LD-85-021, Withdraws Encl Info Re Proposed CESSAR Changes to Reduce Min Required LPSI & HPSI Flows.Reanalysis of CESSAR Events Provided on plant-specific Basis.Evaluation of Events for HPSI & LPSI Changes Impractical

Withdraws Encl Info Re Proposed CESSAR Changes to Reduce Min Required LPSI & HPSI Flows.Reanalysis of CESSAR Events Provided on plant-specific Basis.Evaluation of Events for HPSI & LPSI Changes Impractical

ML20116D854
Person / Time
Site:	05000470
Issue date:	04/26/1985
From:	Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	Thompson H Office of Nuclear Reactor Regulation
References
LD-85-021, LD-85-21, NUDOCS 8504300068
Download: ML20116D854 (20)

Text

,

C-E Power Systems Tel. 203/688-1911 Combustion Engineering, Inc. Telex: 99297 1000 Prospect Hill Road Windsor, Connecticut 06095 M POWERSYSTEMS STN 50-470F April 26,1985 LD-85-021 Hugh L. Thompson, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

High and Low Pressure Safety Injection Flow

References:

(A) C-E Letter LD-84-070, A. E. Scherer to D. G. Eisenhut, dated December 5, 1984 (B) C-E Letter LD-84-071, A. E. Scherer to D. G. Eisenhut, dated December 5, 1984

Dear Mr. Thompson:

Included in Reference (A), which transmitted several proposed CESSAR changes for NRC Staff review, was a change which reduced the mininum required Low Pressure Safety Injection (LPSI) flow. Justification for this reduction was also included in Reference (A). Reference (B) forwarded a proposed CESSAR change to reduce the minimum required High Pressure Safety Injection (HPSI) flow and included justification for that reduction.

Since that time, treetings have been held with the Staff concerning the justification of tie aforementioned changes and to what extent re-analysis of CESSAR events is r' quired. Combustion Engineering (C-E) believes that adequate justification for the LPSI and HPSI flow reductions was provided in References (A) and (B), and, therefore, no further analysis is required. The Staff, however, feels that re-analysis of several safety analyses is required to provide a clean record of the events described in CESSAR.

Since re-analysis of affected events is currently being provided on a plant specific basis on the Palo Verde docket to address these and other minor deviations from CESSAR, C-E believes it to be impractical to evaluate the same events solely for the HPSI and LPSI changes. C-E has, therefore, reluctantly reconsidered our request to reduce the required HPSI and LPSI flow in CESSAR and hereby withdraws the information previously submitted to identify and justify those changes. The pages from References (1) and (2) being withdrawn are provided as Attachments (1) and (2), respectively.

8504300068 850426 0 PDR ADOCK 0500 y I

r Mr. Hugh L. Thompson LO-85-021 April 26,1985 Page 2 If you have any coments or questions on this subject, please feel free to call me or Mr. G. A. Davis of my staff at (203) 285-5207.

Very truly yours, COMBUSTION ENGINEERING, INC.

Director Nuclear Licensing AES:las cc: P. Moriette (NRC) l l

i

l l

LO-84-070 1 Attachment Page 1 of 2 !

i

SUMMARY

OF CHANGES Chapter 5: Reactor Coolant System and Connected Systems Section 5.2.2.10 and Figure 5.2-2 are changed to reflect a lower maximum AT across a steam generator. This change is made to allow additional operational flexibility and limits the maximum AT at which a Reactor Coolant Pump can be started to 100*F. Additionally, clarification is added to more accurately reflect the Shutdown Cooling System relief valves used in construction.

Figure 5.2-1 is modified to credit the typical difference in height between the pressur.izer and the shutdown cooling system. This change was also made to provide operational flexibility.

Section 5.4.2.4.1 is changed to correct the Section numbers referenced. This change is considered editorial in nature.

Section 5.4.10.3 is revised to indicate that pressurization rate testing is not performed during Hot Functional Testing. Pressure control setpoints are determined analytically and checked during Power Ascension Testing.

Chapter 6: Engineered Safety Features Tables 6.3.3.3-1 and 6.3.3.3-2, and Fi jgur.3r3. .- , are modified to ease Low Pressure Safety Injection flopwtrements. These flaw changes were found to be sufficient to preser s etTpresent ECCS performance results. Comparison of the revised f th the figure currently in CESSAR demonstrates the

& e, insi ce of this change on the worst case postulated break.

Q;Ntm Appendix 6B: Iodine Removal System Section 7.16.4 is revised to remove unwarranted restrictions on the transfer of hydrazine and to clarify the qualifications of the arrangement used to transfer hydrazine to the Spray Chemical Additional Tank.

Chapter 7: Instrumentation and Controls Section 7.1.2.10 is revised to clarify conformance to IEEE 384-197a as augmented by Regulatory Guide 1.75 (Rev. O, 2/74). It shows that the commitrw!nt to perform specific analyses have been completed, i

i

TABLE 6.3.3.3-1

~

SAFETV INJECTION DUMD5 MINIutM DEL!vERED cL:W 70 RC5 (Assuming One Emergency Jererator Fai'ec)

C10 w Rate Per Injectier -cint" (;::m )

RCS Pressure A A B B (psic) 1 2 1 2 1775.0 0 0 0 0 1650.0 50.0 50.0 50.0 50.0 1440.0 -

100.0 100.0 100.0 100.0 1270.0 125.0 125.0 125.0 125.0 1095.0 150. 150.0 150.0 150.0 865.0 17 .0 175.0 175.0 175.0 605.0 0.0 200.0 200.0 200.0 310.0 225.0 225.0 225.0 225.0 200.0 234.0 234.0 234.0 234.0 t%o M l.0 ' ' ' * . 0 694o ' ' ' ? 0 Jk

• M 3>

• M 100.0 241.e'7t!.082S d'750.0 243.0 243.0 50.0 sag 4.e 2'00.0808 4 0'00.0 246.0 246.0 0 25r1 2500.029673 2500.0 250.0 250.O a

Injection Point is assumed to be attached to the broken pump discharge leg.

bk es @ (a h

TABLE 6.3.3.3-2 IE'.E:1,. $YSTEM PA:AME~EP. A'iD !.'!!TI AL CC'C:~::'.5 S??ALL 3REAK ECCS ERFOR"A' ICE ;*.ALv5:5 Ouantity value . nits Reactor Power Level (102", of flominal) 3876 ftWt Average Linear Heat Rate (102" of Nominal) 5.6 kw/ft Peak Linear Heat Rate -

15.0 kw/ft Gap Conductance at Peak Linear Heat Rate 1497 btu /hr-ft 'F Fuel Centerline Temperature at Peak Linear Heat Rate 3681 *F Fuel Average Temperature at Peak Linear Heat Rate 2319 *F Hot Rod Gas Pressure 1187 psia Moderator Temperature Coefficient at Initial Density 0.0 _:/*F 6

System Flow Rate (Total) 164.0x10 lbs/hr 6

Core Flow Rate 159.1x10 lbs/hr-Initial System Pressure 2250 psia Core Inlet Temperature 565 F Core Outlet Temperature 623 'F Low Pressurizer Pressure Scram Setooint 1600 psia Safety Injection Actuation Signal Setpoint 1600 psia Safety Injection Tank Pressure 608 asia High Pressure Safety Injection Pump Shutoff Head 1775. Osig Low Pressure-Safety Injection Pump Shutoff '

Head I  % psig bb 4 h (6 M

10 , , , ,

I i

• o * -

8 -

X d

R '

8 6

N h d

5 g 4 -

5

=

$ 2 -

Di i l I I I g

O 20 40 60 80 100 120 t

Tif1E AFTER RUPTURE, SECONDS

( (A M 1

c-E 1.0 x DOUBLE ENDED CUILLOTINE BREAK Figur.

IN PUMP DISCilARGE LEG 6.3.3.

SAFETY INJECTION FL0ll INTO INTACT DISCHARGE LEG 2-5L

C-E Power Systems T3 203/688-1911 Combustion Enginsnng. Inc. 41:x 99297 1000 Prospect Hill Road Windsor, Connecticut o6095 POWER m SYSTEMS STN 50-470F December 5, 19 LD-84-071 Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

High Pressure Safety Injection Flow

Dear Mr. Eisenhut:

In an effort to provide a suitable technica specification margin for High Pressure Safety Injection (HPSI) pump perf rmance for the first System 80' plant, a re-analysis of the most limitin small break Loss Of Coolant Accident (LOCA) has been performed. This small eak (0.05 ft cold leg) was selected 2

ds the basis for determining the effec of reduced HPSI pump delivery for the following reasons.

(1) Large break LOCAs are not nfluenced by HPSI flow.

(2) This break size and loc tion (0.05 ft2 cold leg) is the most limiting small break.

(3) Reduced HPSI pump p formance has no Impact on the consequences of the non-LOCA Chapter 1 safety analyses.

A comparison of the previ us peak clad temperature and two-phase mixtiJre height in the core is attached Figures 1 and 2). Also attached is a CESSAR change that is provided for y r review. It will be incorporated into CESSAR in the next amendment.

A review of Figures 1 and 2 indicates that the maximum peak clad temperature for this break si increased from 1557*F (from previous CESSAR analyses) to 1630*F. This in ease is attributed to the slightly longer period of core

_ uncovery result g from the decrease in HPSI flow delivered. This small break analysis is st 1conservativelyboundedbythegostlimitinglargebreakLOCA peak clad tem erature (2169*F occurs in a 1.0 ft double-ended cold leg guillotine b eak).

In surinary, a CESSAR change is forwarded to reflect a reduced HPSI pump flow.

This cha e was necessary due to as-built conditions in the first System 80 plant. re-analysis of tne most limiting small break LOCA demonstrates that system erformance remains well within the acceptance criteria of le.N e r Yi f. 7 4 W A

l 1

Mr. Darrell .G. Eisenhut LD- - 071 '

December 5, 1984 P e2 10 CFR 50.46. Additionally, the nigher resulting peak clad te erature remains at least 500*F below the limit case large break LOCA.

The attached change will be included in a future amendment o CESSAR. If you have any questions or comments, feel free to call me or . G. A. Davis of my staff at (203) 285-5207.

Very truly y rs, COMBUSTI ENGINEERING, INC.

7 A. . Scherer 4 D ector uclear Licensing AES:las Attach, QNer dib b #

cc: P. Moriette .

FIGURE 1 0.05 FT2 BREAK - REDUCED HFSI PUliP DELIVERY PEAK Cl.AD TEllPERATURE _

~ -

I6a* - .

$j O

lb - ,

, 7%& -

j'

/*

__ ps psy.

4 -

l l

\

A

/ g ,

j \ .

I \

Np ip -

/ \

t i/ \

4 1/ \ l 4 /88* -

fI I j j[ l '

l i -

i-

\

I l I I i

' 3 .

i '

5 _ . Lt i i i 4 i i 4 + ine 24. >

frw, ter, ,

Lc.%e r -

.. I

& &hrAv)k .

~

FIGURE 2  !

2 0.05 FT BREAK - REDUCED HPSI PUfir DELIVERY '

TWO-FHASE HIXTERE HEIGHT IN THE CORE

[  !

48 000

)..

f

. 40.000

/gened 4 mc -

l k .

/_ =

s' 32 000 r -

O u.J ..

24.000 " i '

N ,

S '

~

M" Q j r 16.000 f i

. 8.000 00 a o a o d d E E  !

o 8  ?

8 m ~

E m

o - -

• ktN e v-

.tJ A .Ar ,a n TIME IN SEC-

O 9

=

CESSAR CHANGES

l The four safety injection tanks (SITS) are piped so that each SIT feeds a

! single cold leg injection point. Thus:

a. for a break in the pump discharge leg, the SIT flow credited is 100%

of the flow from three SITS. The remaining SIT is assumed to spill out the break.

b. for breaks in other locations, the SIT flow credited is 100% of fo SITS.

Table 6.3.3.3-1 presents the high and low pressure safety injectic pump flow rates assumea at each of the four injection points as a func on of reactor coolant system pressurey gggg gng 6.3.3.3.3 Core and System Parameters The significant core and system parameters used in the sa break calcula-tions are presented in Table 6.3.3.3-2. The peak linear eat generation rate (PLHGR) of 15.0 kw/ft was assumed to occur 15% fro the top of the active core. A conservative beginning-of-life modera e temperature coeffi-cient of 0.0 cc/*F was used in all small break calcu tions.

The ECCS performance analyses as performed, do no account for steam generator tube plugging which may occur over the plant's i fetime.

The initial steady state fuel rod conditions are obtained from the FATES U) computer program. Like the large break, th small break analyses employed a hot rod average burnup which ma'ximized t e amount of stored energy in the fuel. Since the small break analysis used higher PLHGR than did the large break analysis (15.0 kw/ft vs 14.0 kw/f the fuel rod parameter values given in Table 6.3.3.3-2 differ from ose on Table 6.3.3.2-2.

Because the large break results are always more limiting than the small .

break results,*the small break an ysis is run at a higher PLHGR to prevent requiring a reanalysis should t large break results improve. Since the small break results are govere d mainly by the core liquid level transient (see Results Section below) ich is a function of the total core decay heat generation rate, the h her PLHGR does not significantly affect the l

small break results.

6.3.3.3.4 Contain nt Parameter.s ,

The small break anal is does not credit any rise in containment pressure.

Therefore, other t n the initial containment pressure, which is assumed to remain constant, containment parameters are employed for this analysis.

The initial cont inment pressure was assumed to be 0.0 psig.

6.3.3.3.5 Break Spectrum Six breaks areanalyzedtocharactgrizethesmgilbreakspectrum. Five breaks, r nging in size from 0.5 ft tg0.02ft were postulated to occur in the discharge leg., The 0.5 ft break was also analyzed for the large eag3 octrum (Sect { ion 6.3.3.2) and is defined as the transition brea size One break, 1 equal in area to a fully open pressurizer safety

]

In.sr87(B)(wtjeft.

Ne 3 6.3-zs Q&bx

)

INSERT A for the six break spectrum analysis identified in ragraph 6.3.3.3.5. Table 6.3.3.3-1A presents the safety injection (SI) p flow rates used in an alternate analysis of the limiting small bre LOCA. the 0.05 ft8 break in the i reactor coolant pump discharge leg. This eak was reanalyzed to demonstrate -

the acceptability of a small reduction the SI pump flowrate.

INSERT B 1

i The 0.05 ft8 brea'k which w determined to be the limiting break size and the

, most sensitive to the S ump flow capacity was also analyzed using the reduced SI pump flow scussed in paragraph 6.3.3.3.2.

i

.i t

, a; LJ<wn J

- - - - _ - _ - , - . - - . . - ..--._._.-..m, ,nn nnv w _ ,_-_ ,,,, , ,, _ - _.- ,- , _ -,--+ --m,_,,, w--. ew-, e,.-

valve, (.03 ft )2 was postulated to occur in the top of the pressurizer.

Table 6.3.3.3-3 lists the various break sizes and locations examined f this analysis.

,6.3.3.3.6 Results -

The transient behavior of important NSSS parameters is shown ni s he figures listed in Table 6.3.3.3-4. Table 6.3.3.3-5 summarizes the imoNetant results l of this analysis. Times of interest for the various breaks adalyzed are presented in Table 6.3.3.3-6. Aplotofpeakcladtempergtde(PCT)versus break size is presented in Figure 6.3.3.3-7. The 1 05 ft bieak results in .

the highest clad temperature (44633&& of the small break analyzed ( w vem 73NJEtTIC S : :r 500*" ? n:: th:n th:t 7:;;rt:d '- R;ti:n 5.2.!; Sr th; 'initin;; */ *"hy 12 ; 5- " The break resu breakspectrumisthe0.2ft}tinginthenexthighest/CTofthesmall break with a PCT of 1030*F.

/

It is important to note the differences in the tr sient behavior of these twobreaksizes,becauseeachcharacterizesdiffeegtcontrolliggfeatures of small breaks. The larger breaks (between 0 t and 0.5 ft ) temperature transientsareterminatedbytheactionofthy/.safetyinjectiontankg(SIT) whereas the temperature transients for the spaller breaks (<*0.05 ft ) are terminated solely by the high pressure saf9ty injection pump (HPSIP) prior to the2 actuati 0.2 ft to n jteSITs. For the inpraediate break sizes (approxiestely 0.05 ft ) both the SITS and HPSIP. play an important part in j terminating the transient, with the HP (P becoming more important as the i break size decreases.

e a function of break size remains AsshowninFigure6.3.3.3-7,PC{greak.

fajrlyconstantuntilthe0.2ftj Then the PCT rises for the 0.05 ft and then falls for the 0.02 ,(t break. This rise and fall in PCT can be adequately predicted 2 by obsetving the transient behavi r f r breaks less than or equal to 0.2 ft The peak clad temperature predictably affected by:

1) Time of initial cor uncovery,

2) Depth of core un very, and

3) Duration of core uncovery.

2 As the break six becomes progressively smaller than 0.2 ft , the inner vessel two phas level follows a definite pattern:

1) The time of initial core uncovery is later,

2) The d th of core uncovery is less,

3) The ime of core uncovery becomes longer, and,

4) T e actuation of the SITS is later during the period of core uncovery nd eventually does not occur.

9L wo_ '

6.3-2s

INSERT C The .05 fts case yeilds a peak clad temp ure of 1557'F based on the SI pump flow capacities of Table 6.3.3.3-1 an 30*F based on the SI pump flow capacities of Table 6.3.3.3-1A. either case the result is more than 500*F higher than the other small br cases presented yet more than 500'F below the limiting large breaks orted in Section 6.3.3.1.

b h c.

g)$ AVA

\.

1 This trend continues until the core does not2 uncover at ajl. For System 80 this occurs for a break size between 0.05 ft anc 0.02 ft (and for all smaller breaks). l As the. break size decreases, both the later time of initial core uncovery and its shallower depth tend to mitigate the temperature transient. However, )

In 1 theincreaseddurationofugcoveryactsinthgoppositedirection.

progressing from the 0.2 ft break to 0.05 ft breakation the increased dur/

dominates and therefore the peak clad temperatures risg. This trend' continues '

until a break size is reached, typified by the 0.05 ft break, where the l three parameters are balanced. For breaks smaller than this, the' increase in time to initial core uncovery and the shallower depth dominate c'ausing less

, s: vere temperature transients. Thistgendcontinuesuntilthe/coredoes

nst uncover as typified by the 0.02 ft break. Thus, by analyzing se$eral bre k sizes over this range, the behavior of PCT versus break size can be adequately determined. / i

/

To demonstrate the conservatism associated with the snail break ECCS perfor- '

2 mance results provided herein, the 0.05 ft break was/ reanalyzed using a more realistic measure of the decay heat generationjrate. As required by App:ndix K to 10CFR50, the spectrum analysis employed a decay heat generation i

rageequalto1205ofthestandardANScurve. The reanalysis of the 0.05 ft break used a decay heat generation rate equal to 1005 of the ANS curve.

This one change reduced the peak clad temperature t M"?" O CT.

l by here. $Gm 500*F 6.3.3.3.7 Instrument Tute Rupture e a l

/

• In addition to the.snm.small breaks discussed above, the rupture of an in-corainstrumenttube'wasconsidergd. A break, equal in size to a completely I severed instrument tube (0.003 ft ) was postulated to occur in the reactor vessel bottom head. /

/

Following rupture, the primary system depressurites until a reactor scram signal and safety injection actuation signal (SIAS) are generated due to low pressurizer pressure at 1600 psia. The assumed loss of offsite power causes the primary coolant pump and the feedwater pumps to coast down. ,

1 After the 30 second delay required to start the emergency diesel and the high pressure safety injection pump, safety injection flow is isitiated to the reactor vessel. At this time an emergency feedwater pump is also started, providing a source of cooling to the steam generators. Due to the assumed failure of one diesel, only one high pressure safety injection pump cnd one emergency feedseter pump are available. (Four SITS and one low pressure safety injection pump are also available but do not inject due to the high RC5 pressure.) The steam generator secondary sides also become issisted at this time.

Th3 primary side depressurization continues accompanied by a rise in sec'ondary sida pressure until the secondary side pressure reaches the lowest set pointofthe/teamgeneratorsafetyreliefvalves. The primary system pressure continues to fall until it is just slightly greater than the secondary side pressure. At this point, the flow from the one operating HPSIP (66 lbs/sec) exceeds the leak flow (26.4 lbs/sec). Therefore the .

l b qe.

• 6.3.30 l

Mbah

i Table 6.3.3.3-1A SAFETY' INJECTION PUMP 5 MINIMUM DELIVERED FLOW T0,Rt3 '

(Assuming one Emergency Generator FailedV Flow Rate Per Injecti Point *(gpa)

RCS Pressure 82 g A1 A2 81 1700 .5 .5 .5 .5 51.25 51.25 51.25 51.25 1581 1483 76.75 76'75 76.75 76.75 1349 102.75 102.75- 102.75 102.75 1199 128.75 128.75 128.75 128.75 993 155.25 155.25 155.25 155.25 782 181.50 181.50 181.50 181.50 605 200.0 200.0 200.0 200.0 310 225.0 225.0 225.0 225.0 234.0 234.0 234.0 234.0 200 130 581.0,- 581.0 240.0 240.0 1282.0 1282.0 243.0 243.0 100 1884.0 1884.0 246.0 246.0 50 2357.0 2357.0 250.0 250.0 0

• Injection Point /Al is assumed to be attached to the broken pump discharge leg. /

l l

l l GkU); 9 A n l

I l

l l

TABLE 6.3.3.3-2 l GENERAL SYSTEM PARAMETER AND INITIAL CONDITIONS SHALL BREAK ECCS PERFORMANCE ANALYSIS Quantity Value Units Reactor Power Level (102% of Nominal) 3876 MWt Av; rage Linear Heat Rate (102% of Nominal) 5.6 kw/f Peak Linear Heat Rate 15.0 ft Gap Conductance at Peak Linear Heat Rate 1497 btu /hr-ft 'F Fu]l Centerline Temperature at Peak Linear Heat Rate 3681 'F Fuel Average Temperature at Peak Linear Heat Rate 2319 'F Hot Rod Gas Pressure 11 psia Moderator Temperature Coefficient at Initial Density 0.0 ao/*F 0 '

System Flow Rate (Total) 164.0x10 lbs/hr '

0 Core Flow Rate -

159.1x10 lbs/hr Initial System Pressure 2250 psia Core Inlet Temperature 565 'F Core Outlet Temperature 623 'F ,

Low Pressurizer Pcessure Scram Setpoin 1600 psia Safety Injection Actuation Signal Se int 1600 psia Safety Injection Tank Pressure 600 psia I High Pressure Safety Injection P mp Shutoff

Head 1775. g) psig

! Low Pressure Safety Injectio Pump Shutoff /

psig I

Head

• g (4 S ~

% 4 <, eA2 sz A,a f A n.= m - l y*t^ fA'='a.# ^ *L o 0 4~ ,

G/s 2. .r. 3 - l A 2

0%es a;Mw -

4

--~, , ---.-.,----,--,.--,_,,m,----n--- -

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

TABLE 6.3.3.3-5 FUEL ROD PERFORMAtlCE

SUMMARY

SMALL BREAK SPECTRUM

~

Maximum Clad (a) Peak Local (b) Hot Rod (C)

Break Size Surface Temperature Zirconium Oxid. . Zircortid Oxid.

(ft ) 2 (*F) ( ",)

f(".)

2 M <.0020 <.0003 O.50 ft /PD 954 2

• 0.35 ft /PO 932 '

<.0015 <.0002 2

0.20 ft /PD 1030 <.0041 <.0007 2

0.05 ft /PD 1557 <.8825 <.1430 2

995 <.0003 0.02 ft2/PD @ <.00) 0.03 ft fgg 1012 <.0011 <.00004 g.gff/)

/430 < t 41,2 6 4 2023' (a) Acceptance, Criteria is 2200*E. /

(b) Ar teptance Criteria is 17. -

(c) Acceptance Criteria isA.0','. Hot rod oxidation values are given as aconservativeindicaIionofcore-wideoxidation.

(d.) Arto K =/ k x= .Se,. .$1 f =f & AsM m 1a*.4/c .1.1. 3 - I a) AroM am*f;ed. J Aj .tRt .it'z M f rh 4%d a 1*e .4/e d'. 3. 3 . 3 - t A .

b qt L)Ae 4

4

, . . , - , , , , - , , --, e __--

TABLE 6.3.3.3-6 TIMES OF INTEREST FOR SMALL BREAKS (Seconds)

~

Break Hot Spot Size . Peak Clad (ft 2) HPSI Pump On LPSI Pump On SI Tanks On Temp. Occurs 2 '

0.50 ft /PD 46.5 158.0 -

142.0 160.0 2

0.35 ft /PD 50.0 244 204.0 235.0 0.20 ft /PD 62.0 445 400:0 442.0 2

0.05 ft /PD 208.0 a.

/ b. 2010.0 2

0.02 ft /PD 492.0 a. b.

437.0 6-0.03 ft2 /HL ) 585.0 a. / b. 540.0 g.gggpp Y 2/2 0 m b /900 i .

. /

/

/ '

/

a. Calculation terminated before time of LPSI pump activation.

/

b. Calculation terminated before initiation of SI tank discharge -

4. drw u s /.ee k A .f? y & r~%,/Ce&4 h %A/e, 6 3.2.3-I

d. J'M w f# h f2 f f $ ~^A= M "

h 14/s. 6 3. 3.3 - /A.

/

e6))<

L L60~

.-----,,---,-n , , . - - ,v,- , - - - - - - , - - - - , - - - - - - , - . - - - - - - - - - - - - _ _