ML20009D217

Summary of 810609 & 10 Meeting W/Utils & Snupps in Gaithersburg,Md Re Review of Mechanical Engineering.Meeting Agenda & Applicant Responses Encl

ML20009D217
Person / Time
Site:	Wolf Creek, Callaway
Issue date:	07/06/1981
From:	Dromerick A, Edison G Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8107230288
Download: ML20009D217 (60)

Text

,

JUL 6 1981

• fg b

Douket Hos.: 15TE 90-48ted f

/M and STN 50-483~

l4

%@ f 4 3

-f APPLICANTS: Union Electric Company SI

[

12 Kansas Gas and Electric Company FACILITIES: Callaway Plant, Unit 1 C>9 f/

/

Wolf Creek Generating Station, Unit l' 4

SUMECT:

SurtlARY OF lEETING I4 ELD JUNE 9 AND 10,1981 WITH CALLAWAY AND WOLF CREEK APPLICANTS TO REVIEW t!ECHMilCAL ENGINEERING A' meeting was held on June 9 and 10,1981 at the Bechtel offices in Gaithers-burg, Maryland with representatives of the tion Electric Comparty, Kansas Gas and Electric Conipany, SNUPPS organization, Bechtel Power Corporation and Westinghouse Electric Corporation. This meeting was held as a result of our

. letter of-April _ 22, 1981 to the applicants transmitting a draft SER and draft questions on the SUUPPS~ FSAR, and requesting a meeting to discuss those ques-tions. The agenda for the meeting is attached as enclosure 1.

The applicants' responses to the draft questions, as revised and agreed to by HRC, are attached as enclosure 3.

The status of the agenda items (enclosure 1) follows.

. Topic III.

(a) Bechtel/ Westinghouse Division of Responsibility The applicants indicated that Westinghouse is responsible for all Class 1 piping stress analysis and Bechtel for all Class 2 and 3 piping stress analysis. Westinghouse is responsible for design of the piping supports for the reactor coolant loop and pressurizer surge line, and Bechtel is responsible for design of the supports for Class 1 auxiliary lines.

(b) Comparison to ASB and HEB Criteria The applicants indicated that in general they are adhering to NRC Branch Technical Posiitions 3-1.

In the case of MEB 3-1, paragraph B.3.b(2),

the applicant stated that Class 1 auxiliary lines are being fabricated as seamless piping,' and that a longitudinal rupture is so low in probability it should be considered incredible. MRC indicated that further review of this position 5;ould be required and the iss;e will remain open. In the case of ASB 3-1, paragraph B.3.b(3), the applicants stated the Auxiliary Feedwater System will be classified the saue as a raoderate-energy system because it will only be used under eccident conditions. A separate svartup/ shutdown feedwater train is being added to the min feedwater sy st%.

Inis was tot nu acceptable oy rmL.

OFFICE) u....................

.a.......--.-

-.nono.a. n.o no.uan~~

o.a.

a.~.-n.u-

-~.~~.n~~.

~ - - ~ - -

h hhh-k hh-b b

$[f$0 sunme) oo 2-~~~~~~~---

---

~~~

.,............P,p,,,,,,,,,,,,,,,,,,,,,,,,

omy.... A

- nne ronu sie noencu oua OFFIClAL. RECORD COPY usom --mm

! (c)

High Energ) Line Break Analysis The applicants indicated to ASB that they would clarify the FSAR to indicate that only the charging systems piping is not considered to have a high energy line break in one train concurrent with a single active failure in a different train. We indicated we would consider it further but did not think we would find it to be a problem since no serious transient could result and it is consistent with previously acceptable plant reviews.

Topic IV. Auxiliary Systems Branch Questions We qJestioned whettier a pipe failure in a non-Safety grade systen routed above 'he Essential Service Water piping night erode the eupport around the ESW piping. The applicants stated that the ESW piping was enbedded in con-crete so this would not be a problen. They will revise the FSAR to clarify this.

Topic Y.

Optional Discussion Items We toured the SNUPPS scale plant model both days to evaluate selected piping runs, supports, and restraints. An impromptu audit was perforr.ed on several piping support calculations.

The calculations were found acceptale by the HRC.

Topic VI. ItEB SER Draft Questions There were 31 agenda items discussed whicn had been identified fron the draf t questions in the April 22 letter.

In addition, 8 additional questions were developed during the necting, for a total of 39 items.

The status of each of these itens is listed below.

Item Status 1.

Pg. 3.2 - 2 Resolved by revising FSAR page 3.2-2 as agreed to in the neeting.

2.

Pg. 3.6 - 3 Resol ved. No change to FSAR required.

(section 3.6.1.1.h.2(b))

3.

Pg. 3.6 - 4 Resolved. FSAR will be revised per

( section 3.6.1.1.j )

words agreed to in neeting.

4., Section 3.6.1.1.k Resolved, pending review of an FSAR revision to be provided in July 1981 which will include a sunnary of results from the Pipebreak Hazards Protection Ano:,31s.

\\

omet >

SURNAME)

...................a

a. a. a.a. a a a

.m.

..a a

a.a onn >

.................a

.a

.....a

.... ~..

a OFFICIAL RECORD COPY usom e-mea unc ronu ais o0-80) NaCM ONO

i 3

b h

7; b#

6 e

W y

3 L

- 5.

Figure 3.6 -1 Resolved, pending review of FSAR re-vision to be provided in July 1981 updating Figure 3.6 -1 and providing results of the pipebreak hazards ana-lysis.

6.

Section 3.6.7.

FSAR will be revised to include a statement that leakage cracks in non-Seismic Category I piping are addressed at all worse case locations.

7.

Table 3.6 - 3 Resolved. No action required.

(Sheets 1 thru 8) 8.

Table 3.6 - 3 Resolved. FSAR will be revised in July, 1981 to incorporate updated sheets 28, 32, 36.

9.

Table 3.6 - 4 Resolved. FSAR will be revised in July, 1981 to reduce the number of "under review" i tecas.

10. General Clari_fieu (Resolved).

No action required.

(Protective fleasures for Jet Impinsement)

11. Pg. 3.7(B) - 7 Resolved. FSAR will be revised as agreed to in meeting to indicate use of Reg. Guide 1.92 Equation 4.

Also, reference to sec-tions 5.1 and 5.2 of BP-TOP-1 will be eliminated.

12. Figures 3.7(B) - 5 Resolved.

'SAR will be revised, using words thru 3.7(B) - 8 agreed to in the meeting, to delete these figures and instead refer to NRC approved topical report BC-TOP-4-A, Rev. 3.

13. Figures 3.7(B) - 9A Resolved. No action required. The seismol-and - 9D oqy area is currently under review by HGEB at at NRC.

14. Pg. 3.7(N) - 14 Resolved per the clarification agreed to in the meeting. FSAR will be revised to correct Equation 3.7(N) - 30.

15. Pg. 3.9(B) - 1 Resolved. FSAR will be revised to clarify (Section 3.9(B).l.1) whether thermal transients or other dynamic events are referred to.

DATEh n:c ronu ais iio,soinacu o24c OFFICIAL RECORD COPY

• * ' + m e24

f.

. 16.. General Resolved. FSAR will be revised (section (Thermal shock to RPV 3.9.5) to incorporate the words agreed to internals from ECCS in the meeting. Thermal shock is analyzed i

injection after t.0CA) under LOCA/ECCS injection conditions.

17. Pg. 3.9(B) - 1 Resolved. FSAR will be revised using words (Section 3.9(B).l.2.1.1) agreed to in the meeting. RELAP4 should be referenced in SEction 3.9(B).1.

18. Pg.'3.9(B) - 3 Resolved. FSAR will be revised (pg. 3.9(B)-3)

(Section 3.9(B).1.3.2) to eliminate " inelastic method or."

19. Pg. 3.9(B) - 4 Resolved. FSAR will be revised per the (Section 3.9(B).2.1) meeting response.

20. Pg. 3.9(B) - 5 Resolved. No action required.

(Section 3.9(B).2.1)

21. General Open. NRC is considering further. The (Assure the functional applicants are preparing to modify their re-capability of Class 1,2, sponse based on input from Westinghouse.

3 piping essential to Small piping needs to be adoressed in more safety under all loads) detail.

22. Pg. 3.9(B) --15 Resol ved. FSAR will be revised using words (Section 3.9(B).3.3.1.g) agreed to in the necting to note there are no instances when a dynamic-load factor 2.0 was used.

23. Pg. 3.9(N) - 33 Resolved. FSAR will be revised using words (Section 3.9(N).2.1 agreed to in the neeting.

'24.. Pg. 3.9(N) - 36 Resolved, pending review of FSAR revision to (Section 3.9(N) - 2.4) sunnarize the basis for SNUPPS plants being classified as non-prototypic Category I in accordance with Reg. Guide 1.20.

Wording used in Commanche Peak FSAR will be considered and is expected to be acceptable.

25.. Table 3.9(H) - 3 Resolved. FSAR will be revised to add the words agreed to in the meeting.

26. Pg. 3.6 - 10 Resolved. Applicants will revise the FSAR to (Section 3.6.2.1.1.9.2(B))

indicate that 2.4 Sm was used.

n omce>

.......... ~... - ~..

- - - ~ ~ -

- - - - ~

sunume >

.. ~ ~ -.. - - -

emp

-. - ~ ~ ~. - ~. -

. - - - - - ~ ~

Nac ronu us oo aoi nacu ano OFFICIAL RECORD COPY usom mmeeo J

.C

.3-2 r

s I,H

27. Pg. 3.6

13 Resolved. Welded attachments are not used (Section 3.6.2.1.1.e) on high-energy piping in containment penetra-tion areas. Applicant comitted to provide

~

location details and advise HRC f f welded attachments are used in the futuro.

i.

- 28. Tables 3.9(N) - 2 and - _4 Resolved, pending review of FSAR revision per the meeting response.

29. Section 3.9.6 Open. A separate submittal will be made by applicant in July 1981 to respond to this question.

L

30. General Open. Test program to be submitted by (Submittal of preservice applicant in July 1981.

and inservice test pro L

for pumps and valves) gram

31. Pre-Oper. Testing of Resolved. Applicant provided response in Snubbers (also Pre-service meeting minutes.

Inspec.)

1

32. RG. 1.121 Resolved. Not an SER. item: However, must be reviewed for-license tech. specs. before granting an operating license.

33.- Exception to McB 3-1 Open. NRC does not accept the response and (No jet _ impingement feels that further justification is required.

effects are considered The applicant believes NRC should further for Class I longitudinal consider their position but will. provide addf-breaks) tional justification.

34. RG. 1.124 and 1.130 Resolved. FSAR will be revised per the words agreed to in the meeting.

35. Section 3.9(N).3.3.A

' Resol ved. FSAR will be revised per the appli-cants' response in the meeting.

36. -Table 3.9(B) - 7 Resolved. FSAR will be revised using words agreed to in the meeting.

37.

t-

. Table 3.9(B) - 3 Resolved per the discussion in the meeting j

and - 5 and the applicants' response.

a l

4 omer>

........... ~. " - - -

- - " " - - " ~ ~

~ " " " " " " " " " " '

~ " " " " " " " " "

sunnme >

.................. ~.

.......... ~. - ~ ~

"~""""""~~"

our>

.................... ~.

- ~ ~ ~ ~ ~ "

• ~ ~"""""""

" " " ' " " " " ~ " '

~""""""""""

OFFICIAL RECORD COPY us =

• wac ronu sia 00-80) NRCM ONO

• T M*

r

, 38. Pg. 3.9(N) - 44 Resolved by revising the FSAR using the words agreed to in the meeting response.

39. Section 3.9 Resolved by FSAR revision using words agreed to in the meeting response.

heiginal signed b3 A. W. Dneerick j A. W. Dronerick, Project Manager Licensing Branch No. 1 Division of Licensing Original signeJ bysf OsNoa 5. Mason G. E. Edison, Project flanager Licensing Branch No.1 Division of Licensing

Enclosures:

As stated cc: See next page i

DL:LB#1

, DL:LBel D

i OFFICE) sunname) GEdf. son.Jy.s....aQr. amer.ick..

B.J.

99.d.

..?!.W............ 7.Y.8.....

.ZL

8.......

EATE)

Nac ronu m com> uncu cao OFFIClAL RECORD COPY usemasei-meo

MEET!'

iUMMARY DISTRIBUTION

..; Docket File ~

G. Lear NRC POR WJohnstgn Local PDR JUL 6 1981 S. Pawllcki TIC /NSIC/ Tera V. Benaroya N. Hughes Z. Rosztoczy LB#1 Reading W. Haass H. Denton D. Muller E. Case R. Ba'llard

0. Eisenhut W. Regan

/

R. Purple D. Ross B. J. Youngblood P. Check A. Schwencer Chief, Power Systems Branch F. Miraglia

0. Parr J. Miller F. Rosa W. Butler G. Lainai R. Vollmer W. Kreger J. P. Knight R. Houston R. Bosnak Chief, Radiological Assessment Branch F. Schauer L. Rubenstein R. E. Jackson T. Speis Project Manager Edison /Dromerick MSrinivasan Attorney, OELD J. Stolz M. Rushbrook S. Hanauer 0IE (3)

W. Gammill ACRS (16)

T. Murley R. Tedesco F. Schroeder D. Skovholt M. Ernst NRC

Participants:

R. Baer C. Berlinger D. Terao (3)

K. Kniel R. Bosnak G. Knighton A. Thadani D. Tondi J, Kramer D, Vassallo P, Collins D. Ziemann bec: Applicant & Service List W

d 3

4 I

6 Mr. J. K. Bryan Mr. Glenn L. Koester Vice President - Nuclear Vice President - Nuclear Union Electric Company Kansas Gas t.ad Electric Company P. O. Box 149 201 North Market Street St. Louis, Missouri 63166 P. O. Box 208 Wichita, Kansas 67201 cc:

Gerald Charnoff, Esq.

Shaw, Pittman, Potts, Dr. Vern Starks Trowbridge & Madden Route 1. Box 863 1800 M Street, N. W.

Ketchikan, Alaska 99901 Washington, D. C.

20036 Mr. William Hansen Kansas City Power & Light Company U. S. Nuclear Regulatory Commission ATTN: Mr. D. T. McPhee Resident Inspectors Office Vice President - Production RR #1 1330 Baltimore Avenue Steedman, Missouri 65077 Kansas City, Missouri 64141 Ms. Treva Hearn, Assistant General Counsel Mr. Nicholas A. Petrick Missouri Public Service Commission Executive Director, SNUPPS P. O. Box 360 5 Choke Cherry Road Jefferson City, Missouri 65102 Rockville, Maryland 20850 Jay Silberg, Esquire Mr. J. E. Birk Shaw, Pittman, Potts & Trowbridge Assistant to the General Counsel 1800 M Street, N. W.

Union Electric Company Washington, D. C.

20036 St. Louis, Missouri 63166 Mr. D. F. Schnell Kansans for Sensible Energy Manager - Nuclear Engineering P. O. Box 3192 Union Electric Company Wichita, Kansas 67201 P. O. Box 149 St. Louis, Missouri 63166 Ms. Mary Ellen Salava Route 1, Box 56 Mr. Tom Vandel Burlington, Kansas 66839 Resident Inspector / Wolf Creek NPS c/o USNRC Eric A. Eisen, Esq.

P. O. Box 1407 Birch, Horton, Bittner & Monroe Emporik, Ka cas 65801 1140 Connecticut Avenue, N. W.

Mr. Michael C. Keener Wolf Creek Project Director State Corporation Commission Ms. Wanda Christy State of Kansas 515 N. 1st Street Fourth Floor, State Office Building Burlington, Kansas 66839 Topeka, Kansas 66612 W

l_ 'l_

enc L oSURS j NRC - SNUPPS MEETING June 9-12. 1981 AGENDA 4

I.

SNUPPS Introduction R.L. Stright II.

MRC Introduction R.J. Bosnak III. Bechtel/ Westinghouse Division of Responsibility C.M. Herbst Comparison to ASB and ME8 Criteria 1

High Energy Line Break Analysts 1

IV.

Auxiliary Systems Branch Questions W.T. LeFave j

V.

Optional Discussion Items

a. stress analysis summary

1. Class I analysts

c. restraint design j

d.model tour i

f VI.

MEB SER Open items (see Ifst attached)

VII. Summary and Conclusions R.L. Stright i

4

~

4 M

ME8 SER REVIEW MEETING AGENDA ITEMS

,4 _

j, g.

e item Reference Summary

Response

comments / notes 1

• I.

1. 1 Page 3.2-2 Non safety-related itene that must C. Herbst j

retain structural integrity

1. 2 3.6.1.1.h.2(b)

Failums of seismic hnd non-seismic C. Herbst PI ing P

i I

1. 3 3.6.1.1.j Pipe Whip effects C. Herbst i

i

1. 4 3.6.1.1.k Line mstrictions in pipe break C. Herbst analysis i

l f5 Fig. 3.6-1 Pipe break analysis figums M. Kalyanan j

1. 6 3.6.2 Breaks in non-seismic Category I C. Herbst P1 ing P

1 j

1. 7 Table 3.6-3 Calrification of FSAR Table N. Kalyanam ga Table 3.6-3 Update FSAR Table N. Kalyanam i

1. 9 Table 3.6-4 Update FSAR Table C. Herbst s

,s Itco Raference

$ranary Rasp:Oso ccamezts/ notes l

1. 10 3.6 Jet 1spingement effects on instru-C. Herbst mentat*on i

1. 11 Page3.7(B)-7 Use of BP-TOP-1 N, Kalyanam i

i

.#12 Fig.3.7(8)-5 Conservatism of response spectra E. Thomas thru 3.7(B)-8

.#13 Fig.3.7(B)-9A Conservatism of msponse spectra E. Thomas and3.7(B)-9D i

j i

1. 14 Page3.7(M)-14 Regulatory Guide 1.92

8. Maumr

[

l

1. 15 3.9(B).1.1 Transients considered in design of N. Kalyanam BOP components l

i

1. 16 3.9(N) i Thermal shock of RPV intemals S. Boyle l

i

1. 17 3.0(B).l.2.1.1 Verification of computer program N. Kalyanam l

1. 18 3.9(B)1.3.2 Inelastic methods in stress analysis N, Kalyanam

MEB SER REVIEW MEETING AGENDA ITEMS-

_c__j,

_p__

i Item Reference Summary

Response

comments / notes r

1. 19 3.9(B).2.1 Thennal expansion and dynamic D. Egan l

effects testing A. Passwater G. Rathbun i

1 l

1. 20 3.9(8).2.1 Thermal expansion and @namic C. Herbst i

effects testing

1. 21 3.9 Functional capabtlity of ASE Class B. Maurer

]

1, 2 and 3 piping systess N. Kalyanas i

1. 22 3.9(B).3.3.1.g Dynamic' load factor less than 2.0 N. Kalyanam

}

.#23 3.9(N).2.1 Vibrstion and dynamic effects B. Maurer j

testing

1. 24 3,9(N).2.4 Regulatory Guide 1.20 S. Boyle l

j

~ -~---

3.9(N).3~~~

--3Pecific paragraphs of ASM

8. Maurer- - -- - - - -

1. 25 Section lit t

1. 26 3.6.2.1.1.9 Pipe break criteria N. Kalyanam 4

1. 27 3.6.2.1.1.e Weided attachments on piping in N. Kalyanam containment penetration areas i

"EL-MR *EMEhLEEE.TlNG=AGEEGA=LTEMI._

' f ---.

a Item Reference Summary

Response

comments / notes

.. ~..

I f

1. 28 Tables 3.9(M)-2 Load conbination

8. Maurer j

and3.9(N)-4

1. 29 3.9.6 Isolation of RCS from low pressure C. Herbst systems C. Hultman

1. 30 3.9 Pump and valve test program C. Hultman

1. 31 3.9 Pre-service examination and H. Borda i

pre-operational testing of C. Hultman j

snubbers l

I i

1 i

.l I

i 3

1 9

s ENCLOSURE 2 NRC Meeting at Bechtel Offices in Gaithersburg, MD.

on SNUPPS FSAR For Callaway Unit I and Wolf Creek Unit 1 Nuclear Plants June 9,1981 NAME ORGANIZATION 1.

R. L. Stright SNUPPS STAFF 2.

J. M. Small BECHTEL 3.

P. A. Ward BECHTEL

_4.

John S. Prebula BECHTEL 5.

Kathy Miller BECHTEL 6.

William Poppe BCCHTEL 7.

Charles Herbst BECHTEL 8.

N. P. Goel BECHTEL 9.

John Hurd BECHTEL

10. Bhupesh Shah BECHTEL 11.

N. Kalyanam BECHTEL

12. Rena Lee BECHTEL

13. Hector E. Borda BECHTEL

14. Joseph H. Smith BECHTEL

15. Jim Alzheimer PNL for NRC

16. Godon Beeman PNL/NRC

17. David Terao NRC/MEB 18.

H. L. Branner NRC/MEB

19. Bob Bosnak NRC/MEB 20.

G. E. Edison NRC/DL 21.

Y. L. Li NRC/MEB 22.

R. A. Jaross ANL/NRC 23.

W. T. Le Fave ASB/NRC 24.

A. C. Passwater UE 25.

G. P. Rathbun KG8E 26.

C. W. Hultman SkUPPS 27.

D. W. Capone UE PART-TIME 28.

B. L. Meyers

'BECHTEL 29.

M. Stuchfield

.BECHTEL 30.

W. L. Luce WESTINGHOUSE 31.

J. J. Mc Inerney WESTINGHOUSE 32.

B. Maurer WESTINGHOUSE 33.

E. Thomas BECHTEL 34.

K. Lee BECHTEL 35.

N. Singleton WESTINGHOUSE

36. Deo Ray Bliandum WESTINGHOUSE 37.

R. W. Beer WESTINGHOUSE 38.

L. S. Shockling WESTINGHOUSE

s 4 SNUPPS MEETING ATTENDEES June 10,1981 NAME ORGANIZATION

1. - R. L. Stright SNUPPS STAFF 2.

P. A. Ward BECHTEL 3.

J. S. Prebula BECHTEL 4.

W. A. Poppe BECHTEL 5.

G. E. Edison NRC/DL 6.

K. A. Miller BECHTEL 7.

C. M. Herbst BECHTEL 8.

W. P. GOEL BECHTEL

'9.

Bhupesh Shah BECHTEL 10.

N. Kalyanam BECHTEL

11. Rena Lee BECHTEL 12.

H. E. Borda BECHTEL

.13.

J. H. Smith-BECHTEL

14. Jim Alzheimer PNL for NRC

15. Gordon Beeman PNL/NRC

16. David Terao NRC/MEB 17.

H. L. Brammer NRC/MEB 18.

R. J. Bosnak NRC/MEB 19.

Y. L. Li NRC/MEB 20.

F. C. Cherny NRC/MEB 21.

B. F. Maurer W-SMD 22.

W. L. Luce W-Licensing

23. J. J. Mc Inerney W-Licensing 24.

A. C. Passwater UE 25.

G. P. Rathbun KG&E 26.

C. W. Hultman SNUPPS 27.

B. L. Meyers BECHTEL 28.

D. W. Capone UE

'i u

O

7.-

y b

J 4'

+

u 5

ENCLOSURE 3 p'

4' Callaway and Wolf Creek Plants Applican'ts' Responses-to_NRC Draft Questions as Agreed with

.NRC (MEB) at Review Meetings June 9 and 10, 1981 d

i' i

f.

l:

l r-f '.

I l :-

t I

-.-, -,,..,. -.... ~...

SNUPPS

1. 1.

Page 3.2-2 "Nonsafety-related structures, systems, and components that must be designed to retain structural integrity during and after an SSE, but do not have a function, are seismicallv analyzed."

Assurance should be made that the above items meet the faulted limits.

It is also stated that these above items are not controlled by a 10 CFR 50 Appendix B Quality Assurance Program.

These items should be inclu ed in the Qu lity Assurance Programes & f y W hb U$

W-

RESPONSE

See revised Page 3.2-2.

ee

SNUPPS All components classified as Safety Class 1, 2, or 3 (classi-fications are as defined by Reference 1), are seismic Category I.

Seismic Category I structures, components, and systems are designed to withstand the safe shutdown earthquake (SSE), as discussed in Sections 3.7(B) and 3.7(N), and other applicable load combinations, as discussed in Sections 3'.8.1 through 3.8.5.

Seismic Category I structures are sufficiently isolated or protected from the other structures to ensure that their integrity is maintained.

Radwaste systems and structures are designated as nonseismic Category I.

In accordance with Regulatory Guide 1.143, a simplified seismic analysis is performed for portiens of the gaseous radwaste system (which by design are intended to store and delay the release of gaseous radioactive waste), including isolation valves, equipment, interconnecting piping, and components located between the upstream and downstream valves used to isolate these components from the rest of the system.

In addition, a simplified seismic analysis is performed for structures housing radioactive waste management systems in cccordance with Regulatory Guide 1.143.

Nonsafety-related structures, systems, and component:s that must be designed to retain structural integrity during and after an SSE, but do not have to function, are seismically l

cnalyzed to ensure that faulted stress limits are not exceeded.

I These items (for example:

piping and piping supports for nonsafety-related piping located over safety-related items) whose continued function is not required are nonseismic Category I and are not controlled by a 10 CFR 50 Appendix B Quality Assurance Program (not Q-listed).

The nonseismic Category I Systems Quality Assurance Program is described in Section 17.D of the SNUPPS Quality Assurance Programs for Design and Construction.

3.2.2. SYSTEM QUALITY GROUP CLASSIFICATION The quality group classification for each water 'and steam-containing pressure component is shown in Table 3.2-1. The i

l components are clas sified according to their importance to safety, as dictated by service and functional requirements and by the consequences of their failure.

The quality group classifications and code requirements for the quality of plant process systems meet the intent of Regulatory Guides 1.26 and 1.143.

Clarifications and specific exceptions to these guides are. discussed in Tables 3.2-4 and 3.2-5, respectively.

These tables compare the design to each regulatory position.

The design, fabrication, inspection, and testing requirements of each classification provide the required degree of conser-vatism in assuring component pressure integrity and oper-ebility.

3.2-2

o SNUPPS

1. 2.

Section 3.6.1.1.h.2(b), Page 3.6-3 It is stated that it was assumed the failure of seismic Category I and seismically supported nonseismic category I piping was caused by some mechanism other than an earthquake and, therefore, that nonseismic Category I equip-ment could be used to bring the plant to a safe shutdown.

What mechanisms are postulated for failure of seismic Category I and seismi-i cally supported nonseismic Category I piping?

Assurance must be made that the failed seismic piping does not damage the nonseismic Category I equipment mentioned above.

Assurance must also be made that only seismic Category I equipment will be used to bring the plant to a safe shutdown in the event of an SSE.

FESPONSE Seismic Category I and seismically supported nonseismic Cate-gory I piping systems are assumed to fail nonmechanistically for the purpose of pipe break hazards analysis.

cou/A

- hnseismic Category I equipment 33 be utilized to bring the plant to safe shutdown following a postulated pipebreak event, since a seismic event is not assumed to occur simultaneously with a pipebreak.

i As stated in Section 3.6.1.1.h.2(b), only seismic Category I equipment is assumed to be available to bring the plant to a safe shutdown following an SSE.

e

~..

SNUPPS I

1. 3.

Section 3.6.1.1.j, Page 3.6-4 It is stated that the pipe whip was assumed to occur in the plane defined by the piping geometry and to cause movement in the direc-tion of the jet reaction.

Assurances must be made that this criteria was used only in the design of pipe whip restraints and that failed piping was considered capable of swinging in any direction about a plastic hinge following a pipe rupture and all potential targets were considered.

RESPONSE

Jet impingement targets are identified in accordance with Standard Review Plan 3.6.2, based on the evaluated movement of the pipe. Pipe whip restraints are provided wherever postulated pipe breaks have any possibility of affecting any system or component required for the mitigation of that break or safe shutdown of the plant. Unrestrained pipe breaks are limited only to those areas of the plant that are physically separated from the systems and components required for pipe break mitigation or safe shutdown. In general, whipping ends from a pipebreak are restrained, such that plastic hinge formation is not allowed to occur. Where equipment, piping systems, raceways, etc. were considered to be the targets of pipeoreak fluids, an evaluation of the hazard is performed on an individual case basis. (i f A R Q4 3 6 d 34 w_cC 4 M uob 7b I

W Mi ek j +G du mp ~>.

SNUPPS

1. 4.

Section 3.6.1.1.k All instances where line restrictions or the absence of energy reservoirs were used in the calculation of thrust and jet impingement forces should be listed.

RESPONSE

l The analysis of the effects of each pipebreak event is described

'in the Pipebreak Hazards Protection Analysis.

Line restrictions cnd limited reservoirs have been considered !.n such ccses where they exist.

A summary of the results of this analysis is being included in an FSAR revision which will be submitted in July 1981.

i l

l 1

l 1

l 1

i l

i l

l.

i

)

l l

i t

-__...r,.,

i SNUPPS

1. 5.

Figure 3.6-1 Various sheets indicate that pipe break restraint locations, Class 1 analysis pipe break loca-tions, and effects analysis for high-energy pipe breaks located within containment are all

.under review.

We cannot complete our review until these reviews are completed.

RESPONSE

Figure 3.6-1 was included in the initial submittal of the SNUPPS FSAR in October 1979 to report on the results of piping eystems stress analysis.

Since then, the stress analyses have been updated to include changes in the input information that

'resulted from refinements in the design.

In addition, the results of some analyses that were not complete at the time of the FSAR submittal have since been completed.

All of these 1

updates will be reported in revised Figure 3.6-1.

The Pipebreak Hazards' Protection Analysis program will pro-vide revised pipebreak locations and pipebreak restraint locations.

The Pipebreak Hazards Protection Analysis program also provides high-energy pipebreak effect analysis.

The results of this program are scheduled to be submitted in July 1981 in the form of an FSAR revision.

I i

--w.-

y

-.__,__,_,,_,_,y,

SNUPPS

1. 6.

Section 3.6.2 A statement should be made that breaks and 1eakage cracks in nonseismic Category I piping are postulated in worse case locations and that failure of non-seismic Category I piping will not cause failure of seismic equipment.

RESPONSE

As stated in FSAR Section 3.6.2.1.1.d, breaks in non-nuclear piping were postulated in each run or branch run at terminal ends of the runs and at all intermediate fittings (e.g.,

elbows, tees, reducers, welded attachments, and valves),

consistent w2th Standard Review Plan 3.6.2.

As stated in FSAR Section 3.2.1, seismic Category I equipment is protected against the. failure of nonseismic Category I equipment. Leakage cracks in nonseismic Category I piping ( 4 A.O rse xT4 3-s Ac se loc @ations. F fa g are. postulated i u s p t dnw

%w n.

o SNUPPS

1. 7.

Table 3.6-3 What is the difference between Sheet 1 and Sheet 2, Sheet 3 and Sheet 4, sheet 5 and sheet 6, and Sheet 7 and Sheet 8?

RESPONSE, Sheets 1, 2,

3, 4,

5, 6,

7, and 8 list the stress results for problems 001, 001A, 002, 002A, 003, 003A, 004, and 004A, respectively.

i en

+

.,-..,,m,-rr.---,,-

~_.--------_,r-,.,--,-..-.-

+

n SNUPPS

1. 8.

Table 3.6-3 Sheets 28, 32 and 36 indicate that the stress analysis is under review.

We cannot complete our review of Section 3.6.2 until this infor-mation is furnished.

RESPONSE

Sheets 28, 32, and 36 have been updated to incorporate the latest stress analysis results.

The remainder of Table 3.6-3 is being updated to reflect refinements in the piping system stress analysis and will be provided in the form of an FSAR revision in July 1981.

l 1

l l

l l

l '

S'.UFPS 1

TA31.E 3.6-3 (Sheet 28)

Prob. No. P-119 SYSTDi - CHD:ICAL AND V0'_'.':E CONTROL SYSTEM 1ssue - 2 Pipe Break Stress Limit (psi)

Node Stress (psi)

+ 1.2S )

Prizary Secondary Total O.8 (SA h

4553 5,349 13,958 19,307 37,244 47.':

5,017 19,116 24,133 37,244 49 18,238 15,476 34,395 37,244 60!

15,515 7,989 23,504 37,244 145.:

9,021 19,464 28,485 37,244 s

1603:

10,232 21,632 31,864 37,244 220::

4,066 18,211 22,277 37,244 245E 4,057 19,998 24,0S5 37,244 270*

3,980 3,553 7,533 37,244 Indicates Terminal Erd

l St.Y?PS TABLE 3.6-3 (Sheet 32)

)

CHD:ICAL AND VOLL?tE CO:iTROL Prob. No. P-146 SYSTDi Issue - 2 Pipe Break Node Stress (psi)

Stress Limit (psi)

+ 1.2S )

Prir ary Secondary Total 0.8 (Sg h

5*

5,469 5,201 10,670 37,648 30I 7,194 23,858 31,052 37,648 35; 8,722 13,492 22,214 37,648 40T 8,741 10,360 19,101 37,648 44:

10,526 12,356 22,882 37,648 4S 11,182 8,537 19,719 37,648 807 12,578 6,878 19,456 37,648 1027 12,189 22,635 34,824 37,648 10$

12,263 10,262 22,550 37,648 130!

15,138 6,132 21,270 37,648 202:

11,671 8,697 20,368 37,648 401 ~-

15,9S6 22,849 38,935 37,645 l

315c 7,077 630 7,707 37,648 l

l i

l

• - Indicates Teminal End

S:.T P.'S i

TA RI.E 3.6-3 (Sheet 36)

Q),/

f SY$1 D1 - SiEX: CE::EFAiOR BI O.?DQ:N Prob. No. P-219 Issue Node Pipe Break Strsss (psi)

Stress Limit (psi) -

Prima ry Se conda ry Total 0.8 (SA

+ I 2S )

h 5

4,755 16,362 21,117 37,400 20E 3,055 6,435 9,520 32,400 95 6,559 412 6,971 32,400 170 13,272 9,836 23,108 32,400 175 7,499 5,976 13.475 32,400 175-17,S93 20 S34 38,727 32,400 A70 4,491 33,600 38,091 32,400 A30 3,863 16,605 20,468 32,400 B35 8,?S3 9,143 17,426 32,400 Bio 8,162 11,297 19,459 32,400 C35 9,304 14,945 24,149 32,400 ASD 4,765 1,877 6,642 32,400 A63 7,333 1,011 8,344 32,400 192T 6,677 2S,319 34,996 32,400

< ~s 240c 3,614 353 3,997 32,400

(_,)

203 2,783 4,150 6,963 32,400 260 4, SIS 1,323 6,146 32,400 255 5,450 1,233 6,713 32,400 D15 10,S16 5,151 15,967 32,400 E50 7,237 3,362 10,599 32,400 F20 4,414 9,291 13,705 32,400 D15 S,653 5,076 13,729 32,400 F25-5,665 19,607 25,475 32,400 C4 0 -

12,973 32,495 45,471 32,400

- Indic a tes T( r.fiial End 2

m.

... _. ~

SNUPPS

1. 9.

Table 3.6-4 Data in this table under effects analysis are listed as (under review).

We cannot complete our review of Section 3.6.2 until this infor-mation is furnished.

RESPONSE

High-energy pipebreak effects analysis results are being

-completed room-by-room as part of the pipebreak analysis.

The results of the analysis are being included in an FSAR revision which will be submitted in July 1981.

4

• e

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

t i

{

f i

L

• a dem/

h&

P =!!

k nam x ~

a 1,:, + s p r u m, - n a

a,,, w r J.

/8MNs(1 W& A WfWW

,,J g A f & A r z A n u =<= b b& Ay fH kW mgmamp a+

sg/ w g m dL & sa

&%Amw~~na 46 w%#A'P~e d 4 a-y & y d h k A N. 1A~ 4 A W sf U Q W

~~w.a Aa~4ypgarnu I

M&M.

l

\\

SNUPPS If.

Page 3.7(B)-7 Reference is made to FSAR Section 3.7(B).2.7 which references Sections 5.1 and 5.2 of BP-TOP-1 for the criteria used for combining modal responses for piping systems:

The last sentence in Section 5.2 of BP-TOP-1 (page 14) includes the words "if they do occur in-phase" with regard to when the grouping method or the double sum rethod will be used for closely spaced modes.

Please indicate how closely spaced modes were determined to " occur in-phase" and give an example of when they were deter-mined not to occur in-phase.

RESPONSE

For piping systems, closely spaced modes were determined per NRC Regulatory Guide 1.92, Equation 4.

FSAR Section 3.7(B).2.7 has been revised to incorporate this statement.

Also the sentence, " Sections 5.1 and 5.2 of BP-TOP-1 describe the criteria used for piping systems" has been eliminated from Section 3.7(B).2.7.

1 l

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

SNUPPS Models, typically shown in Figure 3.7(B)-13, were used to perform soil-structure interaction analyses for all four sites.

For each site, the site dependent soil properties were used.

The vertical dimension of each soil element is squal to or less than Cs/5f, where Cs is the lowest soil olement shear wave velocity reached during iterations and f is the highest frequency of interest to be transmitted through the soil profile.

The highest frequency used was 25 Hz.

In_the analyses for the same buildings with site dependent soil parameters, the structural elements remained unchanged.

The site dependent soil properties consisted of strain dependent damping and modulus relationships fer each material.

In general, the soil properties are nonlinear in character.

An iterative process was used to obtain equivalent linear properties which are strain dependent.

The methods generally used for such an analysis are included in the computer program FLUSH.

3i7(B).2.5 Development of Floor Response Spectra Acceleration time-histories obtained from the FLUSH finite element analyses were used in computing the floor response epectra for the major seismic Category I structures.

The i

spectra were generated following the procedures outlined in Section 5.2 of BC-TCP-4-A, using the SPECTRA computer program (see subparagraph 3.8A.12).

3.7(B).2.6 Three Components of Earthquake Motion Procedures for considering the three components of earthquake motion in determining the seismic response of structures, syster.s, and components follow the recommendations of Regulatory Guide 1.92 and are described in Section 4.3 of BC-TOP-4-A and Section 5.1 of BP-TOP-1.

3.7(B).2.7 Combination of Modal Responses Combination is done according to the criterion o.f "the Square-root-of-the-sum-of-the-squares" (SRSS ).

Section 4.2.1 of BC-TOP-4-A describes the techniques used to combine modal responses for structures and equipment.

For piping' systems, closely spaced modes were determined per NRC Regulatory Guide 1.92, Equation 4.

3.7(B).2.7.1 Significant Dynamic Response Modes The static load equivalent or static analysis method involves the multiplication of the total weight of the equipment or

. component member by the specified reismic acceleration.

Multiple degree-of-freedom systems which may have had frequencies in the resonance region of the amplified response spectra curves 3.7(B)-7

SNUPPS ll$.

Figures 3.7(B)-5 through 3.7(B)-8 The response spectra of the synthetic time-history does not envelope the corresponding design spectra for all frequencies.

Please explain this apparent non-conservatism.

RESPONSE

Figures 3.7(B)-5 through 3.7(B)-8 are consistent with Figures 2-13, 2-14, 2-17, and 2-18 of BC-TOP-4-A, Rev. 3, " Seismic Analysis of Structures and Equipment for Nuclear Power Plants."

These figures will be deleted from the FSAR, and reference will be made to BC-TOF-4-A, Rev. 3, when referring to these figures.

1 i

e

>.. - ~=.,.

3.em

--,-,--e--e-u-.

,--i

,r-.-

-.-m--

v

SNUPPS as-60 percent of the SSE.

The values shown are for the site with maximum amplification.

Section 2.5.2 of each Site Addendum and Section 2.5 of BC-TOP-4-A (Ref. 3 ) discuss the effects of focal and epicentral distances from the site, depths between the focus of the seismic disturbances and the site,-existing earthquake records, and the associated ampli-fication of the response spectra.

Earthquake duration influences only the number of loading cycles on equipment because the equipment is designed for the elastic range in accordance with the analytical proce-dures outlined in BC-TOP-4-A.

A 20.48-second duration is considered.to be adequate for the time-history type of analysis used for the structures and equipment.

The design response spectra and earthquake time-histories are applied in the free field at finished grade for all sites, except the Tyrone site where the design response spectra and earthquake time-histories are conservatively applied at top of rock below grade.

For differences between subsurface conditions at the Tyrone site and those at the

.other three sites, see Figures 3.7(B)-11A and B.

3.7(B).l.l.1 Bases for Site Dependent Analysis Section 2.5.2 of each Site Addendum and BC-TOP-4-A, Sections 2.4 and 2.5, describe the bases for specifying the vibratory ground motion for design use.

3.7(B).1.2 Design Time Historv synthetic earthquake time-histories were generated because the response spectra of recorded earthquake motions do not necessarily envelope any of the sites' design spectra.

Figures 3.7(B)-3 and 3.7(B)-4 show the synthetic earthquake

-time-history motions in the horizontal and vertical direc-tions, respectively.

The time-histories shown were truncated j

to 20.48 seconds for use in the FLUSH finite element an,alyses discussed in Section 3.7(B).2.4.2.

Figures 2-13, 2-14, 2-17, and 2-18 of BC-TOP-4-A show that the response spectra of the synthetic time-histories for the horizontal and vertical directions envelope the corresponding design spectra l

for 1 percent, 2 percent, 5 percent, 7 percent, and 10 percent damping.

Section 2.5.1 of BC-TOP-4-A describes the generation of a typical synthetic earthquake time-history.

Typical foundation-level, free-field acceleration response spec'tra for each of the four sites are presented in Figures 3.7(B)-9A through D.

Their envelope is presented in Figure 3.7(B)-10.

All curves overlay the SNUPPS 60-percent design response spectra.

3.7(B)-2

o SNUPPS

-Figures 3.7(B)-5 through 3.7(B)-8 are to be deleted from the Standard Plant FSAR.

• e

SNUPPS W I)

Figures 3.7(B)-9A and 3.7(B)-9D Please explain the significance and conserva-tism of these figures.

RESPONSE

Figures 3.7(B)-9A and 3.7(B)-9D present horizontal SSE free field accelerat'on response spectia computed at the bottom of the auxiliary control building for the callaway and Wolf Creek sites, espectively.

In order to demonstrate the con-cervatism of the seismic input used, the free field spectra cre compared in these figures with the SNUPPS 60-percent horizontal design response spectrum.

Since the free field cpectra for the most deeply embedded power block foundation cre considered, the figures represent worst-case comparisons for all the power block structures.

The design of all power block structures, systems, and components is based on the envelope of responses for multiple sites.

This procedure leads to an enveloping of the 60-percent design spectrum.

Consequently, the seism a input used is conservative with d

. respect to the 60-percent design response spectrum criterion in.all cases.

l l'

i I

SNUPPS

1. /j[

Page 3.7(N)-14 Equation [3.7(N)-29] is not necessarily con-servative with respect to the requirements of Reg. Guide 1.92.

Provide justification for its acceptability.

Equation [3.7(N)-30] is in error.

RESPONSE

The method used by Westinghouse for the combination of closely spaced modes has been accepted previously by the and numerous plant dockets)

NRC (i.e.,RESAR-41, RESAR-414, as an acceptable alternative to the recommendations 7f Regulatory Guide 1.92. The Mechanical Engineering Branch that, on will notify the Strnctural Engineering Branch this basis, the item is considered closed.

Additionally, FSAR Equation 3.7(N)-30 will be revised in a future revision to correct an editorial error.

l D

..,.---,-,y-.,,---..-w,

,s-,

,,---.--...-.,w-----

m i

i SNUPPS W /5' Section 3.9(B).1.1, Page 3.9(B)-1 Reference is made to section 3.9(N).1.1.

Section 3.9(N).1.1 discusses the transients considered in the design of the reactor coolant system (RCS), RCS component supports, and reactor intervals.

Are these the sa:ae transients used in the design of the BOP components?

RESPONSE

Class 1 branch piping and components are designed and analyzed using the design transients used to analyze the RCS, RCS component supports, and reactor internals as described in Section 3.9(N).1.1.

Class 2 and 3 and aca-Sectica III BOP piping systems and components do~not require thermal transient analysis. Class 2 and 3 piping systems and componehts are designed and analyzed for dynamic transients,,

includ4a; #::t 'fe!ve-ciciwm.c as identified in Section 3.9(B).2, in accordance with Section III of the ASME Code for normal, upset, and faulted conditions.

Section 3.9(B).1.1 and 3.9(B).2.1.a have been revised accordingly.

O t

l l

I

SNUPPS 3.9(B)

MECHANICAL SYSTEMS AND COMPONENTS 3.9(B).1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS 3.9(B).l.1 Design Transients Refer to Section 3.9(N).l.1 for a description of the operating I

conditions considered in the design of the RCS, RCS component supports, and reactor internals.

Class 1 piping systems are designed and analyzed using design transients that are compat-ible with those described in Section 3.9(N).1.1.

Class 2 and,3 a"d --- rectier-; : piping systems and components do not requireItransient analysis. '-

N i

%.c f;giab..S b :y' &

3.9(B).l.2 Computer Programs Used in Analyses can Z - A 3 des

.-t si p.s

%.iQN,Q dgagiy For NSS systems, refer to Section 3.9(N).l.2.

i, 7%%A

--C' e5u4 3MO.y 3.9(B).l.2.1 Seismic Category I Items Other Than t e NSSS Table 3.9(B)-1 lists computer programs used in the balance-of-plant system components.

The verification of programs is as follows:

3.9(B).l.2.1.1 ME-632 Program The ME-632 program is used to determine stresses and loads due to thermal expansion, deadweight, and transient force func-tions such as those created by fast relief valve opening and closing, pipe break, or fast activation of high-capacity pumps (water hammer effects).

The results obtained from pipe stress program ME-632 have been compared with a) ASME 3enchmark problem results, b) Pipe Stress Program TPIPE, c) general purpose ptogram ANSYS, and d) long-hand calculations.

The comparison of the results are given in the verification report of the ME-632 program (Ref. 3 ).

A description of this computer code.*

ncluded in Table 3.9(B)-1.

l Appendix 3.9(B)A provides a verification report for the ME-632 program.

3.9(B).l.2.1.2 ME-101 and TPIPE Programs The :ME-101 and TPIPE computer programs are used to determine stresses and loads due to restrained thermal expansion, dead-weight, seismic anchor movement, and earthquake in the folloaing piping:

a.

Seismic Category I ASME Section III Class 1, 2, and 3 piping 2 1/2 inches and larger.

3.9(B)-1

SNUPPS time dependent forcing function, such as fast valve closure, while the second is a constant vibration, usually flow in-duced.

a.

Transient response Dynamic events falling in this category are antici-pated operational occurrences.

The systems are operated in their normal mode (emergency mode fer auxiliary feedwater turbine pump), and measurements are recorded on the systems during and following the event that causes the transient induced vibrations.

The mystems and the associated transients to be included in the preoperational test program to verify the piping system are:

i 1.

Main steam (a)

Main steam turbine stop valve trip (b)

Main steam atmospheric dump valves opening (c)

Main steam condenser dump valvos opening 2.

Pressurizer power-operated relief valve blowdown 3.

Auxiliary feedwater pump turbine stop valve trip Selected snubbers subjected to the above transients are monitored during this preoperational testing to asnure proper snubber operation.

A' \\ A h eb.v u.pde b,% M

o.

• 4 a time l

dr. pendent dynamic analysis is performed,7n the sys-l tem.

Tha stresses thus obtained are combP.ed with system stresses resulting from other operating condi-l tions in accordance with the criteria 1.ovided in l

Table 3.9(B)-2.

b.

Steady state vibration System vibration re7ulting from flow disturbances falls into this category.

Positive displacement pumps may cause such flow variation and vibrations and, as such, will be reviewed.

Such systems will be I

checked, including the charging systems.

9 Sir.ce the exact nature of the flow disturbance is not known prior to pump operation, no analysis is per-formed.

A visual steady state vibration inspection 3.9(B)-5

SNUPPS

' 'N

1. 16 The thermal shock in the RPV internals due to an ECCS injection following a design basis LOCA should be addressed.

I g PoWsr As shown in Table 3.9(N)-1 a LOCA is defined as a faulted design condition. Since a LOCA is accompanied by a,q ECCS injection, the thermal shock from this injection y included in the evaluation of the LOCA transient for the reactor internals. Additionally, other upset and emergency condition thermal transients, such as inadvertent safety injection, are included in the evaluation of reactor internals.

Stresses due to thermal shock following an ECCS injection have been evaluated and shown to satisfy the requirements of the ASME Code, Appendix F, as defined in the SNUPPS FSAR. In summary, peak stresses in the reactor internals due to thermal shock do not cause any loss of function.

A* N1U FSAR Section 3.9.5 will be revised to addrces thermal shock from an CS injecti n following a LOCA for reactor internals M 4

S S

e SNUPPS

1. IT Section 3.9(B).1.2.1.1, Page 3.9(B)-1 This section references Appendix 3.9(B)A which states that ME-632 results were compared with the results of the previously approved Engi-nearing Data System (EDS) computer programs.

Where is a discussion of the verification of the EDS programs and when was it approved?

RESPONSE

.FSAR Appendix.3.9(B)A, Page 3.9(B)A-1 has been revised to eliminate "previously approved."

Also, Section 3.9(B).1.2.1.1, Page 3.9(E)-1 and Section 3.9(B).7, Pa e 3.9(B)-20 have been J/g)g revised. pS Stz kQx

,9,3 m) G_

s c-elou dd 7 it d c.L Te g - f a 4 ; T A <, d 2'i c &d. $ %c-AJ{ L '

un.

l l

l l

SNUPPS l

APPENDIX 3.9(B)A ME-632 VERIFICATION REPORT The following is a comparison of the ME-632 program lesults with the results of the Engineering Data System computer j

program.

The two piping systems chosen for stress checks were:

a.

The Core Spray Piping System - Monticello Nuclear Generating Plant Unit 1 b.

Lines 48223-18-HE, 50056 1.0-HE, and 50057-10-HE-SMUD Rancho Seco Unit 1 These two test cases were chosen because independent piping stress analyses performed by Engineering Data Systems (EDS) under contract to Bechtel were available for comparison purposes.

The EDS (PISOL 3) analysis of the core spray piping system consisted of both deadweight and thermal loading while the SMUD Rancho Sec o piping system was an earthquake response spectrum analysis.

The ME-632 piping stress analyses were performed in the September 18-20, 1972 period on PICC's Honeywell 635 com-puter.

A relocatable binary deck of the program is stored on tape No. 8312 and will be retained indefinitely for documentation purposes.

A comparison of the ME-632 and EDS analyses is shown in Table 3.9(B)A-1.

Due to differing sign conventions, the reactions have opposite signs.

The EDS program prints the effects of the support on the piping system while ME-632 prints the effect of the piping system on the support.

In come cases, the maximum values for the ME-632. analysis, i

occurred at the middle of the bend.

However, since the EDS program does not compute output quantities at.the middle of a bend, these maximums are not shown in Table 1.

The maxi-mums shown in the table occurred at the same physical point on the piping system in both analyses.

In all cases, the maximum difference in output quantities was less than 5 percent, based upon the corresponding peak value for the particular load case.

It is, therefore, concluded that ME-632 correctly performs static and thermal analysis of piping systems, consistent with the assumptionc of the elastic beam theory and appli-cable flexibility and stress intensification factors speci-fied in ASME Section III.

3.9(B)A-1

SNUPPS 3.9(B)

MECHANICAL SYSTEMS AND COMPONENTS 3.9(B).1 SPECIAL TOPICS FOR MECHANICAL COMPONENTS 3.9(B).l.1 Design Transients Refer to Section 3.9(N).1.1 for a description of the operating conditions considered in the design of the RCS, RCS component supports, and reactor internals.

Class 1 piping systems are

. designed and analyzed using design transients that are compat-ible with those described in Section 3.9(N).l.l.

Class 2 and 3 and non-Section III piping systems and components do not require transient analysis.

3.9(B).l.2 Computer Programs Used in Analyses For NSS systems, refer to Section 3.9(N).l.2.

3.9(B).l.2.1 Seismic Category I Items Other Than the NSSS Table 3.9(B)-1 lists computer programs used in the balance-of-plant system components.

The verification of programs is as follows:

3. 9 ( B ). l. 2.1.1 ME-632 Program The ME-632 program is used to determine stresses and loads due to thermal expansion, deadweight, and transient force func-tions such as those created by fast relief valve opening and closing, pipe break, or fast activation of high-capacity pumps (water hammer effects).

The results obtained from pipe stress program ME-632 have been compared with a) ASME Benchmark problem results, b) Pipe Stress Program TPIPE, c) general purpose program ANSYS, and l

d) long-hand calculations.

The comparison of the results are l

given in the verification report of the ME-632, program, (Ref. 3).

i l

A description of this computer code is included in Table 3.9(B)-1.

Appendix 3.9(B)A provides a verification report for the ME-632 program.

_3.9(B).l.2.1.2 ME-101 and TPIPE Programs The hE-101 and TPIPE computer programs are used to determine stresses and loads due to restrained thermal expansion l dead-weight, seismic anchor movement, and earthquake in the following j

piping:

I j

a.

Seismic Category I ASME Section III Class 1, 2, and 3 l

piping 2 1/2 inches and larger.

1 3.9(B)-1

SNUPPS b.

Seismic Category I ASME Section III Class 2 and 3 piping 2 inches and smaller that cannot be analyzed per ME-602.

Piping Systems. ANSI B31.1 Power Piping Included in High Energy c.

h' l

3.9(B)-la

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

__4 SNUPPS 3.9(B).7 REFERENCES 1.

" Program ME-101 and ME-632 Seismic Analysis of Piping Systems, Users Manual," Pacific International Computing Corp., March, 1971.

2.

BP-TOP-1, Seismic Analysis of Piping Systems, Bechtel Power Corporation, San Francisco, California, Rev. 3, January, 1976.

3.

" Seismic Analysis of Piping Systems Program ME-632 I

Verification Report," Version BIO, Bechtel Power Corporation.

l l

~

i 1

3.9(B)-20

SNUPPS

1. 18 Section 3.9(B).1.3.2, Page 3.9(B)-3 It is indicated that inelastic methods are not used in the design of Code or non-Code compo-nents for the faulted condition.

On Page 3.9(B)-4, Section 3.9(B).1.4.2 it is indicated that inelastic analyses were used.

Please clear up the discrepancy.

~

RESPONSE

FSAR Section 3.9(B).1.3.2, Page 3.9(B)-3 has been revised to eliminate " inelastic method or."

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

e SNUPPS The program is based on Welding Research Council Bulletin 107, August 1965.

The program has been verified based upon hand calculations.

3.9(B).l.2.1.6 CE901 ICES /STRUDL-II The_ ICES /STRUDL-II code is used in the design of component supports.

For ASME Section III Class 1 piping support design, the program is used to obtain stiffness properties of the support.

The results of the analyses are incorporated into overall reactor vessel internal models which calculate the dynamic response due to seismic and LOCA conditions and yield dynamic stres.ses.

In the design of ASME Section III Class 2 and 3 piping. supports, models of certain indeterminate support designs are programmed in order to obtain support loads and stresses.

A description and validation of this program are included in

.Section 3.8A.l.10 of Appendix 3.8A.

3.9(B).l.2.1.7 CE800 (BSAP), CE802 (SPECTRA), and CE786 These programs were used to determine the seismic response spectra of the NSSS for reactor coolant loop branch piping analysis, stresses, and displacements of the main feedwater and main steam system in the reactor building, and to deter-mine seismic anchor movements of the NSSS for incorporation into the piping analysis.

A description and validation of these programs are included in Sections 3.8A.l.5, 3.8A l.6, and 3.8A.l.8 of Appendix 3.8A.

3.9(B).1.3 Experimental Stress Analysis 3'.9(B).l.3.1 NSS System Refer to Section 3.9(N).1.3.

3.9(B).1.3.2 Seismic Category I Items Other Tnan the NSSS Experimental stress analysis methods are not used in the design of Code or non-Code components for the faulted condi-tion.

For code components, the stresses will not exceed the limits of the ASME B and PV Code,Section III.

3.9(B).l.4 Considerations for the Evaluation of the Faulted Condition A listing of all seismic Category I safety-related mechanical systems and components is included in Table 3.2-1.

3.9(B)-3 l

SNUPPS W/@

Section 3.9(B).2.1, Page 3.9(B)-4 More information is needed regarding the piping vibration, thermal expansion and dy-namic effects testing programs.

Please list those systems to be monitored for 1) transient induced vibration, 2) steady state vibration and 3) thermal expansion.

Also list the flow modes of operation to be included in the testing program.

List those locations where visual inspection will be utilized and those loca-tions where measurements will be taken and also the associated acceptance criteria.

A commitment should be included that the NRC will be provided documentation of any correc-tive action resulting from the tests and confirmation by additional testing that sub-stantiates effectiveness of the corrective action.

RESPONSE

To provide more information regarding the testing programs gnd y

modesofoperation,FSARSection3.9(B).2,Pages3.9(B)-5[Aaw, 3.9(B)-6, has been revised.

More specific information concerning the locations where visual inspection or measurements are to be taken are ad-i dressed in the applicable test procedures.

Acceptable cri-t teria for the thermal and dynamic tests are addressed in the applicable FSAR Chapter 14,otest abstracts.

Corrective action for any deficiency identified as a result of the test program till be available for inspection at the site. Retesting will be performed in accordance with administrative controls identified in Chapter 14.0.

l l

t l

SNUPPS time dependent forcing function, such as fast valve closure, while the second is a constant vibration, usually flow in-duced.

a.

Transient response Dynamic events falling in this category are antici-pated operational occurrences.

The systems are operated in their normal mode (emergency mode for auxiliary feedwater turbine pump), and measurements are recorded on the systems during and following the event that causes the transient induced vibrations.

The systems and the associated transients to be included in the preoperational test program to verify the piping system are:

1.

Main steam (a)

Main steam turbine stop valve trip (b)

Main steam atmospheric dump valves opening (c)

Ma,in,s, team condenser dump valves opening g g ' er [ow perated relief valve [f M.;.

3rr 3cMu ia 3P.($) Auxil ary fee wa er pump turbine stop valve trip Selected snubbers subjected to the above transients are monitored during this preoperational testing to assure proper snubber operation.

A\\\\ A W ebov.- es wpsd Mkh

o.

• A a time l

dependent d'ynamic analysis is performed on the sys-tem.

The stresses thus obtained are combined with system stresses resulting from other operating condi-tions in accordance with the criteria proviced in l

Table 3.9(B)-2.

b.

Steady state vibration system vibration resulting from flow disturbances falls into this category.

Positive displacement pumps may cause such flow variation and vibrations and, as such, will be reviewed.

Such systems will be checked, including the charging systems.

5 Since the exact nature of the flow disturbance is not 4

known prior to pump operatio.7, no analysis is per-formed.

A visual steady state vibration inspection 3.9(B)-5

SNUPPS is made during system operation.

Measurements above the following guidelines are recorded:

g

1. g h ank g Frequency 210 Hz

"'A For safety-related systems 20.125 inches Y

^

For nonsafety-related systems 20.25 inche Safety-related systems *and high energy systems 4ill be monitored for steady state vibration for all modes of system operation encountered during the preopera-tional test program defined in FSAR Chapter 14.0.

For specifics of this testing, see FSAR Chapter 14.0 of each site addenda.

W.,

pa.

~,.a u insteam e A1hiS l

r l.

l l

l l

I I

l 3.9(B)-Sa

. ~....

.,7..

~

SNUPPS

'N "*% 3 4

• IO

• ps)

E The acceptance criterion is that the maximum measured amplitude shall not induce a stress in the piping i

system greater than one-half the ent..urance limit,as j

defined in Section III of the ASME Boiler and Pres-8 sure Vessel Code, 1974.

l[

When required, additional restraints are provided to reduce the stresses to below the acceptance criterion levels.

During the thermal expansion test, pipe deflections will be recorded at selected locations.

The system will also be visually monitored for hanger and snubber performance and for piping interferences with structure or other piping.

One complete thermal cycle, i.e.,

cold position to hot position to cold position, will be monitored.

Selected portions of the fo lowing systems will be monitored during their normal mode of operation.

Main steam system Main feedwater system Letdown / charging system Residual heat removal system containment spray system (1) i Emergency core cooling system Auxiliary feedwater system Auxiliary turbine system Steam generator blowdown system-3.9(B).2.2 Seismic Qualification Testing of Safety-Related Mechanical Equipment 3.9(B).2.2.1 Safety-Related Equipment in the NSSS lj Refer to Section 3.9(N).2.2.

i' 3.9(B).2.2.2 Safety-Related Mechanical Equipment Other' i

Than the NSSS The criteria used to decide whether dynamic testing or analy-sis should be used to qualify seismic Category I mechanical equipment are as follows:

a.

Analysis without testing (2)i Design characteristics of the containment spray system do not permit actual testing to monitor thermal expansion of the suction piping from the containment sumps, during the recirculation mode.

Verification of this piping will be attained by its similarity to the RHR suction lines from the RCS hot leg which will be monitored.

3.9(B)-6'

, 7; O

SNUPPS

1. 80 Section 3.9(B).2.1, Page 3.9(B)-5 The applicant should indicate whether the listed systems meet the SRP 3.9.2 requirements with respect to the scope of this program.

RESPONSE

Our modified program satisfies SRP 3.9.2 requirements.

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

1. 21 Provide assurance that the functional capability of all ASME Class 1, 2 and 3 piping systems essential to plant safety is maintained under all designated loading conditions.

RESPONSE

For faulted condition analysis of Class 1 branch piping attached to the reactor' coolant loop, Equation (9) of ASME Section III, Subsection NB-3652 is applied with a stress limit of 3.0 Sm. This criterion provides sufficient assurance that the piping will not collapse or experience gross distortion such the the function of the system

. would be impaired. The basis for this position is described in the Westinghouse **sponse to NRC Question 110.34 on the RESAR-414 appli-cation (Docket NL. STN 50-572), which received a Preliminary Design Approval (PDA) in November 1978.

For Class 2 and 3 piping systems 2-1/2" and larger, the MEB Regulatory Position in " Interim Technical Position - Functional Capability of Passive Piping Components" dated 07/19/78 is met. This has been ver-ified through the use of the Bechtel computer code ME-101 which is described in Section 3.9(B).1.2.1.2 of the FSAR.

For small bore (i.e. 2" normal diameter and smaller) Class 2 and 3 piping, a standard Bechtel program us used to assure that the ASME code requirements are met. The results of the program have been shown to be conservative when compared to the results of ME-101. Since Hi-101 assures the functional capability of lareg bore piping and the standard Bechtel program is conservative when compared to ME-101, the functional capability of small bore piping is assured.

Smtll bore piping is designed in the design office and shown on the SNUPPS model. Therefore, the analyzed design is the actual design installed in the field and the analyses properly consider the final design (i.e. routing, hanger locations, temperature, concentrated j

masses,etc).

l All small bore piping has a Do/t ratio less than 50, which ensur,es stability and no local buckling, i

L I

SNUPPS tl Sk Section 3.9(B).3.3.1.g, Page 3.9(B)-15 Please list all instances when a dynamic load factor of less than 2.0 was used and provide the needed justification.

RESDONSE For all systems analyzed by static methods, a dynamic load factor cf two has been used.

A dynamic load factor was not used for those systems which were analyzed dynamically.

?

I' l.

l l

l l

l k

l

--r g

---r..

w

,e.

-m--,w+.--w.sw w.vm w

-w r-,-.-e.-,,. -, -,-

,,w,,-.

-ww w.---w yw, g,,-v

-,-wr..,, +,,

w.+,-*,,

w--,, -

-w,w-

SNUPPS g w, d.

Where more than one safety or relief valve is installed t' y on the same run pipe, the sequence of valve openings which induce the maximum stresses is considered as required by Regulatory Guide 1.67.

e.

The minimum moments to be used in stress calculations are those specified in ASME Code Case 1569.

f.-

The effects of the valve discharge on piping connected to the valve header are considered.

g.

The reaction forces and moments used in stress cal-t culations include the effects of a dynamic load factor (DLF) or are the maximum instantaneous values obtained from a dynamic time-history analysis.

A dynamic load factor of 2.0,b 49 as required by Recula-tory Guide 1.67, is used p A Ad 4 p

M M,* M.'

3.9(B).3.3.2 Closec ulsenarge A closed discharge system is characterized by piping between the valve and a tank or some other terminal end.

Under steady-state condit) as, there are no net unbalanced forces.

The initial transient response and resulting stresses are determined, using either a time-history computer solution or 1/~T a conservative equivalent static solution.

In calculating

( )

initial transient forces, pressure and momentum terms are included.

If required, water slug effects are also included.

i 3.9(B).3.3.3 Operational Qualification for Active Safety-l Relief Valves t

l Active safety-relief valves are subjected to the following l

shop tests, hydrostatic, seat leak tests, and a static loading equivalent to the SSE applied at the top of the bonnet and pressure at the valve inlet increased until the valve mecha-nism actuates.

Periodic in situ valve ins ection is performed to assure the functional ability of the va vec.

During a seismic event, it is anticipated that the seismic accelerations imposed upon the valve may cause it to open momentarily and discharge under system conditions which otherwise would not result in valve opening.

This is of no real safety or other consequence.

3.9(B).3.4 Component Supports 3.E(B).3.4.1 Supports Furnished with the NSSS Refer to Section 3.9(N).3.4.

p V

3.9(B)-15 Rev. 1 9/80

SNUPPS

1. 24 Section 3.9(N).2.1, Page 3.9(N)-33 Please describe the acceptance limits that will be used for visual inspection of vibra-tion.

How will the stresses associated with the vibration be calculated?

What ASME Code stress and fatigue limits will be used?

What measures will be taken to monitor the thermal movement of the primary loop during heat up to ensure that no restraint to thermal growth is encountered?

RESPONSE

3EE Ts.vis,3 g

3 q (y),,33 i

L

'A5 f 0 -l

& Y.

(C p hA u s %

LJ J A TLT

~ h 21 Y.4gct49

- gu ci

~l q

' a u,/2 y As A s a r ? -

u-

,g ny.n l

l l

t i

SNUPPS C:'

c.

Component support buckling allowable load In the design of component supports, member compres-sive axial loads shall be limited to 0.67 times the critical buckling strength.

If, as a result of more detailed evaluation of the supports, the member com-pressive axial loads can be shown to safely exceed 0.67 times the critical buckling strength for the faulted condition, verification of the support functional adequacy will be documented and submitted to the NRC for review.

The member compressive axial loads will not exceed 0.67 times the critical buckling strength without NRC acceptance.

In no case shall the

'{

compressive load exceed 0.9 times the critical buckling n y strength.

I, Loading combinations and allowable stresses for ASME Code,Section III, Class 1 components and supports are given in Tables 3.9(N)-2 and 3.9(N)-3.

For faulted condition eval-uations, the effects of the SSE and LOCA are combined using S

the square-root-of-the-sum-of-the-squares method.

Justifi-cation for this method of load combination is contained in i

References 4 and 5.

3.9(N).2 DYNAMIC TESTING AND ANALYSIS c

3.9(N).2.1 Preoperational Vibration and Dynamic Effects ky (

!s Testing on Piping A preoperational piping vibrational and dynamics effects l

testing program will be conducted for the reactor coolant loop / supports systems during startup functional testing of the s

l SNUPPS rnits.

The purpose of these tests will be to confirm that the systems have been adequately designed and supported l

for vibration as required by Section III of the ASME Code, paragraphs NB-3622.3.

The tests will include reactor coolant pump starts and trips.

If vibrations are observed which, from visual examination, appear to be excessive, either:

1) an instrumented test program will be conducted and the system reanalyzed to demonstrate that the observed levels do not-smee ASME Code C

--r.d f ' ira ' Mitr to te "Wai, 2) the cause of the vibration will be eliminated, or 3) the support system will be modified to reduce the vibrations.

Particular attention will be provided at those locations where the vibrations are expected to be the largest for the particu-lar transient being studied as per the criteria of the ASME Code as referenced above.

It should be noted that the layout, size, etc., of the reactor l

coolant loop and surge line piping used on SNUPPS is,very similar to that employed in Westinghouse plants now in opera-tion.

The operating experience that has beea obtained from these plants ir.dicates that the reactor coolant loop and surge line piping are adequately designed and supported to minimize s

3.9(N)-33 Rev. 1 9/80

SNUPPS

1. 7[

Section 3.9(N)-2.4, Page 3.9(N)-36 Th9 FSAR should clearly state that the SNUPPS plants are classified as non-Prototype Cate-gory I in accordance with Reg. Guide 1.20.

RESPONSE

A correlation between SNUPPS vibration predictions and prototype testing was discussed in detail. The prototype plant for SNUPPS is Indian Point Unit 2. This plant (Indien Point) was fully instrumented and tested during hot functional and initial startup testing. Data applicable to SNUPPS were also obtained from tests on the Trojan 1 and Sequoyah I plants. The significant differences between SNUPPS and Indian Point internals are the replacement of the annular thermal shield with neutron panels, modifications resulting from the use of 17 X 17 fuel, and the change to the UHI-style inverted top hat upper internals.

FiAR Section 3.9(N).2.3 will be revised to address the correlation between SNUPPS and prototype internals vibration testing.

e I

SNUPPS

1. J7 Table 3.9(N)-3 I

ML.

The appropriate eeste+rve of ASME Section III should be re'erenced for the various compo-nents listed.

I

RESPONSE

j 7he opp / cable subsechens of He MME ubich opp es k C/oss

/ domponerr/s

\\

h l

Cafe are as A//ows:

Vesse/s /RwpsylMt N8 3200 var en sw h/ff!'j N8 3600 exl &pcrh NF 3000 T bt

3. 9 ( N) -3 e;ll b e. r t<lsel he ide.44;()

.s

w. s vuh ole. 7-,g4 ma s,.,, m

\\ evd aqqu u.G he CWs I

cc m 43 g

S TR i

l D

s

-e

.--,,----,w---m

--e

-, - - - - - - - +. - - - -

--,e----,-v,---

,-,,--m.,,,,~,-_-.-,-----------,----.-,------n-ve-,---w-----,--n---r-,-

SNUPPS

- # f4 Section 3.6.2.1.1.9.2(B), Page 3.6-10 The pipe break criteria is not in compliance with SRP 3.6.2 in that the 3.0 S,value should be 2.4 5,.

RESPONSE

Ste. esvise.d Sec%

3. 0,. 2.. l. l. A. 2..

o e

l I

i l

I 1

SNUPPS in order to verify the dasign basis break loca-(

tions in the reactor coolant loop noted therein.

At all postulated circumferential break loca-sg [

tions, the maximum loop piping displacements, as determined by the dynamic RCS analysis or the location of pipe restriants, are such that the separation results in a limited flow area.

Longi-tudinal breaks are assumed to have an opening area equal to one flow area of the pipe.

2.

Pipe breaks are postulated to occur in the following locations in Class 1 piping runs or branch runs outside the primary reactor coolant loops and pressurizer surge line as follows:

t.

(a)

The terminal ends of the piping or branch run.

(b)

Any intermedie*-e locationc between the terminal ends where stresses, calculated using equations (12) and (13) of the ASME B&PV Code,Section III, Subsection NB, exceed 2.4 Sm, where Sm is the design stress intensity, as given in the ASME B&PV Code, and the stress range calcu-3 lated, using equation (10) of the ASME

{

B&PV code, exceeds ec$ Sm.

i A 4r (c)

Any intermediate locations be+. ween ter-minal ends where the cumulative usage factor, derived from the piping fatigue analysis, under the loadings associated with the CBE and operationa] plant condi-tions, exceeds 0.1.

I (d)

Additional locations of maximum stress intensity or cumulative usage factor to assure a minimum of two break locations l

between terminal ends.

A complete discussion of the reactor coolant loop break location 5is provided in Reference 1.

b.

ASME B&PV code,Section III - Class 2 and 3 Piping Within Protective Structures l..

Breaks are postulated to occur at terminal ends, including:

(a)

Piping-pressure vessel or equipment nozzle intersection l1 (b)

High-energy / moderate-energy boundary

' nisrS ecNi~n

• 2 p e p t a nc hcT (c)

Pept I

3.6-10

9 O

SNUPPS

1. $7 Section 3.6.2.1.1.e, Page 3.6-13 Please provide details of all locations where welded attachment's were made to portions of piping covered under this section.

RESPONSE

Welded attachments have :not been used on systems falling in

.this category, i.e. high-energy piping in containment penetra-tion areas.

Location details will be provided if welded attachments are used in the future.

l l

I l

l l

E.

l

SNUP."!

1. 28 Clarify how loads are combined (e.g. absolute sum, SRSS, etc.).

RESPONSE

The metnodology of load combinat ons and applicable stress limits were discussed. In particular, the following items were noted:

1.For primary equipment, primary equipment supports, and Class 1 branch lines, LOCA and SSE were combined by SRSS on a load component basis (the LOCA and SSE forces in the x direction were combined by SRSS, the LOCA and SSE moments in the x direction were combined by SRSS, etc.).

2.For RCL piping, the deadweight moments were added to the LOCA moments prior to the SRSS combination of the LOCA and SSE loads.

An evaluation was performed to show that if the deadweight moments wers added to the SRSS of the LOCA and SSE (per NUREG-0484),

the maximum loop stresses would increase by less than 0.2%.

It was noted that the deadweight moments are approximately two orders of magnitude less than either the LOCA or SSE moments.

3.For Class 2 and 3 equipment, the loads identified in

' Table 3.9(N)-4 are combined by absolute sum.

N I

The FSAR will be clarified as necessary for tne applicability of the load combination and stress limit tables and the FSAR will include the load combination methodology applic g.jf 4

j e',Y.$$5Y5$':"k?bf"N$%'A'$A ca, ww wwu m-p.za m S t L Gau1.n.s c h ut a Rmie nus.

l T la PWp~ fLQ mfJ tic W} tgx t4 aL. p p )n (*3 R ) M N - 5 l F-137' u 9

SNUPPS i 29

1. )h(

3.9.6 j There are several safety systems connected to the reac0or coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure. There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure. In order to protect these systems from RCS pres-sure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure sys-tems. The leak tight integrity of these valvas must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA. Pressure isolation valves are required to be category A or AC per IWV-2000 and to meet the appropriate requirements of IWV-3420 of Sec-tion XI of the ASME Code except as discussed below. Limiting Conditions for Operation (LCO) are required to be added to the technical speci-fications which will require corrective ac-tion; i.e., shutdown or system isolation when the final approved leakag? limits are not met. Also, surveillance requirements, which will state the acceptable leas rate testing fre-quency, shall be provided in the technical specifications. Periodic leak testing of each pressure isola-tion valve is required to be performed at l least once per each refueling. outage, after valve maintenance prior to return to service, l and for systems rated at less than 50% of RCS design pressure each time the valve has moved from its fully closed position unless justifi-cation is given. The testing interval should average to be approximately one year. Leak l testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a 4 refueling outage, maintenance, etc. l The staff's present position on leak rate l limiting conditions for operation must be equal to or less than 1 gallon per minute fur each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the

r a SNUPPS redundant pressure isolation function and give an indication of valve degradation over a finite period of time. Significant increases over this limiting valve would be an indication of valve degradation from one test to another. Leak rates higher than 1 GPM will be considered if the leak rate changes are below 1 GPM above the previous test leak rate or system design precludes measuring 1 GPM with sufficient accuracy. These items will be reviewed on a case by case basis. The Class 1 to Class 2. boundary will be con-sidered the isolation point which must be protected by redundant isolation valves. In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more valves provide isolation, only two of the valves need to be leak tested. Provide a list of all pressure isolation v,alves included in your testing program along with four sets of Piping and Instrument Dia-grams which describe your reactor coolant system pressure isolation valves. Also dis-cuss in detail how your leak testing program will conform to the above staff position.

RESPONSE

FSAR Section 5.2.5.2.1 provides a discussion of those auxil-l iary systems that interface with the reactor coolant system. The reactor cooling system is shown on FSAR Figure 5.1-1. yMM eMMyf' /Ze / l mm mM/ YA M n. TAL & l w 444hy & wp~ p p.

a. ~, ~ s. n a.u -

.au & us M\\ d5o k M' 1m 7<t.a.s swe = s s ems -sag w. is l l

P 4 so crMrA41 ffu lJL y/ & 44pn p p m & ~ %, w 6 9 yy=7~ J A. 1AGw's4 p p A h M A paf M. i l e l e t l l l

i SNUPPS

1. JI TO ALL APPLICANTS:

Due to a long history of problems dealing with inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and components, it is requested that mainte-nance records for snubbers be documented as follows: Pre-service Examination A pre-service examination should be made on all snubbers listed in tables 3.7-4a and 3.7-4b of Standard Technical Specifications 3/4.7.9. This examination should be made after snubber installation but not more than six months prior to initial system pre-opera-tional testing, and should as a minimum verify the following: (1) There are no visible signs of damage or impaired operability as a result of storage, handling, or installation. (2) The snubber location, orientation, posi-tion setting, and configuration (attach-ments, extensions, etc.) are according to design drawings and specifications. (3) Snubbers are not seized, frozen or jammed. (4) Adeq'Jate swing Clearance is provided to allow snubber movement. (5) If applicable, fluid is to the recommended level and is not leaking from the snubber system. (6) Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly. If the period between the initial pre-service examination and initial system pre-operational test exceeds six months due to unexpected .i situations, re-examination of items 1, 4, and 5 shall be performed. Snubbers which are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria.

SNUPPS Pre-Operational Testing During pre-operational testing, snubber ther-mal movements for systems whose operating temperature exceeds 250 F should be verified as follows: (a) During initial system heatup and cooldown at specified temperature intervals for any system which attains operating tem-perature, verify the snubber expected thermal movement. (b) For those systems which do not attain operating temperature, verify via obser-vation and/or calculation that the snubber will accommodate the projected thermal movement. (c) Verify the snubber using clearances at specified heatup and cooldown intervals. Any discrepencies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval. The above described operability program for snubbers should be included and l documented by the pre-service inspection and pre-operational test programs. The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion. This test program should be specified in Chapter 14 of the FSAR. Y The concerns of items 1, 2, 4, and 6 will be satisfied under 79-14 walkdown procedure. Item 3 will be demonstrated prior tc the 79-14 walkdown ins 7%,s / h,4 -

• 4 j g ~&. m "pection.n r ~x EL.v.e4.,4p An aa y-OperationalTesting x

.e ~. P.e Dur'1:ng the thermal expansion test, snubber movements w verified by recording the deflections in the pipes. Also, the system will be visually monitored for snubber performance and for piping interference with structure or other piping for one complete thermal cycle. The cause of any deficiency will be evaluated and corrected accordingly. L nr,. A p-<.4 %. % h " k h ) 7..s..; w ~ :~. a '".< a + w[ v.. ~ :-* W d A t w s _g n.e um ,- s awua. ;

e-SNUPPS

1. 32 What is.the SNUPPS position on Regulatory Guide 1.121 ?

~ RESPONSE. The SNUPPS position on Regulatory Guide 1.121 is contained in FSAR Appendix 3A. Reg Guide 1.121 analyses for the Model F steam generator have not been completed. Upon completion, tech spec limits-will be evaluated and the information added to the Reg Guide 1.121 position. i l-I- l t ~-,.... _.. _ _ , _ _ _. ~,. - _ _ _ _ _ _ _ - _ _, _. - _. - - -. -

SNUPPS

1. 33 Justify not analyzing for the effects of longitudinal pipe breaks in Class I branch lines.

RESPONSE-The effects of longitudinal breaks in Class I piping were not analyzed because of the seamless construction of the pipe. SNUPPS~ agreed to provide additional justification for this exception to MEB 3-1. The axial stress to circumferential stress ratio'will be considered. O

%3A %:k S.L yeslk:n . L'hu l.)z A a r A. ~

1. \\*b o.

7t7:.1PootE %gssLN 3A d\\\\ b esdisd 4 7.v;1 =_ -N. p.;b - % u_ Ekss b b 9555 % is 1poiAc-oo kk WG\\ k~ e.ssesMd9. L s e,s A l e G td ~g e m i a. L N7.c RE%ca h\\4 [, tb yldu sn 3.'l(N) -l wil.1 k Ol T & 6-pp a h J A f ?q. Gab. I. / 2 4 ml J. / 30, a%s Tsk.._ 4. %.L. - 5.4.I is 3.9 (N).3.1. A ~.k we c4 is Tcseco sE S ed., - %.9(N).1.1.A w!Il b dvis e d L. 4.s\\s.bu 46 c=Ssms + 4. 5.4.1 aml l sbA% bb b enT se re s_ d l \\ Vi.c G n b ruck +

c. cob d 5

I I l u

a e-

1. 36 ?rt Table 3.9(B) - 7, does the sum of stresses acting during a faulted event exceed 2.4 Sh?

Response

Table 3.9(B) - 7 will be modified for faulted conditions as follows: The sum of stresses due to internal pressure, live and dead loads, and those due to occasional loads identified in the Design Specification as acting during a faulted event will not exceed 2.4 times the allowable stress S.

1. 37 ~ Clarify FSAR Tables 3.9(B) -3 and -5 to distinguish which ASME paragraphs apply to divisions 1 and 2.

Response

Tables 3.9(B) -3 and -5 will be clarified as discussed in the meeting.

1. 38 On page 3.9(N) - 44, 2nd line, clarify the me'. hod by which loads were combined.

Response

Section 3.9(N).2.5 of the FSAR will be revised (page 3.9(N) - 44, 2nd line) to read... " combined (by SRSS)... "

1. 39 What allowable stresses were used for anchor bolts used in B0P?

Response

The following change will be made to Section 3.9. j All ASME Section III, Class 2 and 3, supports are designed as welded attachments to embedded or surface mounted plates'. Bolting

• for the plates is designed according to AISC allowables with increases allowed by the loading case identified in FSAR Table 3.8-5.

In no case do the tensile stresses in bolts exceed the yield stress of the bolting materixl at temperature. t ---...,,,,.,n.-- -r -}}