the operable 3ervice conservatively assure that fuel demap.c MN rods t o. h.'. v e long sc au ti ms.
In the
- 111 cet occur due to rod withdrawal errors when W
analy t ical t rcctr.ent of r 5e t r. nsients, 370 thi, condi. tion exi: ts Amen:'nents 1)/18 and 19/20
.n;;;;3c 3,P: are. lh :d bea. ca e neutre p cc: :nt the results cf an e.aluation of a rod block h_
s en ;o e :g ; e.. i r,,.-
h u r.'.a ovint and the t c.ai;o r f ai luce.
These cun&ents s':w that during gg; stage o;
- c. tion o f t hi* c en t rol rods.
Th i r.
r reacter operation with certain limiting control h:FM is adqu..t.
ar.d conse rvat l w v'a cn cor.pt, red rod
.u te rn., the v:thdrawal of e designated singIc
$nE to the t r pi cal ly ob scre::d tite delav of c,,ntrol rod could result in one er mr.rc fuel rods M
aboit 2 70 e.i ll i n ec oa.P..
A,^ rroxi ately 70 l
uith 1:cpa c less t h.. n 1.07. During :;c of such M
=i l l i r.ce e n d.: a f t e r n e u; r:a : lux reaches :he I,. : : t. a n., it 1:. jiah :d that testing of the in:1 W-,
,, n t. r: prior to withdrawal of sun roIs to assure
.f
(, )
j'r-..lon Station Specir.1 Report !!o.
y i t s..pe rab i li *.y will at:sure that i :- roper with-d ra./.il doe r, ac t occur.
It is the re'.ponsibility
/.9, Supploment 1s", F1curo 1 of the hcicar Eng.i acer to identify these IIniting m 12rna aad the designated re!:: either t.hcn the p.:r t r a, are initially established cr as they 63
, '.n t rol
/cendment No. J, 43
%> dae to the s corrence of ir pera le I
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VS. PLANAR AYLRAGE EXP05URi-nmendment [, 43 81C
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FIGURE 3.5.1-B MAxlMUM AVERAGF. PLANAR LIN'AR HEAT GENERATION R ATE (MAPL.'GR)
VS. Pt. AN AR A'lERAGE EXPOSURE 81C-1
/ Wend:nent No. J#, 43
LI:{ITING CCI:DITIOIT FOR C?2R..TICII SU.'.*ICILL2:CE REQUIRB*.ENT 1
J.
Local LTICR J.
Loent Lygg During steady state power operation,.the The 11tCR as a functio'n of core linear heat generation rate (LHGR) of any hei ht shall be checked daily dur-E rod in any fuel assembly at any axial le-ing reactor operatien at > 25t
~
cation shall not exceed th'c maximu:n allcw-rated thernal po er.
able Lf!CR as calculated by the following equation.
L!'OR LMCR 1-M L
e ma> L T /
LI!Gnd Design LIIGR 17.5 kw/ft, 7x7 fuel assemblies 13.4 kw/ft, 8x8 fuel assemblics l
A p) 8x8 R fuel assemblics P Anax - Maximum oower soikinc ~oenalty
.037 initial core fuel
.026 reload 1, 7x7 fuel
.022 8::9 fuel J
.000
- 8x8 R tuel LT Total Core Length - 12 ft.
T.xial distance from bottom of core L
!f at any tine during. operation it is det-
[@:
ernined by normal surveill:nce that the liciting value for 1.HGR is being execeded, acticn shall 4
be initiated within 15 cinutes to restore operation to within the prescribed lik.its.
If the v
IJICR is not returned to s
within the prescribed licit.,
within tso (2) hours, the reactor shall be brougiit' to the Cold Shutde m condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Su rv e i l-lance and corresponding action 81C-2 shall centir.u: until reactor ope ratico i..ci:'u.. the Amendment No. M, 43
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Limiting conditions for operation nases developed in this refer-3.5 ence, the repair period is found t<> be less than Core Spray and LPCI Mode of the RHR 1/2 the tett 1: *.crval. This assumes that the A-S_yntem - This specification assures core rpray and LPCI subsystems constitute a that adequate emergency coolin9 1 out of 3 system, hov.ete'r. the combined ef-fcct of the two sys: cms to limit excessive clad capability is available.
temperatures must al>o be coasidered. The test interval specified in specificatina
.1,., was Based on the loss of coolant analyses U * " """ -
U"'I" 3 3 3 I'"u W t rp.1i r' incluoed in References (1) and (2) in
- f;
, { '
C[."
accordance with 10CFR50 A6 and Appen-core cooling systems provide and this q,ccification is within this n riod.
I k'
dix K, For rn"Itiple fa ! ares. a et.'r:e r.au rvat is r,ufficient cooling to the core to specificil and to impre c the as,urance that
'6 dissipate the energy associated with the ty'ina'cing nstems will fanetior. a dairy--
'jB the loss of coolant accident, to limit I"' * * #
ed fo r.
Utav it it reco;:nized the calculated peak clad teraporature that the mformati..n given in refeience S pro-6 D
to less than 2200aF, to assure that yi,g s a panHtuire m-bl a % u my N[j to 12.mi.t ah!c n erair times, the lack of on rating data to core geometry remains intact, r&
supp.>ct the analytical approach preiects com-i-ha core wide clad metal-vater reaction P ete acceptan::e of ihn me:h,,d at abis time.
W l
to less than IE and to limit the cal-pienWre, the Mmes Mate.! ia the sp. eific culated local metal-water react ion item:L were est.ibli8hc(! with due regard to M7 juiigment.
to less than lby..
'i'h e allowable repair times are os-Sb sdd <,nc core spray subsystem become in-op: r: bte. the remainir; core spray and the tablished so that the average risk rate entice LPCI system are available should the 2.cr repair would be r.o greater than i.bc basic risk ratc.
The method and (2) UE00-20566, General Electric concept are described in Reference Company Analytical i<odel for Loss-(3).
Using the results o t -Cool im t Analysis in Accordance
'Jith lOCFftSO Appendix K.
(1) " Loss-of-Coolant Accident Analysis Report (3) MT.D " Guide lines for Determining for Dresden Units 2, 3 and Quad-Cities f,afe Test Taf:cevals and Repair for I;nyi ncered Si,f egaards"
'r i a.e s Units 1, 2 NEDO-24146, September 1978 "
April 1%9, 1.11. Jacobs and r.w. Marrio:.t..
82 Amendment No. J f, 43
5 Li.2 ting condition for operation Brtsen (cont'd)_
I,
/.ve-ere Plc.nar IltCR, Thio e,.:cification accures that the roak
~
cladding tosperature follo, sing a Tcntrhted dorign bacio locn-og-cool: int cecidont vill not execcd tho 2200 F limit crocified in 1 CCFR50 Appendix l' cencicerir.c tho poctulated offceta of fuel pilot dcncificction.
4 ONh The p:nk claddnc tens:raturo follo' ting a M,
p ct.ulated losc-cf-ccolant e.eddent 1 :
Y, F:1scrily a functica of the avon ;a iD:Il f
of n)1 the redo in a feel ccr.er.L!y nt nny
{x'd
'n7.lul location und 1u o.dy d. r::. dent vocar.d-cr:1y cn the rol to rcd putr dictri'.: tion N
Q'G, within a feel accenbly, 33nco exp7cted local ye varintienu in po::Or dir.tribution uithin a Mf foal ccco:.bly affcet the calculate;1 r;al: olnd
[
toar ratulo by leuc than i20 7 rahl.1*:
0 0 to the r:nk tenperatum fcr a typical fn:1 design, th-: linit on the cvorato phns: 15:R to cuf ficient to accure that cal ;ulctcd tenp-cr.iturco cro tolcu the 10CFR$'), Appondix K The linit.
v_uin.u.,i cverano planar 1EGRc plotted in The r;e:inun averaco p3rnar IliG:ta choun in Fic. 3.5.1 c.t bicher c:Peure recult in a F1:7to 3.5.1 nro b ced on c,ticulaticn: c. ploy-calce w.ca p ak clad tearcraturo of 1:22 In,i tha nulclo dcccribcd $n Itef aranco (1).
il en M039.
!!cucver the nuinum averago I c :.:r op: ration with IECRa et or 1. Ic? thcce plocar IHC1:a are chcun c:t Fic. 3.5.1 as cr.o:#n in Fic. 3.5 1 ccceros that the resh lir.ita b cauno conforr.:nco calculatica, have cirblin", temporatum follouirr; a pcat : lated nct 'c:en perfor..2d to just.ify opar.itica at Ic:1-of-coolant accident.:131 not er.cced tho
]Ect:a in. As:ccu of thcae 'nhour.
2.:0'J F 31ni t.
Thono va3nis repWCent 11DitG C
for opration to encum cc:.foir nco :ith J.
I.nc.l_T H11 l'.*:-itSO and Agg.cndix l' cely $ f th<.y ni o toro liritini: then oth:r decica prane tars.
This cp:cification annuun that the r;;drun 11near h:at cci.cration rato in (1) " Loss-of-Coolant Accident Analysis Report
,any rci in loco than tho deqica lincur for Dresden Units 2, 3 and Quad-Cities Units A,
2 Nuclear Power Stations, 8$A NEDO 24146, September 1978."
Amendment No. J/, 43
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a.ociate d <!ccay anil strm:turnt sensible heal relca::t d durin;t primary 8ystem Inowstown from 3.7 10n0 psig.
A, Prim,rv Containment - The integrity of the Since all of the gases in the drywell are purged primary contaimnent sad operation of the in'o the pressure -nppression chamber air eme:gency care coolin:: sys:cm ni cmnbination, space d irin:t a los of coolant arcident, the limit the of f-site dores to tatues !cs.ithan thoac pressure resultinn from isothermal compres-sugat3ted in 10 Cie': 100 in the even. of a !,rea',c si.m ;i!us the valnr ju c3sure of the litpti<l must in the primary sys?cm piping. Thns, cont:iin-not execed G2 poig, the suppress:on ch:nober ment integri:p is specified w henever the poten-design pres 3u re. The design volume of the tial for violation of the pri:' tai y reactor :-y. tem suppressiott chan.her (water and ais ) w a.s in:cgritt es i: !s.
Concern n'eout such i troht-obt.uard by consie!ering that the total volume of lion exis!s w het.ever the reactor i:, critica! and re:ictor cenlant to be consh-nseil is dis cha rgeel abot e a:mo.v ib ric prewe re.
An exei pt im is to fl.c -oppre.
ion cham!.er arn! that the eh y-m:ide to 'his regnire.nent du ring nitial co re u cll vnhttue is pu r;;ett to the suppression Cham-lo uhng.u:<l w hile the low pow er test progr.im he r.
1(cf. Section 5. 2. :t Salt, cfg is being comhn ted darin:.;, initi;il em e heading and ulo te the low peexer test progr.im is being U..irg the minimum or maximum water va'unes gm ec.odne:cd and ready access :o the reae:or ves-g.ver ir :he speciheation, cont:iimnent pres-h oc! is required. There w ill be no pi eeu re on
.ttre durirn :he dman basis accident is approsi-ps the.syytem at this time w hich will gicat!y ma:cly IM psig w hich is belmv the de. sign of G.!
r educe the chance.s of a pipe break
': he psig..Tlaxinmn v.ater volume of 115,G55 ft3 Q
reactor toay he taken en inical darm:.: this period; re..ults in a dou ncomer. submergence of 4 feet hav.crer. re-t rictive op.-
ting procedu re:. veill and the mini
,a vo'ame of 112.000ft3 results c- ?
1,.. in eficet.tga:n to nonimize the pr.e i'eility of in a 3ubme:nence appensimate!y 4 inches less.
Q4
.m accident occurring. Proceda:es c.nd the Itod The majority of the I;odega tests (ft) were run g3 Worth.ilin6mizer u vuhl limit cont re.1 worth to with a enhme:ged length of 4 feet and with com-(j d preclude a peak fuel enthalov of 280 cal /gm.
piele conden.ation. Thus, with respect to CW V
In a5dition
.am..cc..:g, t. abr..es;,enec, tnis specificat:on is e
in :1.e unti:.cly event th tt an exc$rsion did accer, ad eot:a t e.
the reae:or bitihhng :mel 3:andby ga.s t reatment system. w hich shall be operational dut ing this Experiment il data indicates that excessive time. offer.s a suf ficient ba rrier lo hecp of f-site stcan condensing loads can be avoi.!cd if dows acil w ithin 10 Cl'It 100 the peak temperatorc of the suppression pa.,1 is naintainca below 160*F during any 36 The pressure suppres> ton paol wa:er provides perio ' of relief valve ope.vation with sonic the heat # for :he reactor primary sys:em com!it ions at the discharge exit.
pecifica-c energy t elease fa!!owirg a posta!ated recture of the systcm. The prensurc suppression chamber water volume must absorb the.
W NG"m Dgy Prc}iminary Hazards Sxaary a c.c o r t, nocond:.x 1, Docket 50-205 naconber 28, 1962.
125 Amendment No. J(, 43