Information Notice 1995-26, Defect in Safety-Related Pump Parts Due to Inadequate Treatment

Defect in Safety-Related Pump Parts Due to Inadequate Treatment
	ML031060167
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	05/31/1995
From:	Grimes B; Office of Nuclear Reactor Regulation
To:	;
References
	IN-95-026, NUDOCS 9505240309
	Download: ML031060167 (19)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.-C, -20555-0001--- -

May 31, 1995 NRC INFORMATION NOTICE 95-26: DEFECT IN SAFETY-RELATED PUMP PARTS DUE TO

INADEQUATE HEAT TREATMENT

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice to alert addressees to a potential for pump failure as a result of

inadequate heat treatment of pump parts. It is expected that recipients will

review the information for applicability to their facilities and consider

actions, as appropriates to avoid similar problems. However, suggestions

contained-in this information notice are not NRC requirements; therefore, no

specific action or written response is required.

Description of Circumstances

On October 19, 1994, Westinghouse issued a written report pursuant to Part 21 of Title 10 of the Code of Federal Regulations (Part 21) (Accession No.

9411010251), regarding a defect found in the "JHF" model safety injection

pumps that were manufactured by Ingersoll-Dresser Pump (IDP) Company. The

defect comprises several axial cracks in the pressure-reducing sleeve locknut, which is made of Type 416 stainless steel. The failure mechanism was

attributed to stress corrosion cracking which primarily was caused by a

martensite phase hardness of 47 Rc (Rockwell Scale C). It is known that 400

series stainless steel with a hardness in excess of 40 Rc is highly

susceptible to intergranular stress corrosion cracking in aqueous

environments. The heat treatment for this pump part was specified to be

27-32 Rc.

Although the problem was at first thought to be limited to the locknut on the

pressure-reducing sleeve within the IDP pump, Westinghouse and IDP have

determined that other pump parts may be affected by the same problem. On

February 20, 1995, Westinghouse issued a final Part 21 report which indicated

that (1) in addition to the subject locknut, other pump parts could be

affected by the same problem and (2)the problem is limited to IDP pump parts

that were made of Type 416 stainless steel, processed under heat treatment

process "HT 21," and that were taken from heat numbers 15899 and 28144.

The detailed Westinghouse final Part 21 report (Accession No. 9503020053) on

this issue is attached (Attachment 1).

PDR Tsk f5'd joI4e ?s-O26 An

9505240309 I'\

)

<_IN * 95-26 May 31, 1995 Discussion

As noted in the attached Westinghouse Part 21 report, the suspect parts are

used in intermediate-head safety injection pumps, auxiliary feedwater pumps, and charging/high-head safety injection pumps in various plants. The defect

in these pump parts, if not corrected, could result in pump failure. The loss

of these pumps during a design-basis accident could affect accident

mitigation.

Some of these affected pumps have been in operation for more than 10 years.

Because of the importance of these pumps to plant safety, Westinghouse has

recommended that the affected pump parts be replaced with the parts currently

recommended by IDP.

An IDP safety injection pump (model "JTCH") failed during a post-maintenance

test at Indian Point Unit 3 on February 19, 1995. Inspection of the internal

components of the pump revealed that the locknut on the outboard shaft had

backed off about 6.35 mm [0.25 inch]. The loosened locknut allowed the pump

impellers to move axially, and allowed them to rub against the stationary

diffusers and the casing, and ultimately resulted in pump failure. Although

the root cause of this failure is unrelated to the problem reported by

Westinghouse, a similar pump failure could occur if the locknut on the

pressure-reducing sleeve, mentioned in the Westinghouse Part 21 report, failed.

This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Project Support

Office of Nuclear Reactor Regulation

Technical contacts: James A. Davis, NRR

(301) 415-2713 Peter C. Wen, NRR

(301) 415-2832 Attachments:

1. Westinghouse Part 21 Report, February 20, 1995

2. List of Recently Issued NRC Information Notices

FEB 20 '95 11:Z? FE ~~51127 FRMI OPL LICENSING TO enzime1ssisi. PAGE.002/012

021 R Attachment 1 IN 95-26 FAXED 2/20/95 301/816-5151 May 31, 1995 NTD-NRC-95403 WevtlaOuse EMr Sms .0t

c, wihf

Bectric Cpoaton lo gn5 P:--JwfP PtmaWlli 15230-M5 Fabnrzy 20, 1995 U.S. Nuclear Regulatory Commission

Atm: Documeat Corol Desk

Washington, DC 20555 Subject: Update to IOCFR21 Report Coutained an WeStitghouse Lae NTD-NRC-94-4320, dazed

10/19194 and Stats Report Contai in Westinghouso Letter NTD.NRC-94-4361. dated

12/21/94 Reftence: 1. Loner NTD-NRC-944320. N. I. Upanulo to Document control Dsk,

10/19194

2. Ier ND-NRC-94-4361, N. J. Liparo to Document Control Desfk,

12/21194

3. Westinghouse Nuclear Safety Advisory Ler, NSAL094-23. lO26)94 The following infmoo is provided as an update to the 10 CFR 21 report that was previously

identified to You in References I and 2.

Reference I identified a defect, as defined under 1OCFR21, regarding the pressure reducing sleeve

locknut of the JHF Model Safety Lnjection Pump that was manfactured by Wersoll Dresser Pump

(DP) Company and supplied to seera nudear power plants by Westinghouse and tlP.

Westinghouse also notified the affeced licensees about the defe via Reference 3.

Reference I Indicated that MIP would perform a review of dte applicable 400 series stainless steel

parts applicable heat amets on other safety related pumps it Supplied to the nuclear power

industry to determine whether ths sisation could apply to other parts on other safety relazed pums.

Tbis review was completed on December IS, 1994.

Westinghouse dffed the NRC about the results of this review via Reference 2. The results of

review indicated tda the failure mechanism appers to be limited to I)p pump par ta consist the

of

416 SS, processe uder IIP heat teatent proces 'HT 21 and taken from [DP heat numbers

115899 and 28144. Aow. the review Indicated that additional pump parts may be susceptible to the

same ilure mechanism. Referece 2 indicted dt IDP would determine whether the failure of the

additional pump parts would prevent the applicable pump from performing its intended safey

funiction.

Mi review has been completed and ndictes tha some of the additional pump paru may also

constitute a defect, as defined in 10 CPR 21, whih could crae a Substantial safetr azard. e

following repo provides more Inormation about thee additionat pump parts.

15rt.Jl-nm INrf.

FEB 20 '95 11:27? LR tPL LICENSIS T I'3018165151 P.9034O12

' ' K~' 'K>

EYALUA

IDP's evaluation results Indicate ho all 416 SS pars processed under EDP beat numbers OI899 and

n8144 are susceptible to the sae failure mechanism as the pressu reducing sleeve locimuts.

These pas hve been dvided ko three categories.

The firt categoy Includes pat which were supplied with the origil pump assembly and whose

fuilure could prevent the pump from peformhg Its ;neded safety fnction. These paut and the

applicable plants are Identified to Table 3. Also, now that dies. parts Include the pressure reducing

sleeve locknuts thar were the parts originly Identified In Referenice ,

Tbe second caegory includes pars which were ulied with the original pump and whose fillure

would n prevem the pump from performing its intended safety union. These pam are dif d

in Table 4. The pans should be replaced as a pruiden manean activity. [t should be exmphasized

dtt the parts do not constitute defects pursuant to 10 CFR 21 since the faiure of the parts should

cot prevent the pump from performnig its Intended safety function. Howev, they are Included In

this letter to identify the results of the evaluations that are mentioned in References I and 2.

The third category Includes the parts which were supplied as replacemet pump pars and whose

failure could prevent the pump from performing its intended safety funcion. Most of the replacement

parts were supplied directly to the utilities by IDP. These parts and the applicable plants are

Identified in Table S. Please note da Table S has been divided into two lists. One list (Parts from

Identified Material') includes the replacement parts that are known to have been taken from beat

IS899 and 02814. The other list (-Paxts from Unidentifled Material-) includes those -wus which

could have been manufaaured from either heam 1IS89 or 121844, based on the time of tanufacture, original maial specification and part size. However, there is no documentation to spifically

identify the bear number from which the pan was taken. Therefore, It was assumed that the parts

were taken from heat #15899 and 21844.

Tbe identification of the parts in Tables 3. 4 and S Is based on several reported part failures, a failure

analysis of failed parts and engineering Judgetnent. Duke Power Company reported two separate

failures (crackig of the pressure reducing sleeve locknuts on JHF model safety Injection pumps.

Both locknuts consisted of 416 SS and were processed under heat 11S899. Duke Power performed a

failure analysi on each locnt. The failure analysis indicated that the allure mechanism was stress

corrosion cracking. The analysis also indicated that each lout's susceptibility to stress corrosion

cracking was increased by a relatively high martenske phase hards which allegedly resulted from

a Insufficient tempering operation. A comparison of the Duke Power Failure analysis report and the

IDP beat trearmea specification an beat material certification Is provided In Table 1.

Table 1: Prauur Reducn Sleree Locknut - Comparison of Dukc Power Compaay Faiue

A&LysIs to MP Set Tretment SpodricatIon and Material C cation

Hardness Mc)

First Locknu Second Locknut

Falure Analysis - Bulk PhMse 35 32-33 Failure Analysis- Microhardness 4549 42-44 Heat Tremen 'H721 Specification 27-32 Heau #15899 MaerIl Certification Bulk Phase Hardness 27

.. £i _u -. - .

FEB 20 '95 11-28 FROM OPL LCENSING T0 813016165151 PMGE.0044012 The existne of two fures and the data InTable I provide a basis to condude that the locknuts are

defects as defined in 10 CFR 21. First, it should be nated diat 400 series SS with a hardness of

grea than 40 Rc Is very susccpdble to Wtergar sares corrosion cracig In aqueous

environment. Table I indicates to for both lockuts die microbardness was higher than 40 Rc.

Also. Table I Indicates th for both lockmuts. the bulk phase hardness measured inthe fallure

analysis was s*ifficandy hgher than tie bulk phase hardness provided on the material certification

sheer. It could n be determned why tese bulk phase hardness values are differect. Since, the

micbardoesisw s aer than 40 Re, the bulk phase hardness diffae ce could not be acribuced to

any specific reason. However, since the cracking was observed on each locknut. It was determlned

thar a1 parts from beat M15899 could be susceptible to the same failure.

In addition to the above, Duke Power Company recatdy discovered cracking on a spacer sleeve in

anther IHF model safety Injection pump. The sc sleeve consisted of 416 SS and was processed

under MDP bea treatment specfficaion HT 21". However, the spacer sleeve was taken from a

differe beat. which was beat 128144. Duke Power performed a failure analysis on the spacer sleeve

and determined that the failure mechanism was also sress corrosion cracking. Also, the failure

anuJysis indicated that the spacer sleeve's hardness made the sleeve marginally acceptable for service

In aqueous environments. A comparison of the Duke Power Failur analysis report and the IDP heat

treatm specification and beat material cerificadion for the spacer sleeve Is provided in Table 2.

Table 2: Inpeller Spac Sleve - Comparison of Duke Pawer Company Failure Analysis

to IDP ealt Treatment Specification and Mterial Certification

Hardness (Rc)

Failure Analysis - Bulk Phase 26-30m

Failure Analysis - Microhardoess 33-39 Heat Treatment 'H12 I Specification 27-32 Heat 2n I" Material Cerification- Bulk Phase Hard 27

'he existence of the caced spacer sleeve ad the data in Table 2 provide a basis to condude dta the

space sleeve is a defect as defined In 10 CFR 21. Fist, the measured microhardness of 33.39 Rc is

higher than the material certification value of 27 Rc. It Is less than 40 Rc. but oonetheless Is

marginally sugtible to strs corrosion cacking In aqueous enviromens. The bulk phase

hardness tinge Is somewhat higher than the material ceruficadon value, but it Is still widhin the heat

treatment specification tange of 27-32 Rc. Based on this information, It was concluded that all parts

used from heat 028144 could be susceptible to the same failure mechanism.

Rially, and as mentioned. Tabes I and 2 IndIca that there Is some difference between the bulk

phase hardness value and the material certification values. There are no apparent reasons for tise

differences. Furthermore, Tables I and 2 indicate that tee are sifican differences between the

bulk hardness and the m icardoes values. tere are no appaeu reasons for these differences;

however, It may be postulated that the differences ar attributable to Insufficient tempering. IDP has

lu This value was determined from the uncracded spacer sleva on the same pump. Both the

ucracked and cacked sleaves were taken from Het M814.

LS47CTWF-.3M0W

FEB 20 '95 11:28 FROM tPL LICENSIN TO 13s611sis1 PFOE.005l102 not received say additional repoq s of pat fillures iavolving 416 SS uader het treatment specification

iHi l'. Therefore. It was concluded that the filures should be limited to only those pars hi were

taken from beas #S899 and 121844.

Table 3 identfie Al pump pars that were orig lly supplied with a pump. aken from heat #IS899 and 28144 ad wbose faillre could prevent the applicable pump from petforming us Ided safety

function. Table 4 Ideifies the pan that were orig ly supplied with a pump, taken from beah

IS899 and 128144 and whose failure , preyeat the pump from performing its Intended

safety function. However, It Is ro ended that the pas in Table 4 be replaced as a prudent

maintace practe. Finlly. Table S identifies the replacement pump parts that were either supplied

or believed to have ben suplied from heat 15899 and 028144 and whose failure could prevent the

applicable pump from pedoazfoig Is inded safety fucion.

The sfety sgniflcance for the failure of each par identified In Tables 3, 4 and 5 Is provided as

follows. The failure of the pats in Tables 3 and S could prevent the pump from operating. For

Table 4. the part failure should ot prevent the pump from operating. More deiled Information for

the pans identified in Tables 3, 4 and S will be provided directly to each utility.

The pumps identified in Table 3 are all JHIF model safety injection pumps. These pumps are used

the intermediate head safety ijection system for the applicable plants. The loss of these pumps in

during the shor torm mitigation period of a loss of coolant accident (LCA) would impair the plant's

ability to mitigate the consequences of the LOCA. The loss of the pump (or pumps) would reduce

the overall flow to the core, which could create a condition that Is a substanzial safety hazard.

The pumps Identified in Table S include the Iltermediate head safety injection, auxiliary feedwater, and charging/safety iWection pumps. The Intermediate head pumps are discussed above. 'he

auxiliary feedwater pumps are used to provide feedwa;er to the steam generazors during certain

accident conditions. The loss of these pumps during a feedwater line break accident would impair the

plant's ability to recover from the break. The loss of the pump (or pumps) would reduce the

available secondary side cooling, which could create a condition that is a substantial safety hazard.

The chargWsafety injection pumps are used Inthe high head safety Injection system for the

applicable plants. 'Me loss of these pumps during the short tamn itgation phase of a LOCA would

impair the plant's ability to mtigate the consequences of the LOCA, especiaUy for a small break

LOCA. The loss of the pump (or pumps) would reduce the overall flow to the core, which could

create a condition tha Is a substanial safety hazard.

The folowing recotmiors are provided for this Issue.

1. Review Tables 3, 4 and S to daermine whether the plant has any parts that could be affected

by this failure meancsm. The parts Identified in Table 3 ad S are considered Wdecs as

defined in 10 CFR 21. Although the pans InTable 4 are not considered defects pursuant to

10 CFR 21, the pats in Table 4 should eventually be replaced as a prudent maintenance

practice since these pam are susceptible to the same fillure mechanism.

2. Compare the information for the part In Tables 3, 4 and S to determine whether the pan is

currea y installed on the pump. In me cae, this pan may have bern changed after the

par was supplied.

.. _. _ . __,..

FEB 20 '95 11:29 FROM: OL LICENSING T0 813018165151 PfiGE. 06/12

3. For paru in Table 3 ad 5, if it isdetermined da the pump part Is currectly insuloed on the

pump, de the folowi shoduld be considered. First. as Indicatd in Table s, if dte affected

part is a shaft sleeve compressio mm or a shaft sleeve collar, the part can be Wnspected for

cracifg without diassemby of the pump.

Alternatively. if the affected part can not be ispected without dissembly of dhe pump ad

it is ot practical to immediately disasumble the pump, then th pump operating history

should be reviewed relative to the edanism for sws corosion crackig. Te mechanism

for nress corrosion crackig is dependent upon several fictors including, but ao limited to.

the amounr of sress placed on the part, the time the part is exposed to that stress, the time

exposed to an aqueous enviromen agd die physica dimensions of the part. By reviewing

these (aors. it may be possible to deaonsate dutt the pat filure is not imminent andlor

dt the part will not fai In a manner hat will preven the pump from performng ts Intended

safety function. However, It Is ultimately recommended that the pan be replace with the

part curreay rommded by IDP.

The above information is being concurrently trasmitted to affected utilities via supplement to

Westinghouse letter NSAL.94-023.

If you have any questions regarding this trmsiltal. please contact H. A. Sepp of my staff on

412/374-SS82.

Very truly yours.

. Lparulo. Manager

Nuclear Safety Regulatory and Licensing Activities

"WF/P

cc: R. E. IoinesrlDP

G. MorrisseyfiDP

LUXAW-1=f95

TABLDE 3 ORIGINAL CONSTRUCTION PUMP PARKS FROM HEAT NISISS

AND M2144 I _a

uF11I5lY VINU Ibntv Mod~d a P -_

.... _ __=____- .. run Year Shipped

KalI gPI2*.

v .v ^ Di~

flia mu eiutgc 'UTmnned=Me Impeller Locmi

Head Safety Injecdon) 1975

49347 Preftwe Reducing

49348

49349

49350

49350=

Sleeve Locknut

of- (Ci

i i_ --- - - - -

-l'WA _w. D J e

I,A wit VW I Ia 2 JHF 10 Stage (lwermedaste Impellro Locknut r

Head Safety Injection) 1975 I

49351

493S2 Prcsswe Reducing

Sift" Lftkm mT

49353

_

493534 SDm%& e.

I-

- I

rMNeb* .. I LADO%.. I A I I. 'r"_

QIIu

APUPW U VWm MUMICu I

C A JHF 10 Stae (IUnw edae I Impeller

Locnut 1975 Head Safety Injectlo)

49355 Pressure Redocing

49356

49357

_49358 Sleeve Locknut

t

I

L- -

nq..L ft- D- 1' _

- _

Caawba I & 2

. l_

_I

LUrw nB 10 Stg (Intermediate Impdler LocKn

Head Safety Injection) 1976

49359 Pres Redoucing

49360 UV

49361 Sleeve Lmckve

49362

_ , . . 9I I

X '_^wfh G_

bWsoU DX^_] l ^ ^ Spacer Sleeves _ _

E hAO l lrfUWOO I & 2 J1F 10 Slage (Intennedlate Impeller Lockt

Head Sarety Injection) 1978

49762 PressMe Reduifng

49763 Sleeve Locknut

49764

4976S Spacef sleeves

L .h

49765 I Spacer Sleeves I

m~l7I~w.wpI:b.O215- I I

C

t

TABLR 3 . contired

ORIGINAL CONSTRUCTION PUMP PARTS pROM

HEAT 1589 AND 928144 I -1 utlrty UdsX Pmnp WMnAyp.&li

I

I %UAUWuu1WVu

~.77 rwsvn

-- I1 a

BYM I &Z

. U.- ."____

, 1JHF P10u ,, "a Ia9.I Pan W97r 1hpped

Stage (Intmerndlate Impelcer Lockhd

Head Safety Injection) 197

49758 PTec Redtcing

49759 Sleeve Lodkn

49760

I ft..6.12- __ in-49761 I Spacer Skee I

I rU1zC ;crnco of MSie tn Ih 2 JHP 10 Stage (Sevice

I1dba tpeller Lockv"

Uhm) r

Presurc Reducing

49754 Sleeve Lockmnt

49755

49756 su-Pa I 52D79 &%__

-- .-- I - -- -I 52079 I

I o

VT U II % ? 1 LR W E L V!CK J1W II Stage (Internediate

I

Spaer Sleeves I-J

Operating Compity Head Safety Injection)

51647 1977 a

S164R

.'

Aortei

. -.- j.v,l

hstofp 4 ! h..AS. A . A i ..- * -_ -- _ _

n L#aIUF u1urf ruwqcr JIff IU sbtge (Service

Lab Impeller Lackut

Unknown)

40746 1976 I~d

_I * - sPac meefts

0

fli A?S . wpf:Ib .6 5 5 -S 2

4.

TABLR 4 ORIGINAL CONSTRUCTION PUMP PARTS FROM 1RAT II59 AND #284q, RACEMENT OF PART IS NOT MANDATORY

1r

I . I _ Y

wlry uits ?P~m Modetsrwia m 777---

1im

-

r I .11 nsit

0.-- , ,

-I I_ -- __

vb_

I YeaoSr pped S.

- .aw. .

I r au owt tgmcnn e Split Rings

Head Safety Iection) 1975 C .

49347

49341

49349

.

49350

YIS

I .D

--

a .

I-.- I --- I1 TI-

a V b Wrms ow a c.

^

A n to tsag (intmete a_

r

S Rings 197S

lead Sukty Injecfto)

49351

49352

49353

493S4

..

i . _

-9354

. . . ^

a

_

WUUR rUlm MUA=n I &Z JHF 10 Stage (lotermevise Split Rings c- Head Safety nJecdon) 1975

49355

49356

49357

49358 w

WL_

- t-

.... I _ _ . ^

fI

I I_

I0

110151 rxywur %aWua I a A JHF 10 Stage (ftnemcast ED&

Split Rings 1976 Head Safty Injection) Id

49359

49360

49361

49362

__ _I

it71 w.wo'f:Ib- Ol2 - 3

. a

TABLE 4 - continued

ORIGINAL CONSTRUCTION PUMP PARTS FROM M13AT 15S199 AND 2h144, R1BPLACEMENT OF PART IS NOT MANDATORY

PA

Wily _Ut Pump Mode/rw R" Year Sh

I

ODIRnOfM2ith Suuidwood I & 2 JtW 10 Stage (lntemedlgc pis

Splitng

Edison Bead Safety Injection)

49762

49763

49764

49765 r

commaweft Byro l &2 )HP 10 Stage (ltefmrvedlae Split Rings

Edison . Read Safety Injection) 1978

',

49738

49759

49760

49761 I _ _ _ ^ ^

_ . . _ _

IHF tO Stage (Service

^ . . . _

Marte Hl) I & 2

^ _] ^ .. ...

Pubfbc S1ecoe Of

_LZ

Split Rings 1978 Inmdi Unknown)

49754

49755

49756

52079 Wolf Creek Nuclear Wolf Creek I JSI II Slage (Intenmediase 8 Split Rngs Id

Operidg Compan Head Safety Injection)

51647

_ __ _ _ _ _ __ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _5 1 64 8 Aeljet Noclear Beals Awomc Power IHF 10 SMage (Servie Split Rings

Lab Unknown)

__ _ _ _ _ _ _ _ _ _4 _ _ _ _ 9 756 mMn7AMwwe.Ib42195" 4

I

TABLES . 0

REPLACEMENT PUMP PARTS FROM

fEAT #35899 AND #2144 19

I a

.

I

. =k".

I Utiniy lUnif u Medeuseri i

Pmp rA Pa

I Pa om Heal Rem 15

..

ad 221d

a

Meemw IN _We

. - l

Cbmmoawalth Z n1I ITCH 10 Stage (Auxion Impelier Lochmt

Pedwater) It118/75

45796 Sha Sleeve

COmpresWon NuM Rod

( i

I

shR Sleeve

TL- &-- -- . - - I Nut nT

rm-m rmm uenmuje Materma] (See Nnfr 1f

A _ _ I ._ _ _ ___ . _ __ __ _ _.I

'aummu row" Futey 2 U II Suage (ChtsrgI4SafM Shaft Sleeve CoUlar I(Vt3n

Ilijection)

AmF)f

- - 1 qgIWt

Zbo I ITCH 10 Stage (Auxily

Edwon Peedwer) Impeller Locl ntad 4sr7S

AV70A

% a &U i' ID -. - I.

Zion I ITCH I( Stae (Auxullly

Edb" Impeller L ,zt _ 9a

ACN

45796 0

CD

Consolidated Edl Idm PoW rn

JMTCH 10 Siage (Safety

InJecdion) nPller Lom Red I VI 1m76 I

43461 Sluaft Sleeve

-- - 43466 tf

co

Duquesnm Ugiw Beaver Valley I U (Charing/Safety Injecton) Shlit Sleeve Collar

LI~I

, __ ..

46351 9W3t6--

nMPower Cook 2 u (ChalgnolStey Injection) b kft Sleeve Collar

_145607 orM76 mffurt36..,,..zInst5 S

e

TABLE 5 - coutwd

RMPLACEMM PUMP PARTS PROM HEAT ISM9 AND P23144 I

vihity Unm Pump M~tgetSeddg R *

a!Ins u pIS&eIpped *

Ii Eectric Cik I & 2 U (chatngtSasc Iet ldon) Stft Skere colta SM.r 78

.___4

_

.566__'

e .J^ ......

4 _9

'd

I7Umcm a ruwc ur Iso I hZ UNI 11 Ste (AUXK~y brpel Locukmt

Ftedwter) 12/19f76

46571

45SX1 prmw~

nlooI dx redu -a

Pud Gemi Tonm I (CIM00S1fety hieefts) haft Sleeve C.Uela

Elemctic

Pdftc Oas & flel:Fti Dobfo Comyn nH

43603 r2 1o Stae (Sdty Shift Sle nw7 I

__________ _________ .epcso JCdC

43489 c a m o !I SK?

PI'm. Serylce Ee: Salc I J~r-H 10 Stage St ShW Sleeve

old an Intitdon)

45493 PUc SUvicebectft Sam 2 ifII Stage (Cltvgarsh y Sftfi Scow

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K>_ <_y Attachment 2 IN 95-26 May 31, 1995 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

94-61, Corrosion of William 05/25/95 All holders of OLs or CPs

Supp. 1 Power Gate Valve Disc for nuclear power reactors.

Holders

95-25 Valve Failure during 05/11/95 All U.S. Nuclear Regulatory

Patient Treatment with Commission Medical

Gamma Stereotactic Licensees.

Radiosurgery Unit

95-24 Summary of Licensed 04/25/95 All holders of OLs or CPs

Operator Requalification for nuclear power reactors.

Inspection Program

Findings

95-23 Control Room Staffing 04/24/95 All holders of OLs or CPs

Below Minimum Regulatory for nuclear power reactors

Requirements and all licensed operators

and senior operators at

those reactors.

95-22 Hardened or Contaminated 04/21/95 All holders of OLs or CPs

Lubricants Cause Metal for nuclear power reactors.

Clad Circuit Breaker

Failures

95-21 Unexpected Degradation 04/20/95 All holders of OLs or CPs

of Lead Storage Batteries for nuclear power reactors.

94-64, Reactivity Insertion 04/06/95 All holders of OLs or CPs

Supp. 1 Transient and Accident for nuclear power reactors.

Limits for High Burnup

Fuel

95-18, Potential Pressure-Locking 03/31/95 All holders of OLs or CPs

Supp. 1 of Safety-Related Power- for nuclear power reactors.

Operated Gate Valves

OL = Operating License

CP - Construction Permit

IN 95-26 K>

< May 31, 1995 Discussion