ML20244B906

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Forwards Draft SER Reviewing Preservice & Inservice Insp Programs
ML20244B906
Person / Time
Site: Seabrook, 05000000
Issue date: 05/18/1982
From: Johnston W
Office of Nuclear Reactor Regulation
To: Tedesco R
Office of Nuclear Reactor Regulation
Shared Package
ML20235T530 List: ... further results
References
FOIA-87-51, RTR-REGGD-01.083, RTR-REGGD-1.083, RTR-REGGD-1.830 OL, NUDOCS 8205260244
Download: ML20244B906 (17)
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Docket Nos 50-4a3/444 s />

MEMORANDUM FOR: Robert L. Tedesco, Assistant Director l' for Licensing Division of Licensing FROM: William V. Johnston, Assistant Director l for Materials & Qualifications Engineering l Division of Engineering

SUBJECT:

PRESERVICE AND INSERVICE INSPECTION PROGRAMS:

SEABROOK STATION, UNITS 1 AND 2, PUBLIC SERVICE CO. OF NEW HAMPSHIRE Plant Name: Seabrook Station, Units 1 & 2 Suppliers: Westinghouse; United Engineers and Construction Docket Numbers: 50-443/444 Licensing Stage: OL Responsible Branch and Project Manager: LB-3, L. L. Wheeler Reviewers: J. R. Gleim, M. R. Hum, and T. Taylor, PNL Requested Completion Date: May 7, 1982 Description of Task: Draft SER Reviewing the Preservice and Inservice Inspection Programs Review Status: Applicant's Response Required SER Section 5.2.4 - Open Issue SER Section 5.4.2.2 - Confirmatory Issue SER Section 6.6 - Open Issue The Inservice Inspection Section. Materials Engineering Branch, Division of Engineering has reviewed the available infomstion in the FSAR and the proposed Technical Specifications related to the preservice and inservice inspections, j

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Our review has determined that the steam generator tube inservice  !

inspection (SER Section 5.4.2.2) is acceptable because the applicant I has comitted to use Regulatory Guide 1.83 Revision 1.Section XI of the ASME Code, and the Westinghouse Standard Technical Specification (STS). l Seabmok has Westinghouse Model F steam generators. However, we are identifying this as a Confirmatory Issue because (1) we expect the STS / I to be revised before licensing based on NRC generic investigations, and (2) the applicant has not completed the evaluation of the Model F design and has used sone infomation from the Model D design as a basis for his g g conclusions. We will close SER Section S.4.2.2 after we review and j I accept the Final Technical Specification. -

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l 1C FORM 318 (10-80) NACM 0240 0FFIClAL RECORD COPY usom imi-m ew j

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Robert L. Tedesco WY 6'&

i The applicant has not provided substantive information about the Preservice Inspection (PSI) Program in the FSAR Sections 5.2.4 and 6.6 which merely reference the Technical. Specifications as the source of infomation.

Unfortunately, the detailed program infomation was deleted fmm the l Westinghouse STS several years ago based on generic instructions to all Itcensees and applicants fmm the Division of Licensing. Standard Review Plans 5.2.4 and 6.6 require that we review the applicant's Preservice i Inspection Program and Inservice Inspection Program. To meet the SER completion date of August 7,1982, we will require two copies of the i applicant's PSI Program before July 7,1982. In the event that the appli-cant has not completed the PSI Program required by 10 CFR 50.55a(g)(3),

we request that he provide an estimated completion date and schedule for conpleting the preservice examinations.

Although the applicant states in the FSAR Sections 5.2.4 and 6.6 that he intends to comply with all requirements of Section XI of the ASME Code, no applicant has been able to meet this objective completely. We request-that the applicant revise the FSAR as appropriate and provide an estimated completion date for identifying limitations to conpliance with the ASME Code with a supporting technical justification.

Our Draft SER Sections 5.2.4, 5.4.2.2 and 6.6, which reflect the status of our review, are provided as Attachment 1. Additional guidance for the preparation of the PSI Program and relief requests for the ASME Code require-ments is provided in Attachment 2. The MTEB reviewers and our consultant,

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t Battelle PNL, are prepared to meet with the applicant to resolve this issue.

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William V. Johnston, Assistant Director for Materials & Qualifications Engineering Division of Engineering Attachments:

As stated DISTRIERION:

DOCKET Files cc: R. Vollmer R. Klecker MTEB Reading Files  !

D. Eisenhut G. Johnson MTEB RE 1-1 Seabrook Units 1 & 2 l E. Sullivan J. Cook, INEL F. Miraglia T. Taylor, PNL  ;

L. Wheeler J. Gleim ,

W. Hazelton M. Hum i C. Cheng

Contact:

M. R. Hum x-28118

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. .. St.FETY E V AL'.-a T I O N REPORT

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S J ttE.E R I 4 G E R Al4 C i4 INSERVICE INSPECTION SECTION  ::

Rese!cr "colant Pressure Boundary Inservice I r. s o c e t i o n_

. 'T . 2.1. .., .a r, 5 and Tes:'ns 11, y Complian_ce with the Standard Review Plan _s , ,.

5 . 2. 4. 9 -

_, ,The July 1981 Edition of the " Standard Review Plan for ~~

the Review.of Safety Analysis Reports for Nuclear Power Plants," (NUREG-0800) incluaes section 5.2.4,

" Reactor Coolant Pressure Boundary I r.s e r v i c e Inspection and Testing." Our review is continuing because the ,

applicant has not submitted the Preservice Inspection completed aLL preservice examinations.

Program and has not i

Our review to date was conducted in accordance with Standard Review Plan (SRP) Section 5.2.4 except as.

discussed below.

Paragraph 11.6, " Accept 2nta Criteria, Inspection Intervals," has not b es:n retvi e w e d b e c au s e this area not to the applies only t o i ns e rv,ics . ins p e c t i on (I S I) , .

This subject wiLL be Preservice I n s p e c t~i o n I'P.51.).

addressed during review d the ISI program after

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Examinati;n F.esu.':s" i has been revieae: :r: :e: a :Licant has stated that sata wiLL be evatiated in accercance with ASME Ccde rection IWS-3000, "Stcr.:s.-d: fer However, ongoing *.AC Generi.o ,'.  :

Examination Evaluation." -

Activities and research proj ects indicate that the '

I always p r e s en t-L y spec.ified ASME Code procedures may not '

be capable of detecting the accept able size flaws . . . .

For. example, ASME , , -

specified in the IWB-3000 standards.

Code procedures specified for volumetric examination

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of reactor vessels, bolts and studs, and piping have not ,,

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proven to be capable of detecting acceptable size flaws

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of improved procedures.and 'wiLL, require that these

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improbed procedures be made a part of the ir. service )

examination requirement 5.

We have not reviewed the applicant's repair procedures based on ASME Code Section IWB-4000, " Repair Procedures" because the applicant has not provided specific information. Repairs are not generally necessary in the PSI program. This subject wiLL be acdressed during 1

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our' review of the ISI Program.

Paragraph 11.8r " Acceptance Criteria, Relief Reovests,"

has not been completed because the applicant has not i,dentified all Limitations to examination.

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5.2.0.2 E . : ' e s t i .e h ntre-+e .

General Design Criterior. 32, "Inscreti:e us i.;;;oe Port Coolant Pressure Soundary," Appendix A of 10 C T F.

50 recuires, in part, that. components chi:h z.re ps-t . , _ ,.

t., e cc:igned';

of the reactor coolant p.ressure boundary .

to permit periodic inspection and testing of icoortant .

structural and areas and' features to assess their '

Leak-tight integrity. To ensure that no deleterious defects develop during service, selected welds and .

weld heat-affected-zones (HAZ) wiLL be inspected '

periodically at Seabrook Station, Units 1 and 2. -

The design of the ASME Code Class 1 and 2 components of the reactor coolant pressure boundary incorporates ..

provisions for access f or inservice inspections, as .

required by Paragraph I'J A-15 00 of Section x1 of the ASME Code.

Section 50.55a (g), 10 CFA Part 50, defines ,

the detailed requirements.for the preservice and inservice inspection programs for light water cooled _

Based upon the nuclear power facility components.

construction permit date of July 7, 1976, this section of the regulations requires that a preservice inspection least the program be developed and implemented using at Edition and Addenda of Section XI of the ASME Code applied to the construction of the particular component.

Also, the initial ISI program must comply with the..

requirements of the latest Edition and Addenda of Section XI of the ASME Code in effect twelve months .

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5 . 2 . 4.,3 Evaluation, of_ Complk ance w i t h 10 .CJ.R. 5 0. 5 5 a (g ) ,

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3 Je have-reviewed the ava.ilable inf ormation in the FW '

and find that the Preservice Inspecti:n Program for

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units,1-and ?. has, not been submitted for review. We wit.L complete our SER input on the Preservice .;

Inspection Program based on Standard Review Plan _,

5.2.4 provided the complete document is submitced by -

J uly 7,198 2. The applicant states ir the FSAR that all components of the reactor coolant p* essure boundary are designed, fabricated and erected in such a .

way as to comply fully with the requirements of .

$ Section XI of the ASME Soiler and Pressure Vessel Code.

u We find this commitment acceptable anc, theref ore, do not anticipate evaluating Request s f or Relief from*

impractical examination requirement. It should be emphasized that no other plant in the OL review has j

met this obj ective of full compliance with the ASME

! Code.

The initial inservice insoection progrts has not been submitted b/ the applicant. We wiLL evaluate the j

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program af ter the applicable ASME Code Edition and

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Addenda can be determined based on Section 50.SSa (b) _

of 10 CFR Part 50, but before the first refueling outage

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- when inservice inspection commences. l a '

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i conclusions l 5.2.4.4 The con 3utt cf perirdic in3DeC?i'~~ * :*: ry:rc:.ta*ic testing of pressure retaining enn: enent s of t ne reactor coolant pressure counoary, in accorsance.with t.h e rc ci r e .ent s< of - Se c tion XI e t ' *. ' e A -:e r i c a n S o c i .e l y -

. . 1 of Mechanical Engineers Boiler a n.! Freccure Vessel ]

Code and 10 CFR Part 50c will p'rsvide- reasonable assurance that- evider.ce c.f structural d eg r a d a t i on o r-Loss of.Leaktight integrity occurring during service . . .. .

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will be detect ed in tiise to permit corrective action  !

bef ore the saf ety functions of a component are Compf omi sed.

Compliance with the preservice and inservice ...

inspections required by the Code and 10 CFR Part 50 constitutes an acceptable basis for s a t'i s f ying the inspection requirements of Criterion 32 of the General Design Criteria.

5.2.6.5 . References

1. NUREG-0800, St andar d Review Plans, Section 5.2.4,

" Reactor Coolant Boundary Inservice Inspection and T esting," July 1981.

2. Code of Federal Regulations, Volume 10, Part 50.
3. American Society of Mechanical Engineers Soiler s . . _ _ . _ _ _ . . _ _ _ _ _ _ _ _ _

and Pressure Vesset Code,Section XI.

5. 4. 2. 2 -. S t e a m G e o p r a t o r_ Iqb e inserviee fnepsc.tica

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5.4.2.f.; ; .eki: nee .ith _

The July ;*S* Idition of the "3tandt.r; 'i e . i c . F l a r.

for the Review of Safety Analysis ?.ecerts f:r ';uclear includes section 5.4.2.2, Powe,r, Plants," (NUREG-0800) .

"St e am. Gen e r a t or- Tub e Ins e rvi c e . I r.s pe c t i o rr." .

Units 1. and 2 was reviewed in accordance with this .

Section of SRP, Howevers our review wilL continue. .

until the plant.^ Technical Specifications governing ..

steam generator tube examinations are completed and .-

are in conformance with the applicable Standard- ~

Technical Specification. .

the I n s p e c t i on P~r og r a er 5.4.2.2.2 Evafuation of .

General Design criterion 32, "Inspe: tion of Reactor of 10 CFR Part Coolant Pressure Boundary,". Appendix A

~50 requires, in part, that components which are part boundary be designed to permit of the reactor coolant areah periodic inspection and testing of important and leaktight and features to assess their structural retaining 0

integrity. The design of all pressure parts of the steam generators at Seabrook Units 1 the Ast4E and 2 have been optionally upgraded to meet Boiler and Pressure Vessel Code reovirements f or A$f4E

Code CLsss 1 components. Provisions sLso have been and 2 made to permit inservice inspection of the class 1

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comoonents, including indivioual st e s- :enerst ar The applicant has committed to following the recommendations

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Revision 1, " Inservice Inspection l oi Regulatory Guide 1.83r .

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of Pressurized Water Rea: tor Steam Generator 'I Specificatier* *or ,

and NUREG-0452r."Standere Technical  !

Westinghouse Pressurized Water Reactors," and comply with the requirements of Section XI of the ASME Code

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t We with respect to the inspection methods to be used. .?

find this com'mitment to be acceptable. .

5.4.2.2.3 Conclusions '

Conformance with Regulatory Guide 1.83, NUREG-0452, anc the inspection requirements of Section XI of the ACME Code constitutes an acceptable basis for meeting, in  :-

part, the reovirements of General Design Criterion 32.

5.4.2.2.3 References

1. NUREG-0800e Standard Review Plans, Section 5.2.4,

" Reactor Coolant Boundary Inservice Inspection and i l

Testinge" Sec-tion 5.4.2.2, " Steam Generator Tube C

Inservice Inspection," and Section 6,6, " Inservice Inspection of Class 2 and 3 Components," July 1981.

2. Code of Federal Regulations Volume 10, Part 50.
3. American Society of Mechanical Engineers Boiler

' and Pressure Vessel Coder Section XI.

4. Regulatory Guide 1.83, Revision ir " Inservice Inspection of Pressurized Water Reactor Steam
  • Generator Tubes."
5. NUREG-0452, Revision 2, " Standard Technical Specifications f or Westinghouse Fressuri:ed Wat er Reactors.

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3 . 4 6.6 Inservice insocitgan of_CLss ,I aad 3 c c r gnpna,s,

.iew,, Plan:

6.6.1 C ee: L i a n c e with the 5:ancar:

The July 1981. Edition of the " Standard Review Plan for the Review of Safety Analysis keport s for Nuclear Power P L an ts.r"'(SRP. NUeEr-0800) includes Section 6.6,  ::

"Inservi ce- In:pe cticn of . C t sc: 2 and 3 Co.ponents."

Our review,is continuing because the applica,nr has not submitted the-Preservice Inspection Program-snd has.not completed all preservice examinations. Our review to

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date was conducted in accordance with Standard Review

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Plan Section.6.6 except as discussed below.

Paragraph II.4r " Acceptance Criteria, Inspection Intervals," has nnt been reviewed because this area ,

applies only to Inservice Inspection (ISI) not to .

Preservice Inspection (PSI). T h i's subject wilL be addressed during review.cf the ISI program after Licensing.

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l Paragraph 11.5, " Acceptance Criteria, Evaluation of 1

Examination Results," has been reviewed and the 1

1 applicant has stated in the FSAR that evaluation of j Class 2 and 3 examination results wilL comply with

~ requirements of IWC-3000 and IWD-3000, respectively,

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of Section XI. ,

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  • However, ongoing NRC generic activities and research i proj ects indicate that the presently specified ASME e

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Code arvcedurca mof not si..y> L 'c.;.LLe ai cs. ;ie.;

the n aximum acceptable size flaws specifivu in t he e' ,

standards. For example, ASME Code crocedores specif ied

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for volumetric ex:mi.-ation of ves:eL:, bett! 2 r. d :tuds, and piping nave.not* proven to be cacaole of cetecting maximum acceptable size f law in alL cases. We will

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continue to evaluate development of improvec procedures anc wiLL require that thest improved procedures De made ,

a part of the inservice examination requirements. We .-

have not reviewed the applicant's repair procedures ~

based on ASME Code Articles IWC-4000 and IWD-4000, ]

" Repair Procedures," because the acclicant has not provided. specific information. Repairs are not ,-

generally necessary in the PSI, program. This subject l witL be addressed during our review of the ISI program.

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Paragraph II.7, " Acceptance Criteria, Augmented ISI to i

Protect Against Postulated Piping Failuress" has not been completed because this subject has not yet been addressed in the applicant's PSI program. We wilL .

review the applicant's augmented ISI program af ter it is submitted.

' - Paragraph 11.8. The applicant has not provided the s

complete Listing of exemptions f rom Code examination requirements as permitted by IWC-1220. We wiLL review i a t \

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th::e er.: pti:r.: :: they :r: :ub-itted i r. the :';*

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Parage:oh 11.9, " Acceptance Criteri:, Relief Rccue:ts,"

  • not has not been compte ed because the.aceticant h a e. ,f a'

identif i ed t he. Limitaitons. t o eaa in. tion.

6.6.2 Examination Requirements

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General Design Criteria 36, 39, 42, and 45, Apoendix A x

of 10 CFR Part 50 require, in part that the Class 2

. ;-y, a'nd 3 components be designed to permit appropriate .-

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periodic inspection of important components to ensure Section 50.55afg)  ;

system integrity and capability. ~*

of 10 CFR Part 50 defines the detailed requirements r

f or the PSI programs f or Light wat er cooled nuctcar .

power facility components.~ '

Based upon the construction permit dat e of July 7, 1976, this section of the regulations recuires that a PSI program for Class 2 and 3 components be developed and implemented using at least the Edition and Addenda of Section XI of the ASME Code applied to the const ruction Also, the initial.

of the particular component.

inservice inspection program must comply with.the requirements of the latest Edition and Addenda of Section XI of the ASME Code in effect twelve conths prior to

.< the date of issuance of the operating license, subject 1

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to tne Limitations eno rooific.tivns listeu i" 5e: tion 50.55a(b) of 10 C F F- F:~~ 50*

o.6.3 valuation of Co stiance w i t h . 10_if.i_5.0 . 5 5.s.( g )

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, re.have reviewed the v silsble ir.f:rmation in the TSAE .

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' and find that tnr Preservice Inspec' tion Program for!! .

Units 1 and 2 has not oeen submitted for review.

We w.itL templete.our SER input on the Preservice Inspection-Program based on Standard Review Plan 6.6 provided the The completed document is submitted by July 7, 1982.

~ ~ applicant has not identified any Limitations to examination. We wi t L .e:uire tha: the aopti: ant i

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identify aLL areas where the preservice examination

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requirements of the applicable edition of Section XI can jf not be met and provice a supporting technical justification.

The initial inservice inspection program has not been submitted by the applicant. We wiLL evaluat e the program after the applicable ASME Code Edition and Addenda can be determined based on section 50.55a(b)

. ;o f 10 CFR Part 50, but before the first refueling outage when inservice inspection commences.

.. . . ..d.4 . 4 CAn 0 53.9li9 0.1 Compliance with the preservice and inssrvice inspections

- -required by the American Society of Mechanical I

Ing.ineers Code and 10 CFR Part 50 constitutes an ts :r; table basis for satisfying applicable requirements i1 i

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ei :: n e .- a t Desi-n criteria 36, 30, l i

6.6.5 ?cf***FC't 1.

NUREG-0800, Standard Review FLsn, Section,6.6, et ctest 2 and 3 Comoonenty.,"

"Ir.s c rvi c e Inspe c t i c. . .

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j July 1981.

50. l
2. Code of Federal Regulations, Volume 10, Part *
3. American Society-of Mechanical Engineers Boiler
  • and Pressure Vessel Coder Section XI. 1

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CUIDA::: FOR THE PREPARATICf. OF PCI PROGRAM AND RELIEF REQUECTS MaTEpfab5 ENGINEERING BoaNCH .. l INSERVICE II;5PECTION SECTIC::

Review of tne Preservice Inspection Program ~

250.1.

Paragraph 50.55a(b) (2) Civ) of 10 CFR Part 50 requires that ASME Code Class 2 piping welds in the residual heat removal'systessi emergency core coolant systeme and .-

l containment heat removat systems shalL ce examined.

The control of water chemistry to minimize stress .

corrosion described in Paragraph IWC-1220 (c) of Sect ion  !

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XI is not an acceptable basis for exempting ECCS, RHN, and CHRS components from examination Decause practical evaluation, review and acceptance standards cannot be clearly defined. To satisfy the inspection requirements of General Design Criteria 36, 39, ard 45, t he pre s6rvi ce inspection program must include periodic volumetric l

1 and/or surface examination of a representative s amp t .-

of welds in the RHR, ECCS and Containment Heat i Removat Systems.

250.2. When using Appendix III of section XI for preservice

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examination of either ferritic or austenitic piping weldse the fotLowing should be incorporated:

  • l Any crack-like indication, 20 percent of DAC or A.

greater, discovered during examination of pipinq

l. welds or adjacent base metal materials should.

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-e:o-dee'ard vetti G3ted by 8 L?v*l II 0'

'O Level II: ? : :- i -I - - :. c the e 7. t e n t nee est!"Y cete'enine hs 12:e,1dentity, and Location of t ne -r e f letn:or . ..~

g ,, T h e O wn e r si.c '.;12. e v a l u.a t e- a n d take corrective ,_-

m ien for. rhe =i.sonsition of any indication invest.igat ed'nnna,;f ound. to be other than geometrical .

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o r aret a L Lur.g.ic.a1.im catu r e . .

250.3. Your PSI program should address augmented ISI to protect  :-

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against postulated.oicing failures in high energy f t.ui d s y st em p i pi ng . High-energy fluid system piping between containment i:clation valve should receive a1

< 1 augmented ISI in a:::~ dance with NtlREG-0800, Standard i;

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Review Plan, Section 6.6, Paragraphs I.7 and II.7. ~

Your preservice and inservice inspection program shculd

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! .j include these augmented examination provisions.

R,e g u e s t for Relief from Impractical ASME Code 1

Requirements

.1 J, ' "2 5 0 . 4 . Provide an approximat e dats when all relief recuests

-t The PSI program "j s wiLL be submitted f or evatustion.

p l a n s h o u l d i n.c is:d e t h ed c L L*o w i n g information:

_f.

For ASME 1 Code Clasv'1 and 2 components, provide a

A.

n8 E1 t a b l. e similar to IWB-26t0 and IWC-2600 confirming j .

t h a t siaAer ite.anidre 73 e e t i on XI preservice j] ,.- -

I h

examimrrhaar.w. forced on t he component or .

r e t M.53. "t:1;u reznew ist a t echnical justifica ion

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y B. Where relief is reauested for pres:ure retaining welds in the reactor vessel, identify the specific welds that did not receive a 100% r preservice ultrasonic examination and estimate the extent of the examination that was performed.. #{ .

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C.

Where relief is reauested for piping system welds C-Fe and C-G),. provide (Examination Category B .L, I .

  • Li st of specific welds that did not receive a .

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~

j- .

' complete Section XI preservice examination including a drawing or isometric identification ".

~

number, system, weld number, and physical ..

(

pipe to nozzle weld, etc.

configuration; e.g.,

Estimate the extent of the preservice examination that was performed. Wh'en the volumetric examination was performed from one side of the weld, discuss whether the entire weld volume, and the heat 1

.l affected zone (HAZ) and base metal on the far side State the primary reason I 4 of the weld were examined.

b; 4

that a specific examination is impractical; e.g.,

support or component restricts access, f'stting prevents adequate ultrasonic coupling on one side, cosponent-to-component weld prevents ultrasonic

' examination, etc. Indicate any alternative or i

11 supplemental examination performed and method (s)

_ of fabrication examination. - -.

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