ML20083P925

Forwards FSAR Page Changes That Will Be Formally Implemented in Rev 15,increasing Diesel Generator Start Time to 12

ML20083P925
Person / Time
Site:	Wolf Creek, Callaway, 05000000
Issue date:	04/17/1984
From:	Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:	Harold Denton Office of Nuclear Reactor Regulation
References
84-0069, 84-69, NUDOCS 8404200266
Download: ML20083P925 (31)

Text

-

S SNUPPS Stenderdised Nue4eer Unit Power Plant System 5 choke cherry need Nicholas A. Petrick Rock ville. Maryland 20050 Exocutive Director taou seeaow April 17, 1984 SLNRC 84-0069 FILE: 0278 SUBJ:

Revision in Diesel Generator Start Time

Mr% Harold 1RCDenton; Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Docket Nos. STN 50-482 and STN 50-483

Reference:

SLNRC 82-01, dated 1/7/82, SNUPPS ECCS Analysis

Dear Mr. Denton:

o-A revision to the emergency diesel generator start time by two seconds (from 10 seconds to 12 seconds) is defined herein. This change in start time is based on providing a more practical method in implementing Technical Specification surveillance requirements in the operation of the Callaway Plant, Unit No. I and the Wolf Creek Generating Station, Unit No. 1.

The effects of delaying the start of the emergency diesel generators by two seconds have been evaluated by examining the consequences on the accident analyses presented in Chapter 15 of the FSAR and the containment pressurization calculations presented in Chapter 6 of the FSAR.

The evaluation considered those accidents presented in Chapter 15 that require safety injection.

The only accidents potentially affected by this change are the loss of coolant accident (LOCA) and the main steam-line break (MSLB).

This evaluation has shown the following results:

The LOCA analysis for the maximum safety injection flow case of the double-ended cold leg guillotine break (CD = 0.6) showed an i_ncrease in peak clad temperature from 2106*F to 2174*F.

It should be noted that the peak clad temperature reported in the FSAR (2105*F-Revision 9) was based on Westinghouse's generic penalty for maximum safety injection flow.

Westinghouse has evaluated the minimum safety injection flow case with a two second delay in the start of the emergency diesel generator and fcund this case to be not limiting (2114*F versus 2174*F).

The increase in peak clad temperature (2106*F to 2174*F) of 8404200266 840417 k

PDRADocK05000g

\\

lL o

~

4 D

SLNRC 84-0069 Page 2.

68*F results from the combination of the changes from consider-ing low head and accumulator injection interaction, the LOCTA modification in the axial conduction model, and the two second delay in the emergency diesel generator start time.

It is estimated that of this 68*F increase in peak clad temperature, the two second delay in the emergency diesel generator start time results in an increase in peak clad temperature of approxi-mately 10*F.

This LOCA analysis was performed using the 1978 model.

Use of the 1981 ECCS evaluation model and consideration of the reduction in initial pellet temperature modelling change would yield a net margin of approximately 200*F to the 50.46 criteria for the SNUPPS plants.

The limiting steamline break analysis is the case with offsite power available, where diesel start is not involved.

A review of the steamline break analysis with loss of offsite power, shows that the two second delay in the emergency diesel generator start would have an insignificant effect, if any, on the results in terms of reactivity and-the DNBR evaluation.

The offsite power available case is still limiting for the steamline break.

3 The effects of the delayed emergency diesel generator start time.

on the effectiveness of the containment sprays and the contain-ment air coolers were examined with respect to the containment

-pressurization calculations for the LOCA and MSLB.

For both of these cases, sufficient margin was allowed in the previous calculations for full functioning of these systems, that the 2 l

second increase in emergency diesel generator start time had no effect on the pressurization analysis.

Attached to this letter are.the page ch_anges in the FSAR that will be formally implemented in Revision 15.

It should be-noted that there are i

' numerous. references to the emergency diesel generator start time in the-FSAR and the FSAR changes presented herein address only those areas that could have been affected by the modification.

j

)

Your expeditious review of the above modification' is requested.

If there are-any questions, please do not hesitate to call us.

3 i

Ve tr ly

urs, Aw Ni olas A. Petricd i:

J0C/nldlla16&l7 Attachment

-cc:

D. F..Schnell UE J. Neisler/8.' Little USNRC/ CAL

-G. L. Koester

=KGE W. Schum/A Smith USNRC/WC-D. T. McPhee KCPL B.'L. Forney USNRC/RIII d

m.

n A.

- J

)

4 SNUPPS i

1 failure of these systems is assumed to be consistent s

j with the mass and energy release data assumptions for each break analyzed.

The total heat removed by each of the containment heat removal systems up'to the time of the calculated peak containment pressure is listed in Table 6.2.1-8.

The design bases of the I

containment heat removal systems are discussed in Section 6.2.2.

i

(

i The functional performance of the containment and the ECCS also rely upon the operation of the containment isolation system, as described in Section 6.2.4.

i i

Required isolation operations are assumed for pur-poses of the containment design evaluation in Section i

6.2.1.1.3.

l d.

Parameters Affecting Capability for Post-Accident l

Pressure ReductioA l

The principal parameters which affect post-accident

{

pressure reduction are 1) the heat absorbed by the j

heat sinks inside tne containment, 2) the heat removed by the. containment air coolers, and 3) the heat i

transferred to the containment sump by the contain-j ment spray system.

1 A conservative amount of heat sink material has been calculated, and its heat Mbsorption capability has been considered in the containment design evaluation 2

in Section 6.2.1.1.3.

The parameters describing the j

j heat sinks credited with heat absorption are provided l

in Table 6.2.1-4.

i The pressure reduction capability of the containment I

air coolers and the containment spray system consider the parameters 7rovided in Table 6.2.1-3.

The assumed start time of tsese active heat removal systems considers a diesel start time of 10 ' seconds, load sequencing times, and the maximum startup time of the i

systems.

I IWsW &

5 e.

Parameters ffecting usat Removal from the containment Heat is transferred from the containment to the 1

.outside environment via the fan coolers and residual l

l t

heat removal heat exchangers through the component i

cooling water and essential service water systems and released to the ultimate heat sink.

A small amount.

of heat is also transferred through the containment i

wall and done to,the outside atmosphere.

e i

The component cooling water system is described in i

section 9.2.2, the essential service water system is i

described in section 9.2.1, and the ultimate heat i

sink is described in Section 9.2.5.

6.2.1-3 i

... ~.

e s*

l l

SNUPPS

" INSERT A" 4

=0.6,MaximumIIwith12 In support of case c, large break LOCA (DECLG CD second diesel generator start) of Table 15.6-10, an evaluation of the assumptions used in the LOCA and MSLB containment pressurization calculations with respect to the full functioning times of the containment I

The evaluation spray system and the containment air coolers was performed.

shows that the containment pressurization calculations for both LOCA and MSLB provided sufficient margin so that a 12-secord diesel gene ~rator start 2

time does not change the assumed full functioning times of the containment sprays and the containment air coolers. Therefore, additional LOCA and MSLB containment pressurization calculations are not required for case c of Table 15.6-10, since this case is bounded by the previously performed containment calculations.

1 t

1 s

J e

S 1

1 i

I 1

-- " ~

i SNUPPS i

l 4

)

The small-break analysis was performed with the October 1975

,/ l 4

version of the Westinghouse ECCS Evaluation Model (Ref.

7',

12e

.f i

13, and 14).

15.6.5.3.2 Input Parameters and Initial Conditions

~

Table 15.6-9 lists important input parameters and initial i

conditions used in the analysis.

The ' analysis presented in this section was performed with a reactor vessel upper head temperature equal to the RCS cold The effect of-using-the-eeld Icg temperature---

leg temperature.

in the reactor vessel upper head is described in Reference 23.

In addition, the analysis in this-section-utilized-the-4apfice barrel-baffle methodology described in Reference 19.

i The bases used to select the numerical values that are input parameters to the analysis have been conservatively determined from sensitivity studies (Ref. 20, 21, 22).

In addition, the requirements of Appendix K regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis.

The assumptions made l

pertain to the conditions of the reactor and associated safety i

system equipment at the time that the LOCA occurs and include such items as the core-peaking factors, the containment pres-sure, and the performance of the ECCS.

Decay heat generated ex throughout the transient is also conservatively calculated.

',j) 15.6.5.3.3 Results i

Large-Break Results Based on the results of the LOCA sensitivity studies (Ref. 20, 21, and 22 ), the limiting large break was found to be the double-ended cold leg, guillotine (DECLG) break.

Therefore, only the DECLG break is considered in the large-break ECCS

/t#5hERT 4;(

performance analysis.

Calculations were performed for a range r

The results of these of Moody break discharge coefficients. calculations are summarized in Tables d"

jf The mass and energy release data for the break resulting inthe hig Section 6.2.1.5.

Figures 15.6-7 through 15.6-3Q present the parameters ofFor principal interest from the large-break ECCS analyses.

all cases analyzed, transients of the following parameters are, presented:

,]h 15.6-22 7-;; 4

Insert A For Page 15.6-22 The worst break in the spectrum of break sizes analyzed was a discharge This worst break was analyzed with a 12 coefficient (C ) of 0.6.

d second diesel generator start time, to study the effect of a 2 second Two cases are presented in delay in the start of the diesel generator.

d = 0.6 DECLG break: The minimum safety Table 15.6-11 for the C injection case and the maximum safety injection case (Reference 27 methodology). The maximum safety injection case proved to be the most limiting.

SNUPPS a.

Hot spot clad temperature b.

Coolant pressure in the reactor core

/

c.

Water level in the core and downcomer during reflood l

d.

Core reflooding rate

[

~

e.

Thermal power during blowdown The containment pressure transient resulting from a LOCA is presented in section 6.2.1.5.

For the limiting break analyzed, the following additional transient parameters are presented:

i a.

Core flow during blowdown (inlet and outlet) b.

Core heat transfer coefficients c.

Hot spot fluid temperature d.

Mass released to containment during blowdown e.

Energy released to containment during blowdown f.

Fluid quality in the hot assembly during blowdown g.

Mass velocity during blowdown h.

Accumulator water flow rate during blowdown i.

Pumped safety injection water flow rate during reflood l

The maximum clad temperature calculated for a large break is 2/7Y'F 2^^^ 7, which is less than the Acceptance Criteria limit of I

i 2200 F of 10 CFR 50.46.

The maximum local metal-water reac-F tion is (,./8 percent, which is well below the embrittlement l

l limit'of 17 percent, as required by 10 CFR 50.46.

The total j

core metal-water reaction is less than 0.3 percent for all breaks, compared with the 1-percent criterion of 10 CFR 50.46, and the clad temperature transient is terminated at a time i

when the core geometry is still amenable to cooling.

As a i

result, the core temperature will continue to drop, and the ability to remove decay heat generated in the fuel for an j

extended period of time will be provided.

i 1

An err as disc e

d in th worst a g

fa ure a tion th lar bre EC anal (see Re 26).

a a

on o the pact o h'

error se Ref. 27) i i

ted

.the tx ' -

mum dding te erature e culate or e

rge b k'cou:

(

be 10

, wh* h less

,a the Ac tsnce C te ia mi o

t t

00 F o 0

FR 50. 6.

he C rev d this L ue an ree at new a ysis res were t

ces ry.

i l

W 15.6-23 4/"2-

SNUPPS 18.

Letter NS-TMA-2014, dated December 11, 1978, Anderson, T.'M.

j (Westinghouse) to Tedesco, R. L. (NRC).

19.

Johnson, W. J. and Thompson, C. M., " Westinghouse Emergency 4

Core Cooling System Evaluation Model - Modified October 1975 Version," WCAP-9168 (Proprietary) and WCAP-9169 (Non-Proprietary), September 1977.

l 20.

" Westinghouse ECCS Evaluation Model Sensitivity Studies,"

c i

WCAP-8341 (Proprietary) and WCAP-8342 (Non-Proprietary),

July 1974.-

21.

Salvatori, R., " Westinghouse ECCS - Plan Sensitivity Studies, WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July 1974.

22.

Johnson, W.

J., Massie, H. W.,

and Thompson, C.

M.,

" Westinghouse ECCS-Four Loop Plant (17x17) Sensitivity Studies," WCAP-8565-P-A (Proprietary) and WCAP-8566-A (Non-Proprietary), July 1975.

i i

t 23.

Letter NS-TMA-2030, dated February 12, 1979, Anderson, T. M.

(Westinghouse) to Denton, H. R. (NRC).

24.

DiNunno, J. J., et al., " Calculation of Distance Factors j

for Power and Test Reactor Sites," TID-14844, Division of Licensing and Regulation, AEC, Washington, D.C., 1962.

j 25.

USNRC NUREG-0409, " Iodine Behavior in a PWR Cooling l

System Following a Postulated Steam Generator Tube j

Rupture Accident," by Postma, A. K. and Tam, P.

S., dated January 1978.

I 26.

Maximum Safety Injection Worst Single Failure, NS-EPR-2538, December 22, 1981, letter from E. P. Rahe of Westinghouse Electric Corporation to R. L. Tedesco, Assistant Director of Licensing, and T. P. Speis, Assistant Director for i

Reactor Safety of the U.S. NRC.

2 SNUP S E S Anal sis, C-8 01,FNe0278 Janu y

19 tter om N. A. Pet ck of ANUPPS H.

t D' rect r, Off e of uclear eacto/ Regula ion,

.S.

\\

I Letter NS-K%-2538, dated December 22,19s1, Rahe y E. P.

l 1y.

(Westinghouse) to Tedesco, R. L. (NRC).

l 6.1. 7~'

15.6-32 A'00 -

l

i TABLE 15.6-10 TIME SEQUENCE OF EVENTS FOR LOSS-OF-COOLANT ACCIDENTS

~

I Accident Event Time (sec),

i Large, break LOCA 0.0 a..DECLG C =0.8 Start D

Reactor trip signal 0.82 4

Safety injection signal 1.0 Accumoliter injection begins ~- ~- ~ 13.9 -- -

b 24.2 End-ofi ypass i

j Pump injection begins 26.0 End-of-blowdown 27.0 l

Bottom of core recovery 37.4

~

Accumulator empty 47.9 0.0 j

b. DECLG C =0.6 Start p

Reactor trip signal 0.83 h'd'd8M Safety injection signal 1.15 i

g)ryg 125Ed.ee83 Accumulator injection begins 15.9 4

• )itsuu 6ermasdaiod ump injection begins fp.6 l

nd-of-bypass 26.

gy.,g7) nd-of-blowdown 29.

1 Bottom of core recovery 40.T Accumulator empty 50.1 ir 5

de DECLG C =0.4 Start 0.0 D

Reactor trip signal 0.82 Safety injection signal 1.42 l

Accumulator injection begins 20.7

/

I Pump injection begins 26.4 End-of-bypass 34.6 l

End-of-blowdown 37.7 j

[

Bottom of core recovery 48.9 Accumulator empty 56.9 l

j Small break LOCA 0.0

a. 3 inch start

~

j Reactor trip signal 29.7 Top of core uncovered 623 Accumulator injection begins N/A l

Peak clad temperature occurs 1351 Top of core covered 2300 l

0.0

- C. DECLG c,=0.6 start Reactor trip signal 0.83 i

( Hs;'H-, 52 witW safety injection signal 1.15 62 Tessad Diesel Accumulator injection begins 15.9 1mrde Sep>

ump injection begins Jlta j

26.

nd-of-b p ass nd-of-blowdown 29.

Bottom of core recovery 40 S Accumulator empty

$Y,3 i

1 t

'I e

-._.m,_w.

-p

+.-

,--~_e

,n-.

,y_.---,

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

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SNUPPS TABLE 15.6-11 l

LARGE BREAK LOCAL RESULTS

• FUEL CLADDING DATA g

Results DECLG C =0.8 DECLG C =0.6 DECLG C =0.4 D

n D

Peak clad temperature, (*.

CAsc g, fm b beC fas C i

N gos 3 c,

~2.1 7 4 - Y 287T.2 I'If Peak clad location, ft I 7.5 7.5 75" 7.5 Local Zr/H O reaction,IInax. (%)

4.38 5 38 d,. /8 0.85 2

l Incal Zr/H O location, ft 7.5 7.5 7 5-7.5 2

i Total Zr/H 0 reaction, %

<0.3

<0.3 40.3

<0.3 Not rod burst time, sec 29.2 27.8 27 8 N/A Not rod burst location, ft 6.0 6.0 g,o N/A

,- Secum is c.c. 3 5 J 4e Reke kt11Me Is.c,--*a far-c. deGihsn of (dses w 44 4 d.

3 i

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3 w

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a., oo.1os..nm..1.-,s, ovu SNUPPS FlauRE 15.67 PEAK CLAD TEMPERATURE -

oEcta (c -o.m o

00'092 00'002

• o m.

d b.

00'091 U

w 3

H 00'001 00*09 I

0*O o

o 9

o o

o d

d o

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8 E

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U S3380TJ 00810H*dM31*OAY OY10 SNUPPS F1GURE 15.6-7 PEAK CLAD TEMPER ATURE DECLG (C D=0.6 - M ed inranAcTrva enewies n.c aA.u ay r.ose 4Axx.xxx.o i.oor.o4o4e 4 I2 $ce 3As $7A

00'0S2

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ar 00'001 i

00*05 O*O o

o o

o o

o 8

8 8

8 8=

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Li 533W030) OOW 10H*dM31*SAY OY13 SNUPPS FIGURE 15.6-97n PEAK CLAD TEMPERATURE DECLG (C D=0.6-M AX S !!.-

gyrggwggg IZsse % svaer)

d 3

ik ww I

1 a

H s

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e o

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m re gho' N

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N N

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uo um um,

FIGURE 15A4 DECLG (Cp = OA) DOWNCOMER AND CORE WATER LEVELS DURING REFLOOD ge=~ e e D * * - 4e as agaup mge ggaspe em egne _

m+ M -

ee*

+ Sk '

S 6

8 8--

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9 0

9 0

9 0

9 O

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O b

W N

O pgy itsee (1.41 73A31 W31YM SNUPPV FIGURE %.6-93 0.6' DOWNC OMER-DECLG (C ANDk= ORE WATER s:m.uErW M.8x?M.o**4 LEVELS DURING REFL00D i 1

00'009 t

00'00v j

4 00'00E e

g

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Q 00'002 00'001 0*O o

e o

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5p u.m,sas,us m SNUPPS 15.6 -581*

FIGURE

-0.6 M AX S DECLG (C DOWNCOMER kND CORE WA WU'Mai?.ExTh LEVELS DURING REFLOOD

3

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8 m

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8 R

S W

8 arv.1 a

a a

a a

e a

e 6

em (33S/NI) 31VW 0001d SNUPPS FIGURE 15,610 DECLG (Co = 0.6) COR E INLET VELOCITY DURING REFLOOD

00'059 00'o0*

.00'00E Gwe w

3w 00'002 00'001 1

O'o 1

8 E

S E

E S

E g,4 ST'.7/6 O

g 103S/NI) 31YW 000"Ld i%d#kp4I 3

SNUPPS

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J FIGURE 15.6-k to I

DECLG (C

-0.69 CORE INLET fELOCITY n.cm.ulYE.EeYd"xx.ExEfeE.o*Ms4 DURIRG REFL000

oo*ccs oo*oot oo'ooE 8we C

oo*ooz o0'001 3lt o*o 3

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(33S/NI) 31YW 0007J SNUPPS i

FIGURE 15.6-9toa.

DECLG (C 0.6 M AX SII CORE INLk=T VELOC o,em.u3r"l# M.8x T M.o we4 DURING REFL00D

, (.

R i

oe

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, = =

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=

a ge=

a

g..,

ntV.1 (do WH gid/018) 1N313!dF.03 W3dSNVW1 AV3H g

SNUPPS i

FIGURE 15A.13 CORE HEAT TRANSFER COEFFICIENT -

DECLG (Cp = OA) j

00'0S2 s

00'002 1

/

00'051

}*

.o 00'001 00*0s e

c' e

y' N'A I

0*0 t%ff Qpsp c

e c

i Ho o o o g Ho o o o e Ho o C.

O o

P" ' "

_pgg g 8

-pS S R 2 r

s 3-WH-213/018 SNUPPS 1N3131dd300*SNYW11Y3H FIGURE 15.6-S G CORE HEAT TRANSFER C OE FFIC IE NT o.cm.uSr"l# M M 5ll.o we4 DECLG (C D:0.

~'

~

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~

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i FIGURE 15.6-Flk

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,d-CORE HE AT TRANSFER C OEFFIC IENT DECLG (C D=0.6 M AX S n.cm.JP.EsfdE8x7.ES94

l A

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SNUPPS

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F60URE 15A 14 FLUID TEMPERATURE -

DECLG (CD = 0.8)

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t e

oo'osz

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_u sasse3013Mnivu3dn31 orn u SNUPPS FIGURE 15.6 - 3 14 4

3 j

, FLUID TEMPERATUR tamArve amics

' DECL G (C.D=0.

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FIGURE 15.6-$ l4g FLUID TEMPER ATURE DECLG (C D=0.6 M AX S 0 inranacTtva enwHics D.cHA.MAYF.0934AXX.XXX.01.Oll.040404

w I

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4 6 8 101 i2 i

TIME (SECONDS)

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