ML20137X272

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SER Supporting Thermal Hydraulic Stability Tech Spec Change, Per Util 840515 Application for Amends 40 & 5 to Licenses NPF-14 & NPF-22,respectively
ML20137X272
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 07/11/1984
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20136A677 List:
References
NUDOCS 8603050420
Download: ML20137X272 (5)


Text

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f SAFETY EVALUATION REPORT FCA P.ROPOSED MEDENT 40 TO LICENSC H0. H9F-14 AND PROPOSED MINDHENT 5 TO LICE.4SE NO. NPF-22 SUSf)UEHANNA UNITS 1 AHD ?

PPLL has proposed (N. W. Curtis (PP&L) to A Schwencer (MRC) dated my 15. 1384) chenges to the Susquehanna Unit 1 and Unit 2 Techr.1. cal Specifications designed to resolve tha Themal-Hydraulic Stability concerr.s outlice:1 -In General Electric Service Informtion Letter No. 380, Nevision 1, dated February 10, 1984. The prin fpal additions P.6de to the Technt:a1 Spec 111:aticos are the following: '

t i 1. W'1en operating with or.e or no recirculation loops, the plant will ,

tw.iedtetely initiate an orderly reduction in thermal power to Jess tt,an a spec'ified lia.it.

2. When la twin loop operatbn at flu rates less than 45 tuillion
ibibr, reduce therral power t.o the specified it;it cr mor. iter APRM ana LPRM neutron flux levels at least or.ce per S hours and I insure thet they art lass than thrae times their established baselina levels.

s f The staff has r.tsfewed these proposed changes and tus found thet they are prudent and acceptably resolva our Ther:nl-Hyttraulic Stsbility cor.cerns for Susquehanna Ur.its 1 and 2, assuming 1cng teen singia Joop operation is r.ot pemitted. Should such operation be requested in the future tiu staff will revaluate-1.his hchr.fcal Specification to determine if additional modif f -

caticos are required.

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, RE01:1CULATION LOOPS (IgIJgNDIT!05 FOR OPEMTION _

3.4.1.1 Two resetor coolant systen recircaJstion loops s' hall be in operation and: ,

a. total ccre flov shall be greater than or equal to 45 c.lllion Ibs/br.

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, b. T!!ERMAL F0VEk ths11 be 1e.ss than or equal to the limit specified in I Tigu.e 3.4.1.t-1.

AFFLIC,tBILIT Y : CPERATIONAL CO.\'DITIONT 1* and 2*.

ACTION: ,

a. With one teactor coolant syrtaa recirmlatice loop rot in operation, 12 mediatGly initiat.e an order 1/ reduction of 7FE4fAI. P0kTR to less than or j equal to the if mit specified in Figu~e '8.4.1.1-1, and be it at least HDT
'. HUT 00VN vitbita the r, ext 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />,
b. 'Jirh no taa: tot coolont system tecirculatic-a it, ops in operation.

1 Medistc?,y initiate an order.ly reduction of UIEMAL PCWIR tc, Icss than or equal to the limit tpecified in Figure 3.4.1.1-1 and inieJata measures to plate *,he tr.it fn a'c least $TI.2 TUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to HOT SHUTDOW, vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 'g 1

c. With tt.o reactor coolant syetem tr.cir.ulation '4 oops in operation and total core fiev less than f.5 militou Jbs/hr c,rd THERMAL POWK greater than tlw 1

,t limit specified in Figure 3.4.1.2-1:

1

1. reduce Te'EPMAL POWER to Jess thsn nr equel to the 11mit specified "in '

Figure 2.4.1.1-1, or

1. ircrease care flow to greater char 45 tail 2 ion Ibr/hr, or

, 3. deterdne ti;e APPd and LPRh*** ueutron flux nedsc levels %ithin 1 i hour, and:

a) if the AyF]! and LPT.M*** r.eutron flux noise .levait are Ices then three ti=an their established baseline levels, centinue to '

determine the noisc Isvels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> sna within 30 minutes after the completion of a 7HEks/C. POWE3 ineteast of

! at least $t of RATED THtPJfAL PCWER, or '

b) if the APPR or LPRM*** neutron flux noisc levels are granter ths.n or equal to three tievs their established bassline Icvels, itentiatel;' initiate corrective action and restere ttic noise

' levels to uichin the reqs.ited limits within 2 hcurs by increasiet core flow to trsater thar. 45 million 1bs/hr, and/or b/ initiating an orderly reduction of THtRHAi,' POWER to Isss than

,,. er equal to the ifcit specific 4 dn Figure 3.4.i.1-1.

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SURVEILLAN_CE REOUTREMESTS

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4 . 4.1.1. .! Each pump discharge valve and bypass valve shall be demonstrated OPERA 31.E by cycling each valve through at least one cetr.plete cycle of full *

' trrxel during each startup** prior to THEA.".AL PCER exceeding 25'. of MTED TF.G.'LtL P0k'ER .

4.4.1.1.2 Each pump discharge bypass valve, if not CPERABLE, shall be verifted to be closed at least once per 31 days.

4.4.1.1.3 Each pump }C set scoop tube electrical and techanical step shall be '

de:renstr:.ted 0? ERA 3LI with e*;er. speed setpoints Icss than or equal to 102.5 and 105%, respectively, ef rated core flov, at least once per 18 months.

4.4.1.1.4 EstablishabaseliceAPP.McndLPRMYutron.fluxnoisevalueata point within 5% RATED THEUtAL POWER of the 1002 rated rod line with total core flow between 35% and 50% of rated total core ficw during startup testing following each refueling outage.

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r*If r.ot performed within the previous 31 days. ' .

. *** Detectors A and C of one LPRM string per core octant plus detectors A 'and C '

of one LPM string in.che center of the core should be monitored,

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SSECIAL TEST DCEPTIONS a

3/4.10. .*- RECIRC':LATION LCCPS L.*M: TING CONDITICN FOR OPE 2A710N -

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ay ::e sustended fer up to 24 hcars for the perf3r:2nce of:
a. DHYSICS TES'TS, previ:ed tutt THER.'iA.' PCWER does nc az: Leg 57. et RATED THERMAL PCWER, er 4

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A;G;_~ C. !ILITY:

OFEf.ATIONAL CONDITIONS 1 an.! 2. during PHYSICS TE3TS and t.ie StirtuO Te:% PrCQTam. "

ACTICA:

a. Wit h the above specified time limit exceeded, insert til control rods. .

. b. With t$v above spesified THERMAL POWER linit exceeded curing PHYSICS TESTS immediately place the resctor mede swit:5 in the Shutc'own pcsition. \

. . T SUEVEILLANCE RE0VincMEN75 ___

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4.10.4 The tfra curint wnich the acve spe:if fed teAviramta- has been sus;endec shall be verifiec to 07 less than 24 het.:rs a:, least ente per hcur curing PM'15:CS TE5TS anc ne Starttp Test Program.

4.10.4.2 THERv.AL POWER Ena n te ceta mined to be less than 5% cf R CED THERMAL PCFEE a: 14ast once per nour during P,9YSICS TESTS.

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Docket No. 50-331 Mr. Lee Liu Chairman of the Board and Chief Executive Cfficer Iowa Electric 1.ight and Power *,ompany Por.t Office lor 361 Cedar Rapids, Iowa 52406 Detr Mr. Liu; The Consnission has issued the enclosed Amenthent Wo.110 to Facility Operating License No. DFR-49 for the Duane Arnold Energy Center. TMs amendment ccnsists of changes to the Technical Speci/ications in response to your application dated December 7,1904 The amendment revises the Technical Specifications to incorporate changes to (1) permit reactor operation with one recirculation loop out of service, (2) provide for detection and sbppression of thermal-hydraulic instabilities during both dual loop and single loop operation, and (3) update some refarences and delete some blank pages.

, A copy of the related Safety Evaivation is also enclosed.

Sincerely, .

d -

Mohan C. Thadant, project Manage Operating Reactors Branch #2 Division of Licensing

Enclosures:

1. Amendment No,119 to License No. DPR-49
2. Safety Evalcation cc w/ enclosures:

See next page i

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Iowa Electric Light and Power tampany I Duane Arnold Erergy Center cc:

Jack Newman, Esquire '

Mr. Thomas Houvenagle Harold F. Reis, Esquire Regulatory Engineer Newman and Holtzinger  ! Iowa Comerce Comission 1615 L Street, N. W. Lucas State Offica Building Washington,'O. C. 20036 Des Moines, Iowa 50319 Office for Planning and Programming 523 East 12th Street . .

Des Moines, Iowa 50319 4

Chairwan, Linn County Board of Supervisors Cedar Rapids, Iowa 52406 Iowa Electric Light and Power Company ATTN: B. L. Mineck Post Office Box 351 Cedar Rapids, Iowa 52406 U. S. Nuclear Regulatory Commission ~

Resident Inspector's Office Rural Reute il Palo, Iowa 52324 James G. Keppler Regional Radiation Representative ,

Region III Off'ce U. S. Nuclear Regulatory Commission '

799 Roosevelt Road Glen Ellyn , Illinois 60137 i

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UNITE 3 STATES

[ g NUCLEAR RECULATCRY COMMISSION

.O 't WASMiteGTOet. D. C. 20085

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IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE

_ DOCKET NO. 50-331 D'UANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 119 License No. DPR-49

1. The Nuclear Regulatory Comission (the Cossnission) has found that:

A. The application for amendment by Iowa Electric Light & Power Company, et al, dated December 7,1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the -

Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; i

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. 1 The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:

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(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.119, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. -

3. The license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch f2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: May 28, 1985 o

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, ATTACHMENT TO LICENSE AMENDMENT NO.119 .

FACILITY OPERATING LICENSE NO. DPR-49

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DOCXET NO. 50-331 Revise the Appendix A Technical Specifications by removing the. current pages and inserting ,the revised pages listed below. The revised areas are identified by vertical lines. .

LIST OF AFFECTED PAGES i 1.1-1 3.3-7 1.1-2 3.3-7a**

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1.1-3 3.3-18 1.1-5 3.3-20**

1.1-6 3.6-7 '

1.1-7  : 3.6-7a**

1.1-8 3.6-7b**

1.1-9 3.6-31 1.1-10 3.6-34 .

1.1-11 3.6-35 1.1-12 3.12-1 1.1-13 3.12-2

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1.1-14 3.12-3.

1.1-15 3.12-4 1.1-16 3.12-7 1.1-17 3.12-8

. . 1.1-18 3.12-9 -

1.1-19(Figure 1.1-1) 3,12-9a*

1.1-20(Figure 2.1-1) .3.12-10 1.1-21* 3.12-11 1 1.1-22* 3.12-13 (blank page) 1.1-23* 3.12-15 '

1.1-24* 3.12-16 3.1-3 3.12-17 -

, 3.2-16 3.12-18 3.12-19* '

i 3.12-20*

l *These pages have been deleted.

    • The'se are new pages .

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- DAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLAODING INTEGRITY v.

Applicability: Applicability:

Applies to the inter-related Applies to trip settings of the

' variables associated with fuel instr uents and devices which thermal behavior, are provided to prevent the reactor systen safety limits from being exceeded.

Objective: Objective: <

To establish limits which To define the level of the ensure the integrity of the process variables at which fuel cladding, automatic protective action is initiated to prevent the fuel cladding integrity safety limits from being exceeded.

Specifications: Specifications: .

The limiting safety system settings shall be as specified below:

A. Reactor Pressure > 785'ssig A. Neutron Flux Trips and Core Flow > 10% of ilated

1. APRM High Flux Scram When The existence of a minime In Run Mode.

critical power ratio (MCPR) ,

less than 1.07 for two For operation with the recirculation loop operation fraction of rated power (1.10 for single loop (FRP) greater than or equal operation) shall constitute to the maxima fraction of violation of the fuel c. adding limiting power density

. integrity safety limit. (MFLPD), the APRM scram trip setpoint shall be as B. Core Thermal Power Limit shown on Figure 2.1-1 and (Reactor P essure 1785 psig shall be:

or Core Flow 110% of Rated S f (0.66W + 54)

When the reactor pressure is with a maximm setpoint of

<785 psig or core flow is less 120% rated power at 100%

Than 10% of rated, the core rated recirculation flow or themal power shall not exceed greater.

25 percent of rated thermal -

power.

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Amendment No. J4(,119 I 1-1

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, , ,---- LIMITING SAFETY SYSTEM. SETTING C. Power Transient Where: S o Setting in percent of rat;d p:wnr (1,658 MWt)

To ensure that the Safety Liaits established in Specification 1.1.A and 1.1.B are not exceeded, W = Recirculation logo flow in each required scram shall be aercent of rated flow, initiated by its primary source kated recirculation loop sianal. A Safety Limit shall be flow is that loop flow recirculation which assumed to be exceeded when scram corresponds to 49x106 is accomplished by a means other 1b/hr core flow, than the Primary Source Signal.

For a MFLPD greater than FRP, the - -

APRM scram setpoint shall be:

FRP S < (0.66 W + 54)

- for two MFLPD

' recirculation loop operation, and D. With irradiated fuel in the reactor vessel, the water level shall not be less than 12 in. S < (0.66 W + 50.5) above the top of the normal MFLPD active fuel zone. Top of the ./or one recirculation loop active fuel zone is defined to be operation

  • 344.5 inches above vessel zero NOTE: These settings assume (see Bases 3.2).

operation within the basic thermal design criteria. These criteria are LHGR < 13.4 KW/f t- (8x8 array) and MCPR 7 values as indicated in Table 3.17-2 times Kf, where Kf is defined by Figure 3.12-

1. Therefore, at full power, operation is not allowed with -

MFLPD greater than unity even if the scram setting is reduced.* If it is determined that either of these design criteria is being violated during operation, action must be taken imediately to return to operation within these criteria.

2. APRM High Flux Scram When in'the REFUEL or STARTUP and '

HOT STANDBY MODE the APRM scre;n l shallbesetatIessthanore to 15 percent of rated power. qual

  • With MFLPD greater than FRP cturing power ascension up to 90% of rated rather than adjusting the APRM power, setpoin ts, the APRM gain may be adjusted such that APRM readinos are greater than or equal to 100% of MFLPD, provided that the adjusted APRM reading does not exceed 100% of rated power and a notice of adjustment is c ontrol panel. posted _on the reactor Amendment No. M,119 1.1-2 l

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DAEC-1

___ SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

3. APRM Rod Block Wwn in Run Mode.

For operation with WLPD less than or equal to FRP the APRM s Control Rod Blo:k setpoint shall be as shown on Fig. 2.1-1 and shall be:

5j[ (0.66 W + 42)

The definitions used above for the APRM scram trip apply.

For a MFLPD greater than FRP the APRM Control Rod Block setpoint shall be:

S < (0.66 W + 42) for two MFLPD recirculation loop operation. .and S --< (0.66 W + 38.5)

MFLPD .

for one r ecirculation loop

operat 1.
4. IRM - The IRM scram shall be set at less than or equal to 120/125 of full scale. -

B. Scram and > 514.5 inches Isolation on above vessel reactor low zero (+170" water level indicated

level) l C. Scram - turbine < 10 percent l stop valve v'alve closure  !

closure l l

D. Turbine control valve fast closure shall occur within 30 milliseconds i

of the start of turbine control valve fast closure.

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Amendment No. JA5'i 119 4 1.1- 3 .

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1.1 BASES

FUEL CLADDING INTEGRITY s.

A.

s Fuel Cladding Integrity Limit at Reactor Pressure > 785 psig and, Core t

Flow > 10% of Rated I

The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel danage are not directly observable during reactor operation. l the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.

Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel. rods, the critical power at 4

which boiling transition is calculated to occur has been adopted as a convenient limit.

However, the uncertainties in monitoring the core operating state and in the procedure used to calculate the critical power" result in an uncertainty in the value of the critical power. Therefore, the fuel cladding [

integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rodss in the core are expected to avoid boiling transition considering the power distribution within the core a~nd all uncertainties.

. \

I The Safety Limit MCPR is generically determined in Reference 1, for two  ;

recirculation loop operation. This safety limit MCPR is increased by 0.03 for I single-loop operation.

Amendment No. J4T, 119- 1,1 5 t

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DAEC-1

8. Core Thermal Power Limit (Reactor Pressure 1785 psig or Core Flow 10%

of Rated)

At pressures below 785 psig, the core evaluation pressure drop (0 powe'r, O flow) is greater 'than 4.56 psi. At low power and all flows, this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core 4

pressure drop at low power and all flows will always be greater than 4.56 psi.

3 Analyses show that with a flow of 28 x 101bs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 103 lbs/hr irrespective of total core flow and independent of bundle power for. the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia ,to 800 psia indicate that the fuel assembly critic'al power at this flow is approximately 3.35 MWt. .ith W the design peaking factors this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reac' tor pressures below 800 psia or core flow less than 10% is conservative. '

t C. Power Transient' P1 ant safety analyses have shown that the scrams causea by exceeding any -

safety setting will assure that the Safety Limit of Specification 1.1.A or 1.1.B will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when

~ a scran is accomplished other than by the expected scram signal (e.g., scram from neutron flux folowing close of the main turbine st'op valves) does not i

necessarily cause fuel damage. However, for this specification a Safety Limit violation will be assined when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.

Amendment No. 119 1.1-6 l

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DAEC-1 The computer provided with Duane Arnold h'as a sequence annunciation prdgram which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc., occur. This program also indicates when the scram setpoint is cleared. This will provide infomation

~

on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information nomally will be available for analyzing scras; however, if the computer infomation should not be available for any scram analysis, Speelfication 1.1.C will be relied on to detemine if a Safety Limit has been violated.

D.

Reactor Water Level (Shutdown Condition)

During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If ' reactor water '

level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.

The core can be cooled sufficiently should the water level be reduced to two-i thirds th'e core height. Establishment of the Safety Limit at 12 inches above I i -

the top of the fuel

  • provides adequate margin. This level will be continuously monitored.
  • Top of the active fuel zone is defined to be 344.5 inches above vessel zero (See Bases 3.2). -

Amendment No. 119

  • 4 g k

I e

-w

.. . 9 r

____.p. -

DAEC-1 o .

%* e

1.1 REFERENCES

1. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A* l
2. "Duane Arnold Energy Center Single-Loop Operation " NEDO-24272 July 1980. l

\

l

  • Approved Revision at tim'e reload analyses are perfomed.

. \

I i Amendment No.119 1.1-8 1

. 1

..- . DAEC-1

2.1 BASES

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Duane Arnold Energy Center have been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 102% of 1658 MWt in accordance with Regulatory Guide 1.49. The analyses were based upon plant operation in accordance with the operating map given in Figure 1.1-1 of the Technical Specifications.

In addition,1658 MWt is the licensed' maximum power level of the Duane Arnold Energy Center, and this represents the maximum steady state power which shall not knowingly be exceeded.

l Conservatism is, incorporated in the transient analysis in estimating the controlling factors, such as void reactivity coefficient, control .

rod scram worth, scram delay time, peaking factorsnand axial power shapes. These factors are selected conservatively with respect to their'effect on the applicable transient results as determined by the current analysis mode. Conservatism incorporated into the transient analysis is doctanented in Reference 1.

This choice of using conservative values of controlling parameters and initiating transients at the d'ssign power level produces more conservative results than wou'd be obtained oy using expected' values of control parameters and analyzing at higher power levels.

I Amendment No. 119 l'l-I l

e 4.

t

,, .._,..%--- --'-~ --" *-*"' "

i-DAEC-1 For analyses of the thermal consequences of the transients the MCPRs stated in Section 3.12 as a limiting condition of operation bound those which are conservatively assumed to exist prior to initiation of the transients.

i As discussed in Reference 2, the core-wide transient analyses for one recirculation pump operation is conservatively bounded by two-loop operation analyses an'd the flow-dependent rod block and scram setpoint equations are adjusted for one-pump operation.

Steady-state operation without forced recirculation will not be permitted, except during special testing. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.

In summary:

s

1. The abnormal operational transients have been analyzed to a 4

power level of 102% of 1658 MWt.

ii. The licensed maximum power level is 1658 MWt.

i iii.

Analyses of transients employ adequately conservative values

~

of the controlling reactor parameters. '

Amendment No. 119 1.1-10 , , l e

+em>+ 6-O&*

DAEC-1 iv. The analytical procedures now used result in a more logical s.

answer than the alternative method of assuming a higher starting power in conjunction with the expected values for -

the parameters.

Trip Settings The bases for individual trip settings are discussed in the following paragraphs.

A. Neutron Flux Trips i

1. APRM High Flux Scram (Run Mode)-
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1658 MWt).

Because fission chambers provide the basic input signals, the APRM-system responds directly to average neutron flug During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses demonstrate that with a 120 percent scram trip setting, none of the . abnormal operational transients analyzed violate the fuel Safety f.imit and there is a substantial margin to the threshold for fuel damage. Therefore, the use of flow referenced i

Amendment No. 119 l l'Il * ~

I

_ _ _ _ , - --r -

r * *~ ~ ' *** ' ' *' ~ ~ ' ' ~ '

.DAEC-1 l .

scram trip provides additional margin. An increase in the APRM l

i scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram l

trip setting was determined by an analysis of margins required to .

provide a reasonable range for maneuvering during operation.

Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety

, because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams.

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of MFLPD and reactor core' thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1, when the -

maximumfractionof.limitingpowerdensityis.(r.eaterthanthe.

fraction of rated power. This adjustment may be accomplished by

. increasing the APRM gain and thus reducing the slope and intercept point of the flow-referenced APRM High Flux Scram curve by the reciprocal of the APRM gain change.

Analyses of the limiting transients show that no scram adjustment 1

is required to assure MCPR greater than or equal to the Safety- l l

\

Limit when the transient is initiated from MCPR 1 values as '

indicated in Figures 3.12-2 and 3.12-3.

[

Amendment No. lig 1.1-12 1

l' 0

J

[ . .. - - . - . - . -. -. - - . -.- ..-. _

DAEC-1

2. APRM High Flux Scram (Refuel or Startup & Hot Standby Mode)

For operation in these modes the APRM scram setting of 15 percent of rated power and the IRM High Flux' Scram provide adequate thermal margin between the setpoint and the Safety Limit, 25 percent of rated. The margin is adequate to' acconnodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the systen, tenperature coefficients are small, and control rod patterns are constrained to be unifonn by operating

~

procedures backed up by the rod worth minimizer and the Rod Sequence Control System. Idorths of individual rods are very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peak , and because several

{ rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is near equilibriun with the fission rate. In an assuned '

uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a seram before the power could exceed.the Safety Limit. The 15 percent Amendment No. 119 g, _

DAEC-1 APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greated 'than 850 psig.

3. APRM Rod Block (Run Mode)

Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given power level at constant recirculation flow rate, and thus prevents a MCPR less than the Safety Limit. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents excessive reactor power level increase resulting from control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin s

to the Safety Limit increases as the flow decreases for the specif,ied trip setting versus flow relationship; therefore the worst case MCPR which could occur during steady-state operatio is at 108f, of rated thermal power because of the APRM rod block trip setting. The actual O

power distribution in the core is established'y b specified control rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction of limiting power density exceeds the fraction of rated power, thus preserving

~

the APRM rod block safety margin. As with the scram sett'ing, this may be accomplished by adjustiag the APRM gain.

! Amendment No. 119 t ,-

\ .

1

{Ih.. -. _ - --- - ' '~

DAEC-1 9

4. IRM "

The IRM system consists of 6 chambers, 3 in each of the reactor protection system logic channels. The IRM is a 5-decade instruisent which covers the range of power level between that covered by the SRM 4

and the APRM. The 5 decades are covered by the IRM by means of a

range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram trip setting of 120 divisions is active in each' range of the IRM. For example, if the instrument were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ~ ranged up to acconnodate the. increase in power level, the scram trip setting is also ranged up. The most significant sources of reactivity change during the power increase are due to gontrol rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that the heat flux is in equilibrium with the neutron flux, and the IRM scram would result in a reactor shutdown well before any Safety Limit is exceeded.

O In order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents has been analyzed. This analysis included starting the accident at various power levels. Ths most severe case involves an initial condition in which the reactor is just suberitical and the IRM system is not yet on scale. This' condition exists at quarter rod density.

Additional conservatism was taken in this analysis by assuming that~

Amendment No. 119 1.1 -

\, _

.l s.

=e*+

6 P Nh*

i - -

DAEC-1 ~

the IRM channel closest to the withdrawn rod is by-passed. The results of this analysis show that the reactor is'scrauned and peak 1- .

power limited to one percent of rated power, thus maintaining MCPR above the Safety Limit. Based on the above analysis, the IRM i

provides protection against local control rod withdrawal errors and

' continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

! 8. Scram and Isolation on Reactor Low Water Level The setpoint for the ' low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. Analyses show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR.

is greater than the Safety Limit in all cases, and system pressure l does not reach the safety valve settings. The scram setting is approximately 21 inches below the normal operating range and is thus adequate to avoid spurious scrams.

, C. Scram - Turbine Stop Valve Closure The turbine stop-valve closure scram anticipates the pressure, I

neutron flux, and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram r.ettin~g at 10 j percent of valve closure, the resultant increase in surface heat. flux

~

Amendment No. )(, g,119 1.1-16 -

, l e

e e ,

e

  • --e * , . _ 9

"" Is __'._**__T.3 Y_ del.Ummm_M_.F _4.g..- _,,4. - . , ,, ,.,_wy-.n.g, g9-n __,--_.g,.s,py,_,_. _qg 9,__%._agsp,.,,,g.,,, ,,,pg.yy.y ,., wyw-+_,, ,y ,g.gw,.,cy,.9,99,9,_9_n_g.,y

DAEC-1 i

is such that MCPR remains above the Safety Limit even during the l ,

worst case transient that asstanes the turbine bypass is closed. This scram is by-passed when core thermal power is below 30 percent of l

rated, as measured by the turbine first stage pressure.

0.

Turbine Control Valve Fast Closure (Loss of Control Oil Pressure Scram)

The control valve fast closure scram is provided to limit the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection. It prevents MCPR from becoming less than the Safety Limit for this transient.

l E. F. and J. Main Steam Line Isolation on Low Pressure, Low Condenser Vacuum, and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 850 psig has been provided to protect against rapid reactor depressurization and '

the resulting rapid cooldown of the vessel. Advantage is taken of -

thescramfeaturethatoccurswhenthemainstbamlineisolation valves are closed, to provide for reactor shutdown so that high power '

operation at low reactor pressure does not occur, thus providing protection for the fuel clad.11ng integrity. Operation of the reactor at pressures lower than 850 psig requires that the reactor mode -

~

switch be in the STARTUP position, where protection of thi fuel cladding integrity safety limit is provided by the IRM and APRM high A

AmendmentNo.JHI,119 1.1-17 l 1

4 s

  • e

--- ,. y4. 9 y7,.-gy- g 9,yp eg .n9c@ w

~rt=+-F --ry,-g , ywr-,*->m-e-#--,,,

. _ . - _ . . ~ _

~

. BAEC-1 L . ,

neutron flux scrams. Thus, the combination of main stee line low pressure isolation and isolation valve closure scram assures Ihe availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In  ;

i

, addition..the isolation valve closure scran anticipates the pressure -

4 .

and flux transients that occur during normal or Insdvertent isolation

' valve closure. With the scrams set at 10 percer.t of valve closure.

j. neutron flux does not increase. To protect the main condenser

~

against overpressure, a loss of condenser vacuun feitiatCs automatic.

closure of the main stean: 1 solation valves. -

G. M. and I. Reactor low Water Leve? Setpoint. for Initiation of HPCI and RCIC, Closing Main Stean Isolation Valves, and Starting _ LPCI and Core Spray Pumps These systems maintain adequate coolant i~a ventory and provide core l cooling with the objective of preventing excessive clad temperatores.

} The design of these systems to adequately perfom the intended .

l function is based on the specified fow level scram setpoint and .

2 ' initiation setpoints. Transient analges demonstrste that these j conditions result in adequate safety mergins for both the fuel ad 4

4 the system pressure,

~

i l

2.1 REFERENCES

l 1. " General Electric Standard A $slication for Reactor Fuel," ,

t NEDE-240ll-P-A* '

2. "Duane Arnold Energy Center Single-Loop Operatios " DEDO-24272, My i

1980. j

  • Approved revision nutber at time analyses are parfarimuf.

Amendnent H3. ) % 119 ,1.148 -

I

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. . b

] ~~

+-

i l$RATEDTHER'CLPbwERo

. ,. a l 1658 Mf6 RATED CC:'.E FLOSi t '

=

' l.

a. e - 1 40 x 10 lb/hr -

120 - . .

1  :

/ "

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/j# APRM FLOW

/ BIAS SCRA!! ,

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- 7 '

,' /

t

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J / NOMINI.L EXPECTED

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D 60 Il g

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40 et0RE THERMAL (ho o

_ POWER IIMIT WHEN REAC' TOR I

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' i i

O I.S I 785 PS1G,' ~~

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OR CORE FLCM i 25 p d10% OF RAT 2Dj

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  • I ,

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/ NATURAL CIRCULATION LINE

  • o -

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a i ,

0 ,_ g 20 40 60 E0 100 1;o CORE PLOW RATE ( % OF RATED) a DUANE ARNOLD ENERGY CEM7fR

' IONA ELECTRIC LICET F. PONER COMPAN e TECHNICAL SPECIFICICIONG - -- -

APRM FLov7 BIAS SCRAM RELATIONSHIP t

TO NORMAL OPF.RATIl1G CONDITIONS (

u FIGUREJ..i-1 a

~

l Amendment No. . 119 '

1.1 19 ,

i 6

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Ct.;ANE MNOt.3 ENE.%C* C2'D7I?.

IOWA E* CCIRIC f IGC G PCWE.R CCM7MY' i TECHNICE SPEC 75CATIWS I

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FIGURE 2.2.-1 .

! ~~~~

i knendment M.119 . ,

1.1-2D .

i a

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4D=---"Gi*M

TABLE 3.1 1 ' '

i

.l REACTOR PROTECTIDff SYSTEM (SCRAM' ) INSTRIENTATION REQu!REMENT '

5 Miniraun k No. of iiodes in Wich

Operable Number of

, Instrurnent functiun Ast be Instresnent

,o Operable Channels Channels - - - - -

Provided by

' for Trip(1 )

System Trip function Refuel Startup Run Design Action-(l)

Trip Level Setting (6) 3 1 Mode Switch in 1

Shutdown X X 1 Itxle Switch A

, (4 sections) 1 Manual Scrant X X X 2 Instr ment A Channels 1Rt1 High Fiux

~

2 i120/125 of Full Scale X X (5) 6 Instrum nt A [

Channels

,40 2 1RM Inoperatlye

{ X X (5) 6 instr went A Chan'nels R APRH High riux for two rectre loop operation X 6 Instranent A or 8 j

< .66WF54)(FRP/MFLPD) (11) (12) Channels or one recirc loop operation 1(.66W&50.5)(FRP/Mr"LPD)(11)(12) 2 APRM Increrative (10) X

, X X 6 Instrement A or 8 Channels

! APRM Downscale ) Indicated on Scale

,,,5 (9) 6 Instrtmest. A or 8 l .

Channels .

2 APRM High Flus < 15% Power X X 5 Instriment

~ A in Startisp Cha;mels 2 II)gh Reactor ~

< 1055 psig X(8) A X 4 Instrweent A Pressure .

Channels .l I -

1 i

?4 TABLE 3.2-C Instr.umentttirn That Initiates Contr:1 Rod Blockh

,l

' ~

Miniman No. kaber of

, of Operable Instrament .

i s Instrueent Channels M Channels Per Provided by j g Trip System Instrcnent Trlp level Setting Design Action g - -

5 2 APRM tipsca.le (Flow Blased) for 2 recirc loop operation

, j @

p 1(0.66 W + 42) (

pD)I) 6 Inst. Channels (1) 3 for 1 rectre loop operation e

f(0.66 W + 't8.5) (pp p h 2 APRM Upscale (Not in Run Pbit) i12indicatedonscale 6 Inst. Channels (1) 2 APRN onwiscale 15 indicated on stale 6 Inst. Channels (1) ,

w 1(7) Rod Block Monitor for 2 recirc loop operation
l 7 (Flow Blased) FRP 2 inst. Channels g, ,39)(WLPD)(2) (1) 5 for 1 recirc loop operation RP

~ 1(0.66 W + 35.5) ( p9)(2) 1 (7) Rod Block Monitor Daatscale 15 indicated on scale 2 Inst. Channals (1) 2 IRM 60wnscale (3) . 15/125 full scale 6 last. Channels (1) .

2 IRM Detector not in (8) 6. Inst. Chancels (1)

Startup Position g y 8

,, - 2 , IRM Opscale i108/125 6 Inst. Channels (1) 2(5) -

SRM Dettetor not in (4) 4 Inst. Channels (1)

Startup Tbsition 5

2(5)(6) SitM Upscale i10 counts /sec. 4 Inst. Channels '(1) <

' 1 Scrm Discharge Volune .< 24 gallons 1 Inst. Chanwel (9)

Water Level-High i

, , DAEC-1 LIMITING CONDITIONS FOR OPERAff0N , SURVEILLANCE REQUIREMENT 3.3.0 Reactivity Anomalies 4.3.0 Reactivity Anomalies s.

The reactivity equivalent of '

During the startup test program the difference between the and startup following refueling actual critical rod outages, the critical rod configuration and the configurations wilt be compared expected configuration to the expected configurations during power operation shall at sele:ted operating not exceed 15 A k. If this conditions. These comparisons . l limit is exceeded, the will be used as base data for reactor will be shut down reactivity monitoring during

.until the cause has been subsequent power operation determined and corrective throughout the fuel cycle. At actions have been taken as '

specific power operating i

appropriate. conditions, the critical rod configuration will be compared to the configuration expected based upon appropriately corrected past data. This comparison will be made at least every full power month. l E. Recirculation Punps E. Recirculation Pump 5 When the reactor mode switch With two recirculation pumps in is in startup or run operation and with core thermal positfoc, the reactor shall power greater than the limit not be operated in the specified in Figure 3.3-1 and

' natural circulation flow total core flow less that 4W mode. . of rated, establish taselire APRM and LPRM* neutron flux -

With te recirculation pianos noise levels within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in operation and with core provided that baseline values thermal power greater than have not been previously. l the limit specified in estabitshed since the last core Figure 3.3-1 and total core l refueling. i

  • flow less than 451C of rated, the APRM and LPRM* neutron flux coise levels shall be
  • detemiced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, l and:

, I l

1

'a) if the APRM and LPRM* f '

i 1-neutron flux noise '

1 1

  • Detector levels A arid C of one -

1.PRN string per core octant pics detector levels A and C of one LPP,M <

i string in the center of the core 1 l , shall be monitoned. -

e l

/imendment No.)% )( 119 3.34 i

'. l

.;; z7;L

_ = _: h . :== a

, , D.iEC-1

. LIMITING C_0N0!TIONS FOR OPERATION  ! _ SURVEILLANCE REQUIREMENT _.

levels are less than or equal to three times ..

their established i)aseline levels, continue to determine the noise levels at .

least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> x. '

ad also within 30

afnetes af ter the ,

completion of a core '

thermal cover increase  ;

of at least 5% of rated t -

core thereal power while operating in this region of the power / flow map.

or b) if the APRM and/or < ,

LPRM* neutron flux noise-levels are greater thin three times their i

established baseline levels,1spediately initfate corrective action and restore the noise levels to within the required limfts '

within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow and/or by initiat1n ,an

  • orderly reduction o -

core thermal power by  ;

, inserting control rods.

see Specifications 3.6.F.2

'for operation with one -

recirculation loop not in operation. -

A recirculation pump shall not be started d ile the reactor is in natural

~ circulation flow and reactor 1' power is greater than 1% of -

rated thermal power.

F. If Specifications 3.3.A through D above cannot be i met an orderly shutdown shall be initiated and the reactor shall be in the COLD SHUTDOWN condition within 24 *

hourss
  • Detector levels A and C of one LPRM string per core octant plus
  • detector levels A and C of one '.PRM string in the center of the core l l shall be monitored. . -

i l

Amendment No.119 3.3-7a

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1 After initial fuel loading end subsequent refuelings when operating abovee950 psig, all control rods shall be scram tested within the constraints imposed by the Technical. Specifications and before the 40% power level is reached. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

4 Reactivity Anomalies i

During each fuel cycle excets operative reactivity varies as fael deplet.es and as any burnable poison-in tupplementary control is burned. The magnitude of

(

this excess reactivity raay be inferred frem the critical rod configuration. As fuel burnup progresses, anomalous behavior in the eAcess reactivity may be i

detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating baf e conditions provide the most sensitive and directly interpretable data relative to core i reactivity. Furthermore, using power operating base conditions perrlits frequent reactivity comparisons. '

E.equiring a reactivity comparison at the specified frequency assures that a camparison will be made before the core' reactivity change'8Ceeds 1% AK.

Deviations in cora reactivity greater than 1% 4K are not espected and require j

thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients

~

i.

exceeding desigo conditions of the recctor system.

, 5. Recirculation Pumps I

APRM and/or LPRM oscillations in excess of those specified in section 3.3.E

, v could be an indication that a condition of thermal hydraulic instabil,ity exists and that appropriate remedial action should be taken. These specifications are based upon the~~ guidance of GE SIL #380, Rev.1, 2/10/84.

i Amendment tio.JWM,119 3.3-1B  ;

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60 per 3.3.E or 3.6.F.2.b r 50 .

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20 30 I 40 . 50 60 70 80

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Total Core Flow (% of Edted) ,,

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i DAEC, l ..

Iowa Electric Light *& Power Company Technical Speckfications Thermal Power vs Cbre Flow Limits for Thermal Hydraulic Stability Survef1Tunce Figure 3.3-1 Amendment ~No. 119

~

3.3-20

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' OAEC-1 f ~

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT _

b. The f adicated value of cor.e flow rate varies from the value derived form loop flow

. measurements by more than 10%.

c. The diffuser to lower plenum differential pressure 4

readino on an individual .iet pump varies from the mean of all ,iet pump differential pressures'by more than 10%.

2. Whenever there is recirculation flow from the reactor in the Startup or i Run mode, and one recirculation pump is operating, the diffuser to lower plena differential pressure shall be checked daily and the differential pressure of an individual

,iet pump in a loop shall not vary from the mean' of all

.iet omp differential pressures in that loop by more than 10i.

F. Jet Pump Flow Mismatch F. Jet Pump Flow Mismatch

1. When ,both recirculation 1. Recirculation pump speeds peps are in steady state shall be checked and lopaed at operation, the speed of the least once per day.

faster pump may not exceed 122% of the speed of the 2. a. Prior to operation with one

. slower pump when core power recirculation pump not in is 80% or more of rated operation and core thermal power or 135% of the speed power areater than the limit of the slower p e p when specified in Fioure 3.3-1, f

core power is below'80% of establish baseline APRM and rated power. . LPRM* neutron flux noise levels, provided that

2. If Specification 3.6.F.1 baseline values have not cannot be met, one recirculation pep shall be tripped. The reactor may
  • Detector levels A and C of one

~

be started and operated, or LPRM strino per core octant. plus operation may continue with detector levels A and C of.one LPRM one recirculation loop not string in the center of the core in operation provided shall be monitored.

that:

^

Amendment No.frf,119 3.6-7 ,

L' --

k:

  • DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT
a. MAPLHGR multipliers as beenpreviouslyestablishiif indicated in section since the last core 3.12.A are applied. refuelino. Baseline values shall be established with
b. With one recirculation one recirculation puno not pump not in oneration in operation and core
'. and core thermal power thermal power less than or areater than the limit ecual to the limit specified specified in Fioure in Floure 3.3-1.

1 3.3-1, core flow must be creater than or equal to b. Prior to operation with one 39% of rated, and recirculation punp not in operation and core flow .

(1) the Surveillance creater than 45% of rated, Re 4.quirements establish baseline core 6.F.2.a haveofnot been plate AP noise levels with satisfied, immediately core flow less than or equal initiate action to , provided reduce core thermal to that 45% of rated, baseline values have power to less than or not been previously ecual to the limit established with one ,

specified in Figure 3.3-1 recirculation pump not in within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or operation since the last core refueling.

(ii) the Surveillance Recuirements of 4.6.F.2.a have been satified, continue to determine the APRM and LPRM neutron flux levels -

at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after the completion of a core -

. thermal power increase of at least 5% of rated core thermal power while operatino in this reofon of the power / flow map.

If the APRM and/or LPRM*

  • neutron flux noise levels are areater than three times their established baseline values immediately infliate corrective action and restore the noise levels to within the recuired limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by
  • Detector levels A and C of one LPRM string per core octant plus
detector levels A and C of one LPRM .

string in the center of the core -

i shall be monitored. -

Amendment No. 119 3.6-7a l .

l .

O .E...~~. r T - _. -

, DAEC-1

. . )

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT increasino core flow and/or initiatino an orderly reduction s. 1 of core thermal power by insertino control rods.

c. With one recirculation punp not '

in operation and core flow areater than 45% of rated, and (1) 'the' Surveillance Requirements of 4.6.F.2.b have not been satisfied, immediately initiate action to reduce core flow to less than or eoual to 45% of rated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or (ii) the Surveillance Requirements of 4.6.F.2.b have been satisfied, continue to -

determine core plate AP noise at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after the

, completion of a core thermal power increase of at least 5% of rated thermal power. If the core plate AP noise level is creater than 1.0 psi and 2 times its established baseline value, innediately initiate corrective action an'd restore the noise levels to within the recuired limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by *

~ decreasino core flow and/or initiatino an orderly reduction of core thermal power by  ;

insertino control rods.

d. The idle loop is isolated i electrically by disconnectino ,
  • the breaker to the recirculation pumo motor oenerator (M/G) set drive motor prior to startuo, or if disabled during reactor operation, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Refer to Specification 3.6.A for startuo of the idle recirculation 1o00,

e. The recirculation system ~

controls will be placed in the ,

manual flow control mode.

Amendment No. 119 3.6-7b ,

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DAEC-1

c. The jet punp flow deviation pattern derived from the diffuser to lower plenun differential pressure readings will be used to further evaluate jet punp operability in the event that the jet punps fail the tests in Section 4.6.E.1 and 2. l Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive jet punps. This bypass flow is reverse with respect to normal jet punp flow. The indicated total core flow is a sunmation of the flow indications for the sixteen individual jet punps. The total core flow measuring instrunentation suns reverse jet punp flow as though it were forward' flow in the case of a failed jet punp. Thus the indicated flow is higher than actual core flow by at least twice the nonnal flow through any'backflowing jet punp.*

Reactivity inventory is known to a high degree of confidence so that ,

even if a jet punp failure occurred during a shutdown period, subsequent power ascension would promptly denonstrate abnormal control rod ' withdrawal for any power-flow operating map point.

O i

t A nozzle-riser systen failure could also generate the coincident failure of a. jet pump body; however, the converse is not true.

  • Note: In the case of single recirculation loop operation, when the recirculation pump is tripped, the flow through the inactive jet pumps is subtracted from the total jet pump flow, yielding the correct value for. the total core flow.

~

3.6-31 AmendmentNo.f.119 .

= * * - - -* ' - . . -#

'_ __ -. _ , , - ,r.--,-----er=~=- -*--+-ww'+ * **e' *-*" "' " ' ' ' ""** * - ' "

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DAEC-1 80% power cases, respectively. If the reactor is operating on one pump, the loop select logic trips that ptsap before making the loop selection. ..

An evaluation has been provided for ECCS performance during reactor operation with one recirculation loop not in operation (Sec. 3s12, Ref.11). Th'erefore, continuous operation under such conditions is appropriate. The reactor may also be operated up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one recirculation loop not in operation without electrically isolating the idle loop. This short period of time permits corrective action to be taken to re-activate the idle loop or to implement the changes for continuous operation with one recirculation loop not in operation.

  • During periods of Single Loop Operation (SLO), the idle recirculation loop is isolated by electrically disarming the recirculation pump. This is done to prevent a cold water injection transient caused by an inadvertent pump start-up.-

It is permissible to leave the suction and discharge valves open during SLO to allow flow thrcugh the loop in order to maintain the temperature.

However, if for some reason the discharge valve is inoperable it should be closed and electrically disarmed. This is done to prevent degradation of LPCI flow during a LOCA. With the discharge valve disarmed, the temperature in the loop can be maintained by opening the bypass valve, as the loop selection logic will close the bypass valve, isolating the loop, prior to opening the LPCI injection valve. ~

i APRM and/or LPRM oscillations in excess of those specified in Section 3.6.F.2 could be an indication that a condition of thermal hydraulic instability exists and that appropriate remedial action should be taken. By restricting core flow to greater than or equal to 39% of rated, whi.ch corresponds to the core flow at the 80% rodline with 2 recirculation pumps running at minimum speed, the region of the power / flow map where these oscillations are most.

likely to occur is avoided.

Individual APRM or LPRM channels exhibiting excessive flux noise may be discounted upon verification that a true condition of thermal hydraulic instability does not exist by observation of the

. remaining available APRM and/or LPRM channels. These speci'ications are based upon the guidance of GE SIL #380, Rev. 1, 2/10/84.

Above 45% of rated core flow in SLO there is the potential to set up high flow-induced noise in the core. Thus, surveillance of core plate AP noise is required in this region of the power / flow map to alert the operators to take appropriate remedial action if such a condition exists. ,

AmendmentNo.[,119 3.6-34 .

" ~e's65M*w yge4% P-@v

< DAEC-1 Requiring the discharge valve of the lower speed loop to renain closed until the speed of faster punp is below 50% of its rated speed provides assurance i

when going from one to two punp operation that excessive vibration of the jet punp risers will not occur.

3.6.G & 4.6.G BASES:

f' REACTOR COOLANT SYSTEM j Structural Integrity A pre-service inspection of Nuclear Class I Components was conducted to assure freedom from defects greater than code allowance; in addition, this served as a reference base for future inspections. Prior to operation, the reactor coolant system as described in Article 15-120 of Section XI of the ASME Boiler and Pressure Vessel Code was inspected to provide assurance that the system was free of gross defects. In addition, the facility was designed such that gross defects should not occur throughout plant life.

, The pre-service inspection progran was based on the 1970 Section XI of the

.ASME Code for in-service inspection. This inspection plan was designed to reveal problem areas ,(should they occur) before a leak in the coolant system could develop. The ' program was established to provide reasonable assurance that no LOCA would occur at the OAEC as a result of leakage or breach 'of -

pressure-containing components and piping of the reactor coolant system, portions of the ECCS, and portions of the reactor coolant associated auxiliary systens.

. l A pre-service inspection was not performed on Nuclear Class II Components because it was not required at that stage of DAEC construction when it would have been used. For these components, shop and in-plant examination records l of components and welds will be used as a basis for comparison with in-service inspection data.

Amendment No. [ 119 3.5-35 eg eg w ce re, -r-- .v--- .- y 9 -

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DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

. , 3.12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS

, Applicability Applicability i The Limiting Conditions for The Surveillance Requirements Operation associated with the app y to the parameters which fuel rods app y to those mon tor the fuel rod operat,ing parameters wh ch monitor the fuel conditions.

rod operating conditions.

Objective Objective The Objective of the Limiting The Objective of the Surveillance Conditions for Operation is to Requirements is to specify the assure the type and frequency of fuel rods. perfomance of the -

I surveillancento be applied to the

! fuel rods.

Specifications Specifications

~

A. Maximum Average Planar Linear. A. -Maximum Average Planar Linear neat uenerauon nate trwtnunJ neat ueneration Mate (martnux)

During reactor ower operation, The MAPLHGR for each type of fuel the actual MAPL GR for each type as a function of average planar of fuel as a function of average exposure shall be determined planar exposure shall not exceed dail during reactor operation at the limiting value shown in Figs. > 25 rated thermal ower and any 3.12-5 7 -8 and -9. For Dange in power leve or sin le, loop, the values by ' distribution, that would cause in hese curvesope, are reration,duced operation with a limiting control multip1 ng by 0.87. If at any rod pattern as described in the i

. time du n reactor power bases for specification 3.3.2.

operation one or two loo

>25% rated thermal' power,p? During operation with a limiting tt isat cont;'ol rod pattern, the MAPLHGR Tetermined normal s'urveillance LAPLHGR) shall be determined at that the li tin value for east once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. -

MAPLHGR (LAPLHGR is being exceeded, action shall then be initiated within 15 minutes to Y restore operation to within the i prescribed limits. If the MAPLHGR -(LAPLHGR) is not returned 1 to within the prescribed limits .'

within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce reactor power to < 25% of rated thermal power, or"~to such a power level that the limits are again bein met, within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. g If the reactor is being operated with one recirculation loop not i

in operation and cannot be returned to within prescribed limits within this 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, the reactor shall be brought tu -

the COLD SHUTOOWN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

For either the one or two loop l operating condition surveillance and corresponding action shall continue until the prescribed limits are again being met. ,

AmendmentNo.)Hf,J,W,J.W,'119- .

+ - - , "

_e

g,Jgf-)j 1

, LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT B. Linear Heat Generation Rate B.

(LHGR)

_ Linear Heat Generation Rate (LNGR)

1. During reactor power operation the linear heat The LHGR as a function of core height shall be checked daily generation rate (LHGR) of any during reactor operation at > 25%

rod in any P8X8R, BP8X8R or ELTA fuel assembly shall not . thermal power and following any change in power level or

, exceed 13.4 KW/ft, while the LHGR of any rod in an LTA-311 distribution that would cause fuel assembly shall not operation with a limiting control rod pattern as described in the exceed 14.4 KW/ft. bases for Specification 3.3.2.

If at any time during reactor During operation with a limiting control rod pattern the LHGR power operation at >25% rated thermal power it is- shall be determined at least once

- per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall then be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce reactor power to < 25% of Rated Thermal Power, or to such a power level that the limits are again being mrt. within the -

next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. sarveillance s and corresponding action shall continue until the prescribed limits are again being met.

O

(

  • Amendment No.) 87 M 119 3.1E-2

[ . - _ _ _

e LIMITING CONDITIONS FOR OPERATION SURVEILLANCEREQUlftEMENTS C. Minimum Critical Power Ratio C. Minimum Critical Power Ratio (MCPR) gg -

  • During reactor power operation, MCPR shall be determined daily MCPR for one ar two during reactor power operation recirculation loop operation at > 25% rated thermal power shall be > values as indicated and Tollowing any change in in Figures 3.12-2 and -3. power level or distribution These values are multiplied 'Jy that would cause operation with

?

Kf which is shown in Figure a limiting control rod pattern 3,12-1. Note that for one recirculation loop operation as described in the bases for Specification 3.3.2. During the MCPR limits at rated flow operation with a limiting are 0.03 higher than the control rod pattern, the MCPR comparable two-loop values. If at any. time during reactor shall be determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

power operation (one or two loop) at >25% rated thermal power, it is detennined by normal surveillance that the limiting value for MCPR is being exceeded, action shall then be initiated within 15 minutes to restore operation t) within the prescribed limits.

If the operating MCPR is not returned to within the prescribed limits within two hours, reduce reactor power to

< 25% of rated thermal power, or to such a power level that -

the limits are again being tret, within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. '

If the reactor is being operated with one recirculation o

loop not in operation, and cannot be returned to within prescribed limits within this 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, the reactor shall be brought to a COLD SHUTOOWN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

For either the one or two loop operating condition surveillance and corresponding action shall continue until the prescribed limits are again -

being met. .

1 Amendment No.g.JNK 119 3.12-3 1

l ..

3 .

DAEC-1

~

3.12 BASES: CORE THERMAL LIMI"5 "-

A. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident (LOCA) will not exceed the limit specified in 10CFR50.46. LOCA analyses are performed using General Electric calculational models which conform to the requirements of 10 CFR Part 50, Appendix K.

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all rods of a fuel assembly at any a'xial location and is only dependent secondarily on the rod to rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than + 20*F relative to the peak

-temperature for a typical fuel design, the limit on tb average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10CFR50.46 limit.

For two recirculation loop operation, the calculational procedures used to e'stablish the MAPLHGR's shown on Figures 3.12-5 thru 3.12-9 are documented in Reference 7 The reduction factors for one recirculation loop operation were derived in Reference 13.

Amendment No. M 119 3.12-4 -

S A m,. - ~ ~ ~

DAEC-1, s.

2.

MCPR Limits for Core Flows Other than Rated Flow The purpose of the K f factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCPR is the product of the operating limit MCPR and the Kf factor. Specifically, the

  • Kf factor provides the required thermal margin to protect against a flow increase transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speed up caused by a motor-generator speed control failure.

For operation in the automatic flow control mode, the K f factors assure that the operating limit MCPR of values as indicated in Figures 3.12-2 and 3.12-3 will not be violated should the most limiting transient occur at less than rated flow. In the manual flow control mode, the K f factors assure that the Safety Limit MCPR will not be violated for the same postulated l transient event.

s The K f factor curves shown in Figure 3.12-1 were developed generically and are applica'ble to all BWR/2, BWR/3 and BWR/4 reactors. The K f factors were derived using the flow control line corresponding to rated thermal power at rated core flow, as described in Reference 2.

The K f factors shown in Figure 3.12-1 are conservative for Duane Arnold operation because the operating limit MCPR of values as indicated in Figures 3.12-2and3.12-3isgreaterthantheoriginal1.20operatinglimikMCPR used for the generic derivation of Kf .

Amendment No.f,JHf' .llg 3.12-7 t

. DAEC-1 4.12 BASES: CORE THERMAL LIMITS s-l C. Minimum Critical Power Ratio (MCPR) - Surveillance Requirement.

At core thermal power levels less than or equal to 25%, the reactor wi,ll be operating at minima recirculation pisnp speed and the moderator void content will be very snall. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin.

With this low void content, any inadvertent core flow increase would only place operation in a more conservative state, relative to MCPR. During initial start up testing of the plant, a MCP'R evaluation will be made at 25%

thermal power level with minimum recirculation pisnp speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been s'ignificant powe'r or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached assures that MCPR will be known following a change in power or pcwer shape (regardless .of magnitude) that could place operation at a thermal limit.

Amendment No.119 3.12-8 -

9 e

- * - t l

, ,, - _ , . ,.- . - ' - - ' - " ~ ~ ~'

  • , DAEC-1 3.12 REFERENCES  !
1. Duane knold Energy Center loss-of-Coolant Accident Analysis Report, NED0-21082-03, June 1984.
2. General Electric Standard Application for Reactor Fuel. NEDE-24011-P-A**.
3. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplanents 6, 7, t.nd 8. NEDM-19735, August 1973.
4. Supplement I to Technical Reports on Densifications of General Electric -

Reactor Fuels, December 14,1973 (AEC Regulatory Staff).

5. Comunication: V.A. Moore to I.S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27,1974. ,
6. R.B. Linford, Analytical Methods of Plant Trans'ient Evaluations for the GE BWR, February 1973 (NEDO-10802).
7. General Electric. Company Analytical Model for Loss-of-Coolant Analysis .in
Accordance with.10CFR50, Appendix K. NEDE .20566, August 1974. '
8. Boiling Water Reactor Reload-3 Licensing Anendment for Duane knold Energy Center, NEDO-24087. 77 NED 359, C1 ass 1. December 1977. .
9. Boiling Water Reactor Reload-3 Licensing Anendment for Duane ' Arnold Energy Center, Supplement 2: Revised Fuel Loading Accident Analysis, NEDO-24087-2.
10. Boiling Water Reactor Reload-3 Licensing Anendment for Duane knold Energy Center, Supplement 5: Revised Operating Limits for Loss of .

Feedwater Heating, NEDO-24987-5.

11. Letter, R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request for Information on ODYN Computer Model," September 5,1980.
12. Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment Methods for Determination of Operating Limits," January 19, 1981.
13. Duane knold Energy Center Single Loop Operation, NEDO-24272, July 1980. l
    • Approved revision nunber at time reload fuel analyses are performed.

i Amendment No g ,J W, 119 3.12-9 l

4i

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K g FACTOR - '

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3 .

AUTONATIC FLOW CONTROL t . ,

1 1.2 . l*

. H M

i 1.1 -

MANUAL FLOW CONTROL -.

w Scoop-Tube Set-P; int Calibration

- positioned such that -

Y 5 Flowmax = 102.5%

= 107.04 1.0 -

.= 112.0% *

= 117.0%

j

  • I I I I e I t' 30 40 50 60 , 70 80 90 100

, CORE FLON, t 4- , DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT & POWER CONPANY t

TECHNICAL SPECIFICATIONS i

K g AS A FUNCTION OF CORE FLOW FIGURE 3.12-1

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DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS MINIMUM CRITICAL POWER RATIO (MCPR)  !

VERSUS T l

FUEL TYPES: BP/P8X8R.and ELTA FIGURE 3.12-2 Amendment No.) #,) Y/, M 119 3.12-11 k

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TECHNICAL . SPECIFICATIONS i

MINIMUM CRITICAL POWER RATIO (MCPR) 1 VERSUS r '

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JJ When. core flow is equal to or less than 70% of rated, the MAPLHGR shall not exceed 95% of the limiting values shown.

  • 1 GWd/t = 1000 mwd /t

- i DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY i TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE- i

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1/ When core flow is equal to or less than 70% of rated, the MAPLHGR shall not exceed 95% of the limiting values shown. Values shown'are for two recirculation loops.

loop operation are given in Section 3.12.A. Reductions factors for one recirculatio

  • 1 GWd/t = 1000 mwd /t DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE A$ A FUNCTION OF PLANAR AVERAGE EXPOS'URE' FUEL TYPE: 8P/P80RB301L FIGURE 3.12-6 Amendment No. g g , 119 3.12-15 -

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-1/ When core flow is equal to or less than 70% of rated, the MAPLHGR shall not exceed 95% of the limiting values shown. Values shown.are for two recirculation loops.

loop operation are given in Section 3.12.A. Reduction factors for one recirculation

  • 1 GWd/t = 1000 mwd /t '

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DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANA,R AVERAGE EXPOSURE' FUEL TYPE: P80P8289 FIGURE 3.12-7 Amendment No. M 119 3.12-16 -

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icop operation are given til Section Reduction factors for ELTA).

3.12. A.,4except one recirculation

  • 1 Gud/t = 1000 fwd /t DUAtiE ARNOLO ENERGY CENTER ICVA El,ECTRIC LIGiT AND POWER COMPANY TECH:ilCAL SPECIFICATIONS l LIMITlHG AVfMGE PLANAR LINEAR HEAT GENERATION t? ATE AS A FUf.CTION OF FLANA: Ayt,VGE EXPOSURE' '

FbEL TYPE: 9P/PORB299 and ELTA i

FIGURE 3.12-8 Amendment No.J W,)ts,) W, y , 119 3. 2-17 ,

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loop operat. ion are given in Section 3.12. A.Redu:tton  %

factors for one recirculatio

  • 1 GWd/t = 1000 mwd /t

-e DUANE ARNOLD ENERGY CENTER ICWA ELECTRIC LIGHT AND POWEG COMPANY TECHNICAL SPECIFICATICNE LIMIT]NG AVERAGE PLANAR LINEAR HEAT GENERATION PATE AS A FUNCTION OF PLANAR AVERAGE EXP050RE' FUEL TYPE: P80RB284H FIGURE 3~.1'2-9 Amendment No. M 119 3.12 18 '

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SAFETY EVAL @ TION BY THE OFFICE OF NUCLEAR REA SUPP0911NG AMEMOWEMT NO.119 TO LICENSE' NO. DPR-49 -

I -

i TOWA ELECTRIC LIGHT A O POWER COMPANY LENTPAL IOWA POWER E% PERATIVE JORM DELT POWER CUWEPATIVC DUANE ARNOLD ENERGY CENTER DOCET Wo. 50-331 1.0 Ih7R000CT10N By a letter dated . December 7,1984, the Iowa Electric Light and Power Compan (CAEC)y (the licensee) requested cha es to the Duane Arnold Energy Center Technical Specifications to (1 pemit reactor oper.tico with one recirculation loop out of service, (2 to include General Electric ,

Company's (GE) Service Information letter (SIL) No. 380, Revision 1  ;

recoca:endatioirs regarding thermal-hydraulic stability conceres for dual  ;

loop and single lecp operations, and (3) to incorporate administrative -

changes dealing with updating references and deleting blank pages.

' Crasently, the UAEC opcrating license requires a unit to be in cold shutdown within the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if an idle recirculation loco carmot be ,

rtturned to servica within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee previcusly requested

  • e euthorization for unlimited single loop cperation of DAEC. Subsequently, i Tennessee Valley Authority's operation of Browns Ferry Unit 1 (a boiling water ranctor sirtlar in design to DAEC) in the single loop vede of  ;

operatica at 59% power lead to cor,cerns related to thermal-hydraulic instability, GE, in SIL No. 330. Revision 1

' . providing the boilirg water reactor licensees, addressed generic guidancethese concerns by for actions which tuspress themal-hydraulic instability indaced neutron flux estillations. The licensee has proposed Technical Specificatiocs in accordance with the guidance provided by GE in SIL No, 380, Revision 1.

Specifically, the preposed chaages requeste1 by the Itcensee consist of (1) deletion of the license condition restricting the single loop operation.

Thf s invcives a revision of the Technicat Spcifications for Average Power Range Monitor (APRM) flux scram trip and rad block . settings, an increase in the safety liatt Minimum Critical Power Ratio (MCAR) value, and a revision to the allowable Average Planer.Lineer Heat Generation Rate (APUGR) values;(2)forsingleandduallocpt,peration,incorporat'ing requirentnts in the Technical Specifications to detect themal-hydraulic i instabilities induced neutron flux oscillatiocs an.1 specifying operator response to the detected instabilities; and (3) g9 dating of some references 1

and deletibu of some blank pages.  :

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2.0 EVALUATION We have evaluated the licensae's proposa? to pemit unlimited operatick cf the CAEC with one recirculation loop cut of service, incorporate the GE SIL Nc. 380, Revisico 1 guidanca regarding them,al hydraulic instabilities are!

1mplement sore ad:ninistrativa changes in the DAEC Technical Speciftettfor,s.

2.1 Sing *e Loop _ Operatics Wa have reviewed the licensee's analysis of accidents and transients which are judged loop mode. to be affected by operatic: of the GAEC in the sin 31e operating

  • The accidents and transients which are of ccncern relate to inadvertest variations in .the cociant flow t5rcugh the core and design bases cf the fuel performance safety limits. The events evaluated include
  • Cee Puap Seizure Accident, Idle Lcop Startup Event, Acd Withdrawal Error Event, and toss-of-Cocient Accident, i

One Pump Stirare Accident, Tt.e Itcensae states that the one pump seizure accident is a relatively mild event durirg tv5 recirculatf :n pump operaticn. Similar analyses were performed to deterxine the impact this accident would have en one recirculation prp operation. These analyses were perfomed usicg NRC approved codels for a large core BM/4 plant, The analyses rere conducted fract steady-state operation at the following initial conditions, with the '

added cccditien of one inactive recirculaticn lot,p. Two sets of init.ial condJtions assumed were:

t.

b. Thems1 Power = 75" and core 71ev = 55% cf rated Thermal hwer
  • 82% ,and ccre ficw = 56% af rated These condi.tions were chosen because they represent raasenabla upper Timits a of single loop cperation within existing Maxixm Average Planar Linear Heat Generation Pate (MAPLNER) and Minisv3 Critical Power Patia limits at the sar:e maximun pwirp speed. Pus 3p seizure was sinulated by setting the single operating pmp speed to zero ir.stanttneously, t

The anticipated scocaca of evets folicwing a recirculation pump set ure which cccars durin.3 plant cperation with the alterraate recirculation loop out of service is as fo11cws: ,

e. The racirculatico loop flo.< in the ir,op in wMch the pump seizure

. occurs drops instantaneously to zero.

2:.

Core voids incresse which result in a negative reactivity insertion and a sharp decrease in re'.; tron flux. ,

c. West flux dro
d. Neutrcn flux,ps roreflux, heat sicwly because ructor wateroflevel, the faal timeflow, steam constant, and feedwater flow all e.xhibit transient behaviors. However, it is not enticipated that the increase in water 1tvel will cause a turbine trip +

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1 it is expected thit the transient will terminate at a condition of naturat circulation and reactor operation will continue. There will also be a small decrease in systan pressure.

The licensee concludes that MCPR for the pump seizure accident for the large core INR/4 plant was determined to be greater than the fuel cledding integrity safet limit; therefore, no fuel failures wera postulated to j

occur as a resu t of this analyzed event, These results are applicable to DAEC, and were obtained with the staff approved methodology. We, therefore, agree with the iicensee's conciosion that fuel cladding j integrity safety margins will oct be reduced.

Idle Loop Startup I

The idle loop startup transient was analyzed, lo the OAIC Final Safety Analysis Report (F5AR) for dual loop operation, For single loop operation, the. licensee proposed to increate the rated condition steady-state MCFR '

i limit by 0.01 to account for increased uncertainties in the cora total flow i; and Traversing In-core Probe (TIP) readings. The staff four,d the MCPR increase of 0.01 to be acceptable, but suggested that the licensee conservatively increase the MCPR by 0.03. The licensee agreed and has now i

' prcposed the MCPR increase of 0.03. The MCPR will also vary depending on l'

flow conditipns, This leads to the possibility of a large inadvertent flow increase which could cause the MCPR to decrease telow the Safety !.imit for a low initial MCPR at reduced flow conditions. Therefore, the required MCPR w:ust be increased at 'raduced core flow by a flow factor, K f derived by assuming both recirculation loops in:rease speed to the maximum pemf ttad .

by the sceop tute position set screws. This condition maximizes tha power

-increase and hence theAMCPR for transients initiated from less than rated conditions.

i When operating on one loop the flow and power increase will be less than associated with two rumps increasing speed, therefore, the K, i

factors loop der.ived from the two-pump assur.ption are conservative for single operation. '

~

f Rod Withdrawal Error t

The rod withdrawal error at rated power is given in the FSAP, for the initial core tM in cycle dependent reload supplemental submittals. Thase Analyses are.perfereed to demonstrate that, eVen if the operator ignores

' all instrument indicatlor.s and the alarm which could occur during the '

i course of the transients, the rod block system will stcp rod withdrawal ,

)

at a sinfaun critical power ratio which is higher than the fuel cladding inte"rity sifety limit. The proposed correction of the rod block equation l' end Iower initial power for single loop operation will assure that .the MCPR i

saf. sty limit will not be violated.

j One pump operation resuh: in backflow through 8 of the 16 jet pumps while flow is baing supplied to the lower plenum frorr the active jet pumps. (

i Aecause of thl5 backflow through the inactive jet pumps the oresent rod-block ewation and APRM settings must be modified. The licensee has preposed modifftd rod block equation and APRM settings in the Technical

  • a

4 i

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5pecificatioa for cne pump operation, and the staff has found then acceptable. ,

i toss-of-Coolant Accident (LOCA)

The licensee has contracted General Electric Company (GE) to perform single loop operation analysis for CAEC LOCA. The licenses states that evaluation of these calculations (that are performed according to the procedure outlined in .NE00-20566-2, Rev.1 indicates that a multiplier of 0.86 (? X 7 fuel'), 0.87 (8 X 8 fuel) 0.87 (8 X 83 fuel) (Ref.: NEDE-24272, July 1980))

should be applied to the MAPLHGR limits for single loop cperation of the DAEC.

We find the use of MAPLHGP. Tultipliers as indicated will be adequate to offset LOCA consequences in the single loop operation mode'. The MAPLHGR fact 6rs are, therefore, acceptable.

2.2 Themal-Hydraulic Stabi1_ity in Dual and $1nole Loop Operation We have etaiuated tre licensee's pecposed Technical Specification changes te assure Fat the changes provide adequate detection and suppression of potential themal-hydrauli.: inttabili ties.

EE recently presented the staff with stetrilf ty test data which demonstrated the occurrer.:e of ?!ait cycle neutron flux oscillatforts et nttural circulation and several percent above the rated rod line. The oscillations

~

werc ob:ervable on the APRMs and were suppressed wit'i cuntrol rod insertion. It was prCdicted that limit cycia asci11 4tions would occur .

It the operating condition tested; however, the r.haracteristics of the

( obstryed oscillations were different freet those prepiously observed during l other stability tests. Ma:nely, the test dtta showed that sota LPRM indications oscillated out of phase with the APR4 signal and at asplitude as great as six times the core averace. GE has prepared and released a

> o sarvic.a information letter. SIL No. 580, to alert the BWR owne.'s of these new dEta and to recomend acticos to ayofd and control abnormal neutron flux oscillatictis.

The General Electric recomendations were reviewed by the staff and found to be pr:.: dent recomendcticas which provide adequeta dettetion and suppression of potential therral-hyd:atlic instabilities as required by General Design Criteria (CD") 10 and 12. The staff crepared these retctnendations with the DAE" TecFnical Specifications for operation with a recir:ulation icop out of service anc feued that the proposed thanges are in comfoma.1ce with the Sil No. 380, Revision 1 recomendations end aru ecceptable to the staff. -

\ .

In adcition, on February a,1985 a single loop test was performed by i Tent.esset. Valley Authority (TVA) on its 3rowns Ferry Unit i reactor during which thermsi-hydraulic stab}11ty decay ratios were measured. The main

, findings cf the test were that the. observed increase in neutron noite ,

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during single loop operation (SLO) is solely ke to an increase in flow ' '

i noise because the inlet flow to power transfer functions during two Incp l operation (TLO) and SLO are not significantly different when test platears j kith similar power and flow conditions are compared. The Browns Ferry Unit i

' 1 reactor has been found to be stable in all Qs of opteation attained dur!ng the present tests. The most unstable test plateaus corresponded to minmum recirculation pump speed in SLO, which has the minimum flow and .

maximum power to flow ratio. The estimated decay ratio at this plateau was >

0.53. The decay ratio decreased as the flow was increased during SLO (down >

F to 0.'J4), This implies that the core-wide reacttr stability follows the l same trends fr. SLO as it does in TLO. Finally' no local or higher mode

~ .

l instabilities were found in the data teken from local power range monitors 1 -

(LPRMs). The decay ratios es'imated from LPWis were not significantly' l different than the ones esticiated from the average power rangs monitors.

4 4

In conclusicn. the measured ecay ratios at Browns Ferry Unit I showed tts plant to have adequete stabf Tity margin over a range of power / flow, -

conditions which are .af concern during single leep operation. Since '

e the Ouane Arnold maximum calculated decay ratio (.84) is similar to i j Brwns Ferry Unft 1 (.37), and since it was shown that the stability

)

chsracteHstics of SLO are similar to TLD this test provides additional t justification to allow single leop operetten at DAEC. 6 i '2. 3 Adr.inistrative Changet _

The licensee has proposed to update some referencst and delete blank pages in the Technical Specifications. The staff finds the proposed admthistrstive charges acceptable. -

l 3.0 ENV_IRO MENTAL CONSIDERATIONS This newndment involves a change in the installation or use of a facility component located within the restricted trea as defined in 10 CF2 Part 20.

  • The staff has detemined that the amendment involves no significant inceense i

In the amounts, and to sionificant change in the types.4 afsany effluents l

that.may be released offslte, and that there fs no signilkant increase in individual or cumulative occupational radiation exposure. 'The Cnmissfon i has prevfously issued a propo:ed finding that this aswndment involves no

! significant hazards consideration and there has haen no public coment on such finding. Accordingly. this amendment meets the eligibility criteria f

! for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 4

CFR 51.22(b) no environmental impact statesent or environmental assessment l need be prepared in connection with the issuance of this aettndment.

I 4.0 CONClH510N ~

Wehaveconcluded,basedentheconsiderationsdiscussedabove,that(1) there is reasonable assurance that the henith and safety of tha.oublic will not le endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with 'the comission's i

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I regulations, and the issuance of this amendment will not be inimical to the j common defense and security or to the health and safety of the public. .

Principal Contributor: George Thomas, George schwenk and hohen Thadani

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Dated: .May 28, 1985-

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        • 7  ;

July 20, 1984 Docket No. 50-334 '  ;

I 1 Mr. J. J. Carey, Vice President Nuclear Division Duquesne Light Company Post Office Box 4

. Shippingport, PA 15077  ;

1

Dear Mr. Carey:

i

SUBJECT:

BEAVER VALLEY UNIT NO. 1 - OPERATION WITH TWO OUT 0F THREE REACTOR COOLANT LOOPS - SAFETY EVALUATION By letter dated October 27, 1978 Du i

of technical specification changes. quesne Light Company requested a number One of these requested changes would permit Beaver Valley Unit No. I to operate with only two of its three coolant loops, with the third one isolated by loop isolation valves. The enclosed Safety Evaluation discu'sses in detail results of our reivew.

l Our review of the request has concentrated on the impact of this proposed change on the performance of safety related systems, the loop isolation valves, on certain thegeneric design-basis safety accidents, issues. as well as our evaluation of the impact i We have sent out several rounds of questions and have visited the site on issues concerning human factors. You 4

20 in the enclosed Safety Evaluation Report. responded to such staff e '

i We conclude that (1) the proposed method of loop isolation provide adequate protntion for the integrity of the primary pressure boundary, (2) the l liktlihood of accidents and transients would not be significantly increased, (3) the isolation of one loop will not seriously degrade the i

perfonnance of safety related systems, the instrumentation and control sy stems or that of the closed loop isolation valves, and (4) no major safety problems would result from two-loop operation (these include core thermal i

nydraulics, pressurized thennal shock, and SG tube damage).

i You have reanalyzed the full spectru'a of transients and accidents using acceptable codes and assumptions.

The results demonstrate compliance with l

the acceptance criteria of the regulations and the Standard Review Plan.

1 As a result of our reivew, some changes to your proposed technical

! specifications will be necessary to ensure safe operation with one isolated t

loop. These changes are sunnarized in Section V of the SER. All except one 4 -- (asymmetric LOCA blowdown load) have already been addressed by you in your

~ letters. We will evaluate the adeauacy of your rcvised technical specification changes against these concerns. Issuance of an' amendment

authorizing two-loop operation.will be contingent upon satisfactcry resolution of these concerns.

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Mr. J. J. Carey July 20, 1984 On the basis of the considerations discussed above, we conclude that N-1 loop operation at Beaver Valley Unit No. I does not constitute a threat to public health and safety. Please discuss with your Project Manager, Mr. Peter Tam, the proposed schedule of submittal of your revised amendment request.

?)i.O . . 4" '!' " t iS(ev~en A'. Varia',' Chie'f Operating Reactors Branch #1 Di'ision v of Licensing' i

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~ Mr. J. J. Carey Duquesne Light Company Beaver Valley Power Station Unit 1 cc: Mr. W. S. Lacey Station Superintendent -

Mr. Thomas J. Czerpah Duquesne Light Company Mayor of the Burrough of Beaver Valley Power Station Shippingport Post Office Box 4 Post Office Box 26 Shippingport, PA 15007 Shippingport, PA 15077 Mr. K. Grada, Superintendent Pennsylvania Power Company of Licensing and Compliance Ray E. Sempler ,

Duquesne Light Company One E. Washington Street Post Office Box 4 New Castle, PA 16103 Shippingport, PA 15077 Ohio Environmental Protection Agency Mr. John A. Levin Division of Planning Public Utility Commission Environmental Assessment Section Post Office Box 3265 Post Office Box 1049 Harrisburg, PA 17120 Columbus, Ohio 43216 Gerald Charnoff, Esquire Office of the Governor Jay E. Silberg, Esquire State of West Virginia Shaw, Pittman, Potts and Trcwbridge Charleston, West Virginia 25305 1800 M Street, N.W.

Washington, DC 20036 Charles A. Thomas, Esquire Thomas and Thomas Karin Carter, Esquire 212 Locust Street Special Assistant Attorney General Box 999 -

Harrisburg, PA 17108 .

Bureau 5th Floor,ofExecutive Administrative HouseEnforcement Harrisburg, PA 17120 Regional Radiation Representative EPA Region III Marvin Fein Curtis Building - 6th Floor Utility Counsel Philadelphia, PA '19106 City of Pittsburgh 313 City-County Building Governor's Office of State Planning Pittsburg, PA 15219 and Development ATTN: Coordinator, Pennsylvania Resident Inspector State Clearinghcuse U.S. Nuclear Regulatory Commission Post Office Box 1323 Post Office Box 298 Harrisburg, PA 17120 Shippingport, PA 15077 Mr. Joseph H. Mills, Acting Commissi'rer Department of Environmental Resources State of West Virginia Department ATTN: Director, Office of Radiolo- of Labnr gical Health 1900 Washington Street Post Office Box 2063 East Charleston, West Virginia 253C5 Harri.sburg, PA 17105 l

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Beaver Valley Power Station

. Unit 1 2-cc: N. H. Dyer, M.D.

State Director of Health State Department of Health 1800 Washington Street, East Charleston, West Virginia 25305 Irwin A. Popowsky, Esquire Office of Censumer Advocate la25 Strawberry Square Harrisburg, PA 17120 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA II406

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SAFETY EVALUATION OFFICE OF NUCLEAR REACTOR REGULATION BEAVER VALLEY UNIT 1 OPERATION WITH TWO OUT OF THREE REACTOR COOLANT LOOPS

  • July 20,1984 PQ4 7pqctOW 99
  • This Safety Evaluation Report summarizes results of the staff's review work perfomed in the last several years. The Unit may operate with '

License has been issued by the staff authorizing such. operatio

TA3LE OF CONTEllTS C

Page I. Introduction 1

II. Proposed Mode of Operation A. Primary Coolant System and Reactor B. Secondary System I C. 2 ,

Instrumentation and Centrol 2 D.

Loop Isolation Valves and Loop Supports 5 E. Miscellaneous Systems F. 6 Initial Test Program And Procedures G. 7 H. Control Room Indicators -- Hanan Factors Evaluation8 Conclusions 10 III. Safety Issues A. Core Thermal-Hydraulics B. 10 Pressurized Thermal Shock 11 C. Steam Generator Tube Damage D. Power Peaking Factors 13' E. Conclusions 13 14 IV. Accident Analysis A.

B.

Loss of Coolant Accident and Asymmetric Blowdown Loads 14 Non-LOCA Transients and Accidents 15 C. Conclusions 18 V. Restrictions to N-1 Loop Operation 18 VI. Summary and Conclusions 19 VII. References 20 i

Tables 22 - 24 O

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1. Introduction Duquesne Light Canpany (the licensee. DLC) has applied for approval to operate Beaver Valley Power Station Unit 1 (a three-loop Westinghouse plant) with only two active coolant loops (Reference 1). Two-loop operation is currently precluded by a license condition. Beaver Valley is one of a class of of plants with loop isolation valves in both the cold leg and hot leg each loop.

in the inactive The loop.licensee proposes to operate with the isolation valves closed The purpose of the application is to allow the plant to continue failure in that loop. one loop out of service in the event of an equipment operating with Our review has concentrated on the impact of this proposed change on the performance of safety related systems. The study included a complete review of design basis accidents, as well as an evaluation of the impact on generic issues.

II. Proposed Mode of Operation .

A. Primary Coolant System and Reactor The inactive loop will be isolated from the primary coolant system by closing loop isolation valves in both the hot and cold legs. Rollowing closure, motive power to the valves will be removed by locking the asso-ciated circuit breakers in the open position, in conformance with techni-cal specifications.

inadvertent opening of these valves (FSAR The 14.1.6).In addition, interlock location of the isolation valves are such that the Pressurizer, charging and letdown, ECC and RHR are still open to the reactor coolant system (RCS),

A water relief valve overpressurization. will be installed in the isolated loop to prevent All high pressure interfaces with'the operating loops, isolated.all drain paths and all interfaces with injection systems will be Exceptions to these operating restrictions will be allowed only during cold shutdown the loop will beorcarried refueling, out.during which time maintenance and repair of operation will be performed only at cold shutdown.Switchover between N loop an Operation 65 percentof of the full reactor power. would be limited by technical specifications to The total allowable peaking factor F (2) at 65%

power would decrease from the normalization curve, K(Z). 3.57 to 2.77 and there would be minor9changes in ONBR during operation from about 1.7 to about 2.3.These changes result in an inc 82 aver Valley 1 SER 1

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There will be'no changes to the technical-specific tten valu;s of delta T, ce & average temperature or loop flow, although the actual loop flow would increase somewhat.

Because of the results ofa m'in steamline break calculations described below, the. required shutdown' margin maintained during N-1 loop operation will loopbe 2.44% delta X/K instead of the 1.77% delta K/K maintained during N operation.

We consider the above status of tha primary system acceptable.

8. Secondary System -

The 1icensee indicated that the secondary side of the isolated loop would be kept filled and the steam generator would be maintained in a wet layup condition as determined by secondary side chemistry requirements. The loop would be isolated from the main and auxiliary feedwater system and the main steam system by closing the appropriate valves. Steam supply from the isolated loop to the turbine-driven auxiliary feedwater pump

' would also be isolated. Pressure and temperature of the isolated loop and the corresponding steam generator would be maintained within the constraints imposed by brittle fracture considerations, and stean generator tube differential pressute timits.

A small amount of stesu leakage on the secondary side of the operating loops to the isolated loop is possible through a 3" check valve on the residual heat removal line. Pressure monitoring of the steam generator would remain available and excessive in-leakage would be vented by manual or automatic opening of the atmospheric dump valve.

Ir. addition. steam floy through the check valve could never be more than that through a stuck open' atmospheric dump valve, performed forwhich is bounded the N-1 case. by the analysis of a stuck open safety valve Steam generator level will be maintained above the top of the SG tubes and within the narrow range level indication. No maintenance on the isolated loop portions would be performed during N-1 loop operation because the 3-inch check valve is not considered sufficient protection of personnel from possible steam hazards.

The isolation valves to the down loop would be recuened for three loop operation only during cold shutdown using approv2d start up procedures to prevent excessive thermal and hydraulic stresses. We consider the above status of the isolated loop acceptable.

Maximum steam flow and oressure rating at 65 percent power during N-1 'oop operation are slightly less than the steam flow and pres >ure rating at 100 percent power during 3-loop operation.

Therefore. the. steam safetv valve capacity on the operating steam generators is adequate to remove the maximum calculated rating from the steam generctor. steam flow at the engineered safeguards design conditions far the main steam and feedwater isolation valves.There is also no' c C. _

Instrumentation and Control System The reactor trip system and engineered safety feature actuation system initiate protective action based on measurements of primary and secondary B2 aver Valley 1 SER 2' _

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coolant syst:a parametdrs, as well cs othar plcnt conditions. The para-ideters and conditions associated with an out-of-service loop and the asso-i

  • ciated protective actions are identified in Table 1. The following discussion includes those aspects of the protection systems which are ,

unique to plant operation with a loop out of service:

1. Primary Coolant Temperature ,

The overpower and overtemperature AT are based in part on a measure-ment of primary coolant hot and cold leg temperatures. Each loep provides initiate aone channel reactor trip.of input signals for the 2 out-of-3 logic to During N-1 operation the channel associated with the out-of-service loop is placed in trip and the logic operates on the basis of 1-out-of-2 with input signals from the two operating loops.

channel In addition,the setpoint for the overtemperature AT trip function associated with the operating loops is readjusted corresponding to the value established for N-1 operation.

In addition to the above reactor trip functions, low average reactor coolant temperature is for two other protective actions. Feedwater isolation is initiated on low T-avg and reactor trip (P-4). On low low T-avg (P-12) steam dump is terminated and a permissive is pro-vided to reopen the cooldown condenser dump valves. These functions also operate based on 2 out of-3 logic, with one channel of the average temperature signals being provided by hot and cold leg tem-

-perature measurements.

The average temperature associated,with an out of service loop will be low such that the logic for these functions loops.

operating will be 1-out-of-2 based on input signals from the two

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2. Primary Coolant Flow 1

The loss of flow reactor trip is' interlocked with permissives based on reactor power such that above 10% power (P-71 a trip, is initiated on loss of flow in any two loops and above 31% power (P-81 a trip is initiated on loss of flow in any loop. During N-1 operation the flow channels state. associated with the out-of-service loop will be in a tripped The licensee has proposed to increase the, P-7 interlock setpoint to 71% power such that the trip on loss of flow is . single loop becomes a high reactor power trip. That is, a high rec tor power will provide the P-7 permissive such that loss of flow trip will are be in ainitiated tripped since state.flow channels for these out of service loop While we do not object to raising the setpoint of P-7 to defeating the single loop loss of flow trip, we find the use of the P-7 permissive to provide an overpower trip to be unacceptable. The annunciation associated with a trip based on P-7 would be an indica-tion of loss of flow rather than high reactor power. Therefore, we require that the setpoint of the power range neutron flux channe:s be reduced to 71% and that P-7 be increased to a value which would not provide misleading information to the plant operator. Fur-ther, operator training should emphasize that the two-loop loss-of--

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flow trip during N-1isoperation.

indicative of a loss of flow in either operating loop The trip on loss of flow in two loops is initiated on sensing either low primary coolant loop flow or by contacts indicating that the reactor coolant pump breaker is tripped.

channel associated loop.

for the out-of-service with the latter will also be in a tripped stateDuring N 3.

Steam Generator Level tiation of auxiliary feedwater is interlocked to blo the RCS closed. isolation valves associated with the out-of-service loop are The steam service /feedwater loop will mismatch be in an untripped state. channels associated with the i

Therefore, the reactor less of steam generator level in the out of-service -

The licensee had' proposed to bypass the hi-hi steam generator level -

channels associated with the out-of-service loop to preclude the potential to the twofor an inadvertent operating loops. turbine trip and isolation of feedwater steam generator level on the out-of service loop does not3 prc

' , challenge of plant safety systems. required safety action and cou We will review the manner.in

  • whichchannelsarebypassedduringthetechnicalspecificptionreview for N-1 operation.

4 RCP Bus Undervoltage and Underfrequency main energized during N-1 operation.The buses used to supp loss of power to any two of the three busses.A reactor trip occurs on the Since the protection conditions on the electrical power grid, we find the these busses no safety significance.not supplying power to an RCP during N-1 operation of

5.

Steam Generator Pressure The logic for the low steam generator pressure initiation of safety

' injection is interlocked to block this trip when the isolation valves associated with the out of service loop are closed.

In Summary the following actions are taken with regard to the protection systems when operating with a loop out of service.

1.

Reduction of the power range neutron flux trip setpoint to 71% .

i. 2.

i Readjustment of overtemperature AT trip setpoint for N-1 cperation Bzaver Valley.1 SER 4 -

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Pitcing the chrnnels of the overtemperature and overpower AT trips for the out-of-service loop in trip.

4.

Increasing the P-8 interlock set point to preclude a reactor trip'on low flow for the out of service loop.

5.

. Bypassing the hi-hi steam generator level channels to preclude the potential for inadvertent turbine trip and feedwater isolation.

These actions will be confirmed by appropriate notation to the limiting conditions of operation incorporated in the plant Technical Specifications.

The effect of N-1 operation with respect to the protection systems are the following: ,

1.

The logic for the overtemoerature and overpower aT trips is reduced from 2 out-of-3 to 1-out-of-2.

2.

The logic for feedwater isolation following reactor trip is reduced from 2-out-of-3 to 1-out of-2 coincident low T avg.

3.

The logic for blocking steam dump to the condenser,is reduced from 2 out-of-3 to 1-out of-2 on low-low T-avg.

4.

The logic for reactor trip on loss of flow in a sicale loop when operating above 31% power (P-8) is changed to a loss of flow in a single operating loop when operating above 10% power (P-7).

Since these changes do not violate the single failure criterion we find that requirements for redundancy to initiate safety actions is maintained and operation'with a loop out of service is, therefore, acceptable.

D. Loop Isolation Valves and Loop Support The Unit reactor coolantemploys two motor-operated stop valves in each of the three loops.

27i" in cold leg. The sizes of those valves are 29'.' in hot leg ano ed together with a by-pass line and a motor-operated by-pass valve Electric interlock circuits will permit the operation of pump and valves in the loop according to acceptable patterns. A comparison of the piping is listed as follows: construction parameters for the stop valves and the react Stop Valves Pipino 1 l

Material ASTM A351 ASTM A351 1 Grade CF8M Grede CF8M l Design 650 Temperature "F 650 8:dver Valley 1 SER 5

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[ Stop Valves Piping

[ Design / Working 2485/2235 2485/2235 Pressure, psig *

! Shop / Pre- 3350/3107 NA/3107 operational test pressure.

  • psig ,
Code / Class ASME III, 68/A ANSI B31.1, 67/NA 4

9 The construction parameters for the loop piping and stop valves are similar. The valves were constructed according to the ASME B&PV Code Section III, 1968 Class A requirements while piping was constructed according to ANSI B31.1, 1967 Standard. The ASME Code requires more 4

rigorous quality assurance than ANSI B31.1. In addition, a shop '

hydraulic ' test of 3,350 psig pressure has been performed on the valves.

This was not a part of the requirements on piping.

The ASME B&PV Code Section III_always considers that the piping system, i not the valve body, is limiting..eThis is because the design and fabrication requirements of the v'a'Ives result in a section modulus 3, greater than that of the piping.

We concur with the licensee's assessment.

4

! Nopal 10 flow rate per loop at the N-1 (2-loop) operation will be 36.g X j lbm /hr. compared. to the normal flow rate per loop of 33.6 X .10 lbm /hr. at ,1ormal (3-loop) operation. Since this increase in f1pw rate j is less than 8%, and since the steady state flow rate contributes only a j' small part of the total load on reactor coolant loop supports, these supports'will provide adequate resistance to the additional loading caused by the N-1 operation flow, .

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E. Miscellaneous Systems .

Loop isolation will indirectly affect the operation of the turbine-driven auxiliary feedwater pump and the pressurizer sprays. -

The turbine-driven auxiliary feedwater (AFW) pump which can receive steam

! from loops all threeN-1 during loops loopwill receive steam from either of the two operating operation.

! Thus there is still' redundancy in steam I sources for the AFW pump turbine and, in addition, the two motor driven

' AFW able. pumps, powered from two separate class 1 power sources would be avail-Therefore, 'the reliability of the AFW system is acceptable for N-1 loop operation.

i There is no change in the design flow and pressure required systems.

of the AFWS or other safety related auxiliary cooling water j Water the three is supplied loops. Iftoone the pressurizer sprays from the cold legs of two of of those loops is inoperable,. flow from the other would still be.available. Although the standard technical specifications

  • require that the pressurizer sprays be operable, there is no requirement

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for redundancy. Therefore, operation of the pressurizer sprays in this manner is acceptable.

F. Initial Test Program And Procedures We have reviewed the Unit 1 Cycle 1 Startup Report to determine if adequate testing was conducted during the initial startup to support two-loop operation. Our review indicates that reactor coolant system flow rate and flow coastdown measurements were conducted for the two-loop config-uration, that acceptance criteria were met, and that no modifications were made to the reactor coolant systems that would invalidate the data.

Since adequate resistance tenperature detector (RTD) bypass loop flow is necessary to assure adequate RTD response time for coolant tenperature input to the reactor protection system, we also verified that the RDT bypass loop flow will not be adversely affected with one loop out of service.

We conclude that, other than the surveillance tests required by Technica.1 Specifications, no additional preoperational or initial operation tests are necessary prior to two-loop operation with the third loop isolated.

We have reviewed licensee submittals. including responses to our requests for additional information regarding two-loop operations. In Reference 13 the licensee stated:

"To support two-loop operation the following steps would be taken:

1. The instrumentation, alarms, bistables and valve positions .,for the out-of-service loop would be made identifiable to the qpera-tor and . administratively controlled to avoid confusion during an event. These items would be a part of the procedure for removing the loop from service.
2. The auxiliary feedwater flow to the out-of-service loop would be isolated as part of the procedure for removing a loop from service. Flow verification could not be made and, tnerefore, would require identification of this instrunent as.being out of service for the affected loop.
3. The surveillance tests would provide for monitoring of the following where necessary:

- verification of the closed position of the out-of-service main steam isolation valve

- instrument channel checks for protectic' and control instrumenta tion

- auxiliary feedwater system alignments."

We agree with the stated need for procedures to control actions to renove a loop from service and to specify surveillance tests as indicated.

The licensee should develop procedures for changing from three-loop to two-loop operation and vice versa. The needed surveillance tests and BEAVER VALLEY l- SER - 7

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, monitoring checks for two-loop operation should be specified and con-trolled by written procedures. All procedures related to two-loop foperation and surveillance, including revised or new emergency operating procedures, should be uniquely identified and placed at locations convenient to where they will be used.

G. Control Room Indicators - Human Factors Evaluation .

2 During a visit to the Beaver Valley control rcom on August 19, 1932, we reviewed some of the instrumentation that would be affected. By letter dated September 2, 1982 Duquesne Light Company supplied us with a list of affected instrumentation and an estimate of the number of affe'cted annunciators and bistable lights.

The list provided by Duquesne Light Company includes 30 ' instruments, most of which will read downscale while the remaining instrument indications will depend on the nature of the work being conducted in the isolated loop. Five recorders will have one of three pens each reading downscale and one recorder will have all three pens reading low. The number of off nomal indications included in instruments and recorders is large enough to be beyond the nonnal memory capability of ar operator. A unique identifier should be provided prominently on each display to remind the operator that the indication refers 4 to the isolated loop. This N-l loop identifier should not be part of the normal mainterance tag-out system unless the indicator .is, in fact, inopera-tive because of a malfunction, calibration or test. '

Since certain indications can have zero as a legitimate value during this mode

. of operation, it is also imperative that the affected instruments fail off-scale, and not at zero.

4 A more subtle, but no less important human factors problem may ' exist with displays 'in t'he operating loops, if system operating ranges change because of the 11-1 loop operation, such that normal operating zones on meters are no longer applicable, or values are different from what appear in procedures, the operator may be presented with conflictino infomation. The licensee should detennine if this condition does exist, what its magnitude might be, and how it will be resolved (e.g. simulation, training, procedure modification.1 unique display identifiers).

~

Two different conditions exist for annunciators. In the first case, approximately 27 annunciators share inputs,from all three loops. The input from the isolated loop would have to be defeated to maintain the operability of the alam function. A rigorous administrative control procedure will be necessary to account for the defeated signals and to ensure return to nomal when the N-1 loop operation is teminated. Analysis will be required on each annunciator in this group to detemine if there will be any change in the BEAVER VALLEY l SER 8 .

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operator response procedure based .on one inoperative loop. If any change in procedure is required, the annunciator will require a unique identifier to renind the operator of the unorthodox mode. If no change in ' response pro-cedure is required, there appears to be no need for unique identifiers on the annunciator tiles.

In the second case, up to 12 annunciators that provide status monitoring on loop components would have to be identified with out-of-service tags because they are associated with the isolated loop. The specific annunciators affected would depend on the system configuration.and the nature of the work being performed on the isolated loop. Unless specific response procedures are different for those not tagged out, there appears to be no requirencnt for unique identifiers.

In addition to annunciators, approximately 36 protection system bistable status lights could be affected, depending on the type of work to be perfonned on the isolated loop. To maintain consistency in uniquely identifying iso-lated loop indications, and to reduce the risk that, af ter testing, bypassed l signals in the isolated loop protection systen are returned to the tripped condition, unique identifiers should be installed on all isolated loop bistable status lights. -

Based on our review of the instrumentation, annunciators, and protection system bistable status lights affected by the N-1 loop operation, we c'onclude that all instruments, recorders, annunciators and status lights providing legitimate, though abnormal, status indications in the isolated loop *should be provided with unique and prominent identifiers to remind the operato/ that the information beingp' resented is not that of a nonnal operating loop. The unique identifier should not be part of the normal tag-out system unless the system or component is actually out of service.

An analysis should be conducted to identify all other indications which, though normal for N-1 loop operations, will not remain within the full-loop normal zones operation.

or will be different from values specified in procedures for full-loop This effort should also include recommendations on how these mode-dependent indications will be brought to the attention of the operators and how they will know what limiting values are actually in effect. Operating Proce-cures, specific to N-1 loop operations should be developed and operator train-ing, preferably through simulation, needs to be conducted.

Finally, adninistrative controls need to be put in place to ensure that maintenance personnel are alerted to, and aware of, any limitations to perform routine ESF checks and tests that could result in unit trips or unavailability of systems.

BEAVER VALLEY l SER 9

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+ l H. Conclusions Operation of the plant with one loop isolated in the manner described above provides adequate protection of the primary coolant boundary and -

does not significantly degrade the performance of safety related systems.

[ We therefore conclude that N-1 loop operation as described above will not significantly increase the occurrence rate of accidents and

3. transients.

III. Safety Issues t

Zn the course of the review, several possible safety issues were considered. A few of them were judged to be of potential significance and were examined in detail. Results of those evaluations ere presented below.

A. Core Thermal Hydraulics Our review of the Beaver Valley Power Station (BVPS) Unit 1 i

! thermal ~ hydraulic design included concerns about the effect of changing from three loop operation (N) to two loop operation (N-1) relative to:

(a) thermal-hydraulic parameters, (b) inlet flow maldistribution, and i

(c) flow instability. The licensee's response to our questions was given in Reference 18. 'i

. . Thermal-Hydraulic Parameters The licensee's response provided Table 2, a thermal hydraulic domparison i

for three-and two-loop operation which is also included with this evalu-i tion. In examining Table 2, it is noted that when operating with one loop i

isolated the values for the system pressure, the percent heat generated in the fuel and the affected flow area for heat transfer remain constant. i i

However, the values for the reactor core heat output, coolant flow, J- coolant temperatures, average temperature rise in the core, heat flux, average linear power and core pressure drop are reduced. Also, when operating with one-loop isolated, the minimum DNBR at nominal design con-j ditions is increased, which is conservative.

The licensee responded to our question on the possibility of temperature differences of few degrees in the active cold legs due to the isolation of one loop causing the possibility of a radial power tilt and increase in the enthalpy rise factor. The licensee stated that the quadrant power tilt has a restriction of 2 percent as stated in the Beaver Valley" Tech-nical Specification (Section 3.2.4) and exceeding this value would require

! the necessary actions described in the Technical Specifications. There-fore there is no difference in the thermal hydraulic methodology in the evaluation of the N and N-1 loop operation, and the safety analysis for N-1 loop operation assumes that with the isolation valves closed, no temperature difference is induced.

Beaver Valley.1 SER 10 l .

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Tha cost limiting design transient was given as loss of reactor coolant pump are 1.5flowandin2.0 both three and two pump operation for which the DNBR values respectively.

N-1 loop operation, primarily because of reduced peak power.The mar Inlet Flow Maldistribution In response to our question, the licensee stated that both flow model flow maldistribution with two loop operation. tests and analytical stu The results of 1/7 scale hydraulic reactor model tests studies using the THINC code (Ref. 5).

(Ref. 3 and 4) were applied in analytical tion with the THINC analyses show that it is adequate to use a 5% reduc tion in inlet flow to the hot assembly for operation with one loop out of service.

The licensee provided flow maps for three loop and two loop operaton show-ing the normalized inlet flow distributions.

normalized flow in the center nine assemblies for both three and two loo operation shows an agreement within approximately 2%. The studies per-formed in Reference 5 show that even with a 10% flow reduction int center less than nine assemblies of the core, the hot channel DNBR is reduced by 0.5%.

The licensee stated that any asymmetries that exist due to N-1 loop opera-tion would fuel have little or no impact on power distribution, ONB limits and integrity.

From Reference 5 it was found that even with extreme inlet flow slightly maldistributions, hot channel enthalpy rise and DNBR are only affected.

Generic radial power distributions and a SK flow reduction into the hot assembly used in the safety analyses account for any inlet flow maldistributions.

The restrictions placed on the fuel to ensure fuel integrity are applicable to both N and N-1 loop operation as stated in Chapter 3 of the BVPS Unit 1 FSAR.

Flow Instability The licensee instability withstated that operation.

two loop Reference 6 was used for the analysis of flow This showed that'the margin to inception exit quality. of thermohydrodynamic instability increases with a decrease in Since the exit quality for N-1 loop operation is less than for loopNoperation.

loop operation, a greater margin to flow instability exists for N-1 We have reviewed the thermal hydraulic information on two loop operation for SVNP-1 pertaining to thermal-hydraulic parameters, inlet flow maldistribution and flow instabilities and has found them acceptable .

B.

_ Pressurized Thermal Shock Failure of the reactor pressure vessel can occur when highly irradiated welds are exposed to both high pressure and low temperature. This pheno-menon, known as Pressurized Thermal Shock (PTS), is thought to be most l

severe for certain small break LOCA events, in which natural circulation Beaver Valley 1 SER 11 r- -

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1 is lost and rcprassurization occurs.

, In such cases, low temperature HPI

!q water would gradually cool the cold leg and reactor vessel downcomer. The resulting combination of low temperature and high pressure.would present a i serious wall. challenge to the circumferential and axial welds in the downcomer

,i of vessel failure from-this and other events (Reference 7).Th Operation with one loop isolated presents a potential additional PTS risk.

With the cold leg isolation valve closed, the HPI water injected at the inactivethe entering loop would experience only limited mixing with hot water before downcomer.

the events calculated in R eference 7.This would aggravate the overcooling effect of Furthermore, the potential would exist for overcooling in the region of the isolated cold leg under all t circumstances when HPI is initiated. To date there have been almost 20 HPI actuations at Beaver Valley unit 1.

The Office of Nuclear Regulatory Research (RES) contracted with S. Levy

Inc. to ~ perform a series of calculations to determine the impact of loop isolation over a range of thermal-hydraulic conditions (reference 8). The i

calculations were intended to answer two questions; (1) does loop isola-tion significantly aggravate the overcooling during accidents which have been identified as PTS events, and (2) does the isolation of a loop create the downcomer can occur?new scenerios (such as spurious actuation of HPI) in wh -

To answer the first question, temperature profiles in the downcomer with no coolant flow were calculated for N-1 loop operation, in order to simu-lato conditions during a SBLOCA with loss of natural circulation. Tempera-tures at the top of the nearest axial weld below the isolated , loop cold leg were other coldfound legs. to be 25'F colder than corresponding locations under the i

an excessive cooldown event were also performed and showed a temperature difference.

4 A temperature decrease of this magnitude is conservatively estimated to increase of magnitude the(X100). risk of crack initiation and propagation by about two orders i

two factors; (1) the increased probability applies to only one of th three cold legs and (2) N-1 loop operation will be an infrequent occur-i renced'uring the lifetime of the plant. Furthermore, the temperature i difference due to N-1 loop operation is lower (< 10*F) at the location of peak neutron flux where PTS risks is greatest.
To answer the second question, temperature profiles in the downcomer below the isolated loop cold leg were calculated for both forced flow (reactor .

coolant pumps on) and natural circulation conditions.

i For all cases ana-lyzed, the temperature of the coldest weld below the isolated loop cold i leg wascold active lesslegs. than 40*F colder than the corresponding location under the '

I The lowest temperature reached by any of the welds was approximately shock is a concern. well above the range in which pressurized thermal 440*F, result from normal HPI actuation at power..These results indicate that no PTS -

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82 aver Valley 1 SER 12 ,

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or. ._

.__.-...n.-.-

t The SLI calculations were reviewed by NRC/RES staff. The validity of the and by comparison with 1/5 scale tests performed Plant by Creare, data used directly fromin the the calculations were taken from the BVP5 FSAR or obtain licensee. .

1 The proposed plates PTSapproaching or axial weld rule places anrequirements RT on all PWR's with forgings, welds approaching 300*F. N f 270*F or circumferential BecauseourcSIculationindicatesanegative

, 25*F temperature difference due to loop isolation, the screening criteria for N-1 loop operation at BVPS will be 245'F and 275'F respectively. This

, limitation loops to be will apply only to materials in the downcomer region below the isolated.

C. Steam Generator Tube Damage We believe that two precautions should be taken to avoid undetected damage to steam generator tubes.

First, the isolated loop must be main-tained in a water solid condition, or any voids in the loop must be nitro-gen inerted, to prevent corrosion in the SG tubes.

Second, the addition of heat to a water solid volume can lead to differ- '

ential pressure levels well beyond the design values for the steam genera-tor tubes.

Prolonged operation in such a mode could produce undetected damage in the tubes, which would represent a safety problem when'the isolated loop is returned to service. ,

isolated loop must bc equipped with an appropriately sized pressure-To pre actuated relief valve and a pressure monitor. The licensee has% proposed a relief valve setpoint of 200 psig.
  • D.

Power Peakina Factors N-1 loop operation will involve power levels at or below 65 percent power probably over extended periods of time (likely in excess of t'wo weeks)

This could lead to the possibility of core power distributions and peaking 4

factor increases beyond those normally considered, if there is a return to full power operation within the same cycle. The peaking factor problem associated with extended part power operation has recently been discussed by Westinghouse in the enclosure of a letter from E. P. Rahe, . Westing-3 house to C. H, Berlinger, NRC, November 8,1983, "NRC Request for Reduced Power Operation - Operating Procedure." ,

A copy of the report, " Extended Reduced Power Operation Evaluation and Recommended Operating Procedure "

was also furnished to Westinghouse utility customers in August 1983. The report suggests procedures to minimize the peaking factor increase effect and to maintain a methodology which can retain power density limits on return to full power with normal (F or F stillunderreviewbytheNRCstaffiYbuttke) surveillance. The report is interim position is that the procedure recommended in the report should be followed for N-1 loop opera-tion (and all extended operation below 85 percent power) if return to full power.is contemplated for that cycle.

i.

~ B2 aver Valley 1 SER

E. Conclusions Although several potential safety issues have been identified, the pro-posed solutions are considered acceptable. The staff concurs that the risk from these issues is small as long as the solutions referred to in this section are implemented.

. I V. Accident Analysis With the exception of a few transients which were analyzed for N-1 loop opera-tion in the original FSAR, a full range of transients and accidents has been reanal.yzed or reevaluated in R eference 1.

These calculations have been re-viewed by the staff, Laboratory (Reference 9). with comments provided by the Idaho National Engineering A.

Loss of Coolant Accidents And Asymmetric Blowdown toade The large Break Loss of Coolant Accident (L8LOCA) was er.alyzed in Reference 1 using the 1975 version of the Westinghouse evaluation model.

Compliance with the acceptance criteria of 10 CFR 50.46 was demonstrated for double ended cold leg breaks in an active loop and in the unisolated segment of the inactive loop. As in the N-loop case, the highest peak clad temperature (2155*F) resulted for a discharge coefficient of 0.4. At the time Reference 1 was submitted, the 1975 Westinghouse model had already by beenmodel.

the 1981 superseded by the 1979 model, which has since been superseded In order to confirm the adequacy of the ECCS

, evaluation, the licensee submitted a reanalysis of the limitiag LBLOCA (C

= 0.4) performed with the 1981 model (Reference 17 ). The analysis showed d compliance with 10 CFR 50-46, with a peak clad temperature of 1882 F.

The Small Break Loss of Coolant Accident (SBLOCA) was reevaluated not reanalyzed. but This is primarily because the SBLOCA was shown to be non-limiting in the original FSAR for N-loop operation, and because lots of one coolant loop is not a significant factor in SBLOCA, particularly since ECCS from that loop is still available. In cases where heat removal is through natural circulation, one steam generator is capable of providing more than enough cooling. Moreover, two significant benefits are derived from the reduced core average power (65%); (1) reduced system pressure allows higher ECC flow and earlier actuation of the accumulators, and (2) the lower steam production rate delays core uncovery. Furthermore. the reduced uncovery.

peak power leads to slower heatup of the hot pin following The licensee has demonstrated satisfactory ECCS performance for both LBLOCA and SBLOCA during N-1 operation.

In the event of a double-ended guillotine break, a decompression shock wave would propagate in the cold leg and impinge on the core barrel .

The amplitude of the wave is proportional to the difference between the (

system pressure and the saturation pressure (P - Psat). With the isolation valve in the cold leg closed, the segment of pipe between the isolation valve and the vessel will experience little or no flow. Consequently, the cold-leg temperature and saturation pressure could be less than BEAVER VALLEY l SER 14 J

. ~ . - . . . .. - _ _ _ _ _ - - - - - _ - - _

~

those in the operating loops. Therefore, the amplitude of a blowdown shock wave in the inoperable leg could be significantly larger than for cases previously analyzed, and could exceed the structural capacity of the reactor internals to . withstand it. The licensee should perform an analysis to evaluate the possible magnitude of such a shock wave. The licensee may, alternatively, submit a fracture mechanics analysis to '

demonstrate that a double-ended guillotine break is not a credible event (see Generic Letter 84-04, dated February 1,1984).

B. Non-LOCA Transients and Accidents Main Steamline Break (MSLB)

The MSLB is more limiting with one loop isolated. The reduced coolant inventory leads to more rapid cooling and a greater reactivity insertion.

To compensate, the technical specification shutdown margin will be '

increased from 1.7% ak/k to 2.4% Ak/k.

required in Westinghouse two-loop plants.The new value is similar to that The steam line break analysis for 2-loop operation was performed with '

approved codes and reasonable assumptions. The results show that reactor -

system pressure remains below operating pressure and that the minimum DNBR is greater than 1.30. Consequently the criteria of the Standard Review Plan are met.

i Beaver Valley Unit I has received approval for diluting tne oorop concentration in the boron injection tank (BIT) from 20,000 ppm to 2,000 ppm. The analysis presented in support of that amensnent did not include the N-1 case. Cor.sequently, the BIT must either be maintained I at 20,000 ppo during N-1 loop operation, or an analysis be submitted '

to justify the 2,000 ppm concentration for N-1 loop operation.

Feedwater Line Broak The feedwater line break accident for N-1 loop operation was analyzed with I significar11y different assu ptions fram the FSAR N-loop analysis. The most notable difference is that the rear or tripped on low-low level in the faulted steen generator, a fact which resulted in significantly less stored energy in the primary system at tne time of reactor trip. A second l Gifferer.ce is tnat safety injection jas assumed to operate. Although significantly different from the FSAR analysis, these assumptions are ,

acceptable and are the same as those used in more recent FSAR's. The {

analysis shows slower pressurization of the primary system, and lower i primary temperatures.

l

'he main r.egative impact of loop isclation on this accident is the l avsilability of one less SG for heat removal. This is. offset to some {

extent Dy the reduced core power. The fact that theafew analys a shows less severe response to the feedline break is due primarily to t<e new assumptions discussed above. j l

l BEAVER VALLEY l SER I5

.. - .. . .~ . , ,,

4 3

  • This accident was analyzed with acceptable. codes and methods, and produced results which generally conform to the staf f's understanding of how the accident should be affectedb ' y isolating one loop. All of the applicable acceptance criteria were met.

Other Class IV Events i

The RCP lecked rotor event with one loop isolated tad to somewhat higher pe ak pressure (2730 psia vs. '2690 psia), but st(ll did not owcred the 110%

of design pressure. The calculated peak clad temperattre is lower in the N-1 loop case because of the reduced core power.

, steam ' Generator Tube Rupture (SGTR) 1 4

The original FSAR for BVPS-1 showed compliance with the radiation dose linits cf JO CFR 100 for a 5GTR during N loop operation. In tce submitt61 under review, the licensee h;as asser'ted that the consequences of a SGTR during N-1 loop operation are bounded by the N loop case. Most of the system paraneters which affect SGTR are not altered by the isolation of 4

one loop. A significant advantage of N-1 loop operation is reduced fis-i sion product inventory due to operation at 65% of full power. Offsite releases would be significantly reduced.

Although one less steam generator would be available for decay heat ,,

removal, the recaining SG would have more than sufficient capacity to '

. - remove decsy beat and ccol the RCS at a rate of 75*f per hour.

Sever.al questions have been raised covering the technical bashs for the

$GTR analysis oresented in the FSAR for BVPS-1 and other plants. The main issues relate to the use of non safety grade PORV's, and to the need for the licensee to provide justification for the assumption that the operator can isolate the aff.ecteed SG within 30 minutes, Although resolution of these is3ues codlc potentially affect the results of the N-1 loop analysis, they are not strictly associated with the question of N-1 loop operation.

The licensee has demcnstrated that the consequences of a SGTR during N-1 Icop operation wculd .be less severe than for N loop operation.

ONBR-Limited Transients For a broad range of transients involving loss.of reactor coolant flow, l depressurizat-ion of the primary system or loss of secondary heat removal.

the principel accept 5nce criterion of the Star.dard Review Plan is that )'

ONER must rensin above 1. 30. The margin to OttB during normal operat3cn

. with N-1 2 cops is significantly higner than for N loop operation, ,

primarily because of the reduced peak power. Consequently the severity of i this class of transients for N-1 loop cperation has been found to be bounced by the N-loop results. In all cases, the peak pressures were ]

within the limit of 110% of design pressure. Typical calculated DNBR ,

' values for these transients are showr, in Table 3. The loss of normal

- feedwater event, r,ct includedf in the tanle, produces only small increasas in primary coolant temperatuee and is significantly more benign for the i N-1 loop case. Loss-of-of fsite power calculations produce results sinflar 4

to the less of reactor-coolant-flow event.'

BEAVER VALLEY l SER 16

n _-:

Boron Dilution Isolationevent.

dilution of one RCS loep would have two major impacts on the toren The reduced RCS volume would allow more rapid dilution, but this effect would te of fset by the higher shutdown margin (2.4 vs. 1.7%). Hand calculations by the staff indicate that tne net effect would action, for operator be a srall (approximately 25%) increase in the time allowed Excessive Heat Removal Excess teat removal due to malfunctions of the steam or feedwater systems are bounded criteria by the for this typemain ofsteamline event. break, which meets the pressure and DNBR 3eac+.ivity Transients The reactivity transi ents include control bank withdrawal at startup and at power. control rod ejection and the control rod misoperation events including rod drop, single rod withdrawal and rod misalignnent. The licensee has subnitted reanalyses of the rod bank withdrawal and rod ejection events with conditions applicable to N-1 loop operation and using, for the most part, cetnods and criteria of the FSAR N loop anaTyses.

For the control rod bank withdrawal at startup, DLC reanaly2ed using the same methocology and criteria except that a newer spatial kinetic 4 code (TWINKLE)

This is Valley (part of)was used instead the currect of the methodology coint kinetics code used

&s used, . on the FSAR Beaver 2 c5AR. for example, in the The transient results should change very little since this reduced flow. event is not very sensitive to the primary parameter change, But this improved methodology provides lower transient power and fuei temperature than with peint kinetics and gives temperatures sufficiently low that there is a large targin to DN8 even with the reduced ficw.

event, This result is te ta expected from more recent analyses of the e.g., Beavtr Va!14y 2, where analysis witn only two pumps a^d no isolation and with TnihC results in large analysis ONS margin as calculated with this methodology of DH6.

t The Control rod withdrawal at power events were reanalysed with N-1 loop parameters including ?. hose for the overtemperature delta T trip setpoint.

The resulting margins to ON8 are largar than for N loop cperation, Targe13 because of the improved initial DN8R state.

The control roc efection events at N-1 loop full ocwer (65 percent) conai-tions were calculated using standard Westinghouse methodology. The criteria. fuel temperatures (and enthalpies) were well within norra) resulting to the parameter This would changes. be expected since the results are not very sensitive The standard :onservative generic results for applicabie to the range, of concitions in N-1 loop operation and new calculations are needed in these areas.

BEAVER VALLEY 1 SER 17

Th s2 reactivity transients .bsva been suitt31y analyzed at N-1 lorp con-  ;

ditions. with acceptable method.s. A11 results .eeet required criteria and g- the changes fram N locp analysis rescits are in acccrdance with staff expectations. .

The events f allf r.g under the centes) ras misoperatian category have not been spacifically analyzed at N-1 loop conditions. Bo ever, the events proceed alen3 paths parcIIel to the N 1000 condition analyses but with the core further removed fran Timiting DNB coccitions because of the improved '

initial GNBR conditions. The rod drop, rod misalignment and the sir,gle roo withdrawal at power would each have the same extrere rod configaration to analyze at h-I loop conditions as nerailly used fcr H ioop conditions, but the icproved initial state *culd resuit in improveo estreme DXBE states.

It is tnus concluded that under the allaved conditions for N-J locp opera-tion the teactivi ty tran.sients normalfy . analyzed are n.o mora severa than for N loop conditions ar.d that all applicable criteria for these events would be met.

Events Not kea_nalyzed Several accide6te s for *nich N-l loop operation is nct juoged to be a sig-nificant factn , were not reanalyras in Reference 1. These incluce acn- ,

thermal-hydraulic events such as f. vel handling accidects and accidental reletses of stored waste. Events <nich are Juaged to be precluded by acsinistrativt controls or automatic intericcks were also cat reevaluated.

These include the inadvartant startup of an inactive loco and inaevertent mis-loading of a fuel assembly. '

4 C. Conclusion The ir. pact of ACS ) cop isolation varies from accident to accicent. In some instances., tnere is a measurahla fors of safety margin. while ic others, it is ir. creased. In all cases the calculated responses to accidents and transients meet the acceptance criteria of tr.e standa rd review plan. J V. Restrictions to N-1 tocp Operation ,

Proposed technical specifications have been submitted by the liter.see (Reference 1). However, due to the concerns discussed in detail above, these prooosed technical specification changes should be revised accord- '

ingly; ctmmitments have teen made by the licensee in his responses to staff inquiry (Aeferences 10 - 20). They are. summarized as follows:

1.

lo avoid undetected danage to the SG tubes, the licensee has suggested 3 raximum pressure of 200 psig for both the primary and seccadary side of the isolated loop. This linitation must ce proposeo as a technical ,

( ' specification (Section III .t).

2.

The isolated loop should be monitored in a water solid or nitrogen-inerted  ;

condition to prevent 56 tube corrosion (Section III .C),

BEAVER VALLE Y 1. SER 18 l LS

J 3.

Because the steamline break analysis was performed assuming that the boron injection tani coitains watar with 20,000 ppm boron, this concentration must be a technical specification for N-1 loop operation, until a revised analf sis based on a Tower concentration is submitted and accepted (IV.8),

4 Isc1ation (II shstdown of .A, a 1.oco or returning one to service is permitted only at cold II .3 .ard II.F). i 5.

The setcoint for the power range neutron flux channels will be reduced to 71% and 9-7 w311 be increased to a value which would avoic misleading infor-mation to the plant operator (Section II.C),

6. As noted in Section II.C aceve, several changes in instrumentation setpoints will have to be made in the plant Technical Specifications.

In addition, the procedures, fo17owing staff concerns should be addressed by analysis, new or both:

?. Procedures to prevent excessiva power peaking factors following return to N loop operation (Section III.D). .

8. Wnn factors cancerns (Section II.G)
9. Asymmetrical LOCA blowdown load (Section IV.A)

By letter dated Apri) 10, 1984 the licensee informed us that due to th'e icng time this review effort has spanned, and the changeover of person'ne?

at the A555 vendor (Vestir.gbouse), work has been initiated to confirm ths documentation being used to sapport the pending aucpdment. Thus the licensee should Haft uniti such confirottery design review is completed before the revised anandrent request is subuitted.

VI. Sumnary and Conciusions W2 conclude that the proposee metnac of loop isolation provice adequate pro-tection for the integrity of the prinary pressore boundary. We also conclude that the Tikelihood of accidents and transients would not be significantly increased.

I The isstation of one locp will not seriously degrade the perform-ance cf safety related systems, 5,ystems, or that of the closed loop that .cf the isolation instrumentation and control valass.

With respect to the safety issues examinac, we conclude that no major safety pecblems would result from N-1 1cep cperation (these include core thermal hydraulics, pressurf red thermal shock, and SG tube damage),

The ifcensee has reanalyzed the fuTT spectrum of transients and accidents using acceptable codes and assumptions. The results demonstrate compliance with the acceptance criteria of the regulations and the standard review plan.

~

SEAVER VALLEY 1 EER 19 -

b __ _ p ,,ygg$ ' .-

As it result by the staff,of the sancanalysu, subrutted by the licepsse and' the revf ew conducted tions will be necessary to ensureisafe operation with one changes a e sacarized in Section V. .

isolat,ed lo These We' w(ll evaluate the adequacy of the '

revised technical specification changes to be propose contingent upon satisfactory resolution of these c.oncerns.

On the ecsis operation of the at Berver consideratioal Valley Unit (11scussed above, we conclude that W-1 lo and safety. I does nut constitute a threat to peblic he61th VII. Refersances L.

C. N. Cunn (Doquesne Light) letter to A. Scwencer WiC), "Recuest for Amanneent to the Operating License - No. 35," Dctober 27, 1978.

3.

l Federal Register, Vol. 49, 37321, August 17, 1983 Opportunity for Prior hearing on N-1 loop operation fer Betver valley Unit 1. '

3, G. Metsteni, "June WCAF-3269-9, Hydraulic 1934. 7ests of.the San Ofiefre Reactor Model."

4.

G. Hetsrcni, WCAP-2761, " Studies Jene 19ES. of the Connecticut-Yankee Hydraulic Model,"

~

5.

L. E. Hochreiter, " Application of the THINC IV Program to pWR sign,"

WCAP-8054, October 1973 (properistary), and WCAP-8195, October 1973 (non pr Sprietary).

6. P. Saha, N.

Ishii and N. 2cber, "An Experimental Investigation of the Transfer, Novembcr 1976, pp. Thermally Induced Flow Oscillations in Two 616-22 7.

SECV-62-465, ' pressurized Thermal Shock," Noverber'23, 1982.

8. J. M. Healzer and J. M. '

For Beaver Valley Power Station," SLI-8310-1 (May 1983).Sorenson, "D

9. 7 E. Lyon, " Evaluation c1 Operation of Beaver Valley Power Station Unit 1 With One Loop Isolated," EGG-EA'5350 (February 1981). t 10.

C.N. Gunn (DLC) letter to A. Schwencer (NRC), August 23, 1979.

11.

C.N. Dunn (DLC) letter to S. A. Varga (NRC), March 24, 1961.

12 J.J. Carey (DLC) 1etter to S.A. Varga (NRC), September 2,1982.

13. J. J.

Carey (DLC) letter to S.A. Yarga (NRC), Octobe.r 8,1982.

BEhVER VALLEY l SER 20 g 9' '

-- - - - .; s 14 J.J. Carey (DLC) letter to S.A. Varga (NRC), Nover.ber 22,19G2 15.

J.J. Carey (DLC) letter to S.A. Varga (NRC), March 4,1983 16.

J.J. Cirey (DLC) letter to S.A, */arga (MRC), July 6,1983 17.

J.J. Carey (DLC) letter to S.A. Varga (NRC), July 29, 1983 18.

J.J, Carey (DLC) letter to S.A. Yarga (NRC), Octcber 21, 1983 19.

J.J. Carey (DLC) letter to S.A. Yarga (NRC), March 27, 1984 20.

J.J. Carey (DLC) letter to S.A. Varga (NRC), Aprfl 10, 1984

_ Principal Contributers R. Barrett, lead Reviewer H. Balukjian H. Richings T. Dunning .

R. Goei H. Shaw W. Long 5.E. Bryan R. Eckenrode P. S. Tao, Project Manager Dated July,1984

\

i i

BEAVER VALLEY 1 SER 21

.n . -

m .

TABLE 1 Protection Systems Parameters and Conditions Assoetated with an Gat of-Service Loop Parameter Safety cune.tions

1. Primary Coolant
a. H3t leg temperature
b. Input signals for Cold leg temperature AT reactor trip. everpower and overtemperature Fee.4 water isolatico and steaa dump interlock:

2a. Primary coclant flow

20. RCP Breaker tripped Input signal for lov flow reactor trip.

Input rigarl for low flow reactor trip (two loops).

3. Staan Generater Level Input signals far icw level coincident with s?.eac/

a feedwater flow mismatch and lou-low level reactor trip and low-low level initiatton of auxiliary feedwater. Input for hi-hi level (P-14) turbine trip and f>2edwater isolacion.

4.

RCP Bus undervoltage and Input signals for reactor trip Underfrequency

4

5. - Stear Generator Pressure Input sigreal for safety injection oh loy pressure and high negative pressure rate. '

f 4

i l

l

)

I l

8: aver Valley 1 SER 22 g . . _ .

~

g .

+-

TABLE 2 BEAVER VALLEY UNIT 1 THERMAL AND HYDRAULIC COMPARISON 3 Loop 1 Loop Desian Parameters Operation

_ Isolated Reactor core heat output (MWt) 2,652 1724 -

Reactor core heat output (10s Btu /hr) 9,051 5884 Heat generated in fuel (%) 97.4 97.4 System pressure, nominal (psia) 2,250 2,250 System pressure, minimum steady 2,220 state (psia) 2,220 Minimum DNBR at nominal design

, conditions Typical flow channel 2.26 2.97 Thimble (cold wall) flow channel 1.83 2.43 Minimum DN8R for design transients >1.30 DNB Correlation - >1.30 "R" (W-3 with "R" (W-3 with modified spacer modified spacer factor) factor)

Coolant Flow Total thermal flow rate (108 lbm/hr) 100.8- 72.1' Effective flow rate for heat transfer (108 lbm/hr) 96.3 68.8 2

Effective flow area for heat transfer (ft ) 41.6 41.6 Average velocity along fuel rods (ft/sec) 14.4 ,

10.1 Average mass velocity (108 lbm/hr-ft2 ) 2.32 1.66 Coolant Temperatures Nominal inlet (*F) 542.5 534.4 Average rise in vessel (*F) 67.5 63.2 Average rise in core (*F) 70.3 65.9 Average in core (*F) 579.4 568.7 Average in vessel (*F) 576.2 566.0 l Heat iransfer Active heat transfer, surface area (ft2 ) 48,600 48,600 Average heat flux (8tu/he-ft2 ) 181,400

~ Maximum heat flux for normal operation 118,000 2

(Btu /hr-ft ) 420,900* 326,700**

Braver Valley 1 SER 23' -

p

_" ,.g

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  • * * ' ' ' ' '~ - , - . - . . . - - -

. l

. . j I

Table 2 (Continued)

Heat Transfer Operation Isolated Average linear power (kW/ft) 5.20 3.38 Peak linear power for normal operation (kW/ft) 12.1* 9.4**

Peak linear power resulting fron overpower transients / operator errors, assuming a maximum overpower of 118% (kW/ft) 18.0 18.0 Peak linear power which would result in centerline melt (kW/ft) >18.0 >13.0 Fuel Central Temperature Peak at linear power for prevention of centerline melt (*F) 4,700 Pressure drop *** 4,700 Across core (psi) 21.3 1 2.1 11.2 t 1.1 NOTES:

  • This limit is associated with the value of F
    • This limit is associated with the value of F0 -, 2. 32.

Table 3

~

5 Calculated Minimum DNBR N loop N-1 loop Loss of RCP flow 1. 5 2. 0 Loss of Load (worst case) 1. 6 2.1 Depressurization of The RCS 1.45 Not calculated

}

Braver Valley 1 SER - 24

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