ML20125D924
| ML20125D924 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/11/1984 |
| From: | Asselstine J NRC COMMISSION (OCM) |
| To: | Bernthal, Gilinsky, Palladino, Roberts NRC COMMISSION (OCM) |
| Shared Package | |
| ML20125D664 | List: |
| References | |
| FOIA-84-740 NUDOCS 8506120514 | |
| Download: ML20125D924 (1) | |
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u UNITED STATES NUCLEAR REGULATORY COMMISSION
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W ASHIN GTO N, D.c. 205:5 s..v /
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.s OFFICE oF THE dune 11, 15E4 i
COMMGSIONER MEf'ORANDUM FOR:
Chairman Palladino Comissioner Gilinsky Commissioner Roberts Comissioner Bernthal FROM:
James K. Asselstine
SUBJECT:
DIABLOCANYdN At the time the Commission voted to reinstate the low power license for
.Diablo Canyon, we had the benefit of an ACRS report, dated April 9, 1984, on the design control measures at that site.
Several of the members, in apperded views, urged the staff to prepare, prior to authorizing pcwer operation above five percent,"...a document discussing in considerable detail how the various relevant issues raised by its ins'p'ectors and others have been handled.".Those members also urged"...a careful examination (by the NRC staff) of a selected sample of actual construction details to help assure that the appropriate quality has been accomplished."
I supported those views at the time but an explicit Comission position was not developed.
It is unclear how, if at all, the staff will address the above.
I believe we should direct the staff to carry out those recomendations through the attached memorandum.
I ask that SECY obtain Comiss,ioner responses by C.O.B. Wednesday, June 13, 1984.
June 13 Jim:
cc: SECY I agree with the thrust of the ACRC members' OGC request.
H o w e v e r'*, I w ot$1'd "exp e c t that the OPE SSER prepared for the full power authorization to contain a' full discussion of the resolution of Mr. Yin's concerns and resolution of the allegations.
I would direct the staff to
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ensure that the SSER contains this documenta-tion.
Regarding the need for staff verification of quality, I believe the inspection program has done this.
I believe that the staff should verify that its previous activities have 85 61 g 4 850201 included confirmation of quality as well as DEVINE84-740 PDR quality assurance.
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If a majority agrees, I would redraft the SRM accordingly..
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NUR EG-0675 Supplement No. 22 Safety Evaluation Report ce a :ec
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Diaa o Canyon Nuc ear 3ower 3an:,
.ni:s ' anc 2 Docket Nos. 50-275 and 50-323 Pacific Gas and Electric Company U.S. Nuclear Regulatory Commission Q
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ABSTRACT Supplement 22 to the Safety Evaluation Report for Pacific Gas and Electric Com-pany's application for licenses to operate Diablo Canyon Nuclear Power Plants, Unit 1 and 2 (Docket Nos. 50-275 and 50-323), has been prepared jointly by the Office of Nuclear Reactor Regulation and the Region V Office of the U. S. Nuc-lear Regulatory Commission.
This supplement provides the criteria that were used by the staff to determine which of the allegations that have been evaluated must be resolved prior to Unit 1 achieving criticality and operating at power level up to 5 percent of rated power (i.e. low power operation).
The supple-ment also reports on the status of the staff's investigation, inspection and evaluation of 219 allegations or concerns that have been identified to the NRC as of March 9,1984, excluding those recently received under 10 CFR 2.206 petitions.
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TABLE OF CONTENTS Pace Ab s t r a c t......................................................... i i i 1.
Introd'uction...............................................
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' Appendix E - Status of Staff Resolution of Allegations or Concerns About Design, Construction and Operation of Diablo Canyon Units 1 and 2.............'......
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I INTRODUCTION The staff of the U.S. Nuclear Regulatory Ccmmission (NRC) issued on October 16, 1974, its Safety Evaluation Report (SER) in matters of the application of the Pacific Gas & Electric Company (PG&E) to operate Diablo Canyon Nuclear Power i
Plants, Units 1 and 2.
The SER has since been supplemented by Supplements No. I through No. 21.
SSER 18, 19 and 20 presented the staff's safety evalua-tion on matters related to the design verification efforts for Diablo Canyon Unit 1 that was the result of Commission Order CLI-81-30 and an NRC letter to PG&E of November 19, 1981. SSER 21 presented the program and the status of the staff review and evaluation of allegations and concerns identified to the NRC as of December 19, 1983.
This is SER Supplement No. 22'(SSER 22) and is based on allegations and concerns identified to the NRC as of March 9,1984.
1 i
This supplement provides the criteria that were used by the staff to determine j
which of the allegations that have been-evaluated so far must be resolved prior to j'
Unit 1 achieving criticality and operating at power level up to 5 percent of i
rated power (i.e. low power operation).
i SSER'22 also presents the staff's safety evaluation of these 219 allegations.
The staff evaluation of allegations and concerns is presented as Appendix E to j
the Safety Evaluation Report, consistent with the format of SSER 21. As of l
March 9, 1984, 219 individual allegations or concerns have been addressed by i
the staff..In addition, submittals were received in the form of 2 206 peti-tions from the Government Accountability Project (GAP) on February 2, 1984 and j
on March 1, 1984 which contain additional allegations. The staff has y t been able to avajuate or,categorir,e these. new submittals in dg.
v j
j The NRC Project Manager for the Diablo Canyon Nuclear Power' Plant is Mr. H.
Schierling. Mr. Schierling may be contacted by calling (301-492-7100) or by writing to the following address:
Mr. H. Schier11ng Division of Licensing j.
U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Copies of this Supplement are available for public inspection at the
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Commission's Public Document Room at 1717 H Street, N.W., Washington, D. C.,
L and at the California Polytechnic State University Library, Documents and Maps a
Department, San Luis Obispo, California 93407. Availability of all. material i
cited is described on the inside front cover of this report.
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Diablo Canyon SSER 22
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APPENDIX E STATUSOFSTAFFRESblUTION OF ALLEGATIONS OR CONCERNS ABOUT THE CONSTRUCTION AND OPERATION OF DIABLO CANYON UNIT 1 AND 2 e
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I n t r od u c t i o n................................................ E-1 l
2.
Diablo Canyon Allegation Management Program.................E-2 l
2.1 Scope..................................................E-2 2.2 Approach...............................................E-2 3.
S ta t u s S umma ry o f S ta f f E f f o rt.............................. E-4 4.
Cr.iteria for Priority Resolution of A11egations.............E-5 5.
Allegations Related to Reactor Criticality Considerations...E-7 i
5.1 Small Bore Pi ping De si gn Adequacy......................E-7 j.2 Anchor Bolt Design Margins and Insta11ation............E-9 5.3 Inspectors Certification...............................E-9
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5.4 Design Change Notice and Drawing Control...............E-10
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5.5 Falsification of Vendor Records........................E-10 5.6 Weld Symbol Implementation.............................E-12 5.7 Cable Spreadi ng Room Platform Adequacy.................E-12 6.
Concerns Relating to Employee Intimidation..................E-13 7.
S umma ry a n d C o n c l u s i o n s..................................... E-14 List of: Allegations or Concerns a : Diagram of Allegation Status' : Table of Allegation Status Individual Assessment Summaries t
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Diablo Canyon 55ER 22 E-111
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i 1.0 Intreducticn j
In early 1982 during the course of the Diablo Canyon Unit I design verifica-
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tion program certain allegations were made to the staff regarding the design and operation of the Unit 1 component cooling water system and certain other design aspects. The staff reviewed and evaluated the allegations on the basis of discussions with the individual expressing the concerns and issued its l
safety evaluation in Supplement No.16 to the Safety Evaluation Report (SSER 16).
l Since then numerous additional allegations have been made and concerns expressed regarding the design, construction and operation of the Diablo Canyon Nuclear Power Plant and the licensee's management of these activities.
In many cases the allegations include some aspect of quality assurance or quality control.
The allegations were received by the NRC staff in the Region V Offices and at i-Headquarters as well as by the Commission. They were made by a variety of t
i sources', including private citizens, former and current workers at the plant I
and at the PG&E and Bechtel Offices, news media, intervences, and Congressional i
j Offices. In some cases the source has remained completely anonymous to the NRC, in some cases the source is known only to the NRC, however, in most cases the source has"been publicly identified.
In many cases one source identified many items in a single submittal. In some cases the same allegation or concern was l
raised by more than one source. However, such same allegations from different sources were not combined in order to maintain a record of each item separately.
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As a result of the numerous allegations the Commission directed the staff on j
October 28,~ 1983 to pursue all allegations and concerns to resolution and re-
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r quested a status report on the investigation, inspection and evaluation effort prior tc its decision regarding authorization of criticality and low power test-ing. The staff subsequently developed the Diablo Canyon Allegation Management Program (DCAMp) which was provided to the Commission on November 29, 1983 in a
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l memorandum ~from~the Execiiiiive Director for Operations. A summarv of the oro-gram and the methodology applied are cresented in Section z of this reoort.
i Ine program was oescripec in cetail in SER Supplement 21.
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The staff is performing its investigation, inspection and evaluation of the J
allegations in accordance with the DCAMP.
In late December the staff provided a status of its efforts in SSER 21 on those allegations that had been received by the NRC as of December 19, 1983. The staff provided the Commission with l
written summaries of its ongoing efforts on January 4,1984 (SECY 84-3) and t
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February 6,1984 (SECY 84-61) and verbally briefed the Commission on January 23 and February 10, 1984.'
SSER 21 included, as an attachment, an Individual Assessment' Summary for each'
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of the al. legations.
In some cases the summary contained sensitive information or was predecisional in nature, in that the disciosure coulcf impair the staff's
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ability to initiate and/or conduct appropriate investigations or inspections.
These summaries were issued separately, with a limited distribution consistent l
with the Commission's' August 5, 1983, Statement of Policy on Investigations i
and Acjudicatory Proceedings (48 Fed. Reg. 36358).
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As of March 9, 1984, 219 individual allecati: s or concerns have been addressed by the staff.
In addition, submittals we e sceivec in the form of 2.206 peti-tions from the Government Accountability Prc.'sc: (GAP) on February 2, 1984 and on March 1,1984, which contain many addi:1cral allegations..The staff has not yet been able to evaluate or categorize these new submittals in depth. This supplement provides the criteria that were used by the staff to determine which I
of the allegations that have been evaluated sus; be resolved prior to Unit I achieving criticality and operating at power level up to 5 percent of rated power (i.e. low power operation). SSER 22 also presents the staff's safety j
evaluation of these 219 allegations.
2.
Diablo Canyon Allegation Management Program j
2.1 Scope i
The Diablo Canyon Allegation Management Prog-am (DCAMP) encompasses all allega-tions or expressions of concern which may be construed as allegations, which per-tain to the design, construction, and o'peration of safety-related structures, l
systems and components at the Diablo Canyon nuclear Power Plant, and which per-tain to the PG&E management of the Diablo Caryon Nuclear Power Plant project.
In this regard the DCAMP also includes concerns raised by the public and media, and provided by members of Congress. The pr: gram requires that all NRC Offices
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receiving new Diablo Canyon allegations forward them to the DCAMP staff in a timely manner.
The DCAiiP maintains as one of its tenets that the desire of a'n alleger for con-fidentiality or anonymity will be protected ay all means available. As a result 4
l of this requirement it'is necessary for some allegations and concerns addressed to be provided in a separate, limited distribution document. The assessment in l
this report, however, does include consideration of such items. :
2.2 Approach The fundamental approach in addressing the allegations to date' has been to focus j
on two basic quqstions.
Firstly, does the allegation present a technical problem which could affect safety of the plant?
Secondly, does the allegation reveal ary significant defects in the licensee's or his contractor's management or quality systems?
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The genera 1' sequence of steps was as follows:
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Confirmation of Allegation:
4 As each allegation or concern was received an efi' ort was normally made to contact the alleger to confirm our understanding of the matter.
In many cases confirmation was through a sponsor due to j*
the alleger's desire for anonymity.
In some cases meetings were l
held with the alleger to confirm our u*derstanding of the allega-tion. When requested, the alleger's icentity has been withheld from public disclosure.
In those cases where the alleger is un -
l known, the staff has made an effort to be reasonably broad in i
understanding the general deficiency er cencern provided by the i
alleger.
l Diablo Canyon SSER 22 E-2 l
i Site Insoections Many of the allegations required onsite inspections to verify con-struction practices, records, procedures and personnel qualification.
These were handled by teams of staff personnel with appropriate con-sultants.
In some cases additional, independent measurements and j
evaluations were performed where appropriate.
f Technical Reviews Consideration of allegations in technica-1 areas previously reviewed by the staff included detailed evaluations using licensing documents, regulations, standards, additional information provided by the licensee, and independent analyses as necessary.
In some cases additional audits were performed at the site or in the offices of l
the licensee and its contractors as necessary.
Interviews:
i Interviews with site personnel (crafts, quality assurance personnel, engineers and management) were carried out as required to resolve the issues.
1 Public Meetinos:
Where significant technical meetings were held, verbatim transcripts 4
were generally taken to maintain an appropriate record.
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Feedback from A11eoers:
When practical,--the staf' attempted to* discuss with the alleger the f
approach and findings of the staff's evaluation related to their allegation. The purpose was to assure that the staff properly understood the concern and to demonstrate how the staff dealt with l
the concerns.
Allegation Management Inst'ruction:
Region V's instruction on allegation management was used as guidance for this process. The draft instruction (entitled " Management of Allegations") was provided.as Attachment 4 to SSER No. 21.
The staff examined in detati al
" 211 of the_first, J8Q allegations.1 The purpose in doing this was in an ovtral'1'iierspective of not only the technical aspects of the problems raised but also to use i
the specific allegation as a vehicle for assessing whether the licensee and its major contractors acted responsibly over the years.
Considerable insight was developed on the licensee's and contractor's management control and quality control activities.
1#The allegations were not addressed in the same sequence as presented in i.
Diablo Canyon SSER 22 E-3 1
As the ' picture' began to develep, :ne staff started using more discretion en which individual alieca.icns merited a detailed review.
The staff elected net to review about 30 allega icns in detail.
These are issues which are.either very similar to those already reviewed in detail or, based on an assessment review, do not relate to significant safety issues.
The reasoning was that te do so would not add significantly to the management or quality performance issue.
The staff either has or plans to request the licensee to address most
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'of these from a technical standpoint with the staff auditing the licensee's response.
Allegations in this category are identified on the individual sheets in~ Attachment 4.
The staff continued to look into those allegations which appear to be unique, or which seem to present management control or quality issues not previously considered or those where alleger confidentiality was an issue.
The staff plans to use this more discretionary approach in reviewing the unaddressed and future allegations.
3.
Status Summary of Staff Effort The. staff review has to date involved more than 40 NRC technical staff (inspec-tors, engineers and investigators) from all NRC Regional 0.ffices and Headquar-ters irreluding contractor personnel.
Collectively, these individuals have expended in excess of.18,000 manhours since early November 1983 examining and evaluating the allegations or concerns.
During its inspection and evaluation of allegations the staff did not restrict itself to the allegation itself, but expanded its efforts beyon.d.the. original scope of the allegation whenever it considered this to be necessary.
These efforts provide the staff with a sub-stantial. basis for understanding the technical concerns raised and also the
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, perspective necessary for. making conclusions regarding the effectiveness of the management and qualify systems employed at the site.
In summary,.of the 219 allegations addresseci 146 items are considered resolved, j
73 are unre's'olved. Of the 73 unresolved items the staff has determined that none require a resolution prior to criticality and operation up to 5 percent power (see also Section 5 of this report),16 must be resolved prior to q
exceeding 5 percent power, the resolution of 57 items does not impact low or full power operat. ion, and there are no items for which the resolution status has not been determined.
Attachments 2 and 3 provide an overview of the status in a diagram and table, respectively.
The staff action for allegations or concerns is summarized in the Individual Assessment Summaries, Attachment 4 As discussed in Sectfon 1 of th.is report, in some cases the Individual Assessment Summary contains sensitive information or is predecisional in nature.
These summaries are not included in Attachment 4, but are 'provided to the Commission separately, consistent Qith the Commission's August 5, 1983, Statement of Policy on Investigations and Ad, judicatory Proceed-
'ings (48 Fed. Reg. 36358).
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Criteria f:r :rierity F.est.luti:: cf 7 die;nt: s l
Oe a :liec :: ice-tfy ::se allegatic:s whien need : :e :ursue: at esciiad urin; tr.e suff evaluatie :f : e first I~5 alle;a-!:ss crite-ia er:ive: ::
l wita the nignest rierity :;e :: : eir signift:ance rega-:ing :rt icaitty arc l
iev ;:=er ::erati:. Farticular :::sidernice i.as give: as :: w 4:ner er :::
.an issue caused ::erability te te crawn 1::: ::estica er weet 4r a significant i
ceficier:y is na.agese:: er cuality was incicatec. Curit; ce : elinica y re-l view the f 11: wit; cesideratices were a:: lied:
Is tre allegatics a s:e:ift: safety er :bality issue cr a generali:ec
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l Eas the staff previcusly as:resse this isstei L
Has'the issue been ; evicusly coalt with er is 1: :v bei:; cealt wiu by the 11:essee?
l Is : e allegatie: reas::a:1e as cces it scure cos etent?
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Dces t'4 allegatica re: resent a significant safety er nasageneet ::sce- ?
l It additten : 15ese :: sidera:1:ss the staff :: stee e: twc specift: aspects in maki:; its ceteminatics as :: wtether the allegati sus te satisfacterily res 1vec er :: res:1vec rt:r :: criticality are 1:w :cwer ;eratice. Tr4 twc as:ects are experien:e gaine: ae: fissier :-sca:t tevert: y resultic; frca low power operation. 5cta are accressec 1mel:..
The coeratic: ef Diable Canyon U-it I at_lw_:ceera:LlizeiJest of Ae same
_ systems as at fcH ocwer. Furuem:-e, systems and coaccrests will :;erate
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l arc ce ex:csec := cesic :-essu-e and teece-atu-e. Cte-a W -_m: 1:w mewer l
wcule inerefere ;- vice a seaes :: cetemine ase evaluate ne :lar ;erf:rw-l a::e under acre realisti: : ecitices. I ;ar*.icular, such ::eratics wcul:
erwse the :la:t te actual thersaLA:: asses,ar4_waild_ result in ast identify i
anyteterfaregtesbetwee: :T3U~ andlu;; -ts_ att restrai-ts. uscer._:perati g
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ste*efere, a systematic 1:w power ;eratice :- gram w uld icee-j tify deficies:tes er,c:ents azafyst: ally ceterzi ed cefteren:fes7tf ary, I
that subsecuently eculd be ::r-e: ed.
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At this time the Diablo Canyon Unit I reactor is fueled completely with new, unirradiated fuel without any fission products. During low power operation the amounts of fission products in the reactor would be approximately propor-tional to the power level for short-lived radioisotopes and to the total energy produced for long lived radioisotopes.
Even after several months _of low power _ operation, the fission product inventory._w_ou,1c, s_tt.ll_,p_e_one,_to_two orders.of magnitude. less.,gan_ the amount assumed in our_ safety._ evaluation.
Possible accident consequences would be further reduced since the decay heat is also decreased, not only in the rate at which it is released but also in the total amount available. The energy required to damage the reactor. in a postulated accident and the capacity of the plant heat removal systems and safety features are not reduced during low power operation.
Therefore, postu-lated accidents involving a failure of these systems would require much longer times to evolve and could be contained by equipment operating at only a few percent of its design capacity.
In summary, the possible consecuences of a reactor accident dur_ing. low power operation. are limited to a ve_ty sma.LLfrac-tion of those possible at full power.
Taking these f actors into consideration the staff applied the fo11ovino cri-teria' for assessing _which allegation and concern requires resolution prior to
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criticality:
1.
Prior to criticality those allegations or concerns must be resolved which offer specific new information. not neavinud v available tn the staf,f, inunhie a discrepancy _5etween design, criter_,1a,,oesign, agd,which_appame tn construction or_pperation of a safety-related component, system, or situc-
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tura 'of such map'itude..so as'to' cayledhe operabilitYto'be~dia'wn into question'.~I~n addit. ion, sufficien't' technical info ~rstToEregardWg7hese
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allegations or concerns is not presently available to the staff, or pro-grams have not been developed or. implemented to assure that regulatory concerns related to reactor safety wil1 be resolved prior to c'iticality.
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Prior to criticality those allegations or concerns must be resolved which offer definitive new information. riot _qrtyhusly,,available to the. staf f, and which indicate a potential, significar1t deficiency _in the licensee.'.s.
manag'eme,ntl.or,_qudity, Asturaru;_e_of_nf.ety rel ated act.ivi.ti.e..s.
In addition.
l sufficTent technical information regarding these allegations or concerns is not presently available to the staff, or programs have not been developed or implemented to assure that regulatory concerns related to
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reactor safety will be resolved prior to criticality.
In addition, the etaff applj,ed_a third criter. ion as_f.qlloys to determine which allegations or co.ncerns must be resolved prior to exceedino 5 carcent Dower:
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3.
Prior to exceeding 5 percent ooxer_those allegations or concerns must be i
resolved which offer specific new information not previously available a
i to the staff, and which_ may reasona31y be expecIEG[fdydlieheable f a ilu_r_e s_ o.f sy. stems..tb4_t._c,on,ta_i n_r.Af iracti.y,i_ty o rff the ECCS_ systems.
In addition, sufficient technical information regarding these allegations i
or concerns is not presently available to the staff, or programs have not l
been developed or implemented to assure that regulatory concerns related to reactor safety will be resolved prior to exceeding 5 percent power.
)
In formulating these criteria the staff emphasi_zesLabat the new information must be slefinitive specif_i.g,,.and credible. As the staff has gained experience in evaluatini the first 200 allegations addressed in this report it developed
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i reasonable confidence to conclude that the licensee and its contractors have i
acted responsibly over the years. Although there have been some lapses the quality and management systems related to construction have worked reasonably
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j well. As a ca mit of this nersoective gained the staff feels that the burden i
j has shif ted somewhat such that allegations _,of_,a,,. genera 1 or circygnantial l
nature should not be " assumed true untWoroven otherwise".
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5.
Allegations Related to Reactor Criticality Considerations 4
j In SSER 21 and SECY 84-61 the staff identified seven areas of. concern (involving i
21 allegations) which required resolution prior to reactor criticality and low power operation. Since early of this year the staff has pursued the resolution i
of these issues with the highest priority and has devoted extensive. effort to i
the inspections and evaluation _of.these_ matters. As a_ result the staff reviews have pr.ogressed to the, point that the issues are either completely resolved or~
i resolved to the point where they no longer warrant full resolution prior to -
s reactor criticality considerations. The status of each of these issues is i
provided below.
3 5.1 Small Bore Pinino Design Adeouacy ( Allegation: 55, 79, 82, 86, 87, 88, 89,89,95,97):
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In the course of investigating the numerous allegations concerning the design of small bore piping supports the staff reviewed a large quantity of material l
concerning general design pract' ices, implementation of design control measures i
and the conduct of specific analyses. These efforts included inspections at i
the On-Site Project Engj.,neering,Groun (OPEG), the essentially self-contained engineering group responsible for small bore piping design and analyses at the p
Diablo Canyon Site, and inspections at the San Francisco offices of PG&E and j
the Bechtel Corporation.
As a result of these inspections a number..of tbeWgatign.L,tp,].p_t.e( to the administration _of the.0 PEG.we.r.e_.s.ubs14ntiated in whole or in nart.
Specifi-l cally, al' egations related to deficiencies in" doc'uiseWcontrol at the site, site specific training and effective use of deficiency reports were substan-4 tiated.
Th,e_ principal technical finding,i_Ltha_t the analvses nerformed by computer for h
smallbor_e,_p1,pfnjsupportshavebeen__de.tarMaad in have an unernecteg.ly large l
error, rate.,p,n the,,orce;r_of tiwe.nly_pg. cent = ~- r.ed-to ten nr lann_pg.r.cen,t that experiencT'has,sFowz l like.ly. On the other hand the error rate in the l"
Diablo Canyon SSER 22 E-7
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hand calculations for small bore piping supports was acceptably low.
In light
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of these findings the staff will require that PG&E. establish a program to j
review all.compu.ter analyse.s. for.s_ mall bore piping supports.
In partial response to those staff findings the licensee has reported the results of a review of approximately 130 small bore piping support computer analyses including the analyses in which the staff had previously identified errors. The licensee reported that, with errors corrected where necessary, all completed calculations showed final acceptability of the supports. The staff conducted a special inspection to evaluate the process used to re-review the small bore piping calculation packages. We found with minor exception, that the review process was comprehensive, was being carried out by qualified individuals, and was conducted in a manner to assure that the results could be accepted with high confidence.
Analyses of the type and significance of the deficiencies seen to date has led the staff to conclude that, although the design QA program for the OPEG is not up to acceptable standards, the impact.in terms of design adequacy, has not been significant.
Based on the results of the staff's review to date and the types of errors that have been identified it is very likely that moctrications,1T any, would bi minor anc only to fully ~meifspisliiiFiiYitef a"lifth Tittle or no" impact on
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operabifity of~ systems u5deF iihTfiill~rangegr plant}op' era'tjons.
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Since some_pjping_suppo_rt opf.jfjcation.s.,a_r.1_a9ImAllY_.rsqvi.r.ed as a result of inittar pTant op_er.at. ton._due to unnaicteskthermalin.tt_ons or operating requirementsJf_ attached nc suppo.csed,,,3,guipment, there is so~urd lenic in conductina the raa"4-ad calculation review curing low power operation so that ani resultine =Mifications enuld be iiEluded in inTedeMy a53757IT5Tidated program prior to full newer eneration.
DiabloQanyonS$ER22 E-8
5.2 Anchor Belt Desian Margins _and Installation _(Allegations 25, 58, 96, 142, 154,176):
The concerns raised by these allegations involve the installation and inspec-tion of concrete expansion anchors by the H. P. Foley Company (primary electri-cal contractor and construction completion contractor). A general and non specific concern with anchor bolts was supplied initially to the staff from an anonymous alleger.
Subsequent interviews of onsite contractor personnel re-suited in additional concerns with added detail in some cases. The staff approach to resolution of these isues was to:
(1) review installation proce-dures, audits, nonconformance reports, discrepancy reports, and licensee correspondence relating to concrete anchor bolts; (2) have an independent NRC contract team (Lawrence Livermore National Laboratory) inspect a sample of 124 electrical raceway supports modified in 1982 (involving hundreds of anchor bolts);'and (3) request the licensee to perform torque tests and ultrasonic examination on a sample of 40 installed anchor bolts to verify the adequacy of installation. T_he__staf f...found_that_mone_of the-44-14 gat 4ons iavalved a sub-stantive cuali,ty or =maagamaa+ raatral nenblam-During the course of this review, however,,the_1.taff ident10ed A -number.of,,thety_own _t_echnical conce_rns related to anchor bolt adeaumev-In response to a staff request the licensee undertoR Tiiextensive test and evaluation program. The results of this pro-
~
gram were reported to the NRC, concluding that adecuate marcins of safety were provided in the installed anchor coits.
Based on the results of the test program the staff concludes that there is reasonab' e assurance that installed anchor halts are adeouste. Accordingly, the staff cons 10ers tnts issue adequateTy. resolved for the purpose of licens.
ing dec'1sions.
5.3 Inspector Certification (Allegations 57 and 68):
In response to the allegations concerning certification of ouality cont +ol inspectors employed by both the H. P. Foley Company and by the Pu' lman power Procucts company (primary pipinglnstallation contractor) at the Diablo Canyon project, the staff examined the contractor's programs and their implementation in effect during the companies' activities to assess whether appropriately qualified persons performed quality control inspections of safety related items. The staff concluded from their examination that there is reasonable assurance that 1ndWWals performing'iiidlity contr'o1 tnsneenn were ah'
< fied
~
to perform-~their assignea tasns w' tX the exception ar a emaa snvolvina Du' '==n Power Dead"et F - env durino the ' e73-7a time fr---
f a tM e este certain DC insnectert were found to have.been,p.ttforming inspections prior to comp' etely '
satisfying prescribec grBfication reaui.
ats.
All but,twp of these indl-viduals~had a%quate backgrounds and experience 17i the areas of welding and quality control inspection.
It does not sopear that this mroblem was chronic or widespreads _The licensee has c snaitted ta ca=aiate a==ala rain mn or the inspectors' work prior to tse time that thav_ wore fully certified to,
perform IFrelated7TWO inspections.
This effort will be e-lated by
~
March'3(19#4~
.Diablo Canyon $$ER 22 E-9 6
+
The staff _cencludes that in the overall quality control inspectcrs were i
p6ipdl'y.qualOiid for"the.7. asks they peif.:rmed.rAccoWingTyTthe staff considers that this issue has been adequately addressed for the purpose of i
licensing decisions.
5.4 Design Change Notice and Drawing Control (Allegation 61 and 102):
The staff examined the licensees and contractors programs for the control and i
i issuance of design change notices and related drawings. The staff determined that the controls apajied to these activities were cenerarly adejuate.. At the
~
I thte._of' 4'""ance of SSER 21 the staff had identified a car.1@ arty. complex ~~~~
l desien chance notice and its relm+=d drawines for_ fur +har analysjs. This change notice involved approximately 130 major and minor revisions. At the i
staff's request the resppndble engjaeering personnel met with the staff and l
presented documentary evidence that ea'ch 7eWsTon Wr~eitng.connleted, super-ceded, or voided. The licensee also showed the staff the completed start-up
{
t'est' repoits"for this system which demonstrated that the system operated as intended. Based upon these results and additional programmatic and technical reviews the staff ' concluded that change notices and related drawings were ade-i l
quately controlled and implemented. Thil.j,11ue i s-c.usidnad. Adequately i
resolved for purposes of 1,1 censing, decisions.
5 5.5 Falsification of Vendor Records (Allegation 99):
l This allegation came to the.NRC staff attention through a local San Francisco television reporter. Staff action was initiated at that time.
In addition, the licensee initiated 1.ts investigation of this subject after viewing the television report.
Sfnha the criainal allegations,were received the staff and s
the licensee, through thefr inva"W=, have secaived two aroues of add tional,,a_1lena_tions.
3 The NRC staff response to the allegations includes a combined effort by the
- i j
Office of Investigations, the Licensee Contractor and Vendor Inspection Pro-i gram Branch of the Office of Inspection and Enforcement, and Region V.
T_he staff oosition %as been both one of agni ring,hg the licensee is conducting i
j its investigatian for the Diablo Canyon to ect anFindependently reviewing i
the issues for generic significance (the company has provided products.to multiple nuclear reactor projects),
i The staff has addrelipd_and closed.the.otiginal.. allegation A review of perti-nWtreifirifs"esE11shed that the former inspector (who claims to have docu-L mented inspections he did.not perform) is credited with performing 650 inspec-i tions phile he was employed at the vendor. Fifteen of the.650 inspections-involve safety-related material. These fifteen ' items were found to be i
supplied to Diablo Canyon Unit 2 and involve " stock" material (i.e.-raw. mate-rial items which do not involve welding). As of_this wri,t jin,th_3., staff has inspected 14 of the 15 items and found them to conform with. tequirements. The
)
stiff is f61'1Wf6FGp on thi'lasiiti.s(plat'e washers).
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Otablo, Canyon $5ER 22 E-10
. --. ~.i
i The licensee has selected a 10% sample of the other (non-safety related) inspections related to the inspector and performed a reinspection (involving 940 welds).
Seven of the 9a0 reinspected welds were found to have.,de.v.i.ations from r_equiremelitss tbese are being_ properly addressed._ Based ugon the low I
defect rate the licensee has conc,1uded that.the structures and. components inssa11ed at Diablo d.nyolLiave[not,been_adypr.seJyJmpacted byJhe.,former
~
inspector's. alleged.pe,rformance.. The staff concurs with this_ conclusion based upon a review of licensee actions and independent inspection of the fifteen safety-related items.
Neither the licensee nor the staff can determine conclusively whether the former inspector neglected to do the inspections.
The staff has completed a substantial amount of review on the second and thjrd groups of allegations, and to date has not identified problems of safety signi acance the,_rev.),ewYneyevyr'; atCc_o'pynui'n~g"(~e".~g. the staffMs not completed their review of the operations at the vendors subsidiary). These allegations are mainly general in nature, lacking in specific examples thus requiring extensive interviewing and document reviews.
In a parallel effort the licensee has initiated an inspection of installed hardware to allow a direct assessment of material adequacy, separate from the management and programmatic concerns related to the vendor.
Items that are being reinspected were selected by reviewing all shop drawings _and-selected purchase orders involving the vendor's material shipped to the jobsite since 1969 and includes samples of each material type supplied to Diablo Canyon with particular attention to items which are difficult to fabricate or involve
'~
special materials.
~
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90% of the sampling has been completed and the licensee.reporis that the follow-ing trends and results are apparent:
a)
General inspections are finding that the existing geometries and dimen-siens are in conformance with the shop drawings.
b)
Hardness tests are indicating that correct materials were provided.
c)
Visual weld inspections are indicating that vendor welding meets design requirements.
d)
Records from the NDE' documentation research show that full penetration welds by the vendor are satisfactory.
~
In addition to the licensee's reinspection the staff has ' independently inspected a. small sample (14 types of components) of installed safety related hardwaretoobtainfirsthandevidenceof.productquality. The components were visually inspected for'm'aterial damage, weld location, length, size,
~~
shape, reinforcement, appearance and type. The staff did not identify any discrepant material. Records related to this material were reviewed and appeared to be in order.
Diablo Canyon SSER 22 E-11
Investigations and reviews have been completed on the initial and most alarm-ing allegations. 'This item is resolved.
The reviews are continuing on the other two sets, but, to.date significant. safety problems have not been.identi-fie_d. Based upon Jtaff findings to date and the acceptable results of rein-
' q.
spection of installed hardware it is the. staff,'s opinion _.that.this_is. sue _no longer requires full resolution prjor. to-licensing decisions.
5.6 Weld Symbol Implementation (Allegation No. 126)
The staff received an allegation on December 20, 1983, that alleged that a major problem existed with the licensee's home office and site engineering because no welding symbol' standard (such as AWS A2.4) had been implemented at Diablo Cariyon. 'The'sGff7eviesed the alleger's concern and determined that his concerns had merit.
The_. staff,subse,quent.ly_ reque_sted_that the licensee address this byff,roviding the following information:
Assessment of the safety significance of the inconsistent weld symbol application.
Assessment of the weld symbol interpretations used by organizations engaged in welding activities in the field.
Performance of such field examinations as deemed necessary to establish whether any inconsistencies in interpretations caused a failure of the field welding activities to conform to the designers intent.
On. February 2,1984, the. licensee provided their position on the acceptability of the Diablo Canyon weld design and installation program. The staff's revjew required to comply with AWS A2.4, indicated thal.though_tbe W an uo m an*
the71censee'_s_ pro, gram _ generally _ met the criter,ia of_ LAWS,A.2 Cfor welding sym-
~
bology. AdditionalT9, the licensee did have usable alternate programs for the clarification and interpretation of weld symbols. The staff na+er + %+
- h WRC in.spection and reviews have not identified any instance where the failure by
~
~
the 1.icenste ti.fu11y_3Epleiisnt!.the AWS,.A2_.L.we3 ding 'symbology,"ris~ulted in weldments.yhict would not meet the designer's intentions. This issue is con-sidered resolved.
I 5.7 Cable Sp eadino Room Platform Adecuacy (There is no specific allegation related to this topic. A staff concern was identified in this area while examining documentation related to anchor bolts).
I During a walkdown of cable tray and conduit supports on January 14, 1984, the
'NRC inJpector identified two Class I Electrical Raceway Suppprts attached to the Non-Class I steel supporting a platform in the cable spreading room. The inspector also noticed several deficiencies in the instaiiauon or tne concrete anchor bolts securing the structural steel to the concrete.
Diablo, Canyon SSER 22-E-12 3
A review of records disclosed that the deficiencies in the anchorage of the 3
structural.ste.el had been previously identified by a Foley inspector on 4
October _l,._1983.__ine insoector obseryla__from his review of the recor_ds that the platform steel was not designated Class I (safety-related) despite__the
]
f act thatjhisTstRI'tuRitee t was ceing usec to support class 1E electrical c
panels in the cable spreading _rcom.
1 The condition identified by the NRC inspection was documented in a nonconform-ante report and provided to engineering for assessment of technical adequacy.
This issue was addressed in the licensee's letter to Region V (No. 'DCL-84-047),
dated February 7,1984. The licensee determined the as-built condition of the cable spreading room platform installation. The as-built condition was ana-lyzed by the licensee's engineering verifying that the installed condition was acceptable and conformed with design requirements. In assessing the generic a' +5e implications of this issue it was determined that the unicue na+"ea steel-frame raised-floor conficuration led to the acceptance of the design _and material without the detailed tvoe of as-bui_1_tj.0g_^nd ys's ~ w=c ner-2"=
formed f or the other structures. _. This_ type of conficuratinn eritte nnly in the cable spreading ra= s.
All other platforms which support Class I equipment have been analyzed. Therefore, this installation is not a generic issue.
The staff concludes that the licensee has adequately demonstrated the accept-ability of the cable spreading room platform installation. The staff considers that this issue is resolved and does not require further action.
6.
Concerns Relating to Employee Intimidation A few of the allegations received by the staff related to possible intimidation of workers at the plant. The staff took specific action to assess whether this condition was.a widespread problem or concewn at the facility. The staff effort on Diablo Canyon allegations involved several thousand staff man-hours on-site, where staff members have interfaced with hundreds of licensee and contractor crafts, quality personnel, engineering personnel, supervisors, and managers.
During the course of this effort the staff was instructed to be alert and look for evidence of " corner cutting" or pressure by management that would be counter to good quality practice. The staff interactions with site personnel included informal nne-aa-aae 9 wonions. croup discussions, anc formal meetings. _,1be staff also observed _gr.oups and ind.iv.icual.i 1AteractMg a.mong themselves.iq_lery cTsuaT situations _(such as during plant tours, and lunch room and work area diWDssions). These types of observations have been useful in gathering a sub-jective sense for the overall plant " atmosphere" regarding issues such as freedom to discuss concerns or intimidation.
In addition,_approximately 250 site personnel were.specifically questioned regarding. such ' items as pressures
_to cut corners intimidation, o,r_,_f_.reedom to oring forth_ _ quality _and _ safety -
reiated concerns. These interviews were concucTec, in part, to determine if therT w'aT~a generalized atmosphere to repress problems or safety concerns.
Diablo Canyon SSER 22 E-13 l
Based on the staff work in this area it' appears that a few individuals feel strongly that they have been directly intimidated. Some have offered specific
)
and detailed reports in support of their allegation. These cases are complex.
The staff could not readily tell whether the cases involve intimidation, pro-per exercise of management perogatives, or just poor communication. As appro-priate, these few cases (eight total) are being addressed through the Depart-ment of Labor regulatory process, and/or review by the NRC Office of Investigations. A few additional individuals were concerned about intimida-tion but indicated their views stemmed from events not directly related to them, such as general perceptions that the pressure was on to get the job done, or from the layoff or firing of another employee, or media reports of intimidat-ion. The staff does not detect any widespread company attitude (either deli-berate or inadvert.ent) to suppresJ eE5Tovee enn m ni nr7n%Ipt_diCoy.er.all effectiveness of the Quality Assurance proccam. The staff also found that in the vast niajori{y'of isteractions employees are. net afraid to;comLforward.with
~
reifo'rts of;'and dearwit% qualf.ty_pf.oblens_ in a responsible manner both with the W 5wn organizations ~and with the_NRC.
While the staff concludes that a widespready suppression problem does not exist at Di'ablo Canyon the staff is -concerned with employee perceptions in this area.
Licensee management. shares this concern. The staff has reviewed this subject with licensee management and notes that the licensee has undertaken steps to make imnrolaments. This effort includes such actions as._thg_ development of vI'd U tape presentations for all existing and new employ _e_es regarding surfac-y ing of quality concerns: an "800" telephone number for receiving quality con-j) cerns,; and, a_sy_s.gm fqr_re_ctip.t_ antLcontrol of c_oncerns.
The 11censeYs-acti-vitfes Tn this area will, be monitored by -the staff.
4 7.
Summary and Conclusions
~
1.
-As of March 9, 1984 a total of 219 allegations or concerns have been addressed by the NRC.
2.
The staff has developed criteria that have been used to determine which allegations or concerns must be resolved prior to (a) critical-ity and low power operation and (b) full power, operation.
4 3.
As of March 9, 1984 the staff has concluded that none of these alle-gat 1ons reo_yjy_tasolution prior to a reactor _c_r,i[1_caRty_ decision.
lhe staff has concluded that tne fiInal resolution af 12-sepa.cate
~
allecations relating to two suoaects can_be'de,f r_ red fenm nee-A l
criticality to pre-f6T1 power.
4'.
The staff has concluded on the basis o'f its inve'stigation, inspection, and evaluation, that there have been some lapses in the quality and management systems related to construction, however the systems have worked reasonably well. The staff-has reas6nable confidence that the licensee and its contractors have acted responsibly over the years.
5.
The staff is continuing its' investigation, inspection and evaluation
. I i
of all unresolved allegations and concerns.
i Diablo Canyon SSER 22 E-14 e
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6.
The staff effort is sufficiently complete recarding the._219 allegations to conclude that none of the auegationsJadica.te. problems of such a ma g n i tuc e, e i WeTTn'dTijEalfy. o r c o l l e c ti v e ly.. t ha t s ho ul[ pikc[l Uce auth5FizatWri~fo'r7Efilicality and 1ow power operation.
a Diablo Canyon SSER 22
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items Nos. 25, 58, and 96.
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n s e=-
3.4.1.3 Control and issuance of design change notices and M
!i related drawings.
t."
j As discussed in paragraphs 3.2.1 above there are two allegations or concerns (No. 61, and 102) which lead 7
the staff to seek more information regarding this sub-b ject.
It is the staff's opinion that a preliminary H
assessment of adequacy cannot be made at this time.
4 (g
The collection and evaluation of additional informa-tion is in progress. A preliminary assessment of adequacy cannot be made at this time. The collection 3
and evaluation of additional information is in progress.
s Li s
3.4.1.3 Inspector Certifications
/
9.
As discussed in paragraph 3.2.2 above, inspection of 6
allegations or concerns Nos. 57 and 68 identified
{
several instances were inspections were performed by 2
individuals not certified at the time of the M
r.
inspection.
DI l
At this printing it is the staff's estimate that preliminary assessments regarding the above topics will be completed by j
January 18, 1984. This date is conditioned upon subsequent g
review findings and responsiveness of the licensee.
p!
E 3.4.2 Actions Recuired Prior to Exceeding Five Percent Power It is the staff's position that the following actions p
be completed prior to exceeding 5% power:
?
yi-3.4.2.1 Implementation of a technical specification limit on the operation of the Component Cooling Water System fO whenever ocean water temperature exceeds 64* F.
h This item was a result of staff examination into
('
allegation or concern No. 5 and is addressed in g
detail in the Diablo Canyon Safety Evaluation Report, b
NUREG-0675, Supplement 16.
y n
3.4.2.2 Completion of seismic modifications to the diesel y
generator silencer bracing and pipe supports.
g This item was identified in conjection with the g
staff's examination into allegation or concern No. 8.
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Diablo Canyon SSER 21 E-15 y
V
.e 3.4.2.3 Completion of the inspection and verification of the
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as-built drawings located in the control room. This item was identified in conjunction with staff evaluation of allegation or concern No. 34 4
3.4.2.4 Complete modification resulting from the seismic systems interaction study, in progress,-in accordance with commitments identified in SSER 11.
This item was identified in conjunction with staff evaluation of allegation or concern No. 48.
3.4.2.5 Complete the analyses of significance of coating f
(painting) concerns discussed in concern item No. 6.&
As indicated previously, there were a few allegations or concerns which were received late in the evaluation period and/or sufficient time was not available to effectively evaluated prior to the issuance of this SSER. Five allegations or concerns fall into this category, and are listed below:
. 4 g*
Item No.
Subject
,/-fy 88 Undocumented modifications to small y
bore pipe supports SS 95 Angle members in small bore pipe supports 99 Falsification of Vendor Records (Bostrom 4
l Bergen/Medco)..
101 Welding Qualifications (Foley Company) 103 Welding Qualifications (Pullman Company) i l
All of the above allegations.;or. concerns have been entered into the j
established NRC tracking systems and are scheduled for investigation i
or inspection in a timely manner. The staff will provide the Commission an updated written status at six week intervals and will
)
i.
be prepared to provide an oral status report at any time.
3.5 CONCLUSION
AND RECOMMENDATIONS The allegation management program in place for current and future allegations related to Diablo Canyon has and'should continue to
)
provide a procedure for orderly and thorough yet timely examination of each concern raised.
4 l
Approximately 75% of the allegations currently. received have been
)
examined to a point where it is the staff's opinion that there is no significant safety issue'or: substantial breakdown of management or quality systems. The remaining allegations have been assigned to various elements of tle NRC staff for, evaluation and most have been 3
t Diablo Canyon SSER 21 E.
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UNITED STATES OF AMERICA NUCLEAR REGULATOP.Y COMMISSION
'84 APR -3 P4 :44
[s COMMISSIONERS:
s~ ~"':..., e.
Nunzio J. Palladino, Chairman
'g'RirICE." '
('.
Victor Gilinsky Thomas M. Roberts James K. Asselstine e
Frederick M. Bernthal
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In the Matter of
)
)
PACIFIC GAS AND ELECTRIC COMPANY Docket hrs. 50-275 50-323 (Diablo Canyon Nuclear Power Plant, Units 1 & 2)
)
)
C7 6 0 (CL 4) q
~' This order co'ncerns the issue of the consideration of complicating
. effects of earthquakes on. emergency. planning in the Diablo Canyon licensing proceedings.
In' the San Onofre proceeding, the Comission declared that.
current regulations do not require consideration of the impacts on emergency planning of earthquakes which cause or occur during' an accidental radiological release. Whether or not emergency planning requirements should be amended to include these considerations is a que~stion to be addressed on a generic, as opposed to a case-by-case, basis.
Southern California Edison' Co. (San Onofre Nuclear liendrating Station, Units 2 and 3), CLI-81-33, 14 NRC 1091, 1091-1092 (1981).
In the interim, the Comission precluded consideration of this issue in indi-vidual licensing adjudications. Thus, the boards have properly. excluded i
this issue from this adjudication.
T l
~
I
2 In response to the Come.ission's San Onofre decision, the NRC staff reported its view that generic consideration was neither necessary nor appropriate, but appears to believe that some specific consideration of the effects of seismic events on emergency planning may be warranted for plants located in areas of relatively high seismicity.
See NRC staff memoranda, dated June 22, 1982 and January 13, 1984, attached hereto.
In view of this development, the Commission has decided to address whether to allow such consideration under the circumstances in this case. With respect to low-power operation, however, the Commission is satisfied that, pursuant to 10 C.F.R. 50.47(d), this issue need not be reviewed further because it pertains primarily to offsite emergency planning requiremen.ts which are not essential to low-power license decisions.
To help the Commission with its c'onsideration of this issue, the parties are requested to provide their views on the following issues no later than 30 days after the date of this order.
Issues:
1.
whether NRC emergency planning regulations can and should be read to require some review of the complicating effects of earthquakes on emergency planning for Diablo Canyon; 2.
if the answer to question (1) is no, should such a review
^
be performed for Diablo Canyon on the ground that it presents special circumstances under 10 C.F.R. 2.758.
If so, what are the. pecial circumstances that would permit s
consideration of the effects of earthquakes on emergency planning for Diablo Canyon?
3.
if the answer to (1) or (2) is yes, then the following-inferpation should be provided:
O e
3
3 (a) The specific aspects of eraergency planning at Diablo Canycn en which the icpacts of earthquakes should be considered.
(b) The specific deficiencies in the consideration already given to the impacts of earthquakes on emergency plans _for.Diablo Canyon.
In this regard the NRC staff is directed to serve on the parties to the proceeding a copy of the Licensee's submittal regarding effects of earthquake on emergency planning. However, the Connission is not requesting the filing-of contentions in response to this order.
The matter of contentions will be handled by a Licensing Board if a proceeding is to be held.
(c) The appropriateness of limiting o the Safe Shutdown Earthquake the magnitude of the largest earthquake to be considered.
(d) The substantive criteria for reviewing the effects of earthquakes on emergency-planning..
(e)...The _ necessity for-litigation of this matter, includ-ing the general scope of (i) that should be held, and (ii) proceedings, if any, issues that.should be
~' litigated.
~
The Connission notes that it is not now deciding whether any
^
requirement for further hearings would require that interim operation of the plant be stayed. The stay determination, if and when it is pre-sented, will be a matter for the equitable discretion of the Connission or Appeal Board. See!e.g., Public Service Company of New Hampshire
~
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(Seabrook Station, Units 1 and 2), CLI-77-8, 5 NRC 503 (1977).
Parties j
need not address the stay question at this time.
Comissioner Gilinsky abstained from this decision.
It is so ORDERED.
1 For the Comission g.y.r. n qq, E
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SAMUEL J. CHILK Secretary of th h ission
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h Dated at Washington, DC, this.3 day of April, 1984.
t 1Comissioner Asselstine was not present when this Order was affirmed, but had previously indicated his approval.
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)i UNIT {1L@ WW4ttM i
g3m NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20:53 jg
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%..cWf JUN 2 2 1932 MEMOPANDUM FOR:
Chairman Palladino Comissioner Gilinsky Comissioner Ahearne Comissioner Roberts Comissioner Asselstine FROM:
William J. Dircks Executive Director for Operations SU3 JECT:
EMERGENCY PLANNING AND NATURAL HAIARDS By memorandum dated March 1,1982, the Secretary of the Comission requested the staff to consider several questions with regard to emergency planning.
1.
Should the emergency planning activities of NRC licensees include consideration of the possible effects on emergency plans of a very large earthquake?
It is the judgment of the staff that for most sites earthquakes need not be explicitly considered for emergency planning purposes beca'use of the very low likelihood that an earthquake severe enoug'h to disturb onsite or offsite planned responses.will occur concurrently with or cause a' re' actor accident.
Planning for earthquakes which might have implications for response actions or initiate occurrences of the " Unusual Event" or
" Alert" classes in areas where -the seismic r.isk of earthquakes to offsite structures is relatively high may be appropriate (e.g., for' California sites and other areas of relatively high seismic hazard in the Western U. S.).
2.'
If NRC requirements are to include this consideration, then what-criteria should be applied-in evaluating the adecuacy of such plans in this respect?
~
In view of the staff response to question 1, current review criteria are considered adequate.
Also the staff does not believe that rulemakinc is necessary with regard to this issue based en the analysis conducted." The Hearine 5 cards have read the Comission ruline in the San Onofre case
.(C'_I-El-33) to eliminate consideration of all'earthquaker at California sites.= The interaction of earthquakes less than the SSE with emergency preparedness was considered in the staff SER -for San Onofre.and ultimately was not a matter in contention in the San Onofre proceeding.
Cc=missioner Ahearne requested several actions be taken by the staff and' these recuests were also transmitted in the March 1, 1982, memorandum from the Secretary of the Comission.
These are addressed below.
i "For examole, Pacific Gas 5 Electric Co. (Diablo Canyon-Nuclear Power Plant, Units 1 and 2), Memorandum and Order, December 23,1981(unp0blished),
directed certification denied by Comission Order dated March 5,1982.
-O 4
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, The. Commissioners..
1.
The staff should, in conjunction with FEMA, develop an approach for checking the ability of emergency plans to cope with natural phenomena which would be expected to occur during the life of the plant.
Examples earthquakes, blizzards, tornadoes, hurricanes, tsunamis, and are:
floods that might be expected once every 40 years.
FEMA and the staff should develop guidelines for examining plans for flexibility and should identify measures which can be used to assure flexibility.
As stated in the enclosure, a site emergency plan is expected to address all the site characteristics which may require an emergency response.
Adverse conditions, which generally correspond to once in 20 to 40 yehr events, are considered in the evacuation time estimates called for in staff guidance (Criteria for Preparation and Evaluation of Radiclogical Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, NUREC-0654/
FEP.A-P.EP-1) which was developed jointly by the staff anc FEMA.
The evacuation time estimates are used in the optimization of evacuation and shelter plans as well as being available to decisionmakers in emergency conditions.
Continuing review of plans to assure flexibility is already provided by 10 CFR Part 50, Appendix E and 10 CFR !50.54(t).
2.
The staff should develop a list of the once in a lifetime natural disasters most likely for each plant either holding an operating license or in the OL process.
Because of the relatively high risk, current practice calls for California licensees and applicants.to consider the effects of earthquakes in their; emergency planning and for the Trojan plant to consider the consequences of a Mt. St. Helens eruption in its plan.
Other plants do consider adverse conditions in-developing evacuation time estimates as discussed above but a consolidated listing does not appear to warrant the effort.
3.
Existing emergency plans should be examined to determine whether adequate flexibility is present.
The emergency plan reviews and the onsite implementation appraisals which the staff has been conducting include examinations of the overall flexibility of a licensee's emergency response capability and the adequacy of evacuation time estimates, which include the consideration of adverse conditions.
Therefore, no further review is believed to be necessary by NRC.
(sipaplymitm J.Di$c' ~..
William'U. Dircks Execut_ive Director for Operations.
Enclosure:
Staff Analysis cc:
OPE OGC SECY
swessex~e BASIS'FOR CONSIDERATION OF NATURAL HAZARDS IN EMERGENCY PLANNING A fundamental premise in the approach to emergency planning utilized by the Federal Emergency Management Agency (FEMA) and the Comisison is that the emergency planning basis must be capable of responding to a wide spect rum of accidents.
This was the conclusion reached by th*e Task Force which authored NUREG-03P6 (Planning Basis for the Development of State and Local Government Radiological Emergency Respcase Plans in Support of Light Rater Nucle'ar Power Plants).
That Task Force report was subsequently endorsed by the Comission in its Policy Statement with respect to the Planning.Sasis for Emergency Responses,to Nuclear Power Reactor Accidents (Policy Statement).
44 Fed.
Reg. 61123 (October 23,1979).
The concept is reiterated in NUREG-0654 (Criteria for Preparation and Evaluation of Radiological Emergency Response
' Plans and Preparedness in Support of Nuclear Power Plants).
Consequently, as a single specific accident sequence for ~a ~ light water reactor nuclear-
~
power plant could not be identified as a planning basis, both NUREG-0396 and NUREG-0654' emphasized that the most important element of any planning basis is the distance from the nuclear facility which defines the area over which planning for predetermined action should be carried out.
Not only is this area, termed' the Emergency Planning Zone or EPZ, crucial but the characteristics of the EPZ are _significant.
The need for specification of areas for major exposure pathways is evident.
The location of the p'opiulation for whom protective measures may be needed, responsible authorities.who would carry out protective actions and the means of comunication to these authorities and to the population are all. dependent on the characteristics
~
of'the olannino areas.
(Emphasissupplied).
NUREG-0654, O
lt-is, therefore, inherent in the planning approach utilized by FEMA and the Commission, i.e., the Emergency Planning Zone concept, that the charac-teristics of the Emergency Planning Zones themselves must be factored into l
emergency planning considerations.
For example, if an EPZ is an area with sirigular adverse weather attributes, those attributes must be ~ considered in emargency planning.
This reasoning would extend to all attributes that might adversely affect an Emergency Planning Zone.
Although neither 10 CFR 50.47 nor Appendix E explicitly state that the emergency plans must account for adverse weather conditions or adverse site characteristics, such conditions are covered by NUREG-0654, which the Co= mission has adopted to provide guidance in developing plans for coping with emergencies.
NUREG-0654 calls for required evacuation time estimates to consider adverse conditions which mighi' reasonably be 'e'xpected to occur during the plant lifetime at a particular site and be severe enough to affec,t the time estimates for a particular event.
Two conditions--normal and adverse--are considered in the analyses.
Adverse conditions would deoend on the characteristics
] _,
cf a soecific site and could include floodinc. snow, ice, foc' or rain.
(Emphasis supplied).,NURE,G-0554, pp.- 4-6.
~
Thus, adverse site characteristics of a particular Emergency Planning Zone must be taken into account to satisfactorily implement'the Ccmmission's emergency planning regulations.
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Explici planning for emergency preparedness provides a base capability which can be expanded or contracted to address an actual emergency.
Backup communi-cations and feedback of damage estimates regardine transportation routes to decisienmakers after an earthquake would be generally available with or without specific advance planning.
The general planning be.se would al'lew decisionmakers 1
o chcose specific actions from amen; available alternatives for a spectrum of events.
i There is no explicit guidance in 10 CFR 50.47 or. in Appendix E to Part 50 nor in NUREG-0554 as to the extent to which adverse earthquake conditions are to be taken into account in emergency planning at particular sites..The staff, i
hcwever, believes the answer to this question is dependent upon the nature of the risk and the ' nature of the remedy to deal with the risk. - Except in California and 'other areas of relatively high seismic hazard in the Western U.-S., the staff's I
judgrent is that the nature of the seismic risk is such that no explicit con-sideration of earthquake effects is needed in emergency planning.
(Thisjudgment
, is not based on a quantitative analysis but rather cr. qualitative observations
~
cf the relatively lower seismic risk to rcads, bridges and coc:unications facilities
~
" the east versus the west.). The occurrence of.earthcuakes of.L. nature.that..-_ ___.
ccu',c have implications.for onsite or offsite res:cnse acti ns or initiate o
cccurrences of the " Unusual Event" or " Alert" class is an a~dverse characteristic of the type discussed above.
The NRC staff made requests to California facilities o censider earthcuake effects in their emergency planning, and the NRC staff also recuested FEMA to consider earthquake effects in its evaluation of offsite plans.
On the other hand, the staff concluded that additional requirements such i-I l
L
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as tne design cf adcitional facilities, structures and systems to specifically witnstanc earthquakes was not necessary for the reasons discussed above.
In particular, no special seismic design of public notification systems, environ-mental monitoring capability or communications equipment is contemplated.
Also, explicit consideration need not be given to a seismic event coincident with a significant accident at the plant from.another cause because of the very low likelihood of such a coincidence.
With respect to offsite effects at California sites, the FEMA Radiological Emergency Preparedness staff believes there should be assurance of continued cc munication between the plant and outside agencies.
In addition, the Emergency Operations Centers (EOCs) of each of the jurisdictions involved in the emergency planning effort for.a specific nuclear faci.11,ty should.
have suitably distant backup facilities to permit continued functioning of a jurisdiction's emergency response given the possible failurejof}4ts _
primary EOC.
In adcition, for California site's. the capability should exist to obtain dama;s esticates both to the plant and to transportation and communication facilities offsite to provide a data base to factor into the decisionmaking.
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process.
Finally, California licensees should have availab}e a range of recc=mendations to offsite authorities, taking into account the degree of damage to th'e plant caused by t,he earthquake' and to transportation and ccmmunication facilities offsite.
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V
.- Given an earthquake of magnitude less than or equal to the SSE, while the earthquake could have impacts upon communications and transportation as a consequence of the earthquake, the plant would li,kely not pose an imediate radiological hazard.
If, however, an earthquake substantially in excess of the SSE were to occur, then the potential exists fer a radioidgical hazard cce;1icated by tne nonradiological i= pacts pos'ed by a major earthquake.
In the view of the NRC staff, such a contingency does not warrant specific emergency planning efforts because of the general planning base capabilities discussed.above.
We conclude that this general planning base is adequate because of the remote likelihood of an earthquake substantially in excess of the SSE.
In addition, the characteristics of an accident which could theoretically be created by an earthquake substantially larger than the SSE would not be outside the s'p'ectrum of accident consequences considered in
!40 REG-0596 upon_which the judgment on planning zone sizes and other planning elements was based.
This unlikely sequence would not be unlike the case of a severe accident (not generated by an earthquake) occurring after a winter storm at a site in the ncrthern U. S.
Evacuation may not be a feasible option in such j
a circumsta.nce.
It also should be noted that to provide for a preplanned emergency response in all remote circumstances could require a commitment of 9thstantial societal resources, e.g., to assure that houses-v.d bridges would l
withstand very large earthquakes.
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NUCLE AR REGULATORY boMMtsslO*
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gj wAsmucioN. c. c. 20:55 h
JAN 131954 MEMORANDUM FOR:, Chairman Palladino FROM: '
Nilliam J. Dircks Executive Director for Operations
SUBJECT:
EMERGENCY PLANNING AND SEISMIC HAZARDS On September 9,1953, a meeting was held with you to dis' uss the Staff's c
views on the need for and extent of consideration of the poten'tially compli
~
cating effects of earthquakes in the context of emergency. preparedness, please recall that this issue emanates from the Comission's Memorandum and Order in the San Onofre proceeding, CLI-81-33, issued in December 1981,
~
in which the Comission determined that "its current regulations do not require consideration of the impacts.on emergency planning of eart
.which cause or occur during an accidental radiological release."' h:;uakes The Comission further noted that it "will consider on a generic basis whether regulations should be changed to address the potential. impacts of a severe earthquake on emergency planning" and, a memorandum from the Secretary to the
(
I In the, San, Onofre proceeding, the Licensing Board sought to raise, sua sconte, the. issue of the effects of an earthquake exceeding the Safe Shutdown Eartn-quake on the applicants' and responding jurisdictions' abilities to carry out an evacuation in a timely manner.and/or protect those in the EPZ pending evacuation.
It had been the Staff's and FEMA's positions before the Licensing Board that in that proceeding, while considerati.on of the complicating effects of' earthquakes up te the SSE was appropriate, consideration of the potential of earthquakes exceeding the SSE was not warranted.
The Licensing Boarc rejected this view and instead af. firmed.its prior position calling'for consi-deration of the potential effects of an earthquake exceeding the SSE.
There-after, the Comission, as indicated above, reversed the Licensin; Board's decision.
Parenthetically, based on the Comission's San Onofre decision,.the
~ ~ ' Licensing Board, in the Diablo Canyon proceeding rejectec a contention regarding consideration of the effects of. earthquakes on. emergency. preparedness.
In an unpublished order issued on March 5,1982, the Comission denied the Governor's request for interlocutory review of the Licensing Board's action.
The Licensing
- Board's ruling was affirmed by the Appeal ~ Board in ALAB-7ZB, slip op, at 20-21, (May 18, 1983) and review by the Comission was denied (CLI-E3-32, December 9,
-1983).
~
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Chairman Palladino.
1 Executive Director for Operations, by memorandum of March 1,1952, directed the Staff to undertake such consiceration.
By mencrandum to the Cc=issioners dated June 22,1982 (copy attache:f), the Executive Directgr responded to the questions posed in the Sects a.*y's March 1 memorandum.4 this subject, you requested After our September 9, 1983 meeting with you.onr further technical discussion to provide a rationale.for either including or not including specific emergency planning requirements for seismic events.
The following thoughts are presented to respend to your request:
~
- 1.
Off' site Damace Associated With Extreme hismic Events-Offsite damage generated by earthquakes can signific'antly affect nuclear emergency response.
Tne earthquake hazard and potential for such damage-Varies across the United States.
Severe damage, such as the failure of buildings, bridges, and other engineered structures can typi'cally be associated with large damaging earthquakes and their related ground motion. levels.
For a.large part of the U.S. east of the Rocky Mountains, where most nuclear power plants are located, such ground motion level's would be wel1 beyond the Safe Shutdown Earthquake (SSE).
Fct areas ' associated with higher earthquake hazard, such as the West Coast, these ground motion levels cculd ~be at or even less than the SSE.
Such -high hazard areas may also exist in the east (for example, the New Madrid,' Missouri., area), however, no nuclear. power plants are presently sited within these areas in the ets..
(
- 2_ The Potential ImoacCof Offsite Damace on Emercency Resocnse _
Tne impact on._ emergency response capability 3rdm earthquakes is blearly site region dependent and is generally proportional to the degree of offsite damage.
Tnat is, the higher the intensity of the earthquake, the more extensive and severe is the damage it causes.
For seismic events that result in significant
- and. widespread damage to surrounding areas, the response capability would be
. degraded through extensive disruption of transportation and communication 2To very briefly summarize ~the Staff's position as expressed in its June 22nd -
response,.the Staff concluded.that the Comission's regulations do not require amendment 'since (1) for rrEst sites there is only a very W1ikelihood that an earthquake. severe enough to disturb onsite or offsite planned responses will occur coricurrently with tr'cause a reactor accident, and (2) wnile planning for earthquakes which might have emergency preparedness implications may be warranted in areas where the seismic risk to offsite structures is relativ~ely high (e.g.,
California sites and other areas of the Western U.S.), current review criteria set forth in NUREG-0654 (which are derived from the Commission's regulations in 10 CFR 550.47) are considered adequate.
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networks, and from the failure of major stru:tures.
In this instance the range of protective actions and the capability of the offsite jurisdictions to
-initiate and implement them could be crastically reduced.
The degree of this reduction would vary ba' sed on c'ondi.tions in the region around the site.
For.
example, even with substantial damage to all. bridges, a site might have so few bridges in its vicinity that blockage of roads would.not be significant'.
3.
Plant Damace Associated With Seismic Events-When considering 'the possibilities of plant damage from seismi'c events, it is-l~~
important to understand the severity of seismic events, their--range -of probabi- ---
lities, and the potential for reactor accidents caused by seismic events.
Three classes of seismic events are considered in this discussion.. The first class includes earthquakes of relatively low ground motion, up to the Operating Basis Earthquake (OBE.).
The OBE ground motion' depends on plant location.
These. accelerations vary in the range of about.05g to.109 (higher in areas of high seismicity).
During an OBE all plant systems'would be expected to remain operating.
The second class of events includes earthquakes with ground motion higher than the OBE but equal to or less tnan.the Safe Shutdown Earthquakes (SSE); the ground motion of the SSE is typically about twice that of the OBE.
Probabili-ties of occurrence for the SSE have typically been estimated to be on the order of one in a thousand or one in ten thousand per year.
NRC regulations require that plants be designed to. achieve a safe-shutdown after an SSE.
Given an SSE, all seismically qualified equipment would be expected to function to bring"~
the plant to safe shutdown.
An earthquake up to and includiiig in'SSE hiu~Ta Bi
~
cause for an. alert emergency action level classification.
However, only in the event of a coincident failure of, a safety function (safety systems are designed for the SSE) or some undiscovered cocmon cause failure mechanism (such as a major des ~ign error) would there be a chance 'of an accident which would require offsite emergency response.
The probability of these two events (SSE and ' safety function failure) occurring simultaneously is very much lower than the probability of either one, perhaps on the order of one in a million.per reactor year or less.
The final. class of events includes all earthquakes with grourid motion levels above the SSE.
Fragility analysis is used to estimate-the probability of failure as a function of ground cotion associated with these earthquakes.
l The Ziori, Indian Point, and Limerick Probabilistic Risk Assessments estimated l
that, in general, ground motion on the order of S.5g to 0.75g acceleration would be required to damage a nuclear power plant to the extent 'that significant release of radioactivity could occur.
Of course, some plants, such as those in high seismic regions, are designed to w.ithstand earthqua'kes with ground motion
.this high; they would resist damage to still higher levels of ground motion.
l The probability estimates for such ground accelerations are significantly less l
than the probability estimates for the SSE for these plants (the Zion, IP, and Limerick SSEs are.17,.15g, and.159 respectively).
The absolute
~
I 9
probabilities for earthquakes at and bey.ond the SSE are extremely difficult to estimate and thus have large associated uncertainties.
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Cnair an Palladino 4-4 Current Emercency Precaredness Considerations Seismic events are considered and evaluated to a limited extent as part of i
our current emerg'ency planning reviews.
The following planning standards, some of which explicitly address seismic events', are addr.essed by the licensee, i
state and/or local emergency plans as explained in the following sections from NUREG-0654, " Criteria for Preparation and Evaluation of Radiological Emergency Respense Plans and Preparedness in Support of Nuclear Power Plants."
II.D.4 Emergency Class,1fication System' "Each State and local organization should have procedures in place that provide for emergency actions to be taken which are consistent with the emergency actions' recommended by the nuclear facility. licensee, taking into account local offsite conditions thit exist a't the time of the emergency."
(Emphasis added)
II.H.5.a Emergency Facilities and Equipment "Each licensee shall identify and establish onsite monitoring systems that are to be used to initiate emergency measures in accordance with Appendix 1, as well as those to be used for conducting assessment.
/
l This equipment shall include:
a.
geographical phenomena monitors, (e.g., meteorological, hydrologic, seismic);"
Eme' gency Facilities and Equipment II.H.6.a r
"Each licensee shall make provisions to acquirs data from or for l
emeroency access to offsite monitoring and analysis equipment including:
~ ~" - -
d=phasis acced) a.
geographical phenomena monitors, (e.g., meteorological, hydrologic, seismic);"
~
- I.T.10.k Protective Response "The organization's plans to implement protective measures for the clume expo ~ure pathway shall include:
s k.
Identification of and means for dealing with potential impediments (e.g., seasonal impassibility of roads) to use of evacuation routes, and contingency measures;"
For each of the emergency response classes given in Appendix 1 of NUREG-0654 severe natural phenomena (including seismic events) are included as part of the e
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example initiating conditions. The seismic events 'specifically included in this appendix are the Operating Easis Earthquake, anc the Safe Shutdown Earth-quake as well as "any earthouake felt in-p.lant or cetected on sta-ion seismic
~
. instrumentation."
Tne preceding strow that seismic events are cohsidered in emergency plan'ning but, 'as is evident, these review criteria are not very clear and clarification of them could lead to some improvements in emergency preparedness, perhaps by 1eading to more refined analysis of potential road blockage, etc.
However, it is not clear that such improvements would sub'stantially reduce the impairment' of emergency response caused by seismic damage offsite.
- The Federal Emergency Management Agency (FEMA) reviews offsite radi.ological emergency planning and preparedness to insure the adequacy of Federal, State, and local capabilities in such areas as emergency organization, alert and notif.ication, cc:x::unications, measures to protect the public, accident. assess 1
ment, public education and infomation, and medical sopport.
Detailed, specific assessment of potential earthquake consequences and response are not part of this process related to radiculogical emergencies.
FEMA does, however, have an 1
active program of earthquake preparedness which includes estimates of damage
~
and casualties, planning for Feder.a1 response to a major earthquake, and assistance to State and local governments in their earthquake planning and preparedness activii;ies.
FEMA believes that these separate activities j
would complement each other in the event that a concurrent response to a major earthquake and a serious accident at a nuclear power plant was -
~
i required.
5.
Risk Persoectives _.
Recent pRAs (e.g.,' Zion, Indian Point) have indicated that very large earthquakes (much greater than'the SSE) can dominate the risk from a nuclear power plant.
Such earthouakes can cause massive plant damage leading to imrnediate offsite
.radiologi' cal hazards.
In addition, massive offsite damage was assumed in these'-
analyses which substantially degraded the emergency response.' ' -
Based upon the PRA results, the staff finds that for most earthquakes (including some. earthquakes more severe than the SSE) the~ power plant would not be expe'eted to pose an irxnediate offsite radiologica1 hazard.
For earth' quakes which would cause plant damage leading to irxnediate offsite radiological hazards but for' which there would be relatively minor offsite damage, emergency response capabilities around nuclear power plants would no*. be seriously affected.
For earthquakes which cause more severe offsite damage, such as, for example,-
, disabling.a. siren alerting system, the earthquake itself acts as an alerting system.
For those risk dominant ' earthquakes which cause very severe damage to both the plant and the offsite area, emergency response would have marginal -
l benefit because of its impairment by offsite damage.
The expenditure of additional resources to cope with seismically caused offsite damage is of doubtful value considering the modest benefit in overall risk reduction which
~
could be obtained.
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. Chai rman Falladino 1
~.
.6.
Sum.arv Based on the preceding discussion the following sumary points can be made:
In general, earthquakes up 'to ahd including the SSE. are net expected a.
to pose an inmediate offsite radiological hazard.
b.
Earthquakes beyond the SSE may cause plant damage and radioactive release under. concitions where offsite damage impairs emergency response.
- " c.
Further clarification or' refinemen't of cu'rrent requirements and guidance 1
might reduce the impairmeht of imergency response indicated in b. above, but the value of such reduction is uncertain.
Ci4;nND William iEkoks Nilliam 'J. ' Dircks Executive Director for Operations i
Attachment:
As stated ec:
Co=issioner Gilinsky Comissioner Rober s Comissioner Asselstine' Comissioner Ser--.al -
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wAss moros. o.c. 2csss OFF6CE OF THE CHAtAMAN The Honorable Morris K. Udall Chairman Subcommittee on Energy and the Environment Comittee on Interior and Insular Affairs United States House of Representatives Washington, D.C.
20515
Dear Mr. Chairman:
On March'22, 1984, Dr. Henry Myers of your staff submitted to the NRC staff a list of coments and questions, derived from his review of NRC Inspection Report No. 50-275/83-37 and its draft. This list of coments and questions was revised by Dr. Myers and submitted to the Comission on March 26, 1984 The stiff has reviewed the March 22 and March 26, 1984 lists and notes that all items contained in the March 22, 1984 list are included as a subset of the March 26, 1984 list.
The staff has developed written responses to all of the items contained in the revised list of coments and questions, dated March 26, 1984.. These are attached for your information.
I trust that the staff has been responsive to these inquiries..
' cerely, Nunzio J. Pa adino Chairman b
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1.
Ouestion "GC-1.
A statement to the following effect is made repeatedly with respect to the Region V method used to inquire into the HSC findings:
"The inspector's approach to resolving this issue was to assess the validity of the NSC finding and Pullman response, and evaluate the NRC findings for confomance with the specified Pullman program."
(E.g.
83-37, Item 24.) -This implies that there is a documented Pullman response to the NSC finding.
(E.g. "The licensee conducted an audit of Pullman, during the period of April 2 through June 1,1978, in response to the NSC a'udit and the Pullman response." See Draft 83-37, p. 37.
This statement does not appear in the Final 83-37.) Where is the Pullman response? What interviews were conducted with PG&E, Pullman, and NSC-past and present personnel in the course of preparing 83-377 How were such interviews documented? Where is the documentation?"
Answer The Pullman response to the NSC audit report was submitted to PG&E by Pullman on April 11, 1978.. This response m fomally submitted to the Atomic Safety and Licensing Appeals Board by PG&E as an attachment to the Affidavit of Russell P. Wischow, dated September 21, 1983.-
To better understand the response to this t.uestion, and those which follow, some background may be helpful.
On September g,1983, a filing to the Diablo Canyon Atomic Safety and
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Licensing Appeals Board was made by "(Joint Intervenors" which included an audit of the Pullman Power Products PPP) quality assurance program performed by Nuclear Services Corporation (NSC) and reported on October 24, 1977. The Pullman and Pacific Gas and Electric responses to the NSC audit were dated April 11, 1978 and June 16, 1978, respectively.
One of the significant aspects of the NSC audit was that it was almost exclusively limited to a records and paperwork review as opposed to a hardware or personnel perfo'rmance review.-
The NSC report contained many critical findings and drew far reaching conclusions. Both the Pullman and the PG&E responses took issue with many of the findings and conclusions of the NSC report. Our review of our own inspection reports over the years did not seem to corrob' orate many of the NSC conclusions.-
The Region V Administrator elected to approach this inconsistency by examining, in depth, a large sample of the most significant NSC findings and the associated Pullman and PG&E responses to determine whether the
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NSC conclusions and findings could reasonably be drawn from the QA records which NSC reviewed.- The Region V examination was limited to this sample and did not constitute a comprehensive reconstruction of the entire Pullman activity at Diablo-Canyon.
Consistent with this logic, i
our inspection did not rely, in any appreciable way,-on personnel interviews; consequently no transcripts, tapes,' etc., were made. Summary sheets do exist in our inspection file for three discussions involving five individuals. These discussions did not contribute appreciably to h
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the NRC conclusions.
The discussions are referred to in inspection report 83-37 in paragraphs 20 and 25.
In' summary, as discussed in several Commission meetings and in Report 83-37, the inspection of the NSC report was limited to a sampling of the significant NSC findings to determine if they could be reasonably supported by QA records. The NSC audit was almost exclusively based on review of QA records. The NRC examination was similarly focused.
Except as noted in Report 83-37, we found no reasonable basis to expand the limits of the review.
2.
Question "GC-2.
Inspection Report 83-37 refers to corrective actions taken in response to the NSC audit.
It is unclear in certain instances as to whether the corrective actions were taken with respect-to QA deficiencies that existed prior to the audit; e.g. to what extent did the corrective actions involve activity to insure that inadequate workmanship did not escape detection as a consequence of the QA deficiencies that existed pr'ior to the NSC audit."
Answer.
It is difficult to address this comment in detail in.the absence of 4
specific examples of concern. However, a few general comments addressing this item may be made.
l The vast majority o'f NSC findings involved some kind of paperwork; ie:
program, procedures or instructions rather than workmanship issues.
In writing the report,'and during the conduct of the inspection, the staff made every~ attempt to address _ and assess not only the adequacy of prospective work but also the degree of retroactive back-fitting that was appropriate. Therefore, the staff did consider the applicability of the NSC findings and Pullman responses to previous work.
Thestaffati.emptedtomakeclearthatthemajorityofcorrectiveactions taken as a result of the NSC report were programatic improvements, or amplification of existing program descriptions, and did not necessarily condemn work performed prior to the improvement.
In each case in the inspection report, the staff feels that, whenever.a programatic improvement was made_ subsequent to the NSC audit, the NRC made the.
findingthattheprogrampriortothejmprovementwasadequateorthatno evidence was found to indicate that the.progcam prior to the improvement resulted, or wou'1d likely have resulted, in an inadequa'te implementation-ccndition. ;For example:
a.
In paragraph 8 of the report the foliowing conclusion is stated "The inspector found the QA program elements describing hanger package review and weld preheat were adequate and met the applicable code requirements," even though programatic improvements were effected j
subsequent to the NSC audit.
1 3
b.
In paragraph 15 ~of-the report the following conclusion is stated "Furthermore, there is no evidence in the NSC, PG&E, and Pullman
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3 corporate audits to suspect that any field changes made to pre-1977 i
documents and records impacted adversely on the quality of field construction," even though programmatic improvements were effected subsequent to the NSC audit.
In summary, during the conduct of the inspection the inspectors considered.the effect on previous work during their examination of each item and programmatic improvement.
3.
Question "GC-3.
Inspection Report 83-37 contains several references to the 90 day welders' log. Does the NRC have the log in its possession? If not, is it readily accessible? Where is it? What deficiencies exist in this log vis-~a-vis the ASME code?"
1 Answer The NRC does not hav$.the 90 day welders' log in its possession. The' log is readily accessible and stored in the Pullman QA records vault at the Diablo Canyon site.
The ASME Code does not mandate the use of a 90 day welders' log. The ASME. Code only requires some sensible. method of keeping track of welder.
qualification and welder activity. The 90 day welders' log is the mechanism adopted by Pullman, at Diablo Canyon, to track the welders, employed by Pullmaq,. which were qualified during a particular time period.
It has as its basis"tbe original welder qualification record and the use:of a particular process during a predefined previous period, as derived 'from the weld filler metal withdrawal slips. As discussed in paragraphs 9, 18, 19, 20'and 22 of Inspection Report 83-37, the staff found the Pullman system adequate in fulfilling the requirement of the ASME Code.
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4.
Question "GC-4.
InspectionReport8f-37statesinseveralplacesthatPullman practices were " consistent" with the ASME code. Does " consistent" mean "in compliance with"?:-Is it the NRC position that wherever " consistent" is used that it may be replaced by "in compliance with"?"
Answer Yes, as used in inspection report 83-37.
5.
Question "GC-Sm There is no indication of Region V having sought the views of NSC'
'either to elaborate on the 1977 Jindings or to coment on the findings and conclusions of the Region V inquiry."
Answer Region V has actively sought ihe views of NSC (now named Quadrex).
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On two occasions the staff sought the views of the team leader (who is no longer associated with Quadrex) and of the President of Quadrex by telephone.
In both cases they could not recall details. The details of these calls are contained in the transcripts of the Comission meetings of March 26, 1984 and April 13, 1984.
requested on April 4, 1984 that Quadrex review Further, Region V fonnally(Nos. 50-275/83-37 and 50-323/83-25) and the NRC Inspection Report extended the NRC invitation to appear at the Comission meeting on April 13, 1984.
The President of Quadrex responded by letter dated April 9,1984 and indicated that he didn't believe Quadrex could add substantive i
information regarding the differences in the audits at this time.
The Comission again requested Quadrex to meet with them in a letter dated May 18, 1984, and they have agreed to meet.
6.
Question "GC-6.
Page 3 of the draft states a sample of 25 stainless steel welds were sampled for delta ferrite and that 100 radiographs were selected to verify field weld and inspection review adequacy. What is the basis for selecting these welds? On what dates were these welds produced? Did these welds represent an adequate statistical sample?"
Answer The basis for the sampling done was as stated in the report (83-37) on page 3..."to provide an independent feel for the Pullman work rather than solely relying on information provided by licensee' records."
In the instances cited (dealt with in paragraphs 25 and 33 of report 83-37) the inspector's conclusions were not dependent on the independent sampling.
The sampling was not meant to be, nor was it advertised to be, statistically rigorous but was as stated in paragraph 25 "an additional check...."
7.
Question
" Criterion I, NSC Audit Finding 3.
(Final p.3, Draft, p.2-5.):
Did the fact of QA personnel writing and ' approving Engineering Specifications, performing welding engineering functions; and approving welding. engineering changes constitute a violation of Appendix B requirements?"
Answer No.
8.
Question
" Criterion II, NSC Audit Finding 4.
(Final p.4-5, Draft, p.2-5.):
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5 Is it the NRC conclusion that upper management performed scheduled reviews of nonconformance reports, personnel qualifications, and corrective actions as required by NRC regulations for the time periods addressed by the NSC audit? Note handwritten notation in draft report:
"In conclusion, factual records do not support the NSC finding." The corresponding statement in the final report is:
"The inspector concludes the historical records of corporate management audits do provide evidence that reviews of nonconformance reports, personnel qualifications and corrective actions were performed."
Note comment in final report:
"In addition, Pullman Power Products has since proved programmatic I
improvements..." etc. What was the program pr%r to the improvements?
What was it after the improvements were instituted?"
Answer Yes, it is t'he NRC staff's conclusion that upper management did perform periodic reviews of nonconformance reports, personnel qualifications, and
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corrective actions as required by NRC regulations for the time periods addressed by the NSC audit. This was stated in paragraph 6 of inspection report 83-37.
It is important to keep in focus the purpose of the NRC inspection. As stated previously in answer to GC-1, that purpose did not include a diagnostic evaluation of the entire Pullman QA history. Accordingly, we did not compile a description of the Pullman program for each point of time irrits evolution.
In addition, the NRC inspection did not identify 4
.anything which woul,d indicate a need for such a total reconstruction.
9.
Question "CM terion V, NSC Audit Finding 1.
(Fi'nal p. 5, Draft, p. 39-40.):
NSC stated:
"There is no requirement that activities affecting quality shall be prescribed by documented instructions, procedures, and drawings." Region V states, apparently in reference to fabrication of piping assemblies and erection of pipe in the plant, that KFP-8 established appropriate instructions and procedures.
Region V seems to imply that KFPS-7 established procedures for pipe supports.
KFPS-7, however, was not issued until December 1973. Were.the QA procedures for installation of pipe procedures prescribed by documented instruction, procedures etc. prior to December 1973? Moreover, the draft sta'tes that-ESD-264, dated 9/15/78, provided a specific procedure "to implement precisely the QA program elements of KFP-8 and KFPS-7." -The latter statement does not appear in the final report.
In that the specific procedures for implementing " precisely the QA program elements of KFP-8 and KFPS-7" were apparently.not promulgated until only September 15, 1978 what is the basis for assurance that KFP-8 and KFPS-7 were adequately implemented prior to September 1978."
Answer A brief history of the Pullman QA program applicable to pipe supports would be helpful here. The first pipe support work began during August 1971 with work begun on the first non-safety related pipe support (August
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6 5, 1971) and safety related pipe support (August 16,1971). As a result of a PG&E QA audit, performed during late 1973, it was identified that Pullman did not have a QA program covering the installation of pipe supports and that the QA program for the installation of pressure boundary piping was not fully applicable to pipe support work. A stop work order was issued on pipe hanger / rupture restraint work until an approved QA program covering pipe support / rupture restraint work was implemented.
1 As corrective action Pullman procedure KFPS-7 was issued on December 3, 1973 establishing and implementing a pipe support QA progiam for process planning and control.
In addition, a Pullman Discrepancy Report (Nonconformance Report) was issued on February 11, 1974. This Discrepancy Report recognized that pipe support work was perfonned prior to establishing process planning and control. As corrective action all Class 1 pipe supports installed without process control were identified, reinspected and inspection findings resolved.
In the inspector's judgement, procedures KFP-8 and KFPS-7 were entirely sa'tisfactory~and met NRC requirements. This is documented in NRC Inspection Report 50-275/83-37, paragraph 7 which states, in part, that "The inspector concludes the program elements of KFP-8 and KFPS-7 did establish that documented instructions and procedures were required to be prescribed for control of Pullman's quality related construction activities." The establishment of ESD-264 subsequent to the NSC audit provided a programmatic improvement to an already acceptable system in this area, in that,the details of process sheet completion were more precisely defined and prescribed by the ESD-264.
In addition, the NRC has contracted with Lawrence Livermore National Laboratory-(LLNL)~to provide additional' inspection services, in the area of pipe support inspection, to supplement the regional effort. The laboratory inspectors have already examined a sizeable sample of Unit 1 pipe supports and found a very low discrepancy rate on accepted pipe supports.
For example, the NRC staff and LLNL inspected about 550 safety related pipe supports, out of a total population of about 4300 modified supports, and identified only 5 items of noncompliance. The results of the laboratory inspections provide additional assurance regarding pipe support acceptability. Therefore, the staff feels that the licensee and Pullman have effected a satisfactory pipe support. installation program.
- 10. Question
" Criterion V, NSC Audit Finding 2.
(Final p. 5-6, Draft, p. 40-41):
NSC sta'tes 'that hanger package review was not describ'ed in procedures.
Region V states that hanger package review was described in KFPS-2 dated December 3,1973 and that supplementary requirements ~ were incorporated into ESD-254-dated December 30, 1977. What was the basis for reviews conducted prior to December 3, 1973? The draft, but not the final report, states that ESD-253 provided additional detailed information concerning hanger drawing controls.
What is the date of ESD-253?. Is it NRC's position that hanger package review was described in a manner that s.
7 complied with the Appendix E requirements for all periods covered by the NSC audit?
HSC states that other activities not described in procedures included preheating for welding, use of Note-0-Grams, use of Rejection Notices, and maintenance of Field Quality Inspector Daily Logs.
Is it the NRC's l
position that all such activities were described in procedures in a manner that complied with the Appendix B requirements for all periods 4
covered by the NSC audit?"
Answer Refer to the answer, provided above, to Criterion V, NSC Audit Finding I regarding pipe support QA program history. The Pullman Discrepancy Report was resolved, as indicated previously in answer to the question regarding Criterion 5, NSC Audit Finding 1.
It is the NRC staff conclusion that hanger package review was described in a manner that complied with Appendix B requirements during the time -
periods when hanger package review activities were in progress.
4 The staff found that preheating was appropriately prescribed on the welding procedure specification (see paragraph 28, page-26, of inspection report 50-275/83-37).
It is the NRC staff's conclusion that other activities (such as use of Note-0-Grams, Rejection Notices, and maintenance of Field Quality Inspector Daily Logs) were not required to
.be prescribed and controlled by written procedures, as indicated on page 6 of inspection report 50-275/83-37.
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- 11. Question
" Criterion V, NSC Audit Finding 3.
(Final p. 6-8, Draft, p. 41-42):
NSC found that isometric package review was not sufficiently described.
The draft of 83-37 states that " Field procedure ESD-254 (issued 5/6/75) appears to provide an adequate outline guide for review of isometric drawing packages." The final report adds that May 6,1975 was the earliest date that could be found for ESD-254 and that while most piping installations had been completed prior to May 1975, the inspector found that the final complete document review of isometric drawing pac,kages were performed after ESD-254 was in effect.
Note that draft states that post heat treatment requirements are prescribed in ESD-218. The draft does not indicate that ESD-218 was 4
issued in' October 1977. The-final report states that post weld heat treatment requirements "were always prescribed by weld procedure-specifications." The final report does not refer to specific procedures in effect prior to ESD-218. What is the basis for the conclusion that-post weld heat treatment requirements were in compliance with Appendix B
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prior to issuance of ESD-218 in October 1977?
Is it the NRC staff position that, in the time period encompassed by the NSC audit, non-conformance reporting requirements complied with the requirements of Appendix B?
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8 83-37 states that the " internal audit program, implemented by on-site personnel, (prior to 1978) was determined to be of marginal quality, a redundant program of comprehensive quality was performed concurrently."
The redundant program appears to have been one " directed and conducted by j-corporate management personnel." Did the redundant program find that the t
internal audit program, implemented by on-site personnel to be of
" marginal quality?" Did the corporate audits encompass involve review of weld and welder quality, and Q.A. programs applied to weld and welder i
quality? On what dates were the corporate audits conducted and what were i
their findings?
Region V concluded that, notwithstanding the deficiencies in the internal audit _ program and the failure of the corporate audit program to discover these. deficiencies in a timely manner (e.g. they appear not to have been corrected until 1978),83-37 concludes that "no major breakdown of the Quality Assurance program hao occurred, nor had any significant problems 9one undetected, due to deficiencies identified with the internal i
auditing program." Is it the NRC position that the NSC findings do not indicate that "significant problems (had) gone undetected" until 1974 l
and, to a lesser extent, between 1974 and 1977?"
Answer The basis for the NRC staff's conclusion tha't post weld heat ' treatment l
requirements were in compliance with Appendix B prior to issuance of l
ESD-218 in October 1977 is as stated in the NRC inspectio'n~ report,
" Appropriate post weld heat treatment-requirements were always prescribed by welding procedure specifications."
It'~1s~ the NRC staff's conclusion that nonconformance reporting requirements complied with the requirements of Appendix'B.
Pullman procedure KFP-10 (issued March 19,1971) did provide adequate instructions to establish nonconformance reporting requirements in compliance with Appendix B.
The statement that the-internal audit program (i.e. those conducted by on site' people) was of marginal quality was the opinion of the inspector.
His basis was that the breadth and frequency of the internal audits were not entirely consistent with'today's guidance. The inspector went on to say that the audits conducted by the corporate people compensated for this and the program as a whole met Appendix B requirements. As far as the inspector can remember the frequency.and breadth of the internal audits was not connented upon by the corporate ' audits.- Since the-
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inspector. conc'luded that the Pullman audit program as a' whole met all requirenients, a ' detailed catalog of all findings, dates, and resolutions.
was not made.
As stated in the last paragraph of. item 9 in Inspection Report 83-37, it
'is the staff's position that the Pullman audit program as a whole met the i
requirements of Appendix B.
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- 12. Question
" Criterion VI, NSC Audit Finding 9a. (Final p._8-9, Draft, p. 54-55):
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NSC stated that a Process Sheet for Isometric.2-14-77 was changed approximately 19 months after the work was done.
Region V states that the process sheet (which Region V states should have been in reference to 2-14-47) that "no' evidence could be found to indicate that there had been an attempt to alter the dates or signatures on either or both these documents." Does NRC believe the NSC finding to have been in error? How does NSC explain the apparent discrepancy between its findings and those of the NRC? How does Region V know that the sheets it examined were in fact the same sheets examined by NSC?"
Answer Yes, the NRC staff believes that the NSC finding was in error.
In i
particular, isometric package 2-14-77 did not fit the time frame identified in the NSC finding.
The efforts of the NRC to solicit NSC review and comment on the NRC' inspection report have been dealt with in answering question GC-5. These are precisely the types of details that neither the-former auditor nor the President of Quadrex could recall.
4 The NRC staff examined the available records for the referenced isometric packages and found no basis to conclude that-the records had been altered since the NSC audit.
- 13. Question
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" Criterion VI, NSC ' Audit Finding 9b.
(Final p. 9-11. Draft, p. 61-64):
4 NSC concluded-that FW-1673 was performed without normal controls. Region i
V stated in the draft, that "although it was not the usual practice" the weld was' carried out in accordance with the then existing design change control system.
Is it the NRC position that this departure from the j-usual practice did not violate the NRC's QA requirements? What is NSC's l
response to the NRC finding 7" Answer i
It is the NRC staff's conclusion that the referenced departure from usual i
practice was adequately ~ controlled and did not violate the NRC QA requirements.
- 14. Question
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" Criterion VI, NSC Audit Finding 10.
(Final p. 13-15. Draft, p. 38-39):
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NSC found that no procedu e'or requirement prohibits changing or alteration of records and documents necessary to' track work. Region V.
stated that prior to 1977, insufficient requirements existed to control the changing or alteration of quality records and documents. Region V also concluded in the final report that neither NSC, nor NRC nor Pullman audits had " identified any unapproved technical changes or other substantive changes which would have adversely. affected construction quality." What was done to reach this conclusion? Note that the 3
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i 10 conclusion as stated in the draft was less firm; it said:
" Pullman's corrective action is complete and appears to be effective.
Previous inadequacy of management policy or written instructions in this area is not considered to have resulted in any adverse impact on quality related activities."
Answer As stated in Inspection Report 83-37, the staff examined the results of Pullman audits, the NSC audit and PG&E audits and related QA records.
In addition, the NRC inspection staff was aware of this NSC finding and the documents examined during the inspection were reviewed with particular attention given to unapproved or substantial technical changes.
- 15. Question
" Criterion VIII, NSC Audit Finding 12.
(Final p. 14-15. Draft, p. 59)
NSC stated that ESD-223 did not give adequate instructions for the identification and control of Class I pipe supports.
Region V reviewed ESD-223 and stated that specific revisions were dated November 11, 1975 and May 25, 1976. Region V stated that the procedure revisions contained adequate QA/QC instructions for the control and identification of Class I pipe supports.
Is it the NRC position that the instructions were adequate prior to the 1975 and 1976 revisions? What is the basis for confidence in the adequacy of instructions prior to the 1975 and 1976 revisions in ESD-22,3?."
Answer The Region' considers that instructions iegarding the control and identification of pipe supports were adequate prior to the referenced revisions of ESD-223. The basis for this conclusion was discussed in the answer to the question regarding Criterion V, NSC Audit Finding 1.
- 16. Question
" Criterion IX, NSC Audit Finding 10b.
(Final p. 17-18, Draft p. 6-7.):
NSC found that from August 1972 through December 1972 a ninety day Welders' Log was not maintained nor did a Weekly Qualified Welders' List exist for that time. Region V agreed there was no weekly log but that a 90 tiay log did exist. Did Region V seek to determine the reason for the discrepant findings?"
Answer Yes, the NRC did determine the source of the discrepancy..The void in the 90 day log had been reconstructed by Pullman subsequent to the NSC audit by using the weld rod withdrawal slips for the period in question.
It should be noted that the 90 day log is normally made up using these weld rod withdrawal slips. This seeming discrepancy was dealt with in SSER-22 and was discussed in the March 26, 1984 Comission meeting.
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11 17.
Question
" Criterion IX, NSC Audit Finding 10c.
(Final p. 18-19, Draft p. 7-8):
NSC found that the Ninety Day Welders' Log was not sufficiently detailed to determine if the welder was qualified to perform certain procedures.
The draft stated that the "90 day qualified welders' log was sufficiently detailed to determine whether a welder was qualified to perform certain procedures." The draft did not state that the 90 day welders' log complied with applicable code requirements. The final report, states, depending upon the manner in which a sentence is interpreted, either that the log complied with the code or, alternatively, that the welder complied with the code requirements.
Is it the NRC position that for the period covered by the NSC audit that the 90 day welders' log was in compliance with code requirements?
Note also that Region V based its conclusion in part upon discussions with the former Authorized Inspector.
Does a record of these discussions exist? What was the substance of such discussions?"
Answer The ASME Code does not require a 90 day log - only that some reliable method of determining welder activity be maintained.
It is the staff's position that the 90 day log-based on the weld rod withdrawal slips constitutes a reasonable method of complying with this requirement.
The discussion with' ~the former Authorized Inspector was not relied upon in reaching the conclusions presented in the NRC inspection report because the former Authorized Inspector was, at the time of interview, an employee ~ of Pullman. The discussion re'sults were merely considered another data point recognizing that, while the information was given by an industry professional, the information may be of dubious value. The inspector relied instead on the results of his examination of the 90 day welders' logs.
A record of that discussion exists in the Region V files as discussed in answer to GC.-l.
The substance of the former Authorized Inspector's information was that, in his opinion, Pullman had adequately tracked and documented welder qualifications and had used weld rod withdrawal theets to verify whether t
a welder had used a particular process as a supplement to the ariginal welder qualification record; This is stated in report 83-37.
- 18. Question
" Criterion IX, NSC Audit Finding 10d. (Final p. 19-20, Draft p. 9-10):
NSC states that no procedure stated what the Field Quality Assurance -
Inspector was to use as the primary means to determine welder qualification.
Region V appears to agree that a procedure did not exist but that weld filler metal withdrawal sheets and welder qualification records were used to determine welder qualification and that this method satisfied code requirements?
Is it the NRC staff position that, the absence of a specific procedure notwithstanding, the method used by j
12 ir.spectors to ascertain welder qualifications complied with code requirements?"
Answer Yes, the staff considers that the method historically used by Pullman (i.e., weld filler metal withdrawal sheets and welder qualification records) was sufficient and adequate to document and verify welder qualification, as required by the ASME B&PV Code,Section IX (refer to paragraphs 9 and 21 of inspection report 50-275/83-37).
- 19. Question
" Criterion IX, NSC Audit Finding 10h,101.
(Final p. 22-23, Draft p.
13-16):
The draft focuses on question as to whether auditors' observations need be recorded on the " process sheet or the inspectors' daily work sheet."
The draft does not indicate that the inspector examined the welder audit sh'eets. The final does state that the inspector examined welder audit sheets but does not indicate the period covered by the examination.
The final version of 83-37 states in 10h that the welder audits were "a Pullman program requirement in excess of the ASME code requirements" and twice in.101 that the. program requirements " appeared" in excess of code requirements. The DRAFT did not mention that the code did not require a welder audit.
Draft 101 (p. 15) s'ays "... records of the 9/73 revision and 11/73 implemented procedure are not available."
Final drops this part stating (p. 23) "The November 1973 revision app'arently was issued and implemented beginning'in November 1973.
... welder audit sheets indicate that the 4
required welder audits were performed beginning November 1,1973." The following statement appears in the draft but not the final:
"The welder audit sheets examined indicate the ferrite control measurements were performed on. welds by the auditors." Why was this statement dropped? Is the statement accurate? Was there a requirement to make ferrite control measurements?
What is the significance of failing to adhere to ESD-219 if the ASME code does not require welder ~ audits?
Note fol. lowing statement in draft does not appear in final:
"Since~the record of the 9/73 revision is not available, the inspector could not determine when the procedure was approved for implementation and, thus, was not'able to corroborate the Pullman-statement that the September 1973 revision was made to initiate the auditing of welders." The draft and final state that "the inspector was not able to corroborate the NSC statement that Pullman was in noncompliance with the procedure for about 23 months."
Is the staff's conclusion that neither Item 10h nor Item 101 were identifiable items of noncompliance or deviation rest on the assumption that welders' audits were not required by the ASME code?"
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13 Answer The statement regarding ferrite control measurements was dropped from the final because it did not contribute in any meaningful way to addressing the finding identified by NSC.
The draft version did not maintain a focus on the issue addressed by the NSC finding addressed in paragraph 24 of NRC Inspection Report 50-275/83-37. The statement is, however, accurate.
Ferrite control measurements were included as one of several suggested inspections on the welder audit sheet. The welder audit program was structured around a sampling approach and applied to in process welding activities. There was no requirement that each welder audit performed should address each and every suggested inspection attribute identified on the welder audit sheet.
For example, even though the same welder audit sheet format would have been used for both carbon steel and stainless steel welding activities, a measurement of ferrite level on a carbon steel weldme'nt would be quite meaningless. The intent was that a welder audit should sample the suggested attributes with emphasis on those suggested attributes.which could be meaningfully examined at the time of the welder audit performance.
The NRC staff did not assess the significance of failing to adhere to the welder cudit program of ESD-219 because the NRC found that Pullman did acceptably implement the ESD-219 specified welder audit program.
.The. staff's conclusion that Pullman had acceptably implemented the welder audit program was tih'e basis for the determination, in items 10h and 101, that no items of-noncompliance or deviations were identified. The fact that the Code does not require such a welder audit program has no bearing on the-finding of acceptable implementa' tion, but the Code was mentioned merely to provide additional perspective.
'20.
Question
" Criterion IX, NSC Audit Ffo. ding 10j.
(Final p. 23-24, Draft p.16-18):
Note change from draft which relied on the examination of 25 welds to find that "...there is a high probability that other stainless steel welds installed in the plant comply with delta-ferrite acceptance criteria." The final report cites a " random sample of 25 stainless steel welds" as "an additional check".
Primary reliance for the final report's conclusion that "the inspector was not able,to corroborate that Pullman was in noncompliance with this procedure requirement for-12 months" was based on the assumption that stainless steel welding did not begin until early.1973.
If it is true ' hat on. site stainless steel welding did not begin until 1973, what is the relevance of the examination of the 25 welds since the NSC finding applied to the pre 1973 period?
Is there a documented basis for the statement " Based on discussions with PG&E personnel it appears that stainless steel welding on site began in early 1973?"
Answer 9
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Contrary tc the characterization above, the staff's conclusion was not
~
based on the assumption that stainless steel welding did not begin until 1973.
The staff's conclusion was based on the fact that ESD-219 became effective in November 1973 yet the severin gauges were on site as early as December 1972.
21.
Question
" Criterion IX, NSC Audit Finding 10k.
(Final p. 24-25, Draft p.
120-121.):
The NSC finding that " Hangers are not welded in accordance with Pacific Gas & Electric Company requirements" was not confirmed. Did NSC err in observing that hangers were welded to structural steel on the wrong side of the bracket? What was Pullman's response to the NSC finding? Would NSC agree that an error.of this kind would be made in the audit? Was an effort made to determine whether the hangers might have been modified following the-audit?"
j Akswer Yes, the staff believes that the NSC finding was in error. This conclusion is also stated in the staff response to this NSC finding
(
Reference:
NRC Inspection Report No. 50-275/83-37, page 24).
i The Pullman response states, in part, that, " Pullman inst ection personnel have reviewed Hangers No. 2023-IV and 2039-2V and found that they were welded in accordance with customer drawings."
TheeffortsofthestilfftosolicitNSCreviewandcommentontheNRC inspection report'have been dealt with in answering a previous question (GC-5).
Yes, the staff examined the available records for the referenced hangers and found no. evidence to conclude that the hangers had been modified or reworked after the NSC audit.
- 22. Question
" Criterion IX, NSC Audit Finding 10n.
(Final p. 26-27, Draft p. 20-21.)
N$C found that there was no procedure for preheating weld joints. The draft report (p. 21) states that a series ofsweld procedure specifications was examined and that each contained "ari adequate definition of preheat, postweld heat treatment and interpass.
temperatures." Thedraftalsostatesthat"ESD-218(PostweldHeatand PreheatTreatmentProcedure)wasrevised 12/30/77 to prescribe preheat requirements and indicate preheat applicability." An adjacent handwritten comment -(p'. 21) asks "How about b/f 12/30/77?" Does this mean that the procedures were or were not adequate prior to 12/30/77?
The final report (p. 27) contains an additional statement to the effect that prior to early 1978, compliance with the preheat requirement was dependent upon the welder's knowledge etc.
Did the procedure described
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in the second paragraph on p. 27 comply with Appendix B? What was the basis for the added language? Was there discussion with Pullman or PG&E j
on this point beyond that which occurred during the inspection that ended l'
on December 9, 1983?
The penultimate paragraph on this item states "while no separate and specific procedure for preheating of weld joints existed prior to l
December 30, 1977, preheating requirements were adequately prescribed by the welding procedure specifications and documented by signature on the welding block of the process sheet, which specified the applicable welding procedure." Was this in compliance with Appendix B7" Answer The handwritten coment does not mean that preheat procedures were 1
inadequate prior-to December 30, 1977. The handwritten coment was made by the Region V Administrator to instruct the inspector to make clear the l_
situation that existed prior to December 30, 1977.
\\
The finding that preheating was adequately prescribed is documented in paragraph 29 of the NRC inspection report which states "The inspector concludes that, while no separate and specific procedure for preheating i
of weld joints existed prior to December 30, 1977, preheating requirements were adequately prescribed by the welding procedure specifications and documented by signature on the welding block of the process sheet, which specified the applicable welding procedure."
l This was in complia'nce with Appendix B, hence the finding in paragraph-29 of the NRC inspection report that "No items of noncompliance or i
deviations were identified.."
The added language of the secon'd paragraph on page 27 was to clarify the preheat prescription, implementation and documentation process.. To the p
best of the inspector's recollection, there was no further discussion with Pullman or PG&E on thi.s point beyond that which occurred'during the j
inspection, which ended on December 9, 1983.
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- 23. Question 1
~
" Criterion IX, NSC Audit Finding 10o.
(Final p. 27-30. Draft p. 21-26.)-
NSC stated that the initial results of welding auditing (from November 5,
-1973 to February 1974) indicated the existense of 7 problems which, if i
they did exist, raised question about weld quality. NSC concluded on the basis of a review of these audits 'that "...there is no confidence that welding done prior to 1974 was performed in accordance with welding 3
j specification requirements."
i The NRC inspector said he had " critically examined the records of welder audits performed between November 1, 1973 and April 1, 1974." On the basis of an examination of 183 audit records from this period, the NRC 4
inspector concluded that the " aggregate of problem areas is not so pervasive _such that support can be given to the NSC conclusion" that L
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there is no confidence that pre-1974 welding had been performed in accord i
with requirements.
83-37 states that "It is important to recognize that none of these were NSC findings, but were instead findings of the Pullman welder audit program, which was designed to detect program weaknesses and provide prompt corrective action during the early phases of site welding activity." The problem is that the welder audits referred to by the Region V inspecter (which were found by Region V under 10h and 101 above to be beyond what was required by the code) were not initiated until November 1973.
In addition, the NSC audit states that.its findings were based on a review of Pullman's audits conducted in the period "from November 5,1973 to February,1974). Therefore, how could the audit program upon which Region V relies " detect welding program weaknesses and provide prompt corrective action during early phases of site welding"
'if the audit program was not initiated until November 1974?
4 l
In sum, the NSC finding, based on findings obtained from a review of
~
audits conducted after November 1, 1973, was that "...there is no co'nfidence that welding done prior to early 1974 was done in accordance with welding specifications." Region V, on the other hand, based on a i
review of audit reports prepared during essentially the same period as the reports reviewed by NSC [and ignoring the above noted finding (final,
- p. 23) that "the required welder audits were performed beginning November 1,1973"] concludes "no support can be given the [above quoted] NSC conclusion." Region V does not deal with neither (A) the fact of there. k having been no welder. audits prior to November 1973 nor (B) the question I
j of whether the type's of deficiency discovered in the initial audits j
existed in prior years. -
i
[At the March 19'Consnission meeting, statements were apparently made to the effect that audits other than those that. pursuant to the ESD-219 4
program were conducted prior to November 1973.
If so,'were the findings of such audits discussed in 83-377 Where? Why were these findings, rather than those in the post November 1973 period, used to refute the NSCfindings?]"
Answer j
Again, as discussed in our response to GC-1, the purpose of the inspection was to detennine if the NSC conclusions could reasonably be drawn. from the QA record they reveiwed. We did not undertake to reconstruct the entire quality history'of the Pullman activity.
Dr. Myers correctly points out that both NSC and the staff looked at the record of welder audits from November 1973 through Spring of 1974.
However, as stated in paragraph 30 of report 83-37, -the NRC staff did not feel that many of the NSC conclusions could reasonably be drawn from the QA records they reviewed. Even though the welder audits did not start until November 1973, the Pullman internal audits and corporate audits
+
(previously discussed) routinely examined in-process welding and were implemented from the beginning of work. As stated previously we found the basic audit program to be satisfactory and in compliance with Appendix B.
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As a result of discussion at the March 26 Connission meeting the staff reviewed the Pullman addits and the PG&E audits done in the pre 1974 time period in more detail. The results are reported in Inspection Report i
84-16 and confirm that the audit program met the requirements of Appendix B.
y Many of the following questions deal with very specific hardware items during the pre November 1973 period.
It is important to keep the inspection purpose in perspective.
It was not the.NRC staff's purpose to MM [' purpose was to assess whether the licensee and his contractors were d perform a detailed evaluation of hardware in the plant.
Rather, the j y a responsible job of controlling construction and assuring the adequacy
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of hardware evaluation. As stated previously, the basic system of audits applied to Pullman welding from the start of work was satisfactory and met Appendix B.
The addition of the Pullman welder audit program in November 1973 was beneficial and improved the Pullman system of audits.
- 24. Question
" Criterion IX, NSC Audit Finding 100, Item 1. (Final p. 28. Draft p.
1 23.):
2 The draft, without. citing documents, appears to rely on the gas flow being "near the 20 cfm requirement" for its conclusion that defective i
welds might have resulted from inadequate shielding and purging. The draft states that excessively low flow rates would have been manifest in
. unacceptable porosi.ty which would have-been detected by NDE; the draft does not indicate the extent to which unacceptable porosity was found.
l The final does not state that the flow was near the 20 cfm requirement; it does state that "The vast majority of safety related stainless steel.
welds W re radiographically examined an'd the film was reviewed and
[
accepted by a qualified interpreter for code compliance." How many welds were not radiographically examined? How many were examined? Of those
)
that were examined, what percentage exhibited excessive porosity? What was done to detennine whether shielding and purging deficiencies that might have existed prior.to.,the first welder audit? What was done to correct for such deficiencies?"
Answer 3
i The NRC staff did not consider the specific deficiencies found in the i
purging and shielding area by the welder audits to be of much technical significance.
Consequently, no effort was made to reconstruct the nondestructive examination history of the welds in question nor was it considered worthwhile to do so. The problem here was.one nonna11y encountered in a purge gas, distribution system when welders hook into or drop off of it during the course of the work day.
Pressure and flow variations are introduced in the various distribution outlets. Weld -
quality is not very sensitive to purge gas flow. As long as the purge gas flows are maintained at all, no ASME code violations are involved.
The inspector found in his analysis of the situation that flows were 1
i maintained and in most all cases were reasonably near.20 cfm. As noted' i
in our report, the cr61tical welds in question would have been given i
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radiography and all pressure boundary welds are given a hydrostatic test f
which provides additional assurance of weld quality.
i Prior to the' initiation of welder audits in November 1973, in process l
Welding was audited by the site and corporate audits as discussed in j
paragraph 23.
25.- Question i
" Criterion IX, NSC Audit Finding 100, Item 2.
(Final p. 28. Draft p.
l 23-24.)
i What is the significance of 14 out of 183 audits identifying that welders i
did not have tempil sticks? Region V states that in each case that a j
welder was found not to have a tempil stick, one was provided. What was done to detemine the extent to which welders did not have tempil sticks i
prior to November 1973? Does the code allow interpass temperature requirements to be met by the resumption of welding delayed until the i
welder "can touch the weld?" The draft, but not the final, states that
'j "Tempil sticks were used by welders in the vast majority of cases." What constitutes a " vast majority?" What was done to determine whether there l
was a tempil stick problem prior to November 1973?"
Answer j
The significance is that some fraction,of the welders were not complying with their own internal procedures (1.e. having a Tempil stick in their -
possession). The Pullman audits were effective in identifying this.
i This had no real technical significance since no preheat or interpass temperature violations of the ASME Code were identified. The ASME code 4
does not mandate'the use of tempil sticks. Use of touch to ensure 4
interpass limits were not exceeded would be allowed by Code. The inspector saw no reason to pursue the tempil stick issue back prior to i
November 1973.
- 26. Question l
" Criterion.IX, NSC Audit Finding 10o Item 3. (Final p. 28. Draft p. 24.)
The draft states that in 4 out of 183 instances where amperages were not l;
within the welding procedure specification limit, the welder corrected l
his amperage setting. The draft stopped there. The final adds statements to the effect that defects resulting from improper amperages would be found during inspections.
The final also adds a statement that amperage is not an essential variable: specified by the ASME code...."
Does this mean that a. welds produced with improper amperages could still be in compliance with the code? 'What about improper amperages that might have been used prior to November 1973?"
Answer Yes, welds produced.with improper amperages could still be in compliance with the code. Amperage is but'one variable used by the welding engineers to obtain the proper welding heat input.
Other variables are 4
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Each of these variables are normally specified in a welding procedure specification using fairly wide limits and a change in one. variable is usually compensated for, by a journeyman welder, by a slight change in another.
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I In the judgement of the inspectior, it did not appear to be a necessary or particularly fruitful exercise to attempt to assemble amperage data for l
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the period prior to November,1973.
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- 27. Question
" Criterion IX, NSC Audit Finding 100. Item 4.
(Final p. 28-29. Draft p.
24.)
83-37 states concludes the " vast majority" of welders used welding j
procedures and knew where to obtain them.
Those that did not have them were told to get them. Those that did not know where they could be found i
were given "an explanation of the location from where they could be obtained." This finding was based on welder audits conducted after i
November 1973. What is Region V's p6sition with regard to those not a
members of the " great majority?" What.is Region V's position with regard.
to the availability of procedures and welders' knowledge of where l
procedures could be obtained in the period prior to November 1973?"
t Answer V
The NRC staff position is that all welders should know where the 4
procedures are and 'the Pullman audits properly identified and corrected j
the situation. The, inspector noted that the welding auditors did not identify defective welding as a result,of their original findings in this area. *
- k Region V had no reason to believe that the situation was any worse prior to November 1973 and, thus, saw no reason to pursue this issue any further.
- 28. Question i
" Criterion IX, NSC Audit Finding 10o, Item 5.
(Final p. 29. Draft p.
j 24.)
NSC found that the oxygen analyzer was.not available or not operative.
Region V concludes that only one of the 183 audits reviewed " indicated a 4
problem with the oxygen analyzer." What.was done by Region V.to determine the basis for the significant discrepancy between its finding and those of NSC? What documentation was examined?"
t j
Answer The staff's rationale for its conclusion is stated in NRC Inspection Report No. 83-37, paragraph 30, page 29.
As stated previously in the answer to question GC-5 the staff has attempted to solicit review and l
connent from NSC on NRC Inspection Report No. 83-37.
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- 29. Question i
" Criterion IX, NSC Audit Finding 100, Item 6.
(Final p. 29. Draft p.
i 25.)
NSC concluded that " Oven rod temperature was not monitored by the welders." 83-37 states that 14 of 183 audits identified. instances where rod oven temperatures were lower than those which were required.
A note t
on the draft states:
"With this many audit findings the rod oven temperature must have been too low much of the time." The NRC concludes that "The NSC finding that rod oven temperature was not monitored by the wel.ders is not supported by the audits, although isolated instances of ovens being below temperature were identified by the audits." Is it correct that 14 out of 183 constitutes " isolated instances?" What is the NRC position with regard to temperature control during the period prior l
to the initial welders' audit?"
Answer
+
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Ahstatedin'thereport(page29),thetechnicalsignificanceofthis finding is minimal.
Further, there was no code violation associated with the finding. The audit finding did point out that welders should have i
been more alert in monitoring their rod ovens. Pullman's audits j
recognized this condition and took corrective action.
Region V found no reason to believe the situation was any different prior to. November 1973 and.saw no reason to pursue this issue further.
- 30. Question i
" Criterion ^IX, NSC Audit Finding 10o, Item 7.
(Finalp.29.Draftp.
25.)
I The NSC stated that "Many welders did not understand their duties and responsibilities." Region V states that "Of the 183 audits received, five welder audits indicated that the welder in question did not understand their (sic) duties and responsibilities." The final, but not the draft, contains a sentence:
"The NRC considers that the reason these welder audits were done was to identify such instances and provide 1
corrective action." The draft and final report state that "In each case l-the welder was reinstructed by.the.QA inspectorLauditing:the welding...."
83-37 does not. address the pre-November 1973 period during which audits i-were not conducted. What mechanism existed prior to November 1973 to identify situa'tions where welders did not understand their duties and 3
~
responsibil.ities? What is the basis for assurance that, prior to j
November 1973,: welders understood their duties and responsibilities?"
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Answer As discussed in paragraph 23, there was an active audit program in existence prior to November 1973 which routinely examined in process welding.
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- 31. Question 4
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" Criterion X, NSC Audit Finding 5,6.
(Final p. 30-31.
Draftp.26-28.)
NSC found that the inspection process is generally inauditable on the ground that there were acceptance signatures that did not permit a determination of.whether the individual inspection requirements were fulfilled.
Region V stated that acceptance process sheets identified the procedures necessary to perform a particular inspection and the acceptance signatures were sufficient documentation of these procedures having been followed. The final report, but not the draft, states that this practice was "in accordance with standard industry practice, and in compliance with ASME code requirements...." Was this practice employed at other plants under construction during this period? Did NSC consider this practice in compliance with the ASME code? What was Pullman's response?"
Answer The staff has observed similar documentation practice at other nuclear plants under construction in that each inspection attribute is not contained as a separate line item on the work process traveler.
The staff feels that to list each attribute in detail would unnecessarily complicate the process traveler system. The staff's conclusion that the Pullman practice was in compliance with ASME Code requirements is still valid.
The_NS.C audit did not address the issue of ASME Code compliance here.
Pullman stated thattheir program complied with the ASME Code and regulatory requirements and that their program was acceptable and auditable.
- 32. Question
" Criterion X, NSC Audit Finding 7.
(Final p. 31. Draft p. 28-29.)
NSC found that a "large number of welds...were accepted for visual examination and thereafter accepted on surface NDE inspection.... Visual examination of those welds indicates that the surface is not acceptable for performance of surface NDE inspection." The final report, but not the draft, states "The inspector concludes that the NSC finding (that the surface of the welds was not acceptable for surface NDE inspection) was in error." What is the basis for these contradictory conclusions? Did NSC and NRC inspect the same surfaces? What-evidence exists to demonstrate that remedial work was not carried out in the time betwetn the NSC and NRC inspections?"
Answer The staff's basis and rationale for its conclusion is stated in NRC Inspection Report No. 83-37, paragraph 32 - page 31. The staff cannot state what NSC's response to the NRC finding would be, though as stated previously the NRC has offered NSC the opportunity to review and coment on NRC Inspection Report No. 83-37.
e
22 The staff examined the wele surface of :ne welds centained in the referenced iso etric package as indicatec in the h5C Aucit Re;crt.
The staff exacined the available records for the reference welds and fcend no decurent referencing rework er repair after er since the NSC audit.
- 33. Question
" Criterion X, NSC Audit Finding 9.
(Final p. 31-32. Draft p. 28-29.)
The NRC disagreed with the NSC i= plied finding that inking "R1" cnto a radiograph was not permitted by the code.
NRC also disagreed with NSC that FW-83 centained a surface defect "that is questionable for acceptance under visual standards." Does NSC agree with NRC's findings?"
Answer Refer to the answer to question GC-5 regarding NSCs views.
- 34. Question "Criterien X, MSC Audit Finding 10a.
(Final p. 32-33. Draft p. 28-29.)
NSC found that " Records cf welder qualification prior to 1972 are not available." Thus, the inspector was not able to verify the validity cf the_Pullaan response.to the NSC audit-finding." Region V fcund that 20 -
welders were qualified prior to 1972. Region V also found that the 90 day qualified welders log was started "at the beginning of 1972." The draft report, but not the final, states:
"The inspector was not able to determine Men the first production weliling was performed or on what system the first weld was accomplished." The final report, but not the draft states: "The inspector concludes that records of welder qualification prior to 1972 were available and in acceptable order."
Does Region V now knew when the first production welding was performed and on what system?
In light of NRC having found records for 20 welders, has NSC been asked why they found that records were not available? Does Region V believe that the welder qualification records for this period are cceiplete? How many active welders are shown en the initial 90 day qualified welders log? Is this log consistent with Region V's findings regarding the 20 welders?"
Answer Note: 'This'should have been titled Criterien IX, NSC Finding 10a, in the final report.
Yes, the first class 1 production pipe weld was performed en Dececter 23, 1971, on the Component Ceoling Water System.
The issue of NSCs views was addressed in the answer to questien GC-5.
F
23 l
Yes, the staff believes that the recores for the referenced period are complete.
As stated in f(RC Inspection Report tio. 83-37, paragraph 34, page 33, "The inspector cencludes that records of welcer cualification prior to 1972 were available and were in acceptable order."
The NRC did not catalog and itemize the welders shown on the first 90 day log and saw no particular reason why this would have been necessary or desirable.
- 35. Question a
" Criterion XIII, NSC Audit Finding 5.
(Final p. 35-38. Draft p. 31-37.)
Note that last paragraph'on Draft, p. 37 was dropped. The dropped paragraph mentions a PG&E audit of Pullman which identified prograsmatic and hardware discrepancies? What is the nature of these discrepancies?
Was there a requirement that they be reported to the NRC? Were they reported to the NRC? Does Region V have a basis for concluding that appropriate corrective actions were taken. Note the reference to the inspector having discussed this matter with Pullman and PG&E personnel.
What was the nature of these discussions? Do written summaries of these discussions exist?
It it Region V's position that for the entire period covered by the NSC audit, Pullman was in compliance with applicable NRC i
requirements pertaining to handling procedures?"
Answer The' nature of the discrepancies identified in the PG&E audit of Pullman
~
are described in attachment 5 to the Affidavit of Russell P. Wischow, dated September 21, 1983, to the ASLAE.,
4 These discrepancies were not reported to the NRC. The reporting requirements are described in 10 CFR 50.55(e).
It is the NRC staff's conclusion that none of the identified discrepancies met the threshold defined in the regulation; thus, reporting these discrepancies to the NRC was not considered necessary.
As indicated ~on page 37 of the draft inspection report referenced by Dr.
Myers, the inspector selectively examined the discrepancy resolutions and i
based upon those examinations obtained assurance that appropriate corrective actions were taken. This is documented on page 3 of'the draft inspection report. The whole subject was dropped from the final report because it was irrelevant to the stated purpose of the-inspection.
During the course of this selective examination discussions were held with PG&E personnel regarding the location of the discrepancy reports and the corrective actions. Thpse discussions were essentially an attempt to obtain the necessary documents for review. These discussions were not documented on written summaries because they contributed little to the overall NRC conclusion.
The NRC inspector did not attempt to reconstruct a history regarding compliance with handling procedures, nor apparently did NSC. The NSC -
]
statement was that " handling procedures do not exist" and the NRC's examination addressed whether or not such procedures did exist. The l
2?
inspector found that " appropriate anc adequate handling requirements were in place."
- 36. Question
" Criterion XIV, NSC Audit Finding 1.
(Final p. 38-39. Draftp.55-60.):
HSC stated that the Field Process Sheet _was inadequate. Region V reached a contrary conclusion. What is the basis for the discrepant findings?
Is it the NRC staff position that the Field Process Sheet adequately controlled and specified required work activities?"
Answer The staff's rationale for reaching the conclusion that the Field Process Sheet was adequate is contained in NRC Inspection Report No. 83-37, paragraph 40, pages 38-39.
Yes, the staff states in NRC Inspection Report No. 83-37 that, "...The inspector concludes that the use of the field process sheet adequately controlled and specified required work activities."
- 37. Question
" Criterion XV. NSC Audit Report, p. 36:
NSC.found, among other things, that " Systems that circumvent the nonconformance system have been established." This finding was not addressed in 83-37. What is the NRC's response to this finding?"
Answer During the inspection planning process the staff read the NSC finding, the Pullman response and the PG&E response. The staff determined that this item did not meet the selection criteria of paragraph 4 of report 83-37.
- 38. Question
" Criterion XVI NSC Audit Finding 2.
(Final p. 39. Draft 60-61.):
NSC stated that it appeared that a corrective action system had not been operative.
Region V cited examples where corrective actions had been taken in response to audits.
Was NSC's reference to a' corrective actions system intended to encompass corrective actions in response to nonconformance reports? Are the samples cited by. Region V sufficient to demonstrate that the Pullman did have an operative corrective action system?"
Answer.
Refer to the answer to question GC-5.
k
i 25 Yes, the staff feels that the examples cited in the NRC inspection report were sufficient to show the breadth of the contractor's and licensee's l
-corrective action system.
39.
Question
" Criterion XVIII, NSC Audit Finding 3.
(Final p. 39-40, Draft p.
45-46.):
NSC states that management audits were ineffectual.
Region V stated that "there is no basis to suggest these audits were ineffectual." Why did NSC and Region V reach such disparate conclusions?"
Answer The staff's rationale for reaching the ~ conclusion that "There is no evidence to suggest these audits were ineffectual," is stated in NRC Inspection Report No. 83-37, paragraph 42, page 40.
Refer to the ar.3 der to question GC-5 regarding NSC.
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