TXX-9200, RHR Line Summary Rept

RHR Line Summary Rept

ML20094M480
Person / Time
Site:	Comanche Peak
Issue date:	03/23/1992
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20094M479	List: ML20094M477 TXX-9200, RHR Line Summary Rept
References
TXX-92009, NUDOCS 9204010007
Download: ML20094M480 (21)

Text

.

~

E bJ C Lo S LJ P E ~

i >

pgy 3/73(9V 9 zoo 9 l

RHR UNE

SUMMARY

REPORT B

1 i

I J

4 9204010o07 920323 pDR ADOCK 05000445 PDR WPP0749J/021Wz:xu

?

'. 0 thT_ RODUCTIOll AND BACKGROLWQ The United States Nuclear Regulatory Commission issued Bulletin 88 08 Supplement 3 (Reference 1) following the discovery of a valve leakage induced fatigue crack in the residual heat removal (RHR) suction piping at Cenkai Unit.1 nuclear power plant. This bull: tin requested utilities to identify susceptible piping systems, inspect potential crack locations and provide continuing assurance of piping integrity for the life of the unit.

TU Electric has bee.n very responsive to NkC Bulletin 88-08 requests. An initial evaluation of the Unit 1 RHR piping was completed in April 1989 (Reference 2 -

original issue). A second evaluation was completed in August 1989 (Reference 2 -

Supplemont 2). This evaluation considered a variation of th.: Strati 6 cation loading, i.e.

stratification initiating in the horizontal piping upstream of the first isolation valve.

As a result of the evaluatica performed in Reference 2, temporry temperantre momtoring locadons and critena were established, and TU Electric has been continuously monitoring the Unit 1 RHR piping to provide continuing assurance that the RHR suction piping is not subjected to combined. cyclic and static thermal and other stresses that could cause fatigue failure during thiremaining life of the units.

As a result of succeaful data collection for the drst fuel cyde, a review has been conducted to determine if valve leakage is caurring. -In addition, an evaluation has been. perior:wd to determine augmented inservice inspection intervals (based on fatigue usage and fatigue crack grcwth methodology, and assuming enutinuous cyclic valve leakage), thus satisfying NRC Bulletin 88 08 requirements without continuous momtoring. The purpose of this report is to dccument the results of the tuonitoring data review and to evaluate the postuired valve leskage condition. -

2 -

wpm 74f/022992:t0- 1-1

m 2.0 DVEPAli EVALUATION APPROACH 1 11 GMU .

TU Electric has placed temperature monitoring devices at severallocations oa the RHR suction piping to detect adverse thermal ttansients as described in item 3 of Reference 1.

After reviewing the monitoring results, it was found that there was no evidence of any 1

cyclic leakate in th: valves. Bis report addresses NRC Bulletin 88 08 requirements, evaluates monitoring data and presents technical justification to eliminate temporary monitoring dev;ces on Unit 1.

2.2 Technic;d20proaches While no temperature distributions which would indicate cyclic valve leakage were observed from the monitoring dat?., it was conservatively postulated that the out teakage-from the hot leg would occur through the isolation valves (8702A and 8702B). During plant operation, such leakage is postulated to cause stress cycles betwe:n leakage ano no-leakage. (The phenomena of leak /no-leah is considered b'erein as a postulated

, condition and should not be treated as a design conditiort.)

De steps in tht sttuctural evaluation of such postulated condition are listed below:-

Definition of stratified transients from po6mlated valve leakage

+

De6nition of stratified transients from monitored data (other,than valve

... =,-

)- .

Stress calculations from all cases of thermal strati 6 cation Fatigue usage calculation including postulated valve leakage, .nonitored transients and design transients-Fatigue crack growth calculation including' postulated valve leakage, monitored transients and design transients WFP(n48J/0n$P72:10_ 21

1 d

p Augmented ISI determination based on fatigue usage and fatigue crack Browth calculations.

Comanche Peak Unit I began commercial eperation August b,1990. Monitoring data from the Unit 1 piping has shown no evidence of cyclic valve leakage. Given that the units are new and have erperienced little or no fatigue cycles, it is highly unlikely that cracks are present in the RHR piping of Comanche Peak Unit 1. Fatigue usage and fatigue crack growth have been calculated assuming that cyclic valve leakage occurs, resulting in stress cycles. Fatigue usage provides a minimum time required to initiate the crack. Fatigue crack growth provides a minimum time required to propagate a craelt to 60% of the wall thickness, assuming an initial crack size of 10% of the wall thickness.

Augmented inservice inspection intervals based en conservItive fatigue usage ard fatigue crack grov,tn calculations provide a strong technical justi5 cation to eliminate temperature monitoring of the RHR piping, while still satisfying the requirements of NRC Bulletin 88-08.

o u

4 e

t Wrfc74al/03tu92:to 22

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

- . . - .. -. _ _ . . _ . ___J

, .m ~

\

3.0 MONTT10BJNG DAT 5. REVmW Monitoring Data tor the Comanche Peak Unit 1 RHR loops 1 and 4 suction piping ,

(Reference 3) were reviewed to determine if signif2 cant thernial stratification and/or cycling had occurred. These data were reviewed for the period frora 3/16/90 to 7/14/91.

The Loop 1 and Loop 4 RHR suction lines were instrumented on the pipe outer wall with resistance temperature detectors (RTD's) as shown in Figures 3-1 and 3 2. '"he purpose of each monitoring location is as shown below.

= _ _ . _ __

, RTD fD (13A, B) Pugose 6.7 Monitor temperature of verticalleg to establish boundary l

condition temperature, and provide a qualitative measure of i turbulent peuetration.

1, 2, 5 Monitor stratification magnitude, profile and frequency of cycling.

4 Monitor valve leakoff temperature (provide rcot causa information - packin:; leak) ,

3 Monitor bypass line temperature (provide root cause L.--

, information - bypass valve leakage) s In addition to the tetnporary sensors shown above, the followuig pirnt infmnu. tion was also reviewed. (Tois information was obtained from the plant cotcputer and operator i 1o 25-)

Hot leg temperature for Imps 1 and 4 RCS flowrate for J. coos 1 and 4 RCP Operation RIIR Gy. ration Safe:y injection Operation wpm 74al/075n10 3-1

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

a.

)

Dermal stratification was ub!ctved in both RHR lines ( onnecting to loops 1 & 4) during beamp and cooldown operations that involved lineup and operation of the RHR systems. The strati 6 cation events ' vere directly caused by operdog of the RHR isolation valves and relative:y low 6ow in the lines. This strati 6 cation has characterized by low delta Ts (less than 2(PF) and no signi6 cant cycling was observed.

During nonnal operations (reactor the mal power >95%) a mort. significant observation was rnade. Temperature measurements on the unisolable side of the loop 1 RHR isolation valve were hot, close to that of the loop 1 hot leg temperatures. Tnis result wmpares favoraoly to results frcm Gow model testing which suggest that turbulent penetration from primary loop flow thould penetrate approximately 22 pipe diameters.

The RrIR isoletion valve is approximately 14 pipa diameters from the loop pipe. Except i for cettain test condidons, RHR operations, and one reactor trip (in which all RCP's tripp:d) there were no unexpected thermal events in laop 1 RHR line. Ho vever, loop 4 RHR moniaring data displayed a significantly different response to normal operating conditions (reactor thermal power >95%). During nonnah operations temperature measurements on the umsolable side of the loop 4 VLHR isolation valve were cold, >

between 95 and 1207. It should be noted that during this time no sigui6 cant stratification was observed during power operations. Since the two RRR lines were for all practical intents identical in layout, fur:her investigation into the cause of the colo.

temperature readings during normal operatiott was merhed. It war eventually conchtded that there was an insufficien turbulent penetration eHect (heat transfer by a maas transport mechani:m) to tseatup all of the irreentory in thc umsolable section of s

loop 4 RHR. It is further postulated that the reduced turbulent penetration ef6.cr is the _

result of having two branch pipes in very close proximity to each other on the primary loop. In thir. case, the loop 4 RHR line (a 12 inch line) is 15 inches away from the pressurizer surge line connection (a 14 inch line). It is postulated that the two branch pipes in close proximity to each other result in reduced turbulent penetration energies available to either line. Therefore, the total mass exchange that occurs in the loop 4 RHR line is less than that of the loop 1- RHR tiae and hence the line cimls to ambient after some period of time, wPr<n48vo2199230 - 3-2 4

_ .. _ _ _ _ _ _ _ _ _ - . - - - - - 1

l l

Details of the observed stratification are shown below:

+ _

! RHR Extent Stratification Max No. of Probable i_ Line Stratified Duration AT Occurrences Cause i

Loop 4 Upstream of 3 days 391*F 6 Plant 8702B Cocidown/RCP l _

Operation >

Loop 1 Upstreain and c hours 17C*F 1(Loop 1) RHR Operation

&4 downstream of 1 (Loop 4) RHR Operation 8702A and 8702B Loop 1 Upstream of [ hours 112 F 1 (Leop 1) Reactor Trip,

&4 8702A and 8702B 1(Loop *) RHR Operation Loop 1 Upstream of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 90*F 2 Reactor Tr:p ,

8702A i Loop! Upsueam and 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> 94'F 1 Test Condit.on  !

downstream of j--- 8702A l It should be noted that the above transients from,the monitored data do not reflect any cause from valve leakage, ruher from plant operation. These transiena. have been conservatively includec in the fatigue and fatigue crack ;rowth analyses.

9 WPRT/4al/ tea 39a0 33

. . _ = - - - _ _ _ l

, . = . .

l L

1 i

a?: .XAt::=$On N ' PN

+' /-  % ,

'Doj LOCATE NEAR ~

EL30W WELO AT l ANY CCNVEN!!NT' 138-*

l CIRCUMFERENTIAL LCCATION N'

y l .y 12*-RC-1 007 150lR tu 3 LOCATE NE

9g4 ELECW .iE;.: -

' D* #

g ANY CCNHN 1V LOCATE Ai ANY CONVENIENT M *'

30, CIRCLMFE::i :

LOCAT!!N 1"* 1 7'

"" i C;RCU;4FERENTIAL

• QCATION CN

[EAKOFFLINE (M Nof GO' LOCATE NEAR Gunin 2 f. e t?02A VAmt wgLo LCAKOFF of volve tr A REF: SAP-RC I R$ 005 auttet) pp gap.RM 1 RB-001 TOP tu-t -

30* '

iM-S MTE: INSULATE LEAK 07F11NE FROM VALVE TO A MINIMLM OF 2 FEET 00UNSTADM OF IMTALLED RTD.

l l Figure 31. Unit 1, Loop 1 RHR Monitoring Locations l

l WPF074a1/12:291:10 31

HOTE: MSUI. ATE LEM 0FF LINE FR0r4 VALVE 70 A MIM[Mi.m 0F 2 FEET 00VMSTREM 0F INSTAtt.!O RTO.

s:

-/ g$

ara .:c u::=5 ra Nt 8:p:4 ,g9 L

- b. .S '

LOCATE NEAR #

EL30W 'aEL3 AT

'** , 12* AC 1 069 2501R 3

ANY CCNVENIENT 0;tCUMFERENTIAL

] CAT 10fl

/j LOCATE AT

/

ANY CONVENIENT CIRCUHFERENTIAL , 1 (ccg7g sgA; LOCAfl0N Oh j g(ggy wg;g ;,-

LE Ax0FF L1HE (witnin ; 't. ,,,

8 3 x.y Axy .:asy(s;t ,-

/' C;RCUMFEREN-'...

of valve

?

in 4 outt't) r- LOCATICM M g4 A

124 1 w

._/ '$g W LOCATE NEAR - 4 d

VALVE wELO T*.48 x' ,

MN 3 # 1,..,

c 1,

. /

y,, TOP gy V b E7028 REF: SAP.RC 1-RR 006

'*S 4,gy Rif: CAP. 5 \ R8 002 Figue 3 2. Unit ). Loop 4 RHR Mordtoring I ocatiora v/PHr74clu691do 35  !

l

- - - . . . _ .-m. _ _ _ _

o 4.0 TRANSIENT DEVF101YdSI 4.1 Cyner11 Discussig;taf Postulated Valve Leakage Transtents The NRC Bulletin 88-08 requires licensees to postulate that valve leakage may occur in the RHR Nlation valves. In general, the only leak scenanos that can be applied to the ,

RHR lines se out leakage (due to the pressure differences between the primary side and the downstream portion of the lines). The out leakage could either be through the leak off line of the isolation valve or to the downstream side. In either case, if a s

periodic (cyclic) leak occurred ir. the loop i RHR isolation vahe there would be no

xpact on the unisolable portion of the piping. This conclusion is true since the loop 1 unisolable portion of the RHR line is already hot due to the turbulent penetration. De introduction of primary :oolant into that region of the piping would not change the

\

thermal sta:e. in addition to postulated valve leakage, the possibility of stratiEcation in tne enholable piping was also addressed.

4.2 Developmcfdof the Onerational ReicitJ Stratifintion Transients 4.2.1 General St- 'incation was observed in bodt loop 1 and loop 4 RHR lines dering lineup and a operation of the RHR syttem. as discussed in Section 3.0. Stratifiention was not c

observed in loop 1 during operating modes 1 through 3.

During hot standby operations v.rati6 cation was observed in the loop 4 RHR line. The strati 5 cation that was o! served was the direct result of the existing condition (ccid water approximately 100*F in the unisolable side of the RHR line and hot water approximately 557 F in the loop) and RCP operations that resulted in loop 4 prunary coolant flow increasing to approximately 110% of normal flow. With the, loop 4 line cooled to t

ambient near the isolation valve during hot standby ope ations and the increase in primary loop Os. the already depleted turbulent penetration depth was increased. This

{ %7Rntal/02199M0 4-1 .

t

resulted in a higher mass transfer rate with the primary loop coolant that resulted in the introduction of hot water near the isolation valve. His cotidition stratified the horizontal section of the pipe for some period of time. The maximurn pipe delta remocrature observed curing these e"enu was 391T. The total number of these events observed during the monitoring period was 6, and all were associated with increased primary loop flow.

4.2.2 Behavior of Loop 4 The presence of cold water in the unisolable section of loop 4 RHR during normal operauons raises the question of what the interface is like between the primary coolant and the isolated MIR inventory. It should be noted that there is insufficient data at this time to conclusively support any wie hypothesis; however, there are at least two possible scenarios. One, the interface between the hot primary coolant and thr. cold wa:er is a gradual temperature gradient that is stable and non-cyclic and testricted to the serneal section of the Ic op 4 RHR line. Therefore, the only adverse loadig wud be those events already am- mted for by considering 240 cycles of increased primary loop flow in loop 4. The load condition for this scenario (vertical temperature gradient) is described in Table 4-1 as transicut number 1. A venical temperature gradient restricted to the vertical section vould resuit in an axial teroperature distribution that dropped off quickly from the primary loop temperature at the top of the vertical septient to ambient a temperature at the bottom.

The second scenario is sugges',ed from resiew of the Comanche Peak monitoring data.

From Figure 4-2 it can be seen tat the relauve circumferential locations of RDT 6 and RTD 7 are 180 degrees apart. He monitoring data from these locations suggest that a current exists 'n the vertical section of the pipe. Tais condition is highlyapeculative-since location 6 did display erratic datat at some times the readings were negative.

However, at other times th: readings were normal and within acceptable engineering

- ranges. Durmg periods of normal power operations (reactor thermal power > 95%) the S

temperatur: n.adings at loation 6 are colder than location 7. If the readings from WPFM4J/02f9R:10 - 4-2 l

.n s s ~

~

e location 6 are to be believed, this condition could only be explained by a verticai eurtent that is driven by the turbulent penetration. Turbulent penetration at the 45 degree bend would provide the pumping action by establishing an entrainment region that pulled cooler water from the' lower region of the pipe. Since the ov$rall turbulent penetration is low. tne rate of mass transfer of the RHR line inventory with the primary loop inventory is low. Hence, the introduction of heat through the raass transport mechanism is not sufficient to heat the entire line: this is consistent with the observed data at location 1. This condition is iilustrated in Figure 4 3. Transients that consider these s effects are described in Figure 4 2 as Case 1, Case 2, and Case 3.

i 4.2.3 Operational Transients that Apply to Both Loop 1 and Loop 4 Thermal stratification was observed in both RHR lines (connecting to loops 1 & 4) during heatup and cooldown operations that involved line up and operation of the RHR systems. The stratificatica events were directly caused by opening of the RHR isolation valves and relatively low flow in the iines. These events were characterized by delta Ts less than 200 deg F and no significant cycling.

Transients that envelope these observed conditions are listed in Table 4-1. Transients 2 througn 5 are applicable to both loop 1 and loop 4 RHR lines. Transient I applies to loop 4 only.

4.3 Dc_elcoment of the Po<tuinted Valve I.eaknee Transiema T

The transient at Genkai was due to intermittent valve leakage, which provided a path for bot water to be drawn into the RHR line from the main loop. In the horizontal piping downstream of the second elbow from the RCS connectic , a stratified flow was-established, with hot water filling the top of the pipe to a depth of 10 percent of the inner diameter.

WPM 41/030592:to 43

_ _ - _ _ - - - _ - _ _ . A

l l

To establish a postulated valve leakage transient for this scenario, stratification was assumed to exist in the horizontal piping upstream and downstream of the isolation valve (87028 on loop 4). The same portion of the pipe as at Genkai was assumed to be filled -

k with leakage Dow (i.e'.10 percent of the inner diarneter with a' leakage rate of 1.0 gpm). [A The bulk Guid was assumed stagnant, and therefore its temperature declines quickly with axia! distance, since heat transfer is primarily conduction. The leakage at the top of the 2 pipe does not cool as quickly, due to its Sow as shown in Figure 4-1. This creates a rather large temperature differential between the top and bottom of the pipe, which maimizes at an axial distance of approxitnately 4 feet from the second elbow, and

< dimmishes at an axial distance of approximately 20 feet from the second elbow. At the ~

first isolation valve inlet weld, Figure 4-1 provides 565'F at top of the pipe and 360'F at the bottom of the pipe with .iT = 205'F.

As mentioned in Section 4.2, the actual monitoring data from the unisolable section of loop 4 reveried an alternate interpretation of the axial temperature distributica that could be present du ing valve leakage conditions. In order to account for the alternate interpretation, several additional independent load cases were postulated for vr.lve leakage. These cases assumed that the vertical current existed (Scenario 2 in Sec; ion 4.2.2). These load cases are considered u. alternate states that are independent of and replace ioading conditions during hot standby and normal operating conditions that did not assume a vertical current (Scenario 1 in Section 4.2.2). They were analyzed for their impact on fatigue life and fatigue crack growth. The postulated valve leak transients for these cases are showa in Figure 4-4.

WPFU748J/021992:10 4-4

O TiBLE 4 4 LOAD CONDITION FOR THE VERTICAL TEMPERAT1lRE GRADIENT .

c g l Max Top.to Bottom Temperature Total Number of u

g Tranrad Pipe g Prof 4e 4i M u H _ Cscles Extruel;ued I

1 391*F Step Change at 3.3" from top I.D. 240 2 170*F Step Change at 1.6' from top 1.D. 200 3 112*F Linear d6 4 90*F ,, Linear 400 y 5 94*F Step Change at 7.3' f*om top 1.D. N/A' i

• Test condition Note: Tranient numbers 2 through 5 represent a3 other chsened cems with cycles curapolated for the life g of the plw. Transients 2 through 5 are appli:able to both loop 1 and loop 4 lines. The observed outer wall temperatue di'inbutions from the monitoring data wee utilind to develop these transients, a

WPFU748J/030302:10 45

,f~, , . , . _ _

!11il e

i n

- L m - ' R o j

- 0 H t 2 R t

p o h o B c T n e

p i

e p i Z i P I P t 2 t a

- a d d ' n w i 5 a.

o u - "

1 1

l '

l F f s w 7

e g

k l

t i

l a o n i k B U

- a f

- /l e k L o a e e

. ! f r P

?. o e t u 1 -

y ) e a l T t.

r r c t

u e (F n a

p a t W m r s 0. o e e p

m e

T

% M 8

E t

C w

\,A o

a Tl -

p e

v R.

F E

l f

d A

n e

l a E i

\ A T

f i

VN S

I D

t a

r h -

. t S

\ f o

11 ,

e r

'u t

n e

p m

e T

1 4

e

~

~

r

1. ~

u g

0 0 0 0 i 0 0 0 0 f 6 $ 4 F 3 2 T

y .

sdr

\l l!t! l' i

. - . . ., , ,m -

l POSTULATED LOAD CASES C4SE1 ancAn rwarxwarstenaus; ll L

• * *  ; cross sECDON 1 Cross SECDON 2 CaCis sECDON2 LOO 9 $3DE ' ThR sloe LOOP SIDE FAR SIDE 109 BOTTO$t p_ _ . . - .

-\ __

.\ N Caou no sao Srs der 4e:

LOCA g - \ $dC? SON 1 l {soo (1 MID) \Q Q ., !.i J- C4SE2 LOCA110N7 -.

cross cross Cnoss.- - - - --- 1 SEC11CN 1 SECDON 2 sECitON 3 l

[/

,s (s nio) 1oOes On rAn soDE LaCr ssor }rAn SIDE _

TOP BOTTOM y Cross y sua sus sus 420 soo L-100 -

M sEc110k 2 O CASE 3 proutmese inausmouavstenow>>

CRO$s CROSS Cross Cross SEC11GN 3 SEC110Ns SECDON2 SEC1sCN 3 l

g m nogg _

j (5 RfD1) LOOPsIDE 7AR stDE LOOP stDE FAR *snDE 109 BOTTCht 300 aso 1:s j 488 491 j l _100

[ A->re ses OrmmaanL7zaivor.wruacciar4e 7Emeris4F13fDEsREIUT30W040 FIGURE +2

_ ;n -

U O . . . . . . . . . . . . , . . . .. - ,.

e LOOP 4 N'

N e \x' HOT WATER W ,/

ENTRAlNMENT DUE TO s.

run.uuurPaNemArlON PROVIDES PUMPING ACTION Nw

• p hp /

..y 8 3 I ' '

COLD WATER > 8 0 x w 3

~

> l si s RHR ISOLATION VALVE FIGURE 4-3

POSTULATED VALVE LEAK CASE

• • Y Cross Cross cnoss 1 SECDON t (MAK) SECDON 2 iMAX) . St C TION 2 f M.tv)

LOUP $iDE FAR SIDE lOOPSIDE {FARSIDE 10P BUTTOM CROSS 225 ets sto ats era sto l l '

LOCAfDON 3ECDONI isMTD)

A ~

LOCADON 7 ,-

Cross Cnoss CuOss

\

.IsRID)  ;

SEC slON 1 (MIN) SECTION 2 (M6N) ~ ~. SEC DON 2 iMIN)._.__n-m y LOOP SIDE FAR SIDE LOOP SIDE FAR SiOE lOP BUTTOM Cnoss' '

$ im - - - - - -

225 sta sto 4st sto 1

sto

{' SECDON 5 NOTE: THE MAXIMUM CASE IS ASSUMED TO RANGE MTH THE MINIMUM CASE c:?os2 SEC M 2 gggypgg, PEP:OO FOR EACH CYCLE e 83 blN l p RTD'S)

I i

i flGURE 4-4 i

a: _-

e

f.0

SUMMARY

ANILCOFCI.UsiOF.S s A detailed evaluation of the residual heat removal suction lings for Comende Peak Unit I has been enmpleteci in response to concerns raised by a pipe cmek incident which occurred at Genkai Unit 1 in Japan and subsequent NRC Bulletin 88-0E The monitoted data from the Unit 1 RHR suction lines have been reviewed and evaluated. No NRC Bulle. tin 88 08 type of valve leakage was observed in the da'.a. <

liowever, conservative assumptions were made to pavulate a Genk:u type of leakage :n

\

the Loop 4 RHR line, Based on such assumptions, the resulting stratification loacing and associateJ stresses were calculated. Using these calculated stresses and postulated ,

high number of stress cycles, conservative fatigue usage and fatigue crack growth calculations were then performed for 12)op 4 RHR line. I. cop 1 RHR suction line leakage has no impaci on the unisohble portion of the piping as oteviously discussed in Section 4.1.

For loop 4 RHR line, fatigue usage calculation provides an indication of the probability \

of eracking and of the time required to initiate. F.atigue, c-ack growth analysis was performed to deterrnine de time required for a 60 percent through wall crack to occur 3 based un the postulated transient stratification loading, as shown in Figure 4-4. The critical locations are the inlet weld of va.lve 8702 and the weld at the end of the 90' elbow. Due to the extremely conservative assumption in the fatigue usage calculation, a fatigue u age fac.or of less than 1 cound not be obtained within the plant design life at the governing location. Furthermore, results of this analysis indicate that a minimum of 1.5 years of leakage is required for an initial flaw of 10 percent wdl thickness to p opagate to 60 percent wall thickness. Augmented imervice inspection intervals should

- be developed based on this result of 1.5 years for ooth locaticos of loop 4 RHR line. A21 other welds in the loop 4 RHR line should be inspected in accordance with standard ASME Section XI criteria. The first augment inserv!ce inspection should occur one and one half effective full power years (EFPY) after August 31.1991.

%N#4J/021w2:10 51  :

1

_ _ _ . _ _ _ _ J

For locp 1 RHR line, since the postulated leakage has no impact on the unisolable portion of the piping, the fatigue usage calculated, based on only the operational transients described in Section 4.0, is 0.9 for 40 years of design life. Therefore, all welds in the loop 1 RHR lihe should be inspected in accordance with standard ASME Section a XI criteria.

It is thus concluded that the requirements of NRC Bulletin 88 08, Supplement 3, are satisfied based on the following:

Conservative technical evaluation provided in this report, Augmented inservice inspection intervals, and o

Implementation of the CPSES Unit I long term transient and fatigue cycle monitoring prograin

\

t WPRM4&f/021992:10 52 s

(  % .m _ _ . _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _

i l .

. .o A

6.0 REFERENCES

1. NRC Bulletin 88 08, " Thermal Stresses in Piping Connected to Reactor Coolant Systems!' 6/22/88; Supplement 1,6/24/88; Supplement 2,8/4/88; and Supplement 3,4/11/89.

2. WCAP 12258, " Evaluation of Thermal Stratification for the Comanche Peak Unit 1 Residual Heat Removal Lines," W. H. Bamford, April 189; 9 Supplement 1. June 1989 and Supplement 2, August 1989, Westinghouse Proprietary,

3. Texas Utilities letter CPSE3 9120542, b/14/91,'Cotnanene Peak Steam Electric Station Thermal Monitoring Data Reduction," J. W. Muffet to J. L Vota.

t er e

'W WTT074&l/02F19110 6-1 i

. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ 1