ML19325E547

Responds to NRC Bulletin 88-008,Suppl 3 Re Thermal Stresses in Piping Connected to Rcs.Due to Generic Valve Design, Piping Geometry,Penetration leak-rate Data & Current Insp Programs,No Addl Action Required

ML19325E547
Person / Time
Site:	FitzPatrick
Issue date:	11/01/1989
From:	Brons J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
IEB-88-008, IEB-88-8, JPN-89-071, JPN-89-71, NUDOCS 8911070375
Download: ML19325E547 (7)

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1, November 1,1989 j J JPN 89471 l

,- U. S. Nuclear Regulatory Commission j f Mall Station P1 137 i Washington, D. C. 20555  !

ATTN: DocumentControf Desk l i

Subject:

James A. FitzPatrick Nuclear Power Plant  !

Docket No. 50 333 l Response to NRC Sulletin 88-08, Supplement 3,  :

Thermal Streseos in Piping Connected to Reactor Coolant Systems  !

Referonces: 1. NRC Bulletin 88-06, dated June 22,1088, regarding thermal l stresses in p! ping connec'ed to reactor coolant $ystems. (

2. HRC Bulletin 88 06, Supplement 1, dated June 24,1930 ,

concerning the same subject.

3. NRC Bulletin 88 08. Supplement 2, dated August 4s 1962, l r

concerning the same subject.

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4. NYPA letter, J. C. Brons to NRC, dated October 21,1968,  !

providir,g response to NRC Bulletin 88 08. (JPN-88-055).

5. NRC Bulletin 88 08, Supplement 3, dated April 11,1989, conceming the same subject as Bulletin 88-08. I

Dear Sir:

j NRC Bulletin 88-08 (Reference 1) described an event at a nuclear power plant involving thermal fatigue cracking of unisolable piping connected to the Reactor Coolant System (RCS). The bulletin requested licensees to determine whether unisolable sections of piping j connected to the RCS could be subjected to stresses from temperature stratification or temperature oscillations. Reference 2 provided preliminary information about a similar

• event at a foreign plant and emphasized the need for sufficient examinations. Reference 3 emphasized the need for enhanced ultrasonic testing and for experienced examination personnel to detect cracks in stainless steel piping. In Reference 4, the Authority described the review performed at the RtzPatrick plant. This review concluded that the condition described in Bulletin 88-08 is highly unlikely to occur in the FitzPatrick plant. .

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. o Reference 5 provided intomistion about a similar event at another foreign reactor and emphasized the need for sufficient review and the importance of taking action where neoessary, Sinoe the overd described in Referonoe 5 differs somewhat from the events described previously, the Authority has readdressed the systems affected. The review is discussed in detail in Attachment 1. The Authority has concluded that there is negligible risk of pipe failure which would lead to an unisolable leak of primary oootant, and that no additionsJ action is required.

Should you or your staff have any questions regarding this matter, please contact Ms. Sofia M. Toth of my staff.

Very truly rs, John C.'Brons - N Executive Vice President Nuclear Generation Attrchment W STATE OF NEW YORK COUNTY Or %ESTCHE31ER Subscrib9d gnr1 swom to foro me g trAs /d day of , c44n 1089.

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475 Allendale Road King of Prussia, PA 19406 Office of the Resident inspector U. S. Nuclear Regulatory Commission P.O. Box 136 l

l Lycoming, NY 13093 Mr. David LaBarge Project Directorate I 1 Division of Reactor Projects.1-ll U. S. Nuclear Regulatory Commission Mall Stop 14 82 Washington, D.C. 20555

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i ATTACHMENT i

[ TO L JPN 89-071 f,.  !

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Response to NRC Bulletin 88 08  !

Supplement 3 Thermal Stresses in i Piping Connected to Re_setor Coolant Syo_t_etps r

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l i NEW YORK POWER AUTHORITY ,

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50 333 I

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ATTACHMENT I TO JPN-89-071 Response to NRC Bulletin 8408 Supplement 3,  !

Therme! Stressee in Piping Connected to l l

Rescior Coolant Byetems i I

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1. INTRODUCTION h t

NRC Bu!!stin 88-08, Supplement 3 notified licensees of cor.cerns regarding systems  !

, connected to the Reactor Coolant System (RCS). Specifically, these concerns involved  !

l thermal fatigue cracking caused by stresses resulting from tempstature stratification or l temperature oscillations. Such cracking may occur when fluid from a hot system, such as i feedwater (FW) or recirculation, cools in a stagnant leg. This subcooled fluid unseats the i normally shut isolation valve causing leakage to occur. Hot makeup fluid to this valve I causes the valve disk to expand and subsequently stops the leak. This temperature i oscil!ation may lead to fatigue-induced cracking. The NRC requestod licenscos to identify  !

affected piplog a"d to take action if noosssary.  ;

g II. plSCUBSION, j Du:ing normai cf erstion, thr. recirculativi n #d FW cystems conta.n hot prasaurized toactor Vater. It.ola'od from these cystems are various stendby systr.ms which have utsgnant legs i slmlist tr,.te fallad pips configuration citec in Supplemont 3. Syt. tams that interlace VAth o' the hot pteaturtrod systems are High Pressure Cora injection (HPC!), R9 actor Core e isolation Coollog (RCIC), Res!dval Heat Ptymoval (PHR) to the Low Prsssura Coolant i PQection (UCI) mWe, Re'neior Watu Cleancp (RWCU), and Core Sproy (CS). The Acthe/ty evutuated each of twco s;'sterr.s to determine their susceptlWity to thermal

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stratification fatigueriduced etacking as described in Supp!ement 3. The following summarizes the results of this evaluation.  !

l lil. CS/HPCl/RCIC/RHR (L.PCI)  !

During normal plant operation, these four systems are in the standby mode. They each have -

a generic valve arrangement and a unique piping configuration. The generic valve l configuration is shown schematically below (Figure 1). (The second isolation valve shown in the figure is not considered in the scope of this evaluation.)

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i Pressure Stand-by <3 b<3 j .

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Mode MOV (2) MOV (1) AOV '

FIGURE 1

- i o Attachment I i JPN 89-071 i Page 2 of 4  !

i in all ekses the motor operated valvet (MOVs) are solid disk wedge type gate valves, it is l this type of valve that caused the thermal stratification fatigue induoed cracking described in +

SupplemeM 3. The air operated check valve upstream of the MOV offers the piping  !

between these valves some protection from pipe fatigue falture. Additionally, the RHR I (LPCI), HPCI, and RCIC systems have long pipe runs (>25 feet) between the flow of hot  :

fluid and the first motor operated valve. This eliminates the effects of thermal stratification, l thus reducing the probability of pipe fatigue eracking. l The piping susoeptible to failure is between the air operated valve and the MOVs. A f allute  !

in this pipe run, however, would not cause an unisolable leak from the high pressure  ;

system.  !

in order ior thermal cycling to occur as presented in Supplement 3, the following MOVs I would have to :sak. (These valves are designated MOV(1) in Figure 1), f i

14AOV 12A & 128 (Core Spray) f

2?MOV.19 (HPCI) i 13MOV 21 (CICIC) 10MOV.25A & P.S (RHR/LPCI) i An ana>ysis of the spacT,o valves fuliows t'clow

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A CORESPRAY ,

I o Cuc spray Inboa'd lNoction voves 14MOV 12A and 14MOV 12B have bean loNi tert  :

rate iertad dunny each of the teven refvoling outages since in!t;&l p'. ant operat;on.  !

During the tests gST 39B), these valves are tested in combination with penetrations X 16A i and X 16B. These penetrt.tions have never exceeded allowable leakage, and repairs to these valves have never been required as a result of excessive leakage. The piping i welds between these valves and the reactor vessel nozzle are inspected for flaws in j accordance with ASME Section XI.  !

During the 1981 refueling outage, the Authority replaced the *A* loop piping between the  !

reactor vessel nozzle safe end and 14AOV 13A.

• B. HPCI f The HPCI inboard injection valve,23MOV 19, has also been local leak rate tested during  :

each of seven refueling outages since initial plant operation, During the test (ST 399), ,

this valve is tested in combination with other containment isolation valves of penetration l X 9B. Although the total leakage from this penetration has exceeded allowable leakago l- on three occasionc, the leakage has never been attributed to 23MOV 19, nor has this valve ever required repair due to excessive leakage. In addition the piping welds and heat affected zone downstream from 23MOV 19 are inspected for flaw indications in l

accordance with ASME,Section XI.

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The Authority has local leak rate tested RCIC inboard irQection valve,13MOV 21, during  ;

each refueling outage. During the test (ST 398), this valve is tested in combinatit.n with i

other containment isolation valves of penetration X 9A. Although the total leakage itom  !

this penetration has exceeded allowable leakage on three occasions, the leak has never  !

been attributed to 13MOV 21, nor has this valve ever required repair due to excessive  ;

l leakage. Also, the pipe welds and the weld heat affected zones are inspected for flaws in  !

accordance with ASME,Section XI,  !

D. nHR/LPCI f Ti o RN A/l. pol inbobrd liect;t t, veve't,10MOV 25A and 10MOV.250, i,avn bun tes.k- l rate InstW durig exh ref'xtb2 ot lE3t Dwing ilw tos (ST4)D), ihese vaWes we j te.sted in combination with per.ottstions 13A and 130,8 %pectively, Although she utal:

y leakege imm th4t penetrGtion has exceeded allowable icakege on thros cecasions tha  !

ler/< has never been attrit:uted to 10MOV 25A of 10MOV.25B. Valv610MOV 2bA was e' repacked in 1905, but th!5 was not due to excenive leakage. Valve 10MOV 25B had a live tca.1 thm packing 't.st21k'd dutira tt c, %8 retuohng outupe which is designed to p:evant stem leakage disir,0 operation. Auct.ior ally, the pte wolds and heat atfeeted zones are inspected for flaw Ir dcations in accordance with ASME, Sect'on XI.

E.RWCU l i

The RWCU system currently has the B pump in operation and the A pump isolated. The [

pipe from the header tee in the cleanup pump room to the manually operated isolation  ;

valves was evaluated for fatigue cracking. Leakage through the A pump isolation valves, i 12RWC 19 and 12RWC-73 stem packing would be detected readily on the floor. Small leakage through two manually operated gate valves, caused by heat up and expansion, is unlikely. Manually operated gate valves do not have the problem of preset torque i valves or preset stem engagement which m6y cause a valve disk to expand and lift, l causing leakage, when heated up as discussed in Supplement 3. Additionally, the 4*

pipe line is less susceptible to fluid stratification due to the small pipe diameter and the vertical orientation of the valves. ,

No leak reto data is available on these valves, however, live load stem packing has boon installed on 12RWC 73 to prevent valve stem leakage. i l

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Attachment l I JPN 89-071 l Page 4 of 4 )

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F. RWR/ SHUTDOWN C00UNG t

itse shutdown cooling mode, including the reactor vessel head spray,is a function of the l RHR system and ope.'ates during normal cooldown and shutdown.

During reactor cooldown, coolant is pumped from the B recirculation loop, cooled in the RRR heat excungers, and pumped back into the recirculation loops. Valve 10MOV 18 on the suction line is a wedge type gate valve located close to the recirculation line 02- .

WH GE 1 A. Because of poor valve orientation (stem pointing down),10MOV 18 )

experiencNf nwny maintenance problems before 1985. During the 1985 refuellrg  !

optage, tnis va've was replaced along with the pip ng between the drywell penettelica  ;

and 10RHR 66. The ba'ance of pipt betw6en the highanstgy recirculation pipe and i valve 10RHR Ni was tubt.oouently replaceo during the 1987 refueling outage.  !

Uve load stern packing was installed on 10MOV 18 dering the 120 rofuel:ng outago, i which is designe,4 to pres ent s%m leakcge. The inboard suction salve 10MOV 18 was  ;

Is@. raw tested during the 198G refuel!ng outage. At that time, this valve was tested in  !

combinat!on with p6netration X 17:. The penetration did not exceed allowables, and no j repairs weta required due to ucessive leakage. ,

For the RHR head sprty sut> system, the Authodty evaluated valves 10MOV 32 ar.c 10RHR 23. The valve dMign configuration is similar to that shown in Figure 1, but without an t.lr oper6 tor on the firt.t isolation check valve. The isolation check valve protects against steam coming from the reactor head. This prevents fluid stratification.

Leakage past the first isolation check valve, would initially flash and then condense through the pipe run to the downstream MOV. The 4 inch pipe size and extended run of pipe will eliminate the possibility of stratification-induced thermal fatigue cracking.

IV. CONCLUSION The Authority has evaluated FitzPatrick plant piping systems with the potential for thennal stratification fatigue induced cracking as described in Supplement 3. It concluded that due to generic valve design, piping geometry, penetration leak rate data, and current inspection programs, no additional action is required. There is little risk of pipe failure which would lead to an unisolable lead of primary coolant. The most susceptible location was identified as the piping upstream of 10MOV 18. However, the Authority replaced this piping in its entirety in 1985 and 1987.

As a conservative maintenance effort, the Authority is considering installing live load stem packing on valves to preclude the already low probability of cracking in piping upstream of the valves, if oscillating leakage were to occur. Those valves are: 14MOV 12A,13MOV 21, 10MOV 25A,12RWC 19,10MOV 32,10RHR 29.

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