ML20214P502

Forwards Response to Question 4 of 870309 Request for Addl Info Re NUREG-0737,Item II.D.1, Performance Testing of Relief & Safety Valves. Responses to Questions 1,2,3 & 5 Provided on 870408

ML20214P502
Person / Time
Site:	Wolf Creek
Issue date:	05/29/1987
From:	Withers B WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM WM-87-0147, WM-87-147, NUDOCS 8706030413
Download: ML20214P502 (7)

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WQLF CREEKNUCLEAR OPERATING CORPORATION Bart D. Withers Presadent and Ch6ef ExecutNo Officer May 29,1987 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Letter: WM 87-0147 Re : Docket No. 50-482 Ref: 1) Letter dated 3/9/87, from PW0'Connor, NRC, to s to BDWithers, WCNOC

2) WM 87-0112 dated 4/8/87, from BDWithers,WCNOC, to NRC Subj: NUREG-0737, Item II.D.1, Perormance Testing of Relief and Safety Valves Gentlemen:

Reference 1 requested that additional information be provided in support of the NRC Staff review of NUREG-0737, Item II.D.1, "Peformance Testing of Boiling-Water Reactor cnd Pressurized-Water Reactor Relief and Safety Valves", for Wolf Creek Generating Station. Wolf Creek Huclear Operating Corporation's responses to questions 1, 2, 3, and 5 were transmitted by Reference 2.

As discussed with the Staff, additional time was required to prepare the response to question 4 of Reference 1. Wolf Creek Nuclear Operating Corporation's response to question 4 is provided as an attachment to this letter.

If you have any questions concerning this matter, please contact me or Mr. O. L. Haynard of my staff.

Very truly yours, a

__^LY Bart D. Withers President and Chief Executive Officer BDW:Jad Attachment ec: P0'Connor(2)

RMartin JCummins 8706030413 870029 2 DR ADOCK 0s00 qh0 PO. Box 411/ Burtington, KS 66839 / Phone. (316) 364-8831 An Equal Opporturwty Ernployer PAF.HCVET

Attachment to WM 87-0147 Pags 1 of 6 May 29, 1987 WOLF CREEK NUCLEAR OPERATING CORPORATION RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION CONCERNING NUREG-0737 ITEM II.D.1

" Performance Testing of Boiling-Water Reactor and Pressurized-WaterReactor Relief and Safety Valves" Question 4 The submittals from the licensee provide only scant information on the pipe supports. Thus, further information is requested as follows:

A. The load combinations in Tables 2-1 and 2-2 of the January 7,1983 sub91ttal are said to; govern.both the piping and supports. The stress limits identified in the tables, however, are applicable to pip 9 stresses but are .not . typical of stress limits for pipe supports. Thus, clarify the stress limits used in the load combinations for the supports and identify the governing standards or codes. Also, verify that the load combinations listed in Tables 2-1 and 2-2 were indeed used in the support evaluations.

B. Present a table comparing the maximum loads (or stresses) in several representative supports with appropriate allowable loads (or stresses). Provide comparisons for normal, upset, emergency and faulted conditions. Indicate in the Table the support number and support type, and supply sketches to show locations of the supports listed in the table.

Response

In order to be consistent with the stress analysis, pipe supports from the pressurizer to the first seismic anchor downstream of the relief and safety valves were designed in accordance with ASME Section III Class 1, even though the pipe class changes from ASME Section III Class 1 to ANSI B31.1 downstream of the pressurizer relief and safety valves. The applicable support stress limits and the governing codes for normal, upset, emergency, and faulted conditions are provided in Tables A-1, A-2, and A-3.

The load combinations listed in Tables 2-1 and 2-2 of the report submitted by SLNRC 83-002, dated January 7,1983, were used in the support evaluations with the exception that the pipe break-related loads (Loss of Coolant Accident, Main Steam / Main Feedwater Pipe Break, or Design Basis Pipe Break) as defined in Table 2-3 were not used. It was concluded that the Main Steam Pipe Breaks, Feedwater Line Breaks and Loss of Coolant Accident loads do not need to be included, since these loads have a negligible impact on the pressurizer safety and relief valve supports when compared to other loads.

Tables B-1 and B-2 provide a listing of representative supports comparing pipe support load / stress values to their corresponding allowable load / stress values for normal, upset, emergency, and faulted conditions. The support locations listed in these tables are identified in Figures 6-1 and 6-2 of the report transmitted by SLNRC 83-002, dated January 7,1983.

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ArztameIE STRESS IRGITS IUt PER*NIRIZER SAFErr AID NN WIN vat #E SYSTIBE SLFf0IES - UPSTREERIE AaB nrnesaTIEBft (F V5t#ES .

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l Plate and shell design by analysis W-3221 W-3222 W-3223 b

W-3224 W-3225 Linear type supports by a,nalysis W-3231 W-3231 W-3231 W-3231 W-3231 Camponent standard supports design W-3240 W-3249 W-3249 by analysis W-3249 W-3249 Ozaponent suIports design by load rating W-3260 W-3260 W-3269 W-3269 W-3268 i

j j tcIES:

1) Paragraph ntambers refer to ASME Code,Section III 1974, Subsection W, including Winter 1974 Addendtm.

2

2) The supports upstream of the subject valves are classified as ASME Section III Class 1 and designed i

accordingly. The supports downstream of the valves are classified as ANSI B31.1.

1 designed in accordance with ASME Section III Class 1 requirements for consistency. However, they were 1

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NJN S11tESS LIMITS FGt PRESSURIZER SAFEIT AM) yh metrar VALVE sysmos suProms - noummann or vAtsmS AIEil.B31.1 PIPINE

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0 IDhDING CXBDITIGER casommer SUPPORT WPE MERSE. UPSET Bewarv ygLgD APPLICAREE y

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1.94 1.94 1.9S 1.33S h ANSI B31.1 9 Paragraph 129.2.4 $

Plate and shell design by analysis 1.9Sh 1* s 1*23 h h 1.2S ANSI B31.1 h

Paragraph 121.1.2

, Linear type supports by analysis 1.9S 1*ES '

other than supplemental structural h h l.2% 1.24 ANSI B31.1 steel including conponent standard Paragraph 121.1.2

., supports ,

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NOIES

1) Sh= All able Material Stress by Applicable Cbde l

2) Ioading though conditions were provided piping is classified as ANSIby B31.1.

Westinghouse and are based on ASME Section III design requirements even 4

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~N DESIGN IftOING C3EINATIOtG PER SUPPORrS PGt PIESSURIZER SarErr m maar vatve SuePourS .{

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a Upset (1) N + SOIU + DI "

(2) W + CBE + SOIU + Mi Dnergency N + SOIE + Di Faulted (1) W + SSE + SOP + MI (2) W + SSE + U{

i LEGEND:

n1 = 'Ihermal DW = Piping deadweight OBE = Operating basis earthquake SSE = Safe shutdown earthquake SOIU = Relief valve discharge (transient)

SOIE = Safety valve discharge (transient)

Sor = Maximtzn of SOIU and SOTE Note: Use SRSS for cmbining dynamic load responses.

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TM11E IS-1 Attachment to WM 87-0147 SUMARY & PIPE SUPPORY DESIm Page 5 of 6 May 29,1987 to Aset sacrim III e

s TPPORP N). IBS2-IWS4, R/6 IBS2-RS25, R/S EBS2-HSS1, P/4 Aly. Q]DE AMil B31.1 (1) MEI B31.1 (1) Aset cASS I oN NGE 1280 NGE 3338 KIE 3S66 .

DESIGN AIEWABIE DESIQi AIJDSBIE DESIQi IDADIE IDAD , IDO IDAD IDAD IDAD M ITIN ggypg) gggpg) -(KIPS) (KIPS) (6) (KIPS)

N/A N/A (5) (5) 1.4257 2.475

.fbrmal Upsett N/A N/A 9.6799 15.999 1.4518 2.475 Upset 2 N/A N/A 1.7152 15.999 1.4519 2.475

. Dnergency N/A N/A 7.3057 29.990 1.4522 2.475

', Faulted N/A N/A 7.4460 22.199 1.4522 2.475 Faulted 2 N/A N/A 1.4399 22.199 1.4519 2.475 g MAX. S1PESS STEESS MAX. STRESS S'!9ESS 983. S'IRESS SM c: tbrmal 0.08FA 1.0Fg 9.E g 1.9F3 ti/A N/A O

& Upsety 0.20F3 1.0FA 0.02FA 1. F3 N/A WA s

s Upset 2 0.23FA 1.0F3 0.06Fg 1.0Fg N/A N/A va Dner@ency 0.3Fg 1.33F3 0.25FA 1.33F3 N/A N/A O 3 Faulted t 0.3FA 0.25FA F 30 N/A N/A S F 30 Sec Faulted 2 .23F A S F 30 U*

A Sec F 30  !^ "!^

'>. AIEWABLE g MAX. SIPESS (cr. + o" ),

tbrmal N/A 0 tF-3222 N/A g Upset y N/A 0.022Sg tF-3223 N/A l Upset 2

N/A 0.056Sg tF-3223 N/A Dnergency N/A 0.236Sg tF-3224 N/A Faulted N/A 0.241S NF-3225 N/A M

Faulted N/A 0.047sg rF-3225 N/A 2

i totes:

(1) ANSI B31.1 design required but ASME Section III Class 1 designed (see Table A-1)

(2) F = Allowable stress

= Design stress intensity value by ASME Section 111. Table I-1.2

, o(g=membranestressando = berdirq stress (pipe stresses are l also checkiry considelirq c{, og, and pipe operating cordition stresses) .

(3) Para]raph ntscers ref er to ASME Cxle Section III,1974, Subsection l 57 includits 1974 Ai!erdum.

(4) This suaort has a constant spring support. Derefore, there is no l

increase beyard rurmal design cordition.

(5) Dis cmponent standard sutport is a snubber.

• Werefore, the rurmal operating loads are not restraine1.

(6) Allowable Icn3 values f rom Dergen-Paterson Pipesupturt O)tp. IAd Ca, 6 4 Data sheets.

Attachmeat to SM 87-0147 Page 6 of 6

'- May 29, 1987 .m uE s-2

' SED 94ARY & PIPE SUPPORY DESIGN 10 M EI B31.1 SUPPORT 90. BBS3-C992, W 5 EB03-HG91, WS APP. umz ABEI B31.1 MEI B31.1 gg NEIE 3239 NEIE 3235 (1) NEIE 3139 DESIM Alm 0BIE DESIM AlmeBIE DESIGN AIJIMkBIE IDAD IDAD IDAD IDAD IIMD IDAD (KIPS) (KIPS) (2) (KIPS) (KIPS) (2) (KIPS) (KIPS) (2) termal 0.6211 4.500 1.0958 1.50 0 1.2490 4.500 Upset y 0.6211 4.500 1.0963 1.500 1.2530 4.500 Upset 2 0.6316 4.500 1.0965 1.500 1.2536 4 .500 Dnergency 0.6212 5.400 1.1002 1.600 1.2638 5.400 Faulted 0.6387 5.400 1.1927 1.800 1.2643 5.400 t

F Faulted .6213 5.400 1.1911 1.800 cc 2 1.2526 5.400 o A M M M M M M

, MAX.EnRrsS STRESS MAX. S1RESS S1RESS MAX. S1RESS STRESS

& termal

} N/A ,

N/A 0.13Fg 1.0Fg N/A N/A M Upset y N/A N/A 0.13Fg 1.0F A "!^ ^

a. Upset N/A 2 N/A 0.13F3 1.0Fg N/A N/A o

energency N/A N/A 0.13Fg 1.0Fg N/A N/A Faulted y N/A N/A 0.13F3 1.33FA N/A N/A

>i b h Faulted 2 N/A N/A 0.13FA 1.33Fg N/A N/A tJormal N/A N/A N/A Upset y N/A N/A N/A D1 Upset N/A N/A 2 N/A Dnergency N/A N/A N/A Faulted N/A N/A t N/A Faulted N/A N/A 2 N/A tbtes:

(1) Nodes 3230 and 3235 are 7 7/8 inches apart.

(2) Allowable load values fran Bergen-Paterson Pipesupport Corp. Ioad Capacity Data sheets.

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