LD-84-071, Forwards CESSAR Changes Re Reduced HPSI Pump Flow.Reanalysis of Most Limiting Small Break LOCA Demonstrates That Sys Performance Is within Acceptance Criteria of 10CFR50.46

Forwards CESSAR Changes Re Reduced HPSI Pump Flow.Reanalysis of Most Limiting Small Break LOCA Demonstrates That Sys Performance Is within Acceptance Criteria of 10CFR50.46

ML20100K350
Person / Time
Site:	05000470
Issue date:	12/05/1984
From:	Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
LD-84-071, LD-84-71, NUDOCS 8412110071
Download: ML20100K350 (14)

Text

.

1-E Power Systems T1.203/688-1911 Cornbustion Engineering, Inc. Telex: 99297 1000 Prospect Hill Road Windsor, Connecticut 06095 POWER M SYSTEMS STN 50-470F December 5,1984 LO-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 technical specification margin for High Pressure Safety Injection (HPSI) pump performance for the first System 80" plant, a re-analysis of the most limiting small break Loss 2

Of Coolant Accident (LOCA) has been performed. This small break (0.05 ft cold leg) was selected as the basis for determining the effect of reduced HPSI pump delivery for the following reasons.

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

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

(3) Reduced HPSI pump performance has no impact on the consequences of the non-LOCA Chapter 15 safety analyses.

A comparison of the previous peak clad temperature and two-phase mixture height in the core is attached (Figures 1 and 2). Also attached is a CESSAR change that is provided for your 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 size increased from 1557 F (from previous CESSAR analyses) to 1630*F. This increase is attributed to the slightly longer period of core uncovery resulting from the decrease in HPSI flow delivered. This small break analysis is still conservatively bounded by the gost limiting large break LOCA peak clad temperature (2169'F occurs in a 1.0 ft' double-ended cold leg guillotine break).

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

This change was necessary due to as-built conditions in the first System 60 plant. A re-analysis of the most limiting small break LOCA demonstrates that system performance remains well within the acceptance criteria of hhkkD 0 00 4C A

l (I

~

Mr. Darrell .G. Eisenhut LD 071 December 5, 1984 Page 2 10 CFR 50.46. Additionally, the higher resulting peak clad temperature remains at least 500*F below the limit case large break LOCA.

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

Very truly yours, COMBUSTION ENGINEERING, INC.

7 A. E. Scherer 4 Director Nuclear Licensing AES:las Attach, cc: P. Moriette 1

FIGURE 1 .

0.05 FT2 BREAK - REDUCED FiFSI PUtiP DELIVERY PEAK CLAD TEllPERATURE  !

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FIGURE 2 i

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2 0.05 FT BREAK - REDUCED HPSI PuilP DELIVERY TWO-FHASE lIIXTERE HEIGHT IN THE CORE  :

48 000

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. 40.000 h4Mt -

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32.000 o

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F CESSAR CHANGES

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4 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 four SITS.

Table 6.3.3.3-1 presents the high and low pressure safety injection pump flow rates assumed at each of the four injection points as a function of reactor coolant system pressureX WEM% fndy 6.3.3.3.3 Core and System Parameters-The significant core and system parameters used in the small break calcula-tions are presented in Table 6.3.3.3-2. The peak linear heat generation rate (PLHGR) of 15.0 kw/ft was assumed to occur 15% from the top of the active core. A conservative beginning-of-life moderator temperature coeffi-cient of 0.0 ao/*F was used in all small break calculations.

The ECCS performance analyses as performed, do not account for steam generator tube plugging which may occur over the plant's lifetime.

The initial steady state fuel rod conditions were obtained from the FATES U) computer program. Like the large break, the small break analyses employed l

a hot rod average burnup which maximized the amount of stored energy in the fuel. Since the small break analysis used a higher PLHGR than did the large break analysis (15.0 kw/ft vs 14.0 kw/ft) the fuel rod parameter values given in Table 6.3.3.3-2 differ from those on Table 6.3.3.2-2.

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

break results,*the small break analysis is run at a higher PLHGR to prevent requiring a reanalysis should the large break results improve. Since the small-break results are guverened mainly by the cora liquid level transient (see Results Section below) which is a function of the total core decay heat generation rate, the higher PLHGR does not significantly affect the small break results.

l 6.3.3.3.4 Containment Parameter,s l ,

1 The small break analysis does not credit any rise in containment pressure.

Therefore, other than the initial containment pressure, which is assumed to remain constant, no containment parameters are employed for this analysis.

The initial containment pressure was assumed to be 0.0 psig.

6.3.3.3.5 Break Spectrum Sixbreakswereanalyzedtocharactgrizethesmglibreakspectrum. Five breaks, ranging in size from 0.5 ft tg0.02ft were postulated to occur in the pump discharge leg.r The 0.5 ft break was also analyzed for the large brea breaksize[3fectrum(Sectfion6.3.3.2)andisdefinedasthetransition1 One break, equal in are

)

IturM(B) yned/*fe-6.3-28

INSERT A for the six break spectrum analysis identified in paragraph 6.3.3.3.5. Table 6.3.3.3-1A presents the safety injection (SI) pump flow rates used in an alternate analysis of the limiting small break LOCA, the 0.05 ft* break in the reactor coolant pump disrtarge leg. This break was reanalyzed to demonstrate the acceptability of a small reduction in the SI pump flowrate.

INSEP.T B The 0.05 ft2 break which was detennined to be the limiting break size and the most sensitive to the SI pump flow capacity was also analyzed using the reduced SI pump flow discussed in paragraph 6.3.3.3.2.

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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 for this analysis.

6.3.3.3.6 Results The transient behavior of important NSSS parameters is shown in the figures listed in Table 6.3.3.3-4. Table 6.3.3.3-5summarizestheimpNtantresults l of this analysis. Times of interest for the various breaks analyzed are presented in Table 6.3.3.3-6. Aplotofpeakcladtempergture(PCT)versus break size is presented in Figure 6.3.3.3-7. The o.05 ft break results in .

the highest clad temperature '""*") of the small breaks analyzedgM 73MEtT[C 500** S ::r th:n th:^ r g:rt:d '- R :t h: 5.2.2.' nr th: ' tith;; #/""rf i :::

ur;;: W" ThebreakresuftinginthenexthighestPCTofthesmall break spectrum is the 0.2 ft break with a PCT of 1030*F.

'.t is important to note the differences in the transient behavior of these twobreaksizes,becauseeachcharacterizesdifferegtcontrolliggfeatures i of small breaks. The larger breaks (between 0.2 ft and 0.5 ft ) temperature transientsareterminatedbytheactionofthesafattyinjectiontankg(SIT) whereas the temperature transients for the smaller breaks (<*0.05 ft ) are terminated solely by the high pressure safety injection pump (HPSIP) prior For the intermediate break sizes (approximately 2

totheactuationof)theSITs.

0.2 ft to 0.05 ft both the SITS and HPSIP play an important part in terminating the transient, with the HPSIP becoming more important as the break size decreases.

AsshowninFigure6.3.3.3-7,PCjasafunctionofbreaksizeremains l fajrly constant until the 0.2 ft greak. Then the PCT rises for the 0.05 i ft and then falls for the 0.02 ft break. This rise and fall in PCT can be adequately predicted2 by observing the transient behavior for breaks less '

than or equal to 0.2 ft The psak clad temperature is predictably affected by:

i

1) Time of initial core uncovery,

2) Depth of core uncovery, and

3) Duration of core uncovery.

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

1) The time of initial core uncovery is later,

2) The depth of core uncovery is less,

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

. 4) The actuation of the SITS is later during the period of core uncovery and eventually does not occur.

I i

6.3-29

INSERT C The .05 ft2 case yeilds a peak clad temperature of 1557'F based on the SI pump flow capacities of Table 6.3.3.3-1 and 1630*F based on the SI pump flow capacities of Table 6.3.3.3-lA. In either case the result is more than 500'F higher than the other small break cases presented yet more than 500*F below the limiting large breaks reported in Section 6.3.3.1.

e.

I

l This trend continues until the core does not 2uncover at a}1. For System 80 this occurs for a break size between 0.05 ft and 0.02 ft (and for all smaller breaks).

I As the. break size decreases, both the later time of initial core uncovery cnd its shallower. depth tend to mitigate the temperature transient. However, theincreaseddurationofugcoveryactsinthgoppositedirection. In przgressing from the 0.2 ft break to 0.05 ft break the increased duration 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 three parameters are balanced. For breaks smaller than this, the increase in i time to initial core uncovery and the shallower depth dominate causing less I severe temperature transients. This tgend continues until the core does nst uncover as typified by the 0.02 ft break. Thus, by analyzing se9eral break sizes over this range, the behavior of PCT versus break size can be adequately determined.

To demonstrate the conservatism associated with the small break ECCS perfor-2 mance results provided herein, the 0.05 ft break was reanalyzed using a more realistic measure of the decay heat generation rate. As required by Appendix K to 10CFR50, the spectrum analysis employed a decay heat generation rageequalto120%ofthestandardANScurve. The reanalysis of the 0.05 ft break used a decay heat generation rate equal to 100% of the ANS curve.

This one change reduced the peak clad temperature ' x 1557"I O 1000"T.

mere. cgm 500*F 6.3.3.3.7 Instrument Tube Rupture In addition to the.ada.small breaks discussed above, the rupture of an in-l careinstrumenttube'wasconsidergd. A break, equal in size to a completely severed instrument tube (0.003 ft ) was postulated to occur in the reactor vessel bottom head. .

- Following rupture, the primary system depressurizes until a reactor scram '.

signal and safety injection actuation signal (SIAS) are generated due to low pressurizer pressu-o at 1600 psia. The assumed loss of offsite power causes the primary coolant pump and the feedwater pumps to coast down.

After the 30 second delay required to start the emergency diesel and the .

. high pressure safety injection pump, safety injection flow is isitiated to l tha reactor vessel. At this time an emergency feedwater pump is also l started, providing a source of cooling to the steam generators. Due to the l cssumed failure of one diesel, only one high pressure safety injection pump l cnd one emergency feedwater pump are available. (Four SITS and one low

! pressure safety injection pump are also available but do not inject due to l the high RCS pressure.) The steam generator secondary sides also become isolated at this time.

l The primary side depressurization continues accompanied by a rise in sec'ondary

side pressure until the secondary side pressure reaches the lowest set i point of the steam generator safety relief valves. The primary system pressure continues to fall until it is just slightly greater than the sccondary side pressure. At this point, the flow from the one operating HPSIP (66.3 lba/sec) exceeds the leak flow (26.4 lba/sec). Therefore the ,

l 6.3.30 I

Table 6.3.3.3-1A ,

SAFETY" INJECTION PUMPS MINIMUM OELIVERED FLOW TO RCS (Assuming one Emergency Generator Failed)

Flow Rate Per Injection Point * (gpm)

RCS Pressure 81 B2 g A1 A2

.5 .5 .5 .5 1700 51.25 51.25 51.25 51.25 1581 76.75 76.75 76.75 76.75 1483 102.75 102.75 102.75 102.75 1349 128.75 128.75 128.75 128.75 1199 155.25 155.25 155.25 155.25 993 181.50 181.50 181.50 181.50 782 200.0 605 200.0 200.0 200.0 225.0 225.0 225.0 225.0 310 234.0 200 234.0 234.0 234.0 581.0 581.0 240.0 240.0 130 243.0 100 1282.0 1282.0 243.0 1884.0 1884.0 246.0 246.0 50 250.0 0 2357.0 2357.0 250.0

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

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TABLE 6.3.3.3-2

- CENERAL SYSTEM PARAMETER AND INITIAL CONDITIONS Si1ALL BREAK ECCS PERFORMANCE ANALYSIS Quantity Value Units Reactor Power Level (102% of Nominal) 3876 11Wt 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 Lir. ear Heat Rate 2319 *F Hot Rod Gas Pressure 1187 psia Moderator Temperature Coefficient at Initial Density 0.0 tm/ F 6 -

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

6 C@re Flow Rate -

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

Low Pressurizer Pr. essure Scram Setpoint 1600 psia Safety Injection Actuation Signal Setpoint 1600 psia l~ Safety Injection Tank Pressure 608 psia High Pressure Safety Injection Pump Shutoff Head 1775. g) psig l Low Fressure Safety Injection Pump Shutoff ,

/1700

Head g psig I4$2 M 0dI -

Anac A .A sz mp de w[4 l G/< 6 3.3. 3 - t A 4

r , ---,n -- -, ,- - - . , - , , ~ , - , - -ee en e,- - -,-, -, w--- e,-- ,

e TABLE 6.3.3.3-5 FUEL ROD PERFORMAtlCE

SUMMARY

SitALL BREAK SPECTRUM Maximum Clad (8) Peak Local (b) Hot Rod (c)

Break Size . Surface Temperaturs Zirconium Oxid. . Zirconium Oxid.

2 (ft )

('F) (f) (". )

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

0.35 ft /PD 932.

<. 0015 <.0002 2 <.0007 0.20 ft /PD ,

1030 <.0041 0.05 ft /PD N 2 <.1430 1557 <.8825 2 <.0003 0.02 ft /PD 995 <.0011 2

0.03 ft /HL 1012 <.0011 <.00004 g.offf) ff30 < l*4/M l'2083' (a) Acceptance , Criteria is 2200*F.

(b) Acceptance Criteria is 17f..

(c) Acceptance Criteria is 1.0f,. Hot rod oxidation values are given as .

a conservative indication of core-wide oxidation.

(d.) Bru.K pd St .SI f ==fg & Ashd m 'fa.4/c C. 3. 3. 3 - i (g) Avd H 3ed. Sj JR fI M f rA t

.,4;a h d. m Ta.4/e C. s. 3. 3 - t A .

4 e

~ ~. --- - .,--.-.w..n,.--m.,.c ,o ,,,._.~,.,,,-,,,.,n.-,-,.-_,_.,_,--n.. ,-,.. , _,, ,-.~.m,,,-.,, --m.,,,, _ , - , , . , - , ,

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

Break Hot Spot 5123 . Peak Clad

-(ftc) 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 2

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 2

0.03 ft jgt M 585.0 a. b. 540.0 n.ggp/) 212 0 a b /900

a. Calculation teminated before time of LPSI pump activation. .

b. Calculation terminated before initiation of SI tank discharge

4. Brw . a,,}gul .#e JZ M & r~kH'A h %.,4/e. 6 3.3. 3 - /

d.

fruK w}f A f<,

J2 fff%sJ A P d k fod/e. 6 3. 3.3 - /A.

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