ML20024G297

Ao:On 700920,standby Power Source Failed to Close to Energize Bus,During Test of Emergency Power Source When Power Source 4.16 Kv.Caused by Undervoltage Relay Failing to Drop Out

ML20024G297
Person / Time
Site:	Monticello
Issue date:	09/29/1970
From:	Duncanson R, Jacobson G NORTHERN STATES POWER CO.
To:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 9102080503
Download: ML20024G297 (6)

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3.

EXHIBIT A MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 LICENSE AMENItiENT REQUEST DATED AUGUST 4,1975 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS APPENDIX A 0F PROVISIONAL OPERATING L'ICENSE NO. DPR-22 Pursuant to 10CFR50.59 the holders of Provisional Operating License DPR-22 hereby propose the following changes to Appendix A, Technical Specifications.

.His exhibit makes reference to the following documents:

1. Monticello Nuclear Genersting Plant Technical Specifications, Appendix A

11. Monticello Nuclear Generating Plant License Amendment Request dated August 20, 1974 iii. December 27, 1974, order for Modification of License for

- the Monticello Nuclear Generating Plant iv. Monticello Nuclear Generating Plant License Amendment Request Dated March 12, 1975 1.

AVERAGE PIANAR LINEAR HEAT GENERATION RATE (APLHGR)

PROPOSED CHANGE his specification currently exists as APIIIGR limit figures in Reference i (page 108C), in the proposed changes in Reference 11 (pages 189H,189I, and 189J of Exhibit B) and in Reference 111 (Figures A-1 through A-4).

We attached APLHGR limit figures (pages 189H through 189L of attached Exhibit B) are croposed to replace the limits stated in each of the three above ref-eren..s.

The remaining portions of Reference 11 are not affected.

'l NOTE: Three fuel types are considered in References i'and 11.~ h e fourth fuel type currently in use at Monticello was not limited by ECCS considera-tions using the model described in Reference 11.

Because of changes in the ECCS model required by the AEC staff, Reference 111 imposed a restriction on this fourth fuel type also.

The proposed changes shown in the attached Exhibit B include APIllGR limits for the four fuel types currently in use at Monticello as well as a fifth type soon to be loaded in the Monticello reactor.

I f63

}102000503f ab w ^wn

2 REASON FOR CHANGE These proposed changes are provided in accordance with the requirements of Reference 111.

They are the result of calculations using an ECCS mcdel modified to incorporate the changes required by that reference.

The attached Exhibit C discusses the revised ECCS model and presents the results.

2.

BASES 3.11 (page 189F of Reference iv. Exhibit B)

PROPOSED CHANGES This page supersedes page 189F of Reference iv; it includes two minor changes.

The second line of diis page states that an initial operating HCPR using the ECCS calculations was 1.19.

That value should be changed to 1.18.

Reference 5 of page 189F should be changed as indicated in the attached Exhibit B.

REASON FOR CHANGE Exhibit C uses an initial operating MCPR in the calculation which is slightly different than that used in Reference 11 and reported in Reference iv.

The assumed initial condition for the ECCS calculation is well below the operating limit discussed in the paragraph under change.

The change to Reference 5 on page 189F corrects an error in Reference iv.

3.

SPECIFICATION 3.11.C (onge 189K of Reference iv. Exhibit B)

PROPOSED CRANGE Change the page number of Figure 3.11.2 from 189K (as presented in Reference iv Exhibit B) to 189M as shown in the attached Exhibit B.

REASON FOR CHANGE The insertion of additional pages in item 1 above requires this change to arrange figures sequentially, i

i

...~

. =. _..

.... _ ~. -. -

(-

h l

i i

1 EXHIBIT B LICENSE AMENDMENT REQUEST DATED AUGUST 4,1975 Exhibit B, attached, consists of newly prepared pages for the Appendix A Technical Specifications as listed below.

These pages incorporate the proposed changes.

PAGES 189 F 189 H 189 I 1

189 J 189 K 189 L I

189 M l

)

i i

Bases 3.11 (continued)

C.

Minimim, Critical Power Ratio (MCPR)

The ECCS evaluation presented in Reference 4 assumed the steady state MCPR prior to the postulated loss of coolant accident to be 1.18 for all fuel types. The Operating MCPR Limit of 1.41 for 8x8 fuel and 1.33 for 7x7 fuel is detemined from the analysis of transients discussed in Bases Sections 2.1 and 2.3.

By maintaining an operating MCPR above these limits, the Safety Limit of 1.06 (T.S.2.1.A) applicable' to all fuel types is raintained in the event of the most limiting abnomal operational transient.

For operation with less than rated core flow the Operating MCPR Limit is adjusted by multiplying 'the above limit by K.

Reference 5 discusses how the transient analysis

[

g done at rated conditions encompasses the reduced flow situation when the proper K factor j

is applied.

g f

It is recognized that MCPR is a calculated parameter that is not continually monitored and alamed directly during core power distribution and thermal-hydraulic changes.

If at the tine of the evaluation it is fcund that the limits are being exceeded, there is always an action which will return the MC?R to within prescribed limits, namely power reduction. Under most circums tances, this will not be the only alternative. Whenever the limit is exceeded the monitered value will be documented and available for review, audit and inspection cf plant operations.

The only way to violate the Limiting Condition for Operation is to knowingly allow operation beyond the prescribed limits without taking the necessary action to restore the MCPR to within prescribed limits.

Re ferences I.

" Fuel Densification Effects in ";eneral Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEDM-10735, August, 1973.

s 2.

Supplement I to Technical Report on Densification of General Electricleac' tor Fuels, Decmber 14,1974 (USAEC Regulatory Staff) 3.

Communication: V A Moore to I S Mitchell, " Modified CE Model for Fuel Densification,"

Docket 50-321, March 27, 1974.

4 "Monticello Nuclear Generating Plant I,oss-Of-Coolant Accident Analysis Conformance with 10 CFR 50 Appendix K, August 1974," L 0 Mayer (NSP) to J F O' Leary, August 20, 1974.

5.

" General Electric BHR Generic Reload Application for 8 x 8 Fuel,"

NEDO-20360, Revision 1, November, 1974.

3.11 BASES 189F REV

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D311 BIT C MONTICELLO NUCLEAR GD4EFATING PIANT COSPORMANCE WITil 10 CFR 50 APPDiDIX K (JET Plt!P PIANT)

AUGUST 1975 a

F m

o b

b Presented in the following document are the results of the loss-of-coolant accident analysis of the Monticello Nuclear Generating Plant.

The analysis was performed using General Electric calculational models which are consistent with the requirements of Appendix K of 10 CFR part 50.

A complete discussion of each code employed in the analysis is presented in Reference 1.

Between August and December,1974, General Electric and the USAEC worked together to resolve dif ferences in interpretation of Appendix K and to consider additional phenomena in the evaluation models. As a result, the models used in the present analysis differ frcu those used in previous submittals in the following respects:

(1)

The new analysis assumes a fuel assembly planar power consistent with 102* of the MAPLilGR shown in the Figures; (2)

Fission product decay is computed assuming an energy release rate of 200 MeV/ Fission; (3)

Pool film boiling is assumed af ter nucleate boiling is lost during the flow s tagnat ion pe riod; (4) The ef fectr of core rpray ent rainment and counter-current flow liriting arc included in the reflooding calculation.

In addition, there have been a few other minor improvements to the computer codes which individually and jointly have a small ef fect on the calculated

,results.

The figures in this submittal reflect these changes, as well as the four major changes enumerated above.

In the analysis of the break spectrum of this reactor, a range of break sizes was studied, with a range of single failures being considered for each break size.

A list of the single tailures considered for each break size is shown in the lead plant submittal referenced herein.

That list is ap-plicable to the analysis of this reacter.

INPUT 10 THE ANALYSIS A list of the significant plant in pu t parameters to the loss-o f-coolant accident analysis is presented in Tabic 1.

4 g

,7 O

o TABLE 1 SIGNIFICAtlT INPUTS PAPhiETERS TO THE LOSS-OF-COOLANT ACCIDEllT ANALYSIS "Y,

FOR MONTICELLO PLANT PAPAf tETERS:

Core Thermal Power.......................

1703 MWt which corresponds to 102% of licensed core never*

6 Yessel Steam Output.............

s.o11 x 10 lbm/hwhichcorrespondsto 102 % of rated steam flow 1

Vessel Steam Dome Pressure............

1040 psia Design Basis Recirculation Line 2

2 Break Area for Laroe Breaki 3.4 ft 3,n fg Recirculation Line Break Area for Small Breaks 1.0 ft2 0.07 ft2 FUEL PARAMETERS:

PEAK TECHNICAL INITIAL

'SPECIFICATI0!1 DESIGN MINIP.UM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENEPATION RATE PEAKING POWER t-FUEL TYPE GE0 METRY (kw/ft)

FACTOR RATIO Initla t u.re

( 7D225 7x7 17.5 1.57 1.18 ReloaC 1 i

.7D230 1

7x7 17.5 1.57 1.18 Reload 2 80262 8x8 13.4 1.57 1.18 Reload 3 8D250 8x8 13.4 1.57 1.18 Reload 4 8D219 8x8 13.4 1.57 1.18 i

A more detailed list of. input to each model and its source is presented in Section II cf Reference 1.

• This power level equals the Aopendix K requirement of 102%. The core heatup calcu-lation assumes a bundle power consistent with operation of the highest powered rod at 102% of its maxinum (technical specification) linear heat. generation rate.

e,

f o

o.-

RESULTS Of THE ANALYSIS The results of the analysis are presented in the order in Lhich they are calculateu.

The presentati n of the results is divided into four major portiens according to the model from which the output is obtained.

These portions are:

A.

Calculcted by the Short-Term Thernal Hydraulics flodel (LAMD)

B.

Calculated by the Transient Critical Power Model (SCAT)

C.

Calculated by the Long-Tern Thermal liydraulics 11odel (SAFE)

D.

Calculated by the Core Heatup 1:odel (C'1ASTE)

A sumary of the results is presented in Table 2.

At the MPLilGR* employed in the molyris, the mmt severe pipe break yleias a caicuiates peak cladding temperature less than or c:;ual to220PF, a calculated maximum local neta t-water re ction less than or equal to 17% and a calculated core-wide metal-water reaction less than or equal to 1%. Comoliance with the 10CFR50.46 criteria for coolable geometry and long-tern cooling has been shown in Reference 1.

The reactor is, therefore, fully in confornance with 10;FR50.45 and Appendix K with operation at the MAPLHGR used in the anclysis.

These values, if more limiting than other design parameters, represent limits for operation to ensure conformance with 10CFR50.46 and Appendix K.

The peak cladding temperatures as a function of time are shom in Figure D-1 Other parameters relecant to the analysis are shown in the attached figures and are described in subsequent paragraphs.

Results for guillotine severances of a main steam line, a feedwater line, and a core spray line are presented in Reference 2.

• "ninn (Bundle) Average Planar Linear Heat Generation Rate 4

2

y ;.

q.

~ +. n c.

o

,o c

TABLE 2 APPENDIX K RESULTS FOR M0!iTICELLO w

w m

m.

Break Size Core-Wide' Location Peak Local Metal-Water Single Failure PCT (OF)

Oxidation %

Reaction r DBAANALYSIS(I) 3.9 ft2(DBA)

Recirc Suction 2200())

8.7%

0.5 LPCI Injection valve BREAKSPECTRU!! ANALYSIS (3) 3.9 ft2(DBA)

Rectre Suction 2200(j) 8.7%

0.5

~

LPCI Injection Valve 2

1.0 ft Large Recirc Suction Break 1670())

<1 LPCI Injection Methods Valve Small Break 1690(2)

,j Methods 2

0.07 ft Rectre Suction 1430(2)

,)

F HPCI 7

Notes:

CHASTE - large break methods Non-DBA reflood For other breaks in spectrum see lead plant analysis Reference 2.

Fo: justification of selection of lead plant, see Reference 3.

Ik

.5-

)f*

o o

A.

Aprendix K Short-Tern Thermal Hydraulics Analysis General Description of the L/"3 Code 8

The LM13 code is a model which is used to analyze the short-tem thermodynamics and thermal hydraulics behavior of the coolant in the vessel during a postulated loss-of-coolant accident.

In particular, LNG predicts the core flow, core inlet enthalpy and core pressure during the blowdorm prior to the end of lower plenum flashing

(~ 20 seconds).

For a detailed description of the model and a discussion regarding sources of input to the model refer to the "LMiB Code Documentation" portion'of.

Reference 1.

kesuitt of thn iAf'd Annivsis Presented in the section are results of the loss-of-ecolant accident analysis which are calculated by LM-3.

Table 3 lists the figures provided for all the analyses. These results include the followi:g:

Parameter Figure Core Average Inlet Flow Rate Ciormalized to. unity at the beginning of the accident)

Following a Design Basis Accident A-la Following a 1.0 Sq. Ft. Break A-Id Core Inlet Enthalpy Following a Design Basis Accident A-2a Following a 1.0 Sq. Ft. Break A-2d Core Average Pressure Following a Design Basis Accident A-3a Following a 1.0 Sq. Ft. Break A-3d Diese results are input to the SCAT code discussed in Section B. -- _-.

m J

D q'

B.

Appendix K Transient Critical Powr Analysis General Description of the SCAT Code ne SCAT code is used to evaluate the short-tem themal hydraulics respon.ce of the coolant in the core during a postulated loss-of'-coolant accident.

In particular, the convective heat transfer process in the themally limiting fuel bundle is analy:cd during the transient. For a detailed description of the model and a discussion regarding sources of input to the miel refer to' the " SCAT Code Docmentation" portion of Reference 1.

Results of the SCAT Analysis Presented in this section are results of the loss-of-coolant accident analysis which are calculated by SCAT. Table 3 lists the figures provided for all the analyses. These results include the following:

Parameter Figure Mininnn Critical Power Ratio Following a Design Basis Accident, 8x8 B-la-1 Following a Design Basis Accident, 7x7 B-la-2 Following a 1.0 Sq. Ft. Break, 8x8 B-1d Convective Heat Transfer Coefficient Following a Design Basis Accident B-2

.Following a 1.0 Sq. Ft. Break B-2 Rese results are used as input to the OMSTE code discussed in Section D.

{.

o O

C.

Ie - " d i r v terg Terrijfh?rn1 " % 1i:s /,n.1ysis General Descrintion of SAFE Code The SAFE code is a model thich h used to analyze the long-torn ther-odynamic behavior of the coolant in the vessel during both small and large breaks.

Since the calculational procedure of the loss-of-coolant accident analysis dif fers depending on whether or not a break is classified as "snali" or "large," it is appropriate to distinguish bett;cen tuo classifications of breaks.

A small break is defined as that size break for t;hich nucleate, boiling heat transfer exists in the core until the heat fluxes are belta the pool boiling critical po..er condition.

This occurs approximately 20 to 25 seconds after the break.

For small breaks, core heatup is, therefore, based solely on the uncovery and recovery of the fuel and the duration of spray cooling all of which are predicted by the SAFE code.

For the "large" break analysis, the LA"B and SAFE codes are erployed to deternine tb tire of boiling transition and the post-Miling transition convective heat transfor coef ficient curin, the biundvim.

The SAFE ccJe calculatt.3 the uncov ry :nd r:-

f!:: ding ef t% fuel and thn duration of spray coolihg.

The SAFE anaiytical r.odel has been expanded and refined to consider explicitly the following phenomena:

(1) Counter-current flow liniting (CCFL) in the fuel bundles and ir, e

core bypass region, of ECCS water inj cted over the core; (2) Entrainment and loss of ECCS t,ater injected over the core; and (3) Filling of discrete volumes (control rod guide tubes, core bypass and lower plenua) which were previously taken together.

Calculation of these effects is presently external to the SAFE code:

the calculati'onal logic will eventually be incorporated in the SAFE code.

For a detailed description of the nodel and a discussion regarding sources of input to the rodel refef to the "S;FE Code acumentation" portion of Section 11 of Reference 1. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

r n,

9 O

o Results of the SU'E Analysis Presented in this section are results of the loss-of coolant accident analys:

which are calculated by S\\FE. Table 3 lists the figures provided for all tle analyses. These results include the following:

Parameter Figure k

Water Level Inside Shroud Following a Design Basis Accident C-1 (LPCI Inj. Valve Failure)

Following a 1.0 Sq. Ft. Large Break C-1 f

(LPCI Inj. Valve Failure)

Following a 1.0 Sq. Ft. Smil Break C-2 (LPCI Inj. Valve Failure)

Following a 0.07 Sq. Ft. Smil Break C-2 (HPCI Failure)

Reactor Vessel Pressure Following a Design Basis Accident C-1 (LPCI Inj. Yalve Failure)

Followinga1.0@failgre)

. Ft. Large Break C-1 (LPCI Inj. Yalve Following a 1.0 Sq. Ft. Small Break C-2 (LPCI Inj. Valve Failure)

Following a 0.07 Sq. Ft. Smil Break C-2 (HPCI Failure) f 9

1 l.

o o

4 l

0.

Apperdix x core gatun Analysis General Description of CFASTE Code The Transient therral response of the core to a loss-of-coolant accident calculated by CW STE can generally be bro;en down into four stages; (1) fuel pjn te cerature redistributicn; (2) f ;21 roo bundle te..perature redis.

tribution; (3) cctal-water reaction neatur; nd (4) core standby cooling system effects.

Pheno ena occurring durir hose stages that are consicered in the analysis are described below.

Fuel Pin Temperature Redistribution Following a reactor shutdown, a large heat source is still available within the core in the fer.m ci sens) Die heat in the fuel.

This is represented by the tc oer-ature profile in the fuel rod.

Initially, the terparature profile is ;;c:p i::;use of the high ec..er generation rates curing normal operatien.

Folicuing the snut-down, the sensible heat in the fuel will be redistribJttd by thernal CondkctiCn within the fuci and clacding ar.d by convecticn and radiation in the gap between fuel and cicdding. '.d'.h the attunt of neat re oved being dependent cn surface conditiens.

At the end of tnrce er more fuel tire constants (fuel ther 31 tire constant is about 3 to 10 seconcs), the radial tt rcrature profile in tr.? f uM ;in is almost flat, consistent with the low fission procuct cccay po..er generation.

Fuel Red Bundle Tciperature Redistributien As the cladding tc crature incrcases and the core coolant void f racticn t;p cact.cs unity, radiant heat transmissicn tutueen reds and the channel wall tends to equalize the tcmrerature of all rcds at a given axial position.

The total energy in the core cor.tinues to increase during this period due to continuing fissien product decay.

Metal-Mater Reaction Heatup The fuel pin cladding is rade of Zircaloy, which reacts with steam at hich torper-atures.

The zircalcy stear. chcmical reaction rate is exothernic and sttr;ly dependent upon 52 reactico tc perature.

The tecocrature cependence is csecnential and the rate of rea:..icn becomes significant at claading temperatures in tne rahge of 2200*F or higher.

Emergency Core Cooling System (ECCS) Etfccts Redundant emergency core cooling systcms perfor~ance for a given 1.0CA is deperd-ent uptn the cem:1cm of the accident.

I' ccre ccoling systt s will :

.ida sufficient coolir3 to crevent excessive clecting heatup.

The primary purpose of the core neatun analysis is to determine the effectiveness of the crergency core coolir,3 systcms.

For a detailed description of the CHASTE ccdel and a discussion recardir scurces of input to the redel refer to the "CFASTE Cc,de Documentation" portien of secticn II of Ref erc nce 1.

P O

o A break spectrum analysis has been perfomed using the CRASTE code showing that the most limiting (highest calculated) peek clad temperature is associated with the design basis accident.

The conclusion of this analysis is applicable to this plant.

The analysis has been documented in the Quad Cities Station Special Report 15 Supplement C (Docket No. 50-254).

For each submittal of a construction pemit, operating license, or reload license, the DBA peak cladding tenperature, peak local oxidation, and a MAPLHGR is detemined for each fuel type of interest.

For calculational convenience in some cases, the rod-to-rod power distribution 's assumed to be flat and the least favorable exposure is assumed in determining gap cenductance.

Calculation of the results under these conditfor.s conservatively represents the results at all exposures.

The cod 0 application is described, briefly, as follows:

A.

For jet-pump plants a LAMB calculation is performed.

In mixed cores, full-core LAMB calculations are perfomed for 7x7 and 8x8 fuel and the more restrictive of the two is used in the SCAT input.

B.

For jet-pump plants. SCAT calculations are performed for 7x7 fuel and 8x8 fuel, as appropriate.

C.

A SAFE and a DBA-REFLOOD calculation is perfomed, assuming the fuel to be the most predominant type of bundle in the core (7x7 or 8x8).

D. CKASTE calculations are perfomed for each fuel type (which in a given reactor nay include several 7x7 fuel types and several Sx8 fuel types) at several exposure points.

The PAPLHGR, peak cladding temperature and maximum local oxidation variations with exposure for each fuel type are the results of these calculations.

Results of the CHASTE Analysis Presented in this section are results of the loss-of-coolant accident analysis which are calculated by CHASTE. These results include the following:

Parameter Figure Peak Cladding Temperature Following a Design Basis Accident D-1 Following a 1.0 Sq. f t. Large Break D-1 Following a 1.0 Sq. f t. Sns11 Break D-2 Following a 0.07 Sq. Ft. Small Break D-2 Peak Cladding Tenperature and Local Peat Oxidation versus D-3 Break Area Peak Cladding Terperature and Local Peak 0xidattor versus Planar Exposure Initial Core Fuel (7D225)

D-4a Reload 1 Fuel (7D230)

D-4b Reload 2 Fuel (SD262)

D-4c Reload 3 Fuel (8D250)

D-4d Reload 4 Fuel (SD219)

D-4e

F-[

Parameter Figure Maximum Average Planar Linear Heat Generation Rate versus Planar Exposure.

---

Initial Core Fuci (7D225)

D-Sa

---

Reload I ruel (7D230)

D-5b

---

Reload 2 Fuel (8D262)

D-Sc

---

Reload 3 ruel (8D250)

D-5d

---

Reload 4 ruel (8D219)

D-Se Figures D-4 show the calculated peal cladding temperature as a function of exposure if the fuel bundles are operated at the average planar heat generation rate plotted in fig-gure D-5.

Figures D-5 show the average planar linear heat generation rate (APLHGR) as a function of exposure if the fuel bundles are limited to the most restrictive of:

1.

The 10CTR50.46 limits of 2200 F PCT 2.

The 10CFR50.46 liraits of 17'4 Local Metal Water Reaction 3.

The 10CFR50.46 limits of 1% Core-wide metal-water reaction or, 4.

The maximum design linear heat generation rate (LHCR) for the fuel and the peaking factor limits.

The APLHGR values indicated in figures D-5 impose restric-tions to the operation of the reactor in the exposure range where the peak clad temperature (PCT) is 2200 F or the max-imum local metal-water reaction is 177, in figures D-4.

Outside this exposure range the calculation was perforned on the basis of maximum design basis LHGR for the fuel and these Ibnits are more restrictive than those of Appendix K.

._(

TABLE 3 Monticello KEY TO FIGURES LARGE BREAK "FTHon IfRERi1EDIAlf EREAv SMALL BREAK s

Horst Add'l l

1.0 ft' Sn. Brk.

Sn. Brk.

l.0 ffI) large O)

Smil 0.07 ft' Core Main DBA

.80 DBA

.60 DBA Break Br(ak Sinction Spray Feeddater Steam Methods M _thods I.PCI Fail Line Line L i r.e Core Avg.

Inlet Flof A-la A-lb*

A-lc*

A-ld Core Inlet O

Enthalpy A-2a A-2b*

A-2c*

A-2d Core Avg.

Pressure A-3a A-3b*

A-3c*

A-3d Min. Critical l

Power Ratio B-la B-lb*

B-lc*

B-1d Convective Heat Trans. Coefficient B-2 B-2b*

B-2c*

B-2 D-2 0-2 D-2c*

D-2d*

D-2e*

Water Level t

l Inside Shroud &

C-1 C-lb*

C-lc*

C-1 C-2 C-2 C-2c*

C-2d*

C-2e*

C-2f*

i Reactor Vessel Pressure g

l l

l Peak Cladding D-1 D-lb*

D-lc+

Di I D-2a D -2b D-2c*

D-2d*

D-2e*

l Ter:pera ture Break Spectrum D-3 Peak Cladding Temp

& "ax 0xidation vs. Exposure D-4 MAPLHGR D-5 CSEE QUAD CITIES BWR 3 LEAD PLANT ANALYSIS l

7175

y I

,n 3

f e

V<

a SIfiGLE FAILLtE STUDY O'i ECC SYSTE!: I"WUALLY C0!iTROLLED ELECTRIC /,LLy rG%TED VMYES The effects of a sin;12 failure er crerator error that ccuses any manually controlled, electriceil.i crerated valec in the ECC Syste.1 to move to a position that cculd c/crsely aff;;t tne ECCS has been stuJied.

The purpose of tr s evalu;;1c 1 is to utemine that any such ralfunction dces not affect th: ECCS nere thar. the results of the worst single fiilure which is reported in the LCCA calculctions per onned in accordance With ICCFR50 Appendix K.

l The results of the break spectru,ar: lysis short the single failure which results in the axi

, calculated t-clad tc*;erature (FCT).

Fcr any other singio f ailure te ta rcre cipi:1::nt, its effect on tne ECCS inust be greater than tais sitgie failure.

Therefere, a study ras made to detemine if tne aifeccion of a i v,; ally contro11ea, electrically oper-ated valve b;. sc.e ur.cnc.m cause or of an operatcr imprco rly positioning a control switch could affect the ECCS more severely than this failure.

In accordance nith ar"rcrriate IEEE st3ndards, the ECC Systc, valves are electrically a:sicrc to c1ffercnt divisions cf pm er supply.

The effect

. n e i e s.,; 6:. vr, U.c c.c.tcci p;c.:1 or an eno av e u crouu i, at u n o. e is to c: sc cnly a 51";1e vcive # i nve to an incorrect res1 tion, ror the operatte e'ror of :.c:03;is.g a r..'.;1c 3.iit:P cf t' c f 0S 5 c t+1, tno system valves are nct acttated.

Fo lever, the censequances of a malfunction which causes one ADS valve to inadvertently open has been noted.

The sumary cf the ECCS Valve Single Failure Analysis is provided in the attached Tcbie 4 C:~;aring the effects of the single valve failure noted in Table !.ith the re:alts of the l;;cndix K LOCA analysis, it can be seen that these failurcs are not rore severe tnan those reported.

The single failures considered fcr the ECCS analysis are presented in Tacle 5.

-14

TABLE 4 MONTICELLO ECCS SINGLE VALVE FAILURE ANALYSIS POSITION FOR NORMAL PLANT OPERATION CO'15EQUENCES OF VALVE FAILURE ASSLHED TOGETHER WITH SYSTEM VALVE (5)

CLOSED OPENED DESIGN BASIS LOCA Core Spray Suction X

Nerpte use of one core spray loop Injection (s)

X X

Negate use of one core spray loop Test Return X

Negate use of one core spray loop High Pressure Coolant Injection Condensate Suction X

Utilize Suppression Pool Water Suppression Pool X

Utilize Condensate Storage Tank water Suction Valve M

Suppression Pool X

Partial loss of flew due to flow to suppression pool Test Return Injection (s)

X X

Nega te HPI Turbine Inlet (s)

X X

Negate HPCI Low Pressure g

Coolant Injection Injection (s)

X X

Negate use of LPCI W

Minimum Flow X

Partial flow loss in one loop due to flow to suppression Q

pool Cross 'ie.

X No LPCI fix: Negate on LPCI toop (two pumps per loop) i T

Test Return X

No consequence HX Bypass X

Reduce Flow due to HX Pressure Drop Automatic Punp Suction X

Ncnate one loop Depressurization System One Relief Valve X

Vessel depressurizes faster, increases rate of HPCI injection (assuming the failure of a single ADS valve to open does not affect the results because the effects on m.all breaks is insignificant with HPCI in op ation)'

TABLE 5 SINGLE FAILURES C0"SIDEDED FOR ECCS ANALYSIS PLANT SINGLE FAILURE REMAINING ECCS r

h BWR/3 LPCI Injection Valve 2 CS + HPCI + A05 MONTICELLO HFCI 2 CS + CPCI + ADS (Suction Break) f i.(

O O

l' Reference Plant Analysis The lead plant for this product line BWR is Quad Cities 2. ( )

The 60% DBA, 80", DBA analyses, additional Stull Break analyses, Core Spray line break, Feedwater line break, and Main Steam line break analyses for the lead niant are applicable to this plant and are hereby incorporatea by reference (3).

REFERENCES 1.

General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K NED-20566 (drsf t), submitted August 1974, and General Electric Refill /Reflood Calculation (supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L. Gyorey to Victor Stello, Jr., dated December 20, 1974 2.

Quad Cities Station Special Report No.15 Supplement C, Unit 2 and Attachment A (Proprietary information).

3.

Letter, G. L. Gyorey to V. Stello, " Compliance with Acceptance Criteria of 10CFR50.46 " May 12, 1975.

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