ML20087B999
| ML20087B999 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/06/1984 |
| From: | Norton B PACIFIC GAS & ELECTRIC CO. |
| To: | NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| Shared Package | |
| ML20081B795 | List: |
| References | |
| NUDOCS 8403120006 | |
| Download: ML20087B999 (23) | |
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O UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DiNPC BEFORETHEATOMICSAFETYANDLICENSINGAPPEdB 'h:28 0Fria 9 -- ..
) 00CnEi,jsg 73h g In the Matter of ) Docket Nos. 50-275 RANCH
) 50-323 PACIFIC GAS AND ELECTRIC COW ANY )
) (Design Quality Assurance)
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) )
)
PACIFIC GAS AND ELECTRIC COW ANY'S 4 ANSWER IN OPPOSITION TO JOINT INTERVENORS' MOTION TO AUGMENT OR, IN THE ALTERNATIVE, ,
TO REOPEN THE RECORD On February 14, 1984, Joint Intervenors filed a notion to augnent or, in the alternative, to reopen the record on design quality assurance. In support of that motion they submitted two affidavits of Charles Stokes l
(November 17, 1983, and February 8,1984), a fomer pipe support designer at Diablo Canyon; the affidavit of John Cooper (January 23,1984), a fomer instrument and control technician at Diablo Canyon; and several handwritten outlines presented by the NRC Staff at a January 31, 1984 public neeting regarding Mr. Stokes' allegations on small bore piping:
- Pursuant to a February 23, 1984 Order of this Board, Joint Intervenors served a March 2,1984 supplement to their February 14 motion addressing a transcript of the January 25, 1984 neeting between Charles Stokes and representatives of the NRC Staff. In these filings Joint Intervenors request the Board to augment or reopen the record to consider this "new XA_ Copy _Has Been Sent to.PDR I
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! infomation" as part of the reopened design quality assurance hearings. For the reasons set forth below Pacific Gas and Electric Company (PGandE) respectfully requests that the notion be denied in its entirety, i
Background ;
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On June 8, 1982, Joint Intervenors filed a notion to reopen the Diablo Canyon record alleging deficiencies in the quality assurance program.
After hearing argunent on this and other natters on April 14, 1983, the Board issued an Order on April 21, 1983 (unpublished) granting the notion to reopen on the issue of design quality assurance only. In that Order the Board ,
directed Joint Intervenors to refile their motion insofar as it sought reopening on construction quality assurance issues. Joint Intervenors O complied with this directive on May 10, 1983. After an evidentiary hearing on July 19-22,1983, the Board issued an Order on October 23, 1983, followed by a nenorandum opinion on December 23, 1983, denying the motion. Pacific Gas and Electric Conpany (Diablo Canyon Nuclear Power Plant, Units 1 and 2), ALAB-756, 18 NRC (1983).
I Following nonths of discovery, the reopened hearings on design quality assurance were held at Avila Beach, California, comencing October 31, 1983, and ending on November 21, 1983. Following the final filing of proposed findings of fact and conclusions of law, but before this Board's decision on design quality assurance, the Joint Intervenors, through both their new (GAP) and old (Center for Law in the Public Interest) attorneys, have come forward with nunerous eleventh hour allegations which have but one goal: to stop the licensing and operation of Diablo Canyon by obstructing and thwarting the adninistrative process.
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O Argument The Principle of Administrative Finality Requires that the Motion be Denied With almost clockwork precision, Joint Intervenors have conveniently filed yet another notion to present "new information" allegedly not previously considered by this Board in its review of design quality assurance issues for Diablo Canyon. Before addressing the substance' of the affidavits acconpanying the motion, this Board must resolve the larger and more conplicated issue of the doctrine of administrative finality. Briefly ,
stated, that doctrine is that with any administrative proceeding there must come a tine when the evidentiary record is closed. Obviously, this principle must be applied in a reasoned manner if an administrative natter is ever to be brought to a logical and timely conclusion. As the United States Supreme Court has observed:
" Administrative consideration of evidence ... always creates a gap between the time the record is closed and the time the administrative decision is pronulgated (and we might add, the time the decision is judicially reviewed).
If upon the coming down of the order, litigants might demand rehearings as a matter of law because some new circunstance has arisen, some new trend has been observed, or some new fact discovered, there would be little hope that the adninistrative process could ever be consunnated in an order that would not be subject to reopening. ICC v. Jersey City, 322 U.S. 503, 514, 64 S.
Ct.1129,11K 88 L.ed.1420 (1944). Vt. Yankee Nuclear Power v. National Resources Defense Council, 435 U.S. 519 at TES, 96 5. Ct.1197 at 1217 (1975).
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The Appeal Board has had recent occasion to apply this Q
principle. In the Matter of Union Electric Company, (Calloway Plant, Unit 1), ALAB 750A,18 NRC (decided December 9,1983).
In that case the Board was presented with new evidence of an alleged nonconservatism in the analysis of the loads on the manually welded embedded plates as a basis for reopening the record. In declining to reopen the record, the Board observed that:
"Because the Staff has the matter under review, a final resolution of the question of the purported non-conservatism has not been reached. Thus, it is possible that new information bearing on the safety of the manually welded embeds will be forthcoming. But, particularly given the Staff's monitoring on an ongoing basis of the construction and operation of individual nuclear facilities, the potential for new developments affecting litigated issues always exists. Litigation
"" "*'"*' ** * "' aa'"' ca"* ** "" '"d ^"' "*"
O developments can be brought to the attention of either the Commission (if it still has jurisdiction over this proceeding at the time) or the Director of Nuclear Reactor Regulation. See generally Virginia Electric and Power Co. (North Anna Nuclear Power Station, Units 1 and Z), ALAB-551, 9 NRC 704, 707 (1979); Public Service Co.
of Indiana (Marble Hill Nuclear Generat1ng station, Units 1 and 2), ALAB-530, 9 NRC 261, 262 (1979); Public Service Co. of New Ham) shire (Seabrook Station, Units 1 and 2),
ALAB-blJ. 5 NR; 694, 965-96 (1978)." (Calloway, supra, Slip opinion at 4-5).
See also, Pacific Gas and Electric Co., (Diablo Canyon Nuclear Power Plant, Units 1 and 2), ALAB-644,13 NRC 903, 994-95 (1981) refusing to reopen the record on seismic issues for a third time to review a new USGS report which had been recent1/ issued, but which used underlying data that had been previously available and in fact relied upon by expert witnesses of the parties.
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l Joint Intervenors now seek to avoid this result by arguing that O this is "new and tinely infomation" meeting the standards for reopening a closed record. We disagree.
The "new infomation" offered by Joint Intervenors is neither new nor of such significance that it would change the result and it is not being presented in a tinely fashion.1/ At the outset we note that the subject matter--design quality assurance--has been at issue before this Board for alnost two years. As this Board well knows, the design and design quality assurance aspects of Diablo Canyon have been thoroughly reviewed for the past two years. Hundreds of engineers have spent hundreds of thoe. sands of 4
1/ Counsel for Joint Intervenors, and their key "whistleblower", have a keen appreciation for the tactic of leading one to falsely believe that the tenporal relationship between events is other than it is. In Joint Intervenors' motion it is stated that Mr. Stokes' 11/17/83 affidavit was Q not served implying on Joint that the parties until December Intervenors 27,1983 were not aware of(gp. 20-21),infomation t is "new" obviously until that time. In fact, on February 2,1984, Mr. Stokes testified, under oath, before the California State Assembly Utilities and Comerce Connittee, that upon his temination on October 14, 1983, he went to the Mothers for Peace with his " story" and that they irmediately put him in touch with GAP (whom the Mothers for Peace have retained) and Mr. Devine (Mr. Stokes', GAP's and the Mothers for Peace attorney) helped him prepare his conplaint filed with the Labor Department on November 14, 1983, and the November 17, 1983 affidavit. For Joint Intervenors to attenpt to lead this Board into believing they had no knowledge of the affidavit when in fact they obviously knew of Mr. Stokes' " story" either before or during the hearing on design quality assurance is, to say the least, strange oenavlor for a party wno is engaging in a wholesale attack of the integrity of entire organizations and the individuals employed by those organizations.
l Mr. Stokes employs precisely the same tactic in stating that "on October 5,1983, I disclosed three... deficiency reports... which led to ny subsequent layoff, effective October 17, 1983." In fact, Mr. Stokes first filed the three dehciency notices on handwritten DR foms on l
August 8,1983, and was aware of the ongoing investigation and resolution of his concerns during August and September. See Attachnents l B and C.
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man-hours looking at these issues, including a concentrated independent review Q
of the IDVP. The issue of design quality assurance has been fully aired in lengthy hearings after all parties had anple opportunity to conduct full and complete discovery. In addition, the Board may take . judicial notice of the fact that the public at large in the San Luis Obispo area was given prior j notice of the design quality hearings last November. Obviously the affiants for Joint Intervenors were well aware of those hearings, yet they chose not to cone forward at that time. Rather, they conveniently waited until after the hearings were completed and this Board was, presumably, prepared to issue its decision. All of the issues raised in the affidavits acconpanying the motion relate to events and actions which took place prior to the reopened DQA ,
hearings. Indeed, the factual informaticn upon which the allegations are 1
based was available to the parties during the discovery process. For example, Mr. Cooper was last employed at Diablo Canyon in March,1982, and Mr. Stokes
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from November,1982, to October,1983. In their affidavits they recount actions and events purportedly related to design quality assurance issues which, if true, certainly could and should have been discovered during the months of discovery preceding the DQA hearings. In fact, it is difficult to think of a single factual predicate for any of the allegations upon which the motion relies that, if true, was not easily discoverable during the months preceding the DQA hearing.
The only thing "new" about these allegations is that they were cleverly packaged by Joint Intervenors' new set of lawyers and then paraded before the media and Nuclear Regulatory Conmission. Admittedly, the docunents are "new" i.e., they are dated from November 1983, to the present, but the O
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O info mation on which they were based is not.2/ Since aii of the aew :
infomation" was available to Joint Intervenors during discovery conducted in
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the DQA hearings, 'they should not now be all~c wed to manipulate the administrative process by. trotting out through a new set of attorneys their "new infomation" to augment /reope'n the record. This manipulation, if allowed, will effectively'obst's uct the NRC administrative process. In effect,
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intervenors will be pemitied to lay back until the eve of conpletion of the hearing process and thea spring "new information" on a tribunal and demand that hearings be res6ned to consider that infomation. A nore effective impediment to the orderly functioning of the administrative process would be difficult to imagine. Justice and logic demand that such a result be ,
precluded.3/ See, In the Matter of Cincinnati Gas and Electric Company, et.
al. (Wm. H. Zinner Nuclear Power Station Unit No.1), CLI-82-20,16 NRC 109 O
(1982) where the Commission refused to allow new QA contentions to be admitted when' as in the instant case, the standards for reopening had not been met (tineliness) and the QA allegations were currently under investigation by the
-2/ During discovery for the DQA hearings all internal PGandE and Project 00A audits were produced for inspection and copying by Joint Intervenors. These audits docunented certain discrepant conditions at the Onsite Project Engineering Group (OPEG) and corrective action which was implemented. Joint Intervenors chose to ignore or negligently
' failed to pursue this matter further during discovery or at hearing.
Similarly ITRs 60 and 61 dealing with small bore piping and deficiencies in calculations were available to Joint Intervenors for review and further action through discovery and hearing had they so desired. Many other fssues raised by Mr. Stokes were~ also actually litigated by the parties. ' -
3/ Counsel ~for GAP has boasted at NRC neetings that he.can continue filing 100 allegations per month for six months. Under this scenario the Board i could not close the record and render a decision until Joint !
Intervenors/ GAP decide to end their submittals,. the likelihood of which !
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I O a c Staff. The commission enphasized the proper roie of the Boards and the j Staff. The former are to adjudicate issues; the latter to review, monitor, inspect, and take enforcement action, if necessary. This is just such a case. The instant allegations are under review by the Staff and will Presumably be fully addressed and appropriate action,' if necessary, will be taken. Accordingly, further hearings are not only unwarranted but would constitute an absolute abonination of the adninistrative process.
A. Stokes Allegations The vast majority of the Stokes' allegations are addressed substantivelyintheBreismeister,et.al., affidavit ("Breismeister affidavit") covering some 58 false charges (identified by ronan numerals I through LVIII in the Breismeister affidavit) from the 11/17/83 and 2/8/84 affidavits and the 1/25/84 transcript (Attachment A). Responses to O allegations co.acerning Mr. Stokes' perception of why he was laid off are contained in the affidavits of Mr. Tressler, et, al., ("Tressler affidavit")
and that of Mr. Mangoba (Attachments B and C). Mr. Stokes' false charges regarding deceitful assignment of calculational packages is addressed in Attachments A and B and the affidavit of Mr. Schusteman (Attachment D). The allegations concerning qualification of QC inspectors to AWS code, their alleged inability to read weld symbols, and the allegation that welders did not possess copies of welding procedures is addressed in both Attachment A and Mr. Etzler's affidavit (Attachment E), Mr. Stokes' mistaken belief that Bechtel's contract for Diablo Canyon was for a fixed price is addressed in Mr.
i Friend's affidavit (Attachment F). Mr. Stokes' erroneous assertion that Mr.
Curtis could not answer questions from OPEG engineers regarding drawing 049243
.O aad never sot back with answers is addressed in Mr. curtis' affidavit V (Attachnent G).
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The Stokes' allegations, as contained in his affidavits of Q
11/17/83,2/8/84, and the 1/25/84 transcript of his neeting with NRC Staff f
I nenbers, fall into one or nore of four basic categories. Those categories are as follows:
- 1. The factual predicates of the allegation are, as a matter of law, demonstrably incorrect; 4/
- 2. The factual predicates of the allegation are substantially or partially correct from which an inference or conclusion is drawn but, upon examination of all of the relevant facts, the alleger's preferred inference or conclusion is, as a natter of law, denonstrably incorrect. 5/
- 3. The factual predicates of the allegation are substantially or partially correct, but lead to an inference or conclusion of O iittie or no safety significance, 6/
4/ The allegations which fall into these categories are identified by roman nuneral as set forth in the attached Breisneister, et. al., affidavit or by written description as they may be addressed in M1ier affidavits.
The allegations which fall in this category are: VI. VII, XII-XV, XVII-XXV, XXVIII, XXXI, XXXII, XXXVIII, XL, XLI, XLIII L. LI, LIV. LV, LVII. In addition, the allegations addressed in the Mangoba,
, Shusteman, Etzler, Friend, and Curtis affidavits (Attachnents 5-G).
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5/ The allegations which fall in this category are: I-V, VII-XI, XIII-XVI, XIX, XXII, XXVI, XXVII, XXIX, XXX, XXXII, XXXV, XXXVI, XXXIX, XLI, XLIV, XLVI, XLVII-XLIX LI, LII, LIII, LVI, LVIII. In addition, there are some factual predicates in those allegations addressed by Attachents B-G which are at least partially correct but, when viewed in light of all the facts, do not lead to the conclusions proffered by the alleger, i
6/ The allegations which fall in this category are: I, II, III, XII, XV, '
XXIV, XXXII, XXXV, XLV, LI, LIII, LVIII.
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O 4. The allegation is based on hearsay or speculation thereby ,
lacking sufficient foundation.1/ ;
- 1. Demonstrably Incorrect Factual Predicates A large number of Mr. Stokes' accusations fail when the factual predicates for those accusations are examined and are found to be incorrect.
- For example, Mr. Stokes alleges that Bechtel normally would not mind finding errors in the small bore piping review because
"Bechtel has always had a cost plus ten percent basis contract. They don't care if they have to---They are very adamant, if they aren't doing original design for you and you tell them that you want it done right and you keep insisting on it and you have your own people to monitor, well, hell, if they do it wrong they'll cut it out and do it again because it's cost plus ten. .
Diablo Canyon is the first job I know of where they stuck their neck out and bid lump sum to prove that plant was okay, and Nhen they you how they did it.got close to the end at -- 22) and I'll tell i
O (1/25/84 Transcript The fact of the matter is that Mr. Stokes is incorrect. As set forth in the affidavit of Mr. Friend, Bechtel's contract for Diablo Canyon is not a fixed price, but rather, a cost-plus basis contract.
Another example of patent incorrectness is Mr. Stokes' accusation that nanagement hired people it could control by threat of deportation:
"I believe that management helped to enforce questionable design practices by hiring aliens on " green cards" who were afraid to disagree with superiors due to the risk of being dismissed and subsequently deported if they could not maintain their jobs. I personally know of many 7/
The allegations which fall in this category in whole or part are: III, VI-IX, X, XIII, XX, XXI, XXIX, XXXI, XXXII-XXXIV, XXXVII, XLII, XLIII, l XLV, XLVI L-LII, LIV LY LVII. In addition, many of the factual l
Q gredicatesorbasesoffere,dfortheallegationsaddressedinAttachments
-G fall in this category.
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O Indians brought over from the Catawba nuclear plant in South Carolina who felt this way. One Indian who was a friend i
becane so disheartened that he just signed off anything, I whether it was right or wrong, That is unfortunate, since he i was a good engineer. The combination of nanagement I intinidation and the large number of errors sinply were too much, and he lost his spirit." (Stokes, 2/8/84, p. 2)
An examination of the facts shows that Mr. Stokes is again unequivocably nistaken. Green card holders are pemanent residents of the United States and have all the rights of a United States citizen except the right to vote. 8 U.S.C.S. Sections 1101, 1251. Employnent status has no bearing on their being allowed to remain in the United States. As set forth in the Breisneister affidavit, paragraphs 91-92, the neans for the alleged i
intimidation simply did not and do not exist. ,
Two of Mr. Stokes' nore serious and insidious charges, that of purging of records with the intent of covering up dishonest acts, and the O alleged dischonest acts themselves fall squarely in this category. The facts are sinple and straightforward.
r First, the dishonest conduct never occurred. Calculations were never assigned to accomplish qualification outside design criteria.
(Schusterman aff.; Breisneister aff., par. 40-43, 51-53). If dishonest conduct did not occur then obviously no coverup of that conduct is possible.
In this case the factual predicate of purging the records, which serves as the basis for the inference of a coverup, is also simply not true. No files or records have ever been purged. (Breisneister aff., par. 177)
Along nore technical lines is Mr. Stokes' allegation that the Bechtel in-house STRUDL was limited to 80K nemory thereby limiting its ability O
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i (stokes. 1/25/84 Tr., ,,. 27, 38-39) As set forth O ta a= iirr pipe =#9aarts.
in the Breisneister affidavit, the Bechtel STRUDL computer progran allows the analysis of problems up to 262K nenory. In fact, Bechtel has never experienced a case, at Diablo Canyon or elsewhere in which STRUDL nenory l
- limitations prevented the analysis of any pipe support frane. (Breisneister aff., par. 168-173.) In fact, it was apparently Mr. Stokes' lack of knowledge (as opposed to his professed superior knowledge) about STRUDL which led hin to this erroneous belief. (Id.) !
A detailed reading' of Attachments A thru G, and I, are necessary to l
fully appreciate both the large numbe'r and gravity of factual inaccuracies in Mr. Stokes' allegations. We respectfully subnit that a fair reading of the ,
docunents submitted with Joint Intervenors' notion and the attachnents to this response can lead to only one conclusion: Mr. Stokes' concerns are directed O at something other than safety.
- 2. Demonstrably Incorrect Inferences or Conclusions Many of Mr. Stokes' allegations contain one or nore correct, or substantially correct, factual predicates which, when viewed in isolation, lead to what appear to be logical conclusions or, to apparently reasonable inferences. However, when all relevant facts are brought to bear on the allegation, the conclusion or inference becones illogical or unreasonable and the substance of the allegation disappears. By way of example, Mr. Stokes :
alleges that QC inspectors could not read welding synbols because they were not consistently qualified to the AWS code and were not issued the AWS symbols. (Stokes, 2/8/84, p. 6) Mr. Stokes' factual predicates that the QC O
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i inspectors were not qualified to the AWS code and were not issued AWS synbols
'O is correct enough, but the apparently logical conclusion that they therefore could not read welding angle, effective throat symbols, and related i
4 instructions from the design drawings, is simply incorrect. As set forth in Attachnent A, paragraph 135:
The AWS Structural Welding Code did not, when Diablo 4
Canyon started, and does not today, require AWS qualified
- inspectors. Inspectors need not be issued the AWS weld synbol s. Knowledge of these symbols, like nuch other material, is part of an educational, experience or training background. These symbols are connonly available in references and need not be issued to inspectors.
Another exanple of this type of allegation occurs when Mr. Stokes
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correctly states that Mr. Mangoba spent several days approving many calculation packages. This factual predicate is followed by the inference that the calculation packages were not properly reviewed. The additional Q
!' facts that change that inference are that Mr. Mangoba had instructed five i
other senior experienced engineers to perfom a detailed technical content review of the calculational packages over and above the required and nomal checking procedures, as a part of his approval.
Those serious charges of Mr. Stokes' that do not fall exclusively in category 1 above fall in this category. For example, Mr. Stokes' charge of being teminated for filing discrepancy reports (DR) has sone true, or partially true, factual predicates to it, e.g. Mr. Stokes was indeed laid off on October 14, 1983. When all of the facts and documentary evidence are viewed, it is obvious that Mr. Stokes was leid off in a nomal, scheduled reduction of forces. His DRs were, for all material purposes, filed in f
August, not October. (Mangoba affidavit; Tressler aff., par.14-18).
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- 3. Insignificance of Inferences or Conclusions Q
Mr. Stokes' allegations also seen to contain accusations which are either substantially or partially correct but when pursued to a logical conclusion, anount to little or no safety significance. For example, Mr. !
Stokes alleges that engineers were put to work designing pipe supports without first receiving training. (Stokes, 11/17/83, p. 2) As set forth at length in the Breisneister affidavit, only experienced, technically qualified engineers were hired to work in the snall bore pipe support group. (Breisneister aff.,
par. 1 -7 ) In addition, they did receive additional training, albeit not always in as tinely a fashion as would be optinally desirable. More inportantly, there was no correlation between errors made in calculations and the tineliness of training received. (Anderson aff., Exhibit 1, p. 35)
Another example is that of Mr. Stokes' allegation that Bechtel issued out-of-date STRUDL nanuals to engineers in the seisnic design review.
The allegation, while partially true (Breisneister aff., par. 161-166) sinply does not give rise to any significant concern. The basic STRUDL user's nanual has not changed in 16 years. (Id. at 163) The changes that Mr. Stokes clains were late in arriving are sinply changes to make it easier for the user. (Id.
at 164) Mr. Stokes' concern is largely academic. (Id. at 165)
- 4. Speculation and Hearsay While all of Mr. Stokes' allegations fit into one of the three categories described above, many of the factual predicates are speculation l and/or rank hearsay. An allegation cannot, and should not, be used as the basis for any notion when there is absolutely no foundation to support the factual predicates as legally adnissable evidence. For exanple, Mr. Stokes O
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I O specuiates that an operator misht have difficuity operatias the piaat (stokes.
1/25/84, pp.115-116) and then, piling hearsay on top of speculation, Mr.
Stokes clains that "there are other allegations that I have read, concerning whether the operators could indeed operate the plant safely." (Id. at 116)
Other examples of allegations which are rife with speculation and/or hearsay are those concerning " green card" holders (Stokes. 2/8/84,
- p. 2), the Bechtel contract (Stokes,1/25/84, p. 22), why Mr. Stokes was laid off (see Attachnents A and B), and management's motives for various and sundry other alleged evils.
Mr. Stokes has brought a new dinension to these proceedings. After literally a decade of adversarial proceedings with years of discovery, endless days of hearings, and uncountable numbers of pages of evidence, the Joint Intervenors have been unable to show that anything of safety significance is O wrong at Diablo Canyon. The documents provided to thim in discovery alone anount to tens of thousands, if not hundreds of thoust.nds, of pages. Now they have Mr. Stokes, who, while working for the licensee, was able to steal about in the night ransacking files seeking what he perceived to be danaging infornation. Mr. Stokes freely admits:
"I knew I'd better start getting some infomation to back j up my allegations. I knew if I didn't get sonething they would just squash ne like a bug. So during those tines occasionally when my workload was low, like when they had a bomb threat at 9:00 o' clock on Friday night and everybody left because it was approved, so-called, leave time, nobody came to see me from 9:00 o' clock on. I was working until 2:00 o' clock. So I just strolled around the plant and looked at hangers. I worked on my DRs during those times.
I nade up a list of about 200 angle franes which failed just outright. The unbraced angle without even doing a calc.
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O I aiso visited the traiier that ar aaasaba occuP ied l
during the day because the maintenance people who had to open the door tended to leave it unlocked occasionally.
I had free access to all books without anyone looking at ,
i me, other than occasionally the sweeper. It allowed me to take the docunent out of the book, over to the Xerox nachine and make a copy of it, put it right back iri the book and nobody knew I'd ever looked at it. I got those copies of those docunents, and those are the ones I gave Isa and the guys at the site." (Stokes,1/25/84, p. 94)
Although it is most painful to know of Mr. Stokes' conduct, it is reassuring to know that an engineer as obviously dedicated to finding damaging infornation as he was, and who essentially had free access to all records, I cannot identify any safety significant itens. Each and every one of his allegations, when investigated thoroughly, results in the conclusion that either the allegation was false, its conclusion or inference incorrect, or, when correct, of little significance.
O s- cooper ^11 aations k The affidavit of John Cooper (Exhibit F of the notion) seems to have been tacked on to Joint Intervenors' notion as an afterthought, and is given very little attention in the body of the motion. This is not surprising. The information in the affidavit is neither significant nor relevant to design quality assurance, nor is it timely.
Joint Intervenors' notion is premised on the assertion that the supporting affidavits, including the Cooper affidavit, " bears directly on the issue of design quality assurance at Diablo Canyon," (motion at 2) is of
" undeniable relevance," (motion at 3) and "goes to the very heart of the seismic redesign of the plant and the verification program undertaken by the DCP and the IDVP," (notion at 3). When we examine the only two pages in the O
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O motion which discuss the Co.,er affidavit, we find the statement by counsei i
that Mr. Cooper portrays "a seriously flawed design practice." Counsel then i l goes on to refer to the items recounted in Mr. Cooper's affidavits as " quality I
assurance deficiencies," (notion at 18). These statements, as they apply to :
the Cooper affidavit, constitute an egregious and questionable j misrepresentation by Joint Intervenors to this Appeal Board.
When Mr. Cooper worked for PGandE, he was not a design engineer and was not a part of any design engineering organization. He was not a QA engineer. He was an inspection maintenance technician and construction field engineer, and neither his duties nor his activities as described in his i
affidavit bear any relationship whatsoever to design quality assurance. .
Indeed, Mr. Cooper himself does not even claim that his allegations relate to design QA--the relationship seems to be purely an unexplained figment of Joint O Intervenors' imagination.
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As explained in the affidavit of J.D. Shiffer, et al., ("Shiffer -
affidavit") (Attachment H), the Cooper affidavit centers around an interesting and positive aspect of project administration at Diablo Canyon--the encouragenent by management enployees to raise safety concerns, irrespective of the employees' organizational affiliations or job description. (Shiffer aff. , par.18-25) Mr. Cooper, on several occasions, brought to the attention of management and the NRC, concerns related primarily to the design of the residual heat renoval (RHR) system. In each and every case his concerns were actively considered and fomally dispositioned (Id., par. 26-31). Mr.
Cooper's concern--satisfaction of single failure criterion and prevention of spurious closure of valves between the RHR system and the reactor coolant O system by uaanticipated eiectricai sisaais--have been fuiis resoived by esandt v
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t l O in a manaer specificaiis approved by the aaC. The decay heat removai functioa meets the applicable licensing criteria for Diablo Canyon, and the potential ,.
for unwanted valve closings has been eliminated (Id. , par. 3-17).
j Mr. Cooper's complaint is, in essense, that the matter was not resolved to his satisfaction. Certainly he is entitled to his opinions, but J
the concerns he raised were duly taken into consideration by PGandE management in arriving at an acceptable resolution. His concerns, though, clearly show no characterization whatsoever of inadequate design QA.
Mr. Cooper also cites several instances of what he considers to be nanagenent retaliation or punishment for, in his own words, being a "whistleblower." As explained in the Shiffer affidavit, paragraphs 51-70, i
there has been no punishment, retaliation, or even threats of punishment or retaliation against Mr. Cooper.8_/ Significantly, nowhere in his affidavit O
does Mr. Cooper allege that he sought redress at a higher level of managenent at PGandE with regard to any alleged or perceived harassment, intimidation, or retaliation.
Joint Intervenors make a nunber of allegations which they assert are supported in the Cooper affidavit. (notion at 18) However, they offer no explanation of how such support is to be gleaned from the affidavit. In fact, 1
the allegations are extraordinarily misleading, particulary since they are characterized as " quality assurance deficiencies."
8_/ This is not to say that isolated acts of intimidation, harassnent, or retaliation cannot or have not occurred on a job as large as Diablo Canyon. Rather, the question to be asked is -- "Has managenent ever encouraged or condoned such actions once brought to its attention?" As noted, PGandE has had a stated policy for many years of not allowing i harrassing, intimidating, or retaliatory conduct. (Shiffer aff., par.
Q 18-20,53,54).
o 1
O The alleged " failures of corrective action" involved only a failure of the corrective action Mr. Cooper would like to have seen taken. There was ;
indeed corrective action taken, and it did not fail (Id., par.13-14).
Sinilarly, the dramatic accusation of docisient destruction involved legitinate renoval of unofficial documents which had nothing to do with QA or the design process (Id., par. 47). There was no retaliation, intimidation, or threats (Id. , par. 51-70). The " violation" of internal adninistrative controls was a single de mininus lack of a signature of no substantive significance, (Id., par. 48-50) and there was no " refusal" to correct an erroneous FSAR description, (Id., par. 35). The other allegations in the Cooper affidavit are similarly without substance. (Id., see generally par. ,
- 16,32-52).
I In short, the allegations are of little substance, bear no
,O relationship to design QA, and cannot be considered as support for the
~
extraordinary action of reopening or augmenting the record following the conclusion of hearings.
C. NRC Staff January 31, 1984 Meeting Handouts All of the questions which were raised in the NRC Staff handouts at the January 31, 1984 neeting have been addressed in PGandE's February 7,1984 submittal. (Exhibit 1 to Attachnent I) In that subnittal, PGandE responded to each and every question and concluded that there is reasonable assurance that the as-constructed small bore piping neets all design criteria. PGandE believes these answers are fully responsive to the Staff's questions and will assist the Staff in its ongoing investigation.
O o
.. _ . - _ - - .- . _ _ . _.. - _ - . . - - = . ._
Q D. Conclusion Joint Intervenors' notion nust not be considered in a vacuum.
Rather, it must be viewed in light of the hundreds of thousands of manhours of review that the design of Diablo Canyon has received during the past year and the evidence this Board has received on the subject of design quality assurance at Diablo Canyon. It nust also be seen as it relates to the regulatory framework under which plants such as Diablo Canyon are to be i designed and built. As they argued at the DQA hearings, Joint Intervenors are 1 of the view that the regulations require absolute assurance of absolute perfection in each and every instance. Anything less than absolute perfection is fatal. The adoption of such a view by this Board would, of course, prevent Diablo Canyon, or any other plant, from ever being licensed or operated. It is respectfully submitted that is indeed the goal of Joint Intervenors.
O The test here must be whether Diablo Canyon is designed and constructed to reasonably assure protection of the public health and safety.
PGandE is confident that this Board has that reasonable assurance as a result of the evidence presented to it during the DQA hearings. A thorough review of Joint Intervenors' notion, its Exhibits, this response and the affidavits attached hereto should not intrude upon that reasonable assurance. It is respectfully requested that the notion be denied in its entirety.
l lO v
O Dated: March 6, 1984 Respectfully submitted, ROBERT OHLBACH PHILIP A. CRANE, JR.
RICHARD F. LOCKE DAN G. LUBBOCK Pacific Gas and Electric Company P. O. Box 7442 San Francisco CA 94120 (415) 781-4211 ARTHUR C. GEHR Snell & Wilmer '
3100 Valley Center Phoenix AZ 85073 (602) 257-7288 O BRUCE #0RT0a THOMAS A. SCARDUZIO, JR.
Norton, Burke, Berry & French, P.C.
P. O. Box 10569 Phoenix AZ 85064 (602) 955-2446 Attorneys for Pacific Gas and Electric Conpany Dated: March 6, 1984 By' O A W Bruce Norton l
O
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
(])ntheMatterof )
)
PACIFIC GAS AND ELECTRIC COMPANY ) Docket No. 50-275
) Docket No. 50-323 Diablo Canyon Nuclear Power Plant, ) '
Units 1 and 2 ) -
)
CERTIFICATE OF SERVICE The foregoing document (s) of Pacific Gas and Electric Company has (have) been served today on the following by deposit in the United States mail, properly stamped and addressed:
Judge John F. Wolf Mrs. Sandra A. Silver Chairman 1760 Alisal Street Atomic Safety and Licensing Board San Luis Obispo CA 93401 US Nuclear Regulatory Commission Washington DC 20555 Mr. Gordon Silver 1760 Alisal Street -
Judge Glenn O. Bright San Luis Obispo CA 93401 -
Atomic Safety and Licensing Board US Nuclear Regulatory Commission John Phillips, Esq.
Washington DC 20555 Joel Reynolds, Esq.
N Eric Havian
--udge Jerry R. Kline Center for Law in the Public Interest Atomic Safety and Licensing Board 10951 W. Pico Blvd. - Suite 300 US Nuclear Regulatory Commission Los Angeles CA 90064 Washington DC 20555 David F. Fleischaker, Esq.
Mrs. Elizabeth Apfelberg P. O. Box 1178 c/o Betsy Umhoffer Oklahoma City OK 73101 1493 Southwood San Luis Obispo CA 93401 Arthur C. Gehr, Esq.
Snell & Wilmer Jcnice E. Kerr, Esq. 3100 Valley Bank Center Public Utilities Commission Phoenix AZ 85073 State of California 5246 State Building Bruce Norton, Esq.
350 McAllister Street Norton, Burke, Berry & French, P.C.
San Francisco CA 94102 P. O. Box 10569 Phoenix AZ 85064 Mrs. Raye Fleming 1920 Mattie Road Chairman Shall Beach rJ4 93449 Atomic Safety and Licensing Board Panel Mr. Frederick Eissler US Nuclear Regulatory Commission Sc nic Shoreline Preservation Washington DC 20555
, Conference, Inc.
623 More Mesa Drive nnta Barbara CA 93105 I
hairman
- Judge Thomas S. Moore
(])AtomicSafetyandLicensing Chairman 4 Appeal Panel Atomic Safety and Licensing US Nuclear Regulatory Commission Appeal Board Wachington DC 20555 US Nuclear Regulatory Commission Washington DC 20555 Secretary US Nuclear Regulatory Commission
- Judge W. Reed Johnson Washington DC 20555 Atomic Safety and Licensing Appeal Board Attn: Docketing and Service US Nuclear Regulatory Commission Section Washington DC 20555
- Lawrence J. Chandler, Esq.
- Judge John H. Buck Hanry J. McGurren Atomic Safety and Licensing US Nuclea~r Regulatory Commission Appeal Board Office of Executive Legal Director US Nuclear Regulatory Commission WEshington DC 20555 Washington DC 20555 Mr. Richard B. Hubbard Commissioner Nunzio J. Palladino MHB Technical Associates Chairman 1723 Hamilton Avenue Suite K US Nuclear Regulatory Commission" Sen Jose CA 95125 1717 H Street NW Washington DC 20555 Mr. Carl Neiberger alegram Tribune Commissioner Frederick M. Bernthal P. O. Box 112 US Nuclear Regulatory Commission ,
Sun Luis Obispo CA 93402 1717 H Street NW Washington DC 20555 Michael J. Strumwasser, Esq.
Susan L. Durbin, Esq. Commissioner Victor Gilinsky Pcter H. Kaufman, Esq. US Nuclear Regulatory Commission 3580 Wilshire Blvd. Suite 800 1717 H Street NW Los Angeles CA 90010 Washington DC 20555 Maurice Axelrad, Esq. Commissioner James K. Asselstine Newman & Holtzinger, P.C. US Nuclear Regulatory Commission 1025 Connecticut Ave. NW 1717 H Street NW Washington DC 20036 Washington DC 20555 Commissioner Thomas M. Roberts US Nuclear Regulatory Commission 1717 H Street NW ,
Washington DC 20555 j l
IMN l (2) 3 i- Dnte: March 6, 1984 > *& A W l
- Ccpies delivered by Courier.
l O UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD
)
In the Matter of ) Docket Nos. 50-275
) 50-323 PACIFIC GAS AND ELECTRIC )
COMPANY ) Design Quality Assurance
)
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) )
)
AFFIDAVIT OF F. C. BREISHEISTER, D. J. CURTIS, M. J. JACOBSON, M. E. LEPPKE, ,
G. H. MOORE, R. G. OMAN, L. E. SHIPLEY, AND W. H. WHITE STATE OF CALIFORNIA )
O CITY AND COUNTY OF SAN
)
)
ss.
FRANCISCO )
The above, being duly sworn, depose and say:
I, Gary H. Moore, an Project Engineer (Unit 1) for the Diablo Canyon Project.
I, Larry E. Shipley, an Technical Consultant for Piping for the Diablo Canyon Project.
I, Michael J. Jacobson, an Project Quality Assurance Engineer for the Diablo Canyon Project.
I, Robert G. Onan, an Assistant Project Engineer (Unit 1) for the Diablo Canyon Project.
I, Myron E. Leppke, an Onsite Project Engineer for the Diablo Canyon Project.
(($ '[
- lO _3_
l O I, Daniel J. Curtis, am Onsite Plant Design Group Supervisor for the Diablo Canyon Project.
I, Fred C. Breismeister, am Manager of the Research and Engineering / Materials and Quality Services Group for the Bechtel Group.
I, William H. White, an Assistant Project Engineer, Seismic, for Diablo Canyon Project, Unit 1.
I. It is alleged that:
There was only minimal training; initial assignments were received on the first day with an example of Bechtel calculations. (Stokes, 11/17/83, p. 2)
- 1. As Mr. Stokes has acknowledged, pipe support engineers are a select group with specialized knowledge and nationwide experience which nakes them uniquely qualified to do their job. As a result, minimum technical indoctrination and training are necessary.
- 2. Indoctrination and training of pipe support engineers assigned to the Onsite Project Engineering Group (OPEG) began with the process of selecting experienced, technically qualified engineers whose professional qualifications for properly performing pipe support design work were already established.
- 3. To ensure technical competence, pipe support engineers are hired in large part on the basis of interviews, educational qualifications, and previous experience. For both permanent and temporary or " casual" employees, the professional credentials of all are required to be verified by either the Engineering or the Personnel Departments of Bechtel or PGandE. For contract employees, such verification is a 2-
1 contractual requirement for the contract fim. A thorough review of the engineer's work experience is confimed by senior engineering personnel.
- 4. A thorough review of the technical background of the engineers in the small bore pipe support group at the site shows that experienced, technically qualified engineers had been hired,with little or no need for additional instruction in snall bore piping calculations other than that nomally provided to familiarize then with the proper design criteria and project calculational nethodology. Most of the engineers had worked on two or nore other nuclear power projects, with nany having
' worked on five or more plants. All have at least a BS in Engineering or equivalent, and their mininun professional experience is one year; the naxinun professional experience is 14.5 years, and the average professional experience is greater than five years.
In order to indoctrinate aewly assisaed easineers in Pro 3ect Procedurai
, O 5.
requirenents, the Project provides fomal training in the Engineering Manual Procedures (EMP) which implements project QA requirenents. Those i
requirements neet Criterion II of 10 CFR Part 50, Appendix B, and are set forth in the Nuclear Quality Assurance Manual (NQAM), and Bechtel Quality Topical Report, Rev. 3A (BQ-TOP-1) which has been approved by I
the NRC for the Project. Each engineer assigned nuclear safety-related work receives indoctrination and training in EW in accordance with Procedure 2.1 of that manual. This course for the engineers identifies and describes the procedures appifcable to their work. The training enphasizes the procedares on design criteria menoranda, design calculations, design changes, drawing control, discrepancy reports, and j
nonconfomance reports.
_ _ - = . . - - . _ . . . _ . _.
- 6. PEI-15 specifies that the indoctrination and training are to be given within 30 days of assignnent to the Project. Training records indicate that approxinately 70% of all 0 PEG design engineers on the current OPEG roster received Engineering Manual training within 30 days of assignment as required. Approximately 95% received such training within four nonths of assignnent. The majority of those instances where an engineer did not receive training within 30 days of assignment occurred early in the Project. Project Audit 28.4, conducted in February 1983 and closed in May 1983, resulted in the correction of most of those discrepancies.
Since May 1983, only five OPEG design engineers have exceeded the 30-day training requirement by nore than a few weeks.
- 7. In addition to these organized training sessions, working familiarity with DCP calculational procedures and pipe support design criteria was O acquired by aew en9 1 oeer7 through the practical experience gained in originating preliminary calculations. Newly assigned engineers were given copies of conpleted example calculations to use as nodels for calculation fornat. Copies of project procedures, instructions, and criteria were made available for reference and adequate opportunity was given for the engineer to gain familiarity with project calculation format and nethods. Supervisory personnel were available to answer individual questions and provide clarifications for points of.
uncertainty. Newly assigned engineers were assigned more experienced checkers to review their work for adequacy and correctness prior to its being issued.
I lO l
l
I II. It is alleged that:
O Controlled documents were not inmediately received for work assignnents. Field engineers were working to unverified xerox copies which were incomplete.
Management was not responsive to requests for controlled documents. (Stokes, 11/17/83, pp. 2, 4, and 5)
- 8. It is true that not every support engineer had an individual copy of controlled design documents assigned to him. No such requirement exists and such a policy or requirenent would create far nore problems than it night alleviate. However, an adequate number of controlled copies were available in the specific work area for reference use by all engineers.
- 9. Mr. Stokes was assigned to the small bore pipe support group of OPEG as one of 11 engineers in November 1982. At the time of his assignment, three controlled copies of the project piping design criteria were assigned to the support group which was located together in one trailer.
O io. By aanuary 1983. the number of engineers assigned to the pipe support group in the trailer had increased to 35. Steps had already been taken to obtain additional copies of controlled documents for use by the expanded piping group. Additional controlled copies of design documents were requested from San Francisco. These documents were received in Decenber 1982, and were distributed for use by the expanded pipe support group. It was soon realized that the documents received, although identical, were not controlled documents, and therefore a further request was made in January of 1983 for additional controlled documents. Consequently, while there may have been sone inconvenience, copies of controlled and identical uncontrolled design dociments were available within easy reach of every pipe support engineer. Thirteen 0
l l additional controlled copies were received and distributed in February O 1983. Mr. Stokes was assigned his own individual controlled copy in February 1983. In April 1983, all controlled copies were replaced by a conplete reissue of new controlled copies of the design documents.
III. It is alleged that:
Field engineers were working to records of calculations they brought from other nuclear plants. Use of other plants' documents results in assumed load ratings for other nanufacturer's equipment that may not be applicable to DCPP. Assumptions used differed from those on controlled documents. Unique conditions of DCPP were not accounted for. (Stokes, 11/17/83, pp. 2 and 3)
- 11. Questions have been raised as to whether references, such as the following, in the possession of pipe support engineering personnel were used in Ifeu of approved work procedures:
An interoffice memorandum dated March 21,1983, " Guidelines for O o Calculating Design of Skewed Welds" o Westinghouse Nuclear Technology Division Data for calculating double cantilever supports o Bechtel GPD STRUDL II Computer Program Users Manual'CE-901 Novenber 3, 1983 o Bechtel GPD IOM dated Novenber 11, 1980, "GPD Pipe Support Newsletter No. 5, Beta Angle" o Control Data Corporation (CDC) Bechtel National Support Manager to Civil / Structural Projects staff, " Baseplate II User Aids" o Midland " Pipe Deflection Formula" o UE & C Pipe Support Design Standard, August 15, 1979 O
l
l
- 12. Reliance on one's past experience is not uncomon in the profession and O especially for pipe support engineers who, as Mr. Stokes acknowledges, i
have specialized talents based on past experience. Experienced engineers comonly have general reference material as a part of their
- personal and professional library. This type of naterial includes textbooks and handbooks, and typically provides standard formulas and tables, code discussions, example calculations, rules of thumb and other simplified, conservative methods in connon use in the industry. As general reference material, they are not controlled and, nore i
importantly, they do not constitute acceptance criteria.
- 13. Project Engineering Procedures (EMP 3.3) require that calculations be sufficiently detafled so that qualified technical personnel can verify their adequacy without consulting the originator. References such as textbooks, catalogs, monographs, and other such accepted industry Q
techniques nust be documented in the calculation when necessary to provide details of the design sufficient to allow an independent review. Their use then is checked and approved via the calculation review process.
- 14. The above identified docunents are references of the type nomally found
! in an experienced engineer's personal library. We know of no instances where the references were inproperly used. In one instance, a non-project document was referenced as the source of a double cantilever deflection fornula used in a calculation. It was a standard engineering fomula, not unique to any particular project, and need not have been referenced in the calculation.
lsO
- 15. Prior to May 1983, design calculations originated by OPEG were prelininary in nature since they were based on preliminary assumptions due to the absence of final thernal and sef snic design data at th'.t time. All such prelininary calculations have been subsequently reviewed and revised as the final design data have become available. These revisions of the calculations to final status were conpleted using the latest revision of project criteria and were subjected to Independent Design Yerification Progran (IDVP) review.
IV. It is alleged that:
Supplier's ratings for U-bolts were one-third to one-fourth nore stringent than claimed on DCPP drawing 049243. This drawing represents a false statement.
PGandE relied on a series of suspect assumptions in order to exaggerate the load ratings. The 1978 PGandE U-bolt test program was biased by not reflecting actual plant iO coaditions. Stokes was aiiowed to use load ratin9s which failed sone of the U-bolts. Even if load ratings of U-bolts.were accurate, the hangers to which they are attached would not meet design requirements. (Stokes, 11/17/83, pp. 5 to 8)
- 16. A U-bolt is used in conjunction with other structural nenbers to provide lateral restraint to a piping systen. It restrains the piping in directions perpendicular to the pipe centerline and provides both themal and seisnic restraint, The ASME B&PVC,Section III, recognizes i
and provides detailed rules for the qualification of pipe supports by three different nethods. They are analysis, testing, or experinental analysis.
- 17. ITT Grinnell qualified the U-bolt by analysis. To analytically represent the load / deflection relationship between the pipe and the l.v O
. _ _ _ . _ _ _ _ - _ _ _-_____ - -- m - -
- - " ? '--- - - - - + w wYu
1 l
U-bolt becones a very conplex problen. To provide this qualification, Grinnell sinplified the relationship between the pipe and U-bolt to produce very conservative results from a nodel that can be handled analytically.
- 18. Testing provides a more accurate representation of the pipe /U-bolt interaction by including elements such as the distribution of the load
' on the U-bolt, the frictional resistance between the pipe and the U-bolt, and the pipe's influence on the U-bolt's defomation.
- 19. DCP Standard Drawing 049243 for small bore pipe supports uses load ratings that were derived in accordance with the intent of the ASME B&PVC Section III rules for qualification by testing and does indeed give higher load ratings than given by ITT Grinnell. These tests were conducted at the DCP site in 1978. It is true that these two methods, O anaiysis and testing, can yieid a factor of 4 difference. sowever, the
'l test results are closer to reality, whereas the analytical results are only a very conservative approxination.
- 20. ASME Section III, Subsection NF-3260, provides the procedure by which U-bolt allowable ratings were developed. Per NF-3260, the procedure for load ratings consists of inposing a total load on one or nore duplicate full-size samples of a component support. The total load is to be equal j
to or less than the load under which the component support fails to perfom its required function. If a single test sample is perfomed, NF-3260 requires the load ratings to be derated by 10*..
- 21. The tests perfomed for the Diablo Canyon supports were more nunerous
)
than the single test pemitted by the code but were less than the O
1
" statistically significant sample" allowed by the code as an alternate. '
O The conservatisns added in the generation of allowables are considered i
to be at least equivalent to a derating of allowables by 10%. The following is a surriary of conservatisns:
- 22. A ninimum of four U-bolts were tested for three loading conditions for each pipe size. The loading conditions consisted of the application of side loading, tension loading and a conbination of side and tension
- loads (450). The allowables for tension and side loading were based on the' lowest test load of all pipe sizes tested using a 'given dianeter i
U-bolt. The test loads used in the equations of NF-3260 represent the lowest tension and side test loads found for 1/4-inch and 3/8-inch dianeter rod U-bolts, respectively.
- 23. Added conservatism occurs in the interaction fomula with the h application of both tension and side loading because the minimum tension test results and the minimum side loading test results are conbined.
- 24. U-bolt tension failure did not occur for any U-bolts for piping sizes greater than 1-1/4 inches in diameter. The allowables were based on the testing nachine's capacity rather than the U-bolt's capacity.
Therefore, substantial nargin exists for the larger U-bolts.
- 25. In sunnary, the load ratings for U-bolts neet the requirenents of the ASME Code for qualification by type testing. The use of allowable U-bolt ratings detemined by qualification testing will reliably ensure a conservative design and meets all design criteria.
- 26. Interaction equaticus for tension and shear are used in bolting applications. The fom that the equation takes is dependent on the i
hO l
l i
application. In accordance with ASME Section III, Appendix XVII, O paragraph 2461.3, the capacity of a bolt in a bearing type connection is deternined using the following expression:
2 f
f*2 +
s j Z
F, g F s
where fe = conputed tension stress fs = computed shear stress Fe = allowable tensile stress at tenperature Fs = allowable shear stress at tenperature l 27. This is exactly the equation appearing on DCP Drawing 049243 which was used for the qualification of U-bolts. Because no guidelines are given in NF-3261 for the conbination of load ratings established for a O particuiar restrained direction, tension and shear loads were conbined in accordance with ASME Section III, Appendix XVII, paragraph 2461.3(a ) . This equation is used when stresses are calculated for bol ts. Accordingly, it is considered appropriate to use this equation for load ratings as stress and load ratings are directly proportional.
- 28. Although the interaction equation given in Section III, Appendix XVII, 2461.3 may not have been specifically intended to address bolts with conbined tension, bending, and shear, the results of test loading indicate that it is appropriate and conservative for this application.
- 29. The assertion that because Schedule 160 pipe was uF?d in the test, any thinner wall piping could be danaged or " buckle" due to the U-bolt capacity is illogical. The naxinun capacity of the U-bolt and stress 10 _
analysis of the piping at any particular support location are two lO independent issues. Piping stress.at any location in the piping systen is a function of the nonent in the piping component. The magnitude of this nonent is determined b) the seisnic acceleration at the given plant location and is therefore independent of the maxinun capacity of the U-bol t. The U-bolt allowable on the other hand, or naxinun capacity as derived fron he tests, is independent of the location in th'e plant or the piping to which it attaches. 'This concern seems to sten from a lack of understanding of the total design process, both stress and pipe support. and ASME requirenents. The analysis of the piping and 3
subsequent satisfaction of all code requirements ensures that buckling of the piping will not occur. ( ',
- 30. The fact that the tests wereIot performed at elevated temperatures'has no bearing on/ th e load ~ capacities developed in accordance with ASME
'O Section III, Appendix XVII, paragraph 2460. Allowables for bolts are i
derived based on Ultinate Tensile Strength (SU). This value does not change between the ambient test temperatures and 6500F which qualified the U-bolts for all Seismic Category I supports at the Diablo Canyon site. U-bolts have not been used in Seismic Category I ars plications where they would be on lines above'6500F.
. /
- 31. During construction some U-bolts may have been slightly'6ent to align the U-bolt legs with predrilled holes. Any such bending would be of a cold foming nature. It is connon practice to fom material's by cold bending and this would tend .t'o, increase the yield strength properties of the U-bolt, This would create an Ie'ven stronger naterial through cold t
q ,
4 l O
' ,12 -
') .
\
b-
{
l working. It should be noted that the original foming of the "U" shape is done by cold foming during the manufacturing process. In any event, this practice does not reduce the load capacity of the U-bolt.
l V. It is alleged that:
For code breaks, boundaries of Class I seismic systens, there was not enough offset or space between the valves and the large bore piping to avoid unacceptable stress on ,
the sna11 bore pipeline branches. The vendor had not received correct instructions since they were told to
. - install the piping at roon temperature. DCPP requires j seismic supports, and has to endure temperatures in
! excess of 6500F. (Stokes, 11/17/83, pp. 8 and 9)
- 32. The tem " code break" is used to describe t'he section of a piping systen where the safety-related piping (Class I) changes to nonsafety-related (Class II) piping (see figure below). This " code break" section is always located on the Class II piping and starts at the valve which is 4
O the point at which the fluid system class changes from Class I to Class II. Within the " code break" section is a system of supports or an anchor that dynanically isolates the Class I piping from the remainder of the Class II piping. The " code break" section of the pipe ends when dynanic isolation has been accomplished. The criteria used to achieve the desired isolation, as discussed in the PGandE Phase I Final Report, require that the system of supports that provides dynamic isolation be nade up of either: (1) an anchor or (2) at least two lateral supports in each direction and one axial support. The anchor, or supports, are
^
denoted as Class II* supports and are designed to the same criteria that are used for Class I supports.
!O I
l
I Class I = Safety-related Ciass II* = noasafety-reiated O m Code break section
- but supported to achieve val.YE END OF isolation of the Class I piping CODE BREAK (" Code Break" section)
Class I Class II* Class II Class II = Nonsafety-related '
nonseismic design s m 1 1
- 33. In the above schenatic, the length of Class II* piping is not important as long as the code break requirements are net by providing supports or an anchor. If the length of the Class II* section of piping can be
~
shortened by relocating the Class II boundary closer to the Class I boundary, the systen would then require fewer Class II* supports; this relocation is only accomplished by adding supports or an anchor to the code break section closer to the Class I boundary. As an exanple, assume that following the valve, the code break section included five bilateral supports (these provide support in both lateral directions at O one location) and then an axial support. All these supports would require Class 1 qualification. Two alternatives for improvenent of the design that are acceptable and neet all licensing criteria are: (1) to aod an anchor at the location of the first bilateral support, or (2) to add an axial support at the location of the second bilateral support.
Both alternatives reduce the length of the code break and the number of supports requiring Class I qualification and meet all licensing criteria.
- 34. The allegation that the code break boundaries were relocated in violation of, sone engineering precept, project instruction, or licensing criteria is fallacious. While it is true that the length of Class II*
l piping was mininized wherever possible by nodification or addition of supports, there is no reason not to reduce the anount of the Class II*
piping to the minimun.
O l
l
- 35. Ndependent of the vendor procedures for original installation, the O '
rece'nt reverification effort has considered 1007. of the code break issues as well as all systens with high tenperatures. Therefore, we are confident that sufficient offset or space exists between valves and large bore piping to avoid unacceptable stress on small bore pipe branches.
- 36. The allegation that the offset is insufficient to avoid unacceptable stress on the snall bore branch lines evolves from a misunderstanding.
It apparently comes from a belief that ME-101 analysis of offset is less reliable than M-40. IE-101 is a computer progran that perforns static and dynanic response spectra modal superposition solutions. M-40 is a hand calculation technique based upon simply supported spans. Either technique is acceptable.
(0 .
VI. It is alleged that:
Engineers who questioned suspect assumptions were transferred to Unit 2. Cooperative engineers plus new recruits were assigned to Unit 1. (Stokes, 11/17/83,
- p. 9)
- 37. Contrary to statenents in the affidavit, no attempt was made to deternine personnel assignnents on the basis of objections or questions raised regarding Unit 1 activities.
i
- 38. When the OPEG snall bore piping group was established in the fall of 1982, all efforts were directed to Unit 1 activities. At the tine, there was no specifically defined scope of work or schedule for Unit 2 activities. Consequently, the entire OPEG small bore piping group was assigned to Unit 1. By early 1983, the Unit 2 scope and schedule were l
defined and it becane necessary to increase OPEG nanpower to support O Unit 2 work in addition to the ongoing Unit 1 effort. Accordingly, additional trafier space and engineers were obtained for that purpose.
The decision to establish physically separate teams for the two efforts was based on the desire to assure proper managenent of the two activities. The separate teans within OPEG facilitated independent scheduling, production control and output tracking, control of manhour expenditures against separate project budgets, coordination with the two separate and independent Unit 1 and 2 project teams in San Francisco, and prevented intermixing of calculations, calculation files, support drawings, and other potential administrative problens.
- 39. The basic consideration in establishing the makeup of the two teans was to provide each with an essentially equivalent nix of new assignees, easiaeers with more pro 3ect experience and appropriate supervisory
.( O personnel, such that each project effort could be supported equally.
Security clearance for access to the plant was not a consideration in these assignnents since the relaxation of plant security procedures effective in March 1983, allowed all pipe support engineers equal plant access to Units 1 and 2.
VII. It is alleged that:
These Unit 1 engineers redid calculations entirely for all failed systens. The original calculations vanished with no mention of the failure. The calchlation logs l
l were also rewritten and falsified. Unit 1 would have failed the reevaluation progran and required complete reanalysis. (Stokes, 11/17/83, pp. 9 to 11)
! i.s 0
l
- 40. In verifying the adequacy of existing designs at Diablo Canyon, O engineering design practices consistent with both nuclear and non-nuclear applications were followed. These engineering practices utilized iterative engineering calculations to verify a design that is consistent with the acceptance criteria. It is comon practice to do initial calculations using conservative data and simplified methods.
This can save the tine and expense associated with more detailed, tine When consuning, sophisticated calculations (such as conputer analyses).
an initial calculation using conservative data denonstrates one or more ;
acceptance criteria are not met, an engineer performs additional trial calculations that use more precise input data. Input data can be 4
modified by removing unnecessary conservatism or by selecting more appropriate boundary conditions as an alternative to using progressively more sophisticated approaches.
Q J 41 . Typically, engineers are trained to employ the use of more sophisticated analytical techniques if initial conservative analyses are not acceptable. For exanple, a hand calculation might be replaced by a static computer run, then by a dynamic linear-elastic computer run, and finally by an inelastic time history analysis. All of these increasingly sophisticated analytical methods yield results that are entirely acceptable in accordance with the design criteria.
- 42. The net result of this engineering process is a completed analysis which must be in full compliance with the design criteria and which meets all design paraneters. The documentation of such an analysis constitutes support and verification of the final design. Intermediate calculations l
- 0
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i
I which are not part of the final calculations need not be retained.
O Quality procedures do not require retention of these unapproved, internediate calculations.
- 43. ANSI Standard N45.2.9 (1979) does not require retention of internediate calculations. The only calculations required to be retained are the final calculations which reflect the analysis actually relied upon to show adequacy of design. Superseded calculations are not required to be
(
retained by regulation, regulatory guide, standard, or any procedure to which Diablo Canyon is or has been connitted to. Despite this fact, DCP procedures, based on judgment of the analyst and checker, call for retention of superseded calculational records "to the extent necessary
- to support and verify final designs." This allows an accurate i
reconstruction of each calculation. The cover sheet of each calculation O package contains a change sheet which shows the history of all revised calculations. A review of these records indicates tha_t nore than 70 calculations contain Mr. Stokes' signature in one of their versions.
The calculation logs may, however, be changed to reflect only the latest revision and signatory engineers. This normal practice does not constitute falsification of records as alleged.
VIII. It is alleged that:
Management's first approach to nake Unit 1 look good was to reduce code break spans. This was not done because there was no plausible explanation for it. Managenent decided to use new assumptions that would change the l results fron fail to pass by assuning gaps that did not i
exist or vice versa. (Stokes, 11/17/83, pp.11 and 12)
,.O lv L-- _ _ _ _
- 44. Since Mr. Stokes did not perfom any conputer piping stress analyses at Diablo Canyon, he was probably unaware of the applicable specific design requirenents. However, he is correct in noting that actual restraint clearances, or as-built gaps, are sonetimes included in the qualification calculations as described in Piping Procedure P-11 (Section 4.6.2) when perfoming small bore piping stress analysis for themal expansion or themal anchor notion. The gaps that are included are physical clearances that exist between the pipe and a structural element. Themal loads can be elininated by gaps in pipe supports and, therefore, the inclusion of gaps in the qualification analyses is completely appropriate. In each case where gaps are included to reduce themal loads, adequate assurance is available that the gap can be relied on to be present throughout the plant lifetime.
.O 4s. Before any gaps were inciuded in a pi,ing stress anaissis, eiping Procedure P-11 required as-built reverification. Accordingly, a plant I
walkdown was conducted to establish the actual gap configuration. The gap configuration was modeled and included in the documentation of the stress analysis calculation. This practice of including gaps to reduce themal loads is used in the industry as a nethod of accounting for actual plant conditions.
- 46. As a result of the NRC Staff's question emanating from this allegation,
' a review of all small bore piping stress analyses was conducted. The results of the review denonstrated that as-built gaps were included in l
25 piping analyses affecting a total of 64 pipe supports. The 64 supports represent approximately 3% of the supports analyzed. As m
O l
1 l
reported in the Project's supplemental letter to the Staff dated
)
O December 28,1983,16 of 25 piping stress analyses involved piping with service conditions below 2000F. In these 16 analyses, themal novenents are minor and not of technical concern. The 9 renaining pipe stress analyses involve 16 supports which is less than 1% of all the small bore pipe supports analyzed.
- 47. A description of the 9 pipe stress analyses in which as-built gaps were nodeled into the computer analysis and the piping systen tenperature exceeds 2000F for nomal themal load cases was presented in the December 28, 1983, letter. These 9 analyses fall into two categories.
Category 1 gaps were modeled to acconodate themal anchor novenent (TAM) of large bore piping. Since these gaps are caused by the themal novenent of large pipes and equipment expected to have repeatable themal growth, the gaps are expected to be present throughout the sQ plant's lifetine. All but one support falls in this category. Category 2 consists of gaps nodeled to release themal loads and stresses induced by two opposing supports restraining the pipe in the same direction.
Because of the piping configuration '1at exists, it is clear that the as-built gaps will renain throuaN ,; the plant's lifetine.
- 48. The consideration of actual restraint clearances, as described in the supplemental Decenber 28 letter, is a reasonable and adequate technique for the piping geometries involved. This method is consistent with the licensing criteria for Diablo Canyon and has gained widespread use in the nuclear industry where ignoring as-built gaps results in exussive themal loads, uo
IX. It is alleged that:
O Management assuned joint releases for rigid connections which neans that welds which were in place were assumed to be nonexistent. (Stokes, 11/17/83, p. 11)
- 49. " Joint releases" refers to a method of providing an accurate representation of end connections in structural nenbers. An initial calculation of a pipe support frame night conservatively assune that welded ends at structural nenbers are completely rigid. However, it is obvious that no joint is completely 100% rigid. The structural nenber may have very little noment resistance in some rotational axes, and assuning rigidity is not representative of actual behavior. An engineer nay nodel the joint to closely represent its actual physical characteristics. In many instances, the joint is modeled so that no moment resistance is offered by the steel to which the nember is attached (i.e., assune that nonent loads are not transmitted). This O
nethod provides a core realistic nodel of the structural behavior of the frane.
- 50. The weld at the joint is still considered in the computer model and there is no intent or need to remove it since the forces transnitted by the weld and associated stresses are evaluated and verified to be acceptable. This practice is standard in structural engineering j evaluations of frane structures.
X. It is alleged that: .
t Hangers still failed and nanagement requested designers to perforn reverse calculations to detemine the naxinun loads that each hanger could support. After naximum loads were established, results were returned to the stress group. (Stokes, 11/17/83, p.12) l 1
1 1
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- 51. Different methods exist to qualify a piping systen to design criteria.
,O These nethods often require interaction between engineering designers.
An axample of this can be seen in small bore piping qualification where the pipe stress analysis produces reactions or loads on the pipe supports. After obtaining the loads on the supports, the pipe stress analyst transnits results to the pipe support engineer for his use in qualification or design of the supports for these loads. The pipe support engineer reviews existing as-built pipe support drawings. If the support is determined to be inadequate to sustain the given load, the support designer and the stress analyst may well review the systen to deternine if the engineering assunptions in the piping stress analysis have excessive conservatism. An additional series of nore realistic calculations may be perfomed before it can be shown that the
'tO support meets criteria. This process of recalculation may occur several times before the support is qualified. Such an approach is a logical and orderly nethod of qualifying small bore piping systens and does not violate any design or licensing criteria or regulatory requirenent.
- 52. Another method used to qualify a piping systen involves use of the maxinun capacity of the pipe supports for qualification. This nethod can be more efficient than the method discussed above by reducing the nunber of interactions and recomputations between the stress analyst and the pipe support engineer. In this situation, .the pipe support engineer calculates the maximun capacity of a support for each load case. This information is provided to a pipe stress analyst, who compares the computer results of the piping stress analysis to these maximun l.0
allowable loads. If the calculated support loads are in excess of the allowable, the piping analyst may be able to perfom a reanalysis j iteration without requiring the pipa support engineer to recalculate stress in the support. This method does not alter the final result since both the piping and the supports must be shown to be qualified to the applicable licensing criteria. When the piping analysis is complete, all loads are transnitted to the support engineer for final acceptance, or support modification, and documentation. This reverse calculation technique is often used'in the industry and is analogous to calculating an acceptable " load rating" of a support.
- 53. This question also conveyed the implication that intemediate or iterative calculations were being improperly destroyed. Such an inplication is erroneous. Pursuant to procedure 3.3 contained in the O esande engineering aanuai, aii finai (i.e., approved) caicuiation packages are retained and pemanently filed. There is no regulatory or project requirement to retain the intermediate or iterative analyses.
XI. It is alleged that:
Another technique of adding new supports within six inches of failed existing supports was used. The stress group then modeled new support gap assumptions so that the new supports would handle nost of the load. Instead of naking necessary repair for a pipe resting on a unistrut, this unintentional restraint was modeled as a pipe support instead of being removed. The solution was i
to renove the unistrut and add a full-sized support.
(Stokes, 11/17/83, pp.12 to 14) l
- 54. New pipe supports were added to small bore piping for many reasons; e.g., to neet code break, valve acceleration, or themal criteria. In
- O
some cases these new supports were located near existing supports. This approach would obviously have the effect of reducing loads on the !
existing supports. The small bore piping program was explicitly conducted to ensure that all supports net the licensing criteria. In sone cases, conditions were nodeled where a structural restraint that was not a pipe support was present. For example, there are several instances in which a penetration was modeled as a seismic restraint.
When a support was modeled in the final analysis, either a support or restraint physically existed in the plant or, in the case of a design nodification, a new support point was nodeled in the stress analysis calculation. If a new support is added, a documentation number is assigned to the new pipe support and remains with it throughout the design, construction, as-building, and final engineering approval
~O cycie. This documentation trail ensures that the support is constructed in accordance with the design requirements.
- 55. During the course of modifying piping supports, interferences and obstructions were encountered. These were identified to Engineering and dispositions requested. As an example of this process, it was noted in one case that a Unistrut bean for the support of electrical conduit was i constructed near a pipe and subsequently identified to Engineering for ll disposition (Allegation 89 from SSER 21).
- 56. In a case such as the one involving the above-nentioned Unistrut, First, Engineering went through the following process of qualification.
an attenpt was made to requalify the systen with the added restraint of the Unistrut present. In this case it was not possible to protect the O
1
Unistrut so the addition of a support at the location of the Unistrut O was investigated. This investigation showed that the Unistrut was not required and it was renoved fron the plant. All of this was part of the iterative practice of qualifying an installed piping systen and is not unique to this plant. All applicable procedures were followed in this process and all design criteria were net. In fact, it would appear that this situation clearly denonstrates good connunication between Construction and Engineering, sound engineering practice, and a proper solution that resulted in a systen that neets the design criteria.
XII. It is alleged:
There was a coverup of defective naterials fron Pullman associated with a 50,000-pound bracket on a 20-inch line. The bracket was deforned and failed testing.
Management instructed that only visual inspections be 9errorned oa repiaceaeats- (stokes. 11/17/83. 99 14 and
'O 15)
- 57. The alleged material deficiency discussed in the affidavit was investigated. It involved support 1029-5CS, which is a constant rate spring support used as a dead load support on a 28-inch stean line (not 20-inch as alleged in the affidavit). The " cracks" nentioned in the affidavit were in fact laminations as determined by ultrasonic testing which connonly occur in this type of SA-36 plate, and it is not surprising that ultrasonic or nagnetic particle testing would indicate this condition existed. These laminations do not detract from the conponent's load capacity. In addition, the pieces exhibited punching narks which the component manufacturer has certified do not affect the component's capacity or function.
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- 58. To be conservative, the contractor returned four of the ten brackets included in the order while the two on 1029-5CS were scrapped.
XIII. It is alleged that:
Pipe stress and support engineers were nomally not allowed to prepare Discrepancy Reports. Foley and Pullnan, however, regularly prepared these docunents.
(Stokes, 11/17/83, p.15)
- 59. Training is required of all engineering personnel shortly after assignnent to the project which includes indoctrination in the purpose and use of a Discrepancy Report (DR), as well as a Nonconfomance Report (NCR), and a Design Change Notice (DCN). Project training records indicate that Mr. Stokes attended this training on Novenber 8, 1982, i
shortly after his arrival onsite. NCRs are addressed in Engineering Manual Procedure 9.1 and DRs are addressed in Engineering Manual (O
Procedure 10.1.
- 60. Procedure 10.1 provides that any individual can identify a potential discrepancy and bring the natter to the attention of the responsible Engineering Department group leader or supervisor. The supervisor is responsible for detemining, after investigation, whether the identified iten is a non-confomance, a discrepancy, or neither, and directs that the appropriate report be prepared. During the course of the OPEG piping design effort, there were nunerous instances identified by engineers which required discussion and clarification of the design basis for itens which were unclear to specific engineers. This is not i unexpected in the nomal course of design engineering activities where solutions to engineering problems are developed. Identification of
<O
1
" potential discrepancies" which, upon further investigation, proved to )
O be of no concern were not frequent, but did occur from time to tine.
61 . A DCN is a document used by engineering to effect a modification to an approved specification, drawing, or supplier document that results in a plant modification or revises any other design document or license requirement. Contrary to the allegation, it is not a docunent for engineers to initiate modifications in response to QC inspections unless l the inspection should result in a redesign modification. Procedures controlling the use of DCNs are addressed in Engineering Manual Procedure 3.60N for Unit 1 and Project Engineers Instruction No.16 for Unit 2. Contrary to statements in the affidavit, nunerous DCNs have been initiated by OPEG pipe stress and pipe support engineers to nodify pipe routing and pipe support designs as required by their engineering O aa irsis- 8 tweea aaauars =ad octa6*r of 1983. over zoo $=ca oca$ < re initiated by OPEG engineers. How Mr. Stokes could be unaware of this fact and yet have the knowledge of how the process worked that he alleges he has is, at best, curious.
- 62. Numerous controlled copies of the Engineering Manual were provided as reference documents within the various OPEG office spaces and were easily available for use in clarifying any questions which might arise concerning DRs or DCNs.
- 63. Contrary to statements in the affidavit, neither Foley nor Pullman activites are controlled by the PGandE Engineering Manual and they i consequently do not prepare Engineering DRs or DCNs. Documenis used by Foley and Pullman which are called a DR or a DCN are different docunents from those described in the Engineering Manual.
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XIV. It is alleged that:
O In the three DRs written by Mr. Stokes, flare bevel, flare V and other partial penetration groove welds for pipe supports were deficient. However, managenent insisted AWS standards did not apply to DCPP. (Stokes, 11/17/83, pp.16 and 17)
- 64. The effective throat of flare bevel and flare-Y groove welds are in accordance with AWS Dl.1 Structural Welding Code prequalified condition. In the case of flare bevel welds, the effective throat is taken as 5/16R, where R is the corner radius. This approach is very conservative and AWS Dl.1 recognizes the conservatisn of this approach by not requiring qualification. Had the project desired, even larger effective throats could have been justified per AWS Dl.l.
- 65. In accordance with AWS D1.1., Section 2.1.3.1 and documented understandings between Engineering and Construction, dimensions are not tr di=eastaas are aat provided the
~O re9 aired aa fiere sraave welds-neaning of the symbol is to weld the flare groove joint out flush with the corners. ESD 223 Section 6.8.2.6 D requires a visual inspection to ensure that the weld is acceptable. The design of welds does conform to the requirements of AWS Dl.l. The requirement to completely fill groove joints flush provided the most simple and conservative instruction to construction and inspectors. This eliminated the need for a dimension and related field neasurenents.
- 66. Bevel angles are not required to be placed on the design weld symbols as these are included with the Weld Procedure Specification (WPS) which I
provides direction to both the welder and weld inspectors. Flare groove joints do not have bevel angles and bevel angles cannot be shown on the uO
O design weld synbol. It is not necessary to limit the bevel angles to those given for prequalified welds in AWS Dl.1 Figures 2.9.1 and 2.10.1.
- 67. Dimensions, such as the depth of bevel (S) and effective throat (E), are not required to be placed on the weld synbol per AWS Dl.1 Section 2.1.3.1 for complete penetration welds. For partial penetration joints, l
AWS Dl.1, paragraph 2.1.2.1 reconnends, but does not require, S and E dinensions on drawings. In the case of intersecting nenbers creating weld joints which AWS D.1.1 considers partial penetration welds (for purposes of qualification), but which have no weld groove bevel edge preparation, it is meaningless for the designer to specify S (bevel groove depth) because there is no bevel groove preparation. EDS-223 provided an effective and simple alternative to measuring (S) and (E)
O dimnsions. For the joints between skewed intersecting menbers, it is impossible to directly measure dimension (E) (effective throat).
ESD-223 provided an instruction which specified the simple measuring gauge to be used and a conversion table relating the design drawing .
dimension to an easily measured dimension. The use of the gauge and the table neans that the Pullman inspectors did not need effective throat (E) on the drawings, and it was appropriate to take that dimension off drawings because it cannot be measured.
- 68. It is not necessary to adjust the fillet weld leg size to have all the 1
l welds in a joint have the same effective throat. Adjustments are made in the weld calculations to account for the varying effective throats and the consideration of the local dihedral angle has been made in the calculations. Even though fillet weld synbols have been used for
dihedral angles less that 600, calculations are performed to ensure O that the weld qualifies as a partial penetration weld with the proper throat reduction. This reduction is in accordance with the requirenents of AISC and AWS.
- 69. Pullman Power Products procedures reference the PGandE specification to which pipe supports, are to be installed and the codes to which the weld
. procedures specifications (WPS) are qualified. For the WPS which are qualified, it is not necessary, and inappropriate for Pullman QC to inspect the welds to the AWS D1.1 prequalified joints. The weld procedure specification, ESD-223, and the design drawings contain everything needed to inspect the welded joint. Flare groove welds are inspected in accordance with the requirements of ESD-223.
- 70. It is not necessary for Attachment I of ESD-223 to provide Ifnf tations for the mininun dihedral angle for intersecting structural shapes. The Q limitations on the dihedral angle would be governed by the design drawings used. Throat adjustments are reflected in the weld design calculations. The calculation adjustments have taken into account the effect of skewed dihedral angle rather than perpendicular connections, I and have considered that acute angle connections will not have conplete fusion to the weld root, due to possible slag inclusions.
, XV. It is alleged that:
1 The second Stokes DR stated that angle nenbers were two-to-three tines too long for the allowable bending stress standard used under the AISC code. The angles could buckle under pressure. One hundred frames of 300 checked contained violations. (Stokes, 11/17/83, pp.17 and 18)
O The M-9 conputer analysis for angles onitted the relevant provisions of the American Institute of Steel Construction (AISC) code for allowable bending stress, contrary to licensing connitnents. (Stokes,1/25/84, Tr. 15-21 )
- 71. In paragraphs 71 thru 78, the following synbols are used.
List of Symbols B = Length of angle leg t = Thickness of angle leg L = Length of span
- Fy = Mininun Yield Strength bf = Width of Conpressior. Flange
- 72. The criteria for the use of angles as laterally unsupported beans subjected to bending forces were based upon evaluations initiated in 1977. Project-specific criteria were required because the AISC Manual of Steel Construction (Ref.1) does not provide guidance for angles with laterally unsupported spans greater than 76.0f b / W. The tern 76.0 bf/ WE is the allowable span for an unbraced length of a nenber not neeting the requirements of Section 1.5.1.4.6a of Reference 1. However, these criteria were developed for I beans and not specifically for angles. Reference 1 does not provide criteria for laterally unbraced menbers greater than 76.0f b /M. The lack of specific guidance in this area has been recognized in the literature (see Reference 2).
However, AISC recognizes that special investigations are necessary for angles with laterally unsupported spans greater than 76.0 bf / Wy.
This is indicated on page 2-21 of Reference 1 where a statenent is l
l
provided which explains the use of angle load tables. The statement is O as follows:
"The tables are not applicable for angles laterally unsupported or subjected to torsion; for such nenbers a special investigation is necessary."
- 73. Because the AISC did not completely address the design of laterally unsupported angles, PGandE perfomed a literature search in 1977 to detemine if other infomation was available which would be adequate to develop criteria. In late 1977 it was found that a theoretical solution The to the design of laterally unsupported angle beans was available.
theory had also been verified with extensive testing. The theory and the testing were completed in Australia (Reference 3, 4, and 5).
- 74. In the Australian tests, various sites of angles were characterized by O dirrereat s/t ratias- ^asie sectiaas with s/t retias betweea 5 aad is (Reference 5) have been tested. The nafority of angles at Diablo Canyon fall within this range. The only angles at Diablo Canyon not falling into this range have B/t values less than 6. However, at this end of the range (beans with B/t less than 6 are less slender) the data can be used conservatively since the net effect is to allow an increase in acceptable unbraced lengths. Based on the tests and comparison to structural theory, simple fomulas were developed in Reference 5 for use in the design of laterally unsupported angles in bending using several different nethods of load application.
- 75. For all the various angle ss::tions and load cases investigated, Reference 4 reconnends that an allowable bending stress of 0.66 Fy nay
-O
A be used if L/t is less than 300. The Diablo Canyon Project Design O Criteria M-9 limits the maximum bending stress to 0.6 Fy and a naximun L/t ratio of 270. These limits used at Diablo Canyon fall within the reconnendation of Reference 4 and are therefore acceptable,
- 76. DR 83-042-S, written by Mr. Stokes, questioned the acceptability of certain unbraced angle nenbers because the unsupported spans of those members are greater then 76.0f b / K per section 1.5.1. 4.6b of Reference 1.
It should also be pointed out that the 18 pipe supports identified in 77.
the DR 83-042-S as discrepant have been reviewed. All of the angle bean spans are found within the Project Design Criteria.
- 78. It is concluded that the Project Design Criteria on the design of laterally unsupported angle beans has adequately covered the length (Q greater than 76.0 fb / K .
References
- 1. American Institute of Steel Construction (AISC) Manual of Steel 1 Construction, Seventh Edition, AISC, New York.
- 2. B. F. Thonas, J. M. Leigh, M. G. Lay, Civil Engineering Transactions, 1973, The Institution of Engineers, Australia.
l i
- 3. B.F. Thonas and J. M. Leigh, The Behaviour of Laterally Unsupported l
Angles BHP Melb. Res. Lab. Rep. MRL 22/4, Lecember 1970.
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- 4. J. M. Leigh and M. G. Lay, Laterally Unsupported Angles with Equal and O Unequal Legs. BHP Melb. Res. Lab. Rep./ MRL 22/2, July 1970.
- 5. Safe Load Tables for Laterally Unsupported Angles, Australian Institute of Steel Construction, September,1971.
XVI. It is alleged that:
The third Stokes DR stated the distance between the center of Hilti bolt holes was not verified as the same length required and specified on the drawing. QC had neasured the distance between the centers of plates attached to the bolts whereas location of the bolts is supposed to be control for the location of the plates.
As a result, whole packages could be in the wrong location. (Stokes, 11/17/83, pp.18 and 19)
- 79. The capacity of a concrete anchor bolt is a function of the bolt length
.: O (eabed eat) boit teri i . =ad coacrete strea9th- Aachor bolt capacity relates to a shear cone of concrete originating at the end of the anchor bolt enbednent. This cone projects at a 450 angle to the surface. If
' two anchor bolts are placed close enough together that their shear cones overlap, some of the strength of the anchor bolts may be lost. The 10d (bolt diameter) criterion between anchor bolts was established to assure l this would not occur.
- 80. All shell type anchor bolts on Diablo Canyon have an embedment of less than five bolt diameters. Since the anchor bolt center lines are ten bolt diameters apart, the shear cones can never overlap. Hence the anchor bolts retain their full capacity. The capacity of La anchor bolt is determined by test. The test for a shell anchor is normally O
l
perfomed on one anchor at a time. The anchor bolt will develop that O full capacity so long as no adjacent anchor bolt is less than then 10 bolt diameters away. In other words, the criteria that determines the required spacing is solely a function of concrete failure theory and i
' test results which are categorized by bolt dianeter.
- 81. Tests to validate this premise were conducted in 1962 on a Phillips shell type anchor. The results reported no reduction in capacity for ten bolt diameter spacing. It is true that the recornendation in the Hilti catalogue is to space the bolts 10 hole diameters apart. However, when the actual shear cone is developed, the results are bounded by the 10d bolt criterion,
- 82. The allegation as to the measurement of the centers of plates rather than the location of bolts is difficult to understand. The design location of a base plate is defined on the hanger detail and is dimensioned to the building structure, i.e., elevations and column lines. On the other hand, the required anchor bolts are defined with respect to the base plate, not the building structure. During the installation, the design location of the base plate is marked on the wall and an instrument is employed that locates reinforcing bar within the concrete. The rebar locations are also marked on the wall. Anchor bolt locations are then selected that most closely approximate their design locations without cutting the rebar.
- 83. If anchor bolt locations relative to the base plate are within established construction tolerances fron the design location, construction proceeds. If the location is outside of tolerance, the I
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Pipe Support Design Tolerance Clarification (PSDTC) group would be asked r
O for approval to deviate and upon completion of the installation an as-built drawing would be transmitted to Engineering for final approval as rtiquired by procedure. In the manner described above, both the plate locations relative to the building and the bolt holes relative to the plate are known, documented, and receive Engineering approval resulting in all licensing criteria being net.
XVII. It is alleged that:
Access to Quality Control and NRC personnel by employees was restricted. (Stokes,2/8/84,p.1)
- 84. Diablo Canyon Project written procedures stress bringing potential problems to the attention of engineering supervision in a timely manner so that appropriate steps can be taken to identify and inplement any iO
'- corrective action necessary to resolve the concern and prevent future recurrence.
- 85. Engineering Manual Procedures covering Discrepancy Reports (Procedure 10.1, paragraph 3.1) and Nonconformance Reports (Procedure 9.1, paragraph 4.1.1) both specifically state that anyone who believes he has identified a potential engineering discrepancy or nonconformance should bring the natter to the attention of the appropriate Engineering Department group leader or supervisor for resolution.
l l 86. These clearly written project procedures do not restrain or prevent engineers from discussing potential problems with representatives of quality control or the NRC. These procedures recognize that many concerns raised by engineers are of a nature that ne easily be resolved tO
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by the supervisor who possesses a broader knowledge of the project. If O needed, the supervisor nay involve staff specialists or engineers from i other disciplines to assist. In no event does nanagenent discourage engineers, or any other person, from raising legitimate concerns. (See Exhibit 1, dated March 22, 1982, and referencing previous policy statements dating back to the 1970s.)
- 87. Quality Assurance and Quality Engineering personnel have been physically i
i located within OPEG and have been available at any time to discuss and assist with the resolution of quality problems. Training sessions were held in support of the written procedures. The training sessions on the Engineering Manual procedures, which are mandatory for engineering j personnel, specifically include a description of Discrepancy Reports and I
i Nonconformance Reports. Project records indicate that Mr. Stokes
( received this training in November 1982.
- 88. During the course of quality audits or NRC inspections of engineering work, auditors may ask questions about which individual engineers may not be well enough informed to provide accurate, comprehensive responses. An individual engineer night be questioned about work he is not directly involved with and therefore not be specifically familiar with in detail, or about nore general progran aspects of which the individual engineer nay not have an overall perspective. An excellent example is Mr Stokes' lack of knowledge as to the justification
( Australian test data) for use of angle-shaped nenbers (see January 25, 1984 transcript, p.126). To minimize a nis'7 formed response to auditor questions, a knowledgeable supervisor normally participates during O
audits of these kinds. If questions are raised that cannot be answered
.,O by those present during the audit or inspection, they should be presented to a supervisor or soneone else to assure that the responses are complete and accurate. This policy is intended to ensure that accurate information is provided during audit activities and does not restrain or prevent engineers from discussing problems with " quality control" or the NRC.
- 89. Additionally, the Bechtel Power Corporation, San Francisco Power
' Division Instruction 3-17, "10 CFR Part 21, Reporting of Defects and i
Noncompliances," applies to and is inplenented by the Of ablo Canyon Project. This instruction defines responsibilities, establishes requirenents, and provides guidance for actions necessary to neet the reporting requirements of 10 CFR 21. Procedural requirenents to
'(O 'aitiate avaiuatiaa aad r Partias Pursuaat to 1o cra 21 are 1sa contained in this instruction. The instruction is posted in Diablo Canyon Project work areas for reference. Also, PGandE has posted 10 CFR 19 reporting instructions and a copy of Fom NRC-3 which gives guidance for contacting the NRC and the regional NRC phone nunbers and addresses. These documents have been posted in all PGandE facilities (i.e., PGandE headquarters, construction offices, and operating facilities as well as in the OPEG offices).
- 90. The nethods described above have been available to project personnel to process a design issue which they felt could potentially affect the safe design, construction, or operation of the Diablo Canyon Power Plant.
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XVIII. It is alleged that:
Alien engineers (green card holders) were employed and intinidated by fear of dismissal, to approve incorrect design practices. (Stokes, 2/8/84, p. 2)
- 91. Managenent has not and does not practice intimidation in order to supervise engineers in the perfomance of their work. Only U.S.
citizens or Green Card holders were employed as pipe support engineers in OPEG. Further, pemanent residency (green card holders) in the United States allows a person the rights of a U.S. citizen except the right to vote. (8 U.S.C.S. Secs. 1101, 1251)
- 92. If an enployee on green card status is laid off or teminated for any reason, this temination has no influence on their pemanent residency status. They are free to stay in the United States and seek other employnent in exactly the sane manner as a U.S. citizen. It is obvious 1
O that even the means to intimidate an aiien ensiaeer as aile9ed by ar-Stokes simply do not exist.
XIX. It is alleged that:
The Quick Fix or Pipe Support Design Tolerance Clarification (PSDTC) program was not subjected to controlled documents, the engineers and QC inspectors were not provided clear instructions, those instructions that did exist (the June 16, 1983 meno) were insufficient to define the authority of the PSDTC engineers, there was no fomal review of the Quick Fix work, and the Quick Fix progran bypassed the fomal QA reporting requirenents l
which prevented reporting of serious hardware l deficiencies. (Stokes, 2/8/84, pp. 2 to 4)
- 93. In January 1983, a special team of pipe support engineers was established within OPEG whose assignment consisted of direct engineering liaison with General Construction resident engineers and Pullman Power O
i
Products craft personnel. The purpose of this group was to provide O expeditious resolutions of minor construction difficulties in the installation of large and sna11. bore pipe supports in order to mininize construction delays. The responsibilities and authorities of this group were originally provided in Onsite Project Engineering Guide 4.0 on January 7,1983. This guide was superseded by Project Engineer's Instruction (PEI) 12 on March 11, 1983, which defined the PSDTC progran. The practices defined by these two documents were based upon an identical philosophy and intent, and all guidance previously provided to construction under OPEG Guide 4.0 was again reviewed by engineering for compliance with the requirements of PEI 12 upon its issuance.
- 94. As provided in the procedure, field construction prebicas were defined as pipe support installation problems which could not be resolved using the relati$eTy restric ive, construction tolerances explicitly stated in
~
(Q Pullman Power Products docpnt ESD-223, " Installation and Inspection of f +
Pipe Supports". Construction' tolerances contained in ESD-223 were those I r that could be applied to any pipe support in the plant without additional engineering justification.- Changes beyond those tolerances may be pemitted $ased upon the criteria contained in Diablo Canyon Design Criteria Memorandum (DC'1),M-9, " Guidelines on Design of Class I Pipe Supports and Restraints," Field construction problems were referred to PSDTC tean engineers who, based on their engineering l
judgment and kn6wledge of DCM M-9, would, on a case-by-case basis, detemine whether use of expanh'ed ' tolerance limits could be authorized i _
to resolve the constructica p'roblem while maintaining an acceptable support design.
l l ._. . - - -'.- _
~
- 95. Where field resolutions could be made, in the judgment of the PSDTC team O engineer, they were docunented on individual PSDTC foms provided in Attachment A to PEI 12. Field construction problems which, in the judgnent of the PSDTC engineer, could not be resolved without a design change, were returned to General Construction for fomal referral to Engineering as a DP report requesting hanger redesign in accordance with other project procedures. Pre-existing pipe support configurations i found to be in nonconpliance with appropriate design and construction documents were referred for disposition as a Pullman Discrepancy Report or Discrepant Condition Notice in accordance with Pullman procedures.
- 96. The PSDTC engineers were selected from experienced engineers at the jobsite. It was felt that they, Mr. Stokes included, would be in the best position to know whether qualification of the supports could be j (O demonstrated. In no case, however, was the modification made by the PSDTC engineer allowed to be the final design qualification. .
Notwithstanding Mr. Stokes' apparent lack of knowledge, all the PSDTC group's modifications received final engineering review and approval as part of the as-built acceptance, as required by procedures P-10, I-37 and I-40 discussed below.
- 97. When a PSDTC fom was completed, a copy was attached to the pipe support design package and was treated exactly like the original design package in order to assure that standard quality control procedures were applied to all work accomplished by General Construction. Upon completion of construction of the support, the complete as-built package, including any PSDTC foms associated with that support, was forwarded by uO l
l
Construction to Engineering for final acceptance in accordance with O project engineering procedures. These procedures are P-10. "0 PEG Small Bore Piping and Hanger Review Procedure;" I-37, " Instructions for Incorporation of Field Correction Transnittals;" and I-40, " Instructions for the Disposition of As-Builts Associated with Design Change Notices." During the period of Mr. Stokes' enployment, final large bore support as-built acceptance was completed by the project engineering tean in San Francisco, while final small bore pipe support as-built acceptar.ce was completed by OPEG.
- 98. The as-built acceptance process involved review of the revised support design and perfomance of necessary calculations for qualification of the design. Where qualification could not be shown, a new design was prepared and issued for Construction.
The PSDTC program was neither a substitute for nor a deviation from the (Q 99.
fomal design and construction quality assurance processes for pipe supports. As stated in paragraph 2.2 of PEI 12, the procedure was specifically not authorized for use in lieu of a Discrepancy Report or a Design Change Notice. The program was reviewed and approved for use by both Units 1 and 2 project engineering as well as the project quality assurance organization, all of whom signed PEI 12 when it was issued for implementation. In August 1983 an ~ audit was conducted by the PGandE QA Department which resulted in the overall conclusion that the control of design changes by OPEG appeared to be effectively implenented. One finding was identified with respect to use of the PSDTC forms. In response to this finding, speciaI. training sessions were held in October
.O
i 1983 for all PSDTC engineers to emphasize the limitations on the use of j PSDTC foms and to assure that Design Change Notices would be initiated when required by DCP procedures.
100. Uncontrolled documents were not used to promulgate PSDTC program procedures. These procedures were defined in PEI 12 as supplemented in ESD-223, copies of which were provided to the PSDTC team. The details 1
of the program implementation were emphasized with PSDTC engineers in periodic discussions and training sessions. The June 16, 1983 neno, j referred to by Mr. Stokes as an illustration of inappropriate 1
cormunication of program procedures, was, in fact, written by General Contruction to the piping contractor to reiterate construction procedural requirements already well established. Surmarized, the meno states that the PSDTC program is not a corrective action program and may aat be used in 15 u ar coastructiaa discrePeacy r Parts (ons and ocas)-
.( O This neno was not applicable to the PSDTC engineers and as such did not i
receive distribution to them.
1 01 . As stated previously, a discrepancy report rather than a PSDTC form was used to document a pre-existing pipe support configuration which was found to be in noncompliance with appropriate design documents. The PSDTC fom is not a discrepancy report and does not take the place of one. It may, however, be used to provide disposition for a discrepancy i report written by construction. The PSDTC engineer is not, however, I
required to nonitor writing of discrepancy reports by construction.
This would explain why Mr. Stokes did not always see them. Construction l discrepancy reports are produced as required by construction procedures.
l
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l
XX. It is alleged that:
- O Final calculations do not include assunption sheets which would allow specific errors to be tracked. (Stokes, ,
2/8/84, pp. 4 and 5) l 102. In the small bore piping qualification program, important input information for the pipe stress analysis, e.g., nozzle load acceptance, was subject to revision since additional changes in seismic response However, spectra and other related analyses were in progress.
prelininary data were available to allow initial " assumptions" in the analysis to be made. Such calculations were noted as preliminary on the calculation log and in the calculation itself.
103. We believe that Mr. Stokes is referring to these " assumption" sheets that were used to track this preliminary information. As data were l
finalized, the sheets were revised to reflect the updating of I
(O preii=ia ry iara ==tiaa ta a ria=1 re=aiutiaa- wa a 11 d t were The calculation was approved final, the sheet was no longer required.
as final, and these assunption sheets were discarded. When the calculation reached final status, the calculation master log was updated to show that all preliminary assumptions had been resolved by noting the calculation as final in the log.
104. An after-the-fact " paper" trail of all the various changes to l preliminary input data is not required. Final documentation includes I
only the final input data as required by ANSI N45.2.11 (1974). The final input data is retained in the form of input sheets and assunption sheets for all calculational packages.
- O v
- 44 _
XXI. It is alleged that:
Expansion anchor bolts have not been evaluated with respect to ISE Bulletins 79-02 and 79-14. Infomation in PGandE's January 27, 1984 letter is inconplete and of .
suspect accuracy. Expansion anchors would fail during Hosgri and DDE. (Stokes, 2/8/84, p. 5).
105. Initially, design of all expansion anchors installed at Diablo Canyon was in conformance with PGandE's engineering standard drawing 054162.
Subsequently, pipe support base plates and expansion anchors were requalified to conply with the NRC's design reconnendations in I&E Bulletin 79-02*. Expansion anchors used in other applications (e.g.,
i raceway supports and HVAC duct supports) renain in confomance with Drawing 054162 requirenents.
106. The NRC specifically limited the applicability of the sonewhat nore stringent recormendations in ISE Bulletin 79-02 to large bore and conputer analyzed snall bore pipe supports. As stated in the bulletin,
<(
operational problens had been experienced ir, expansion anchors installed in pipe supports. These problens were attributed to factors that primarily apply to pipe support designs (e.g., cyclic loads and flexible baseplates).
1 07. PGandE's January 27, 1984 letter addressed expansion anchors used in applications other than pipe supports. The January 27 letter provided an overview of the basis of the Drawing 054162 design criteria and included tabulations of the factors of safety achieved by using the Drawing 054162 criteria. In addition, this letter addressed testing and
- Mr. Stokes' references to IAE Bulletin 79-14 are erroneous, as the ISE Bulletin does not address expansion anchors, but addresses as-built
, O piping.
l j
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O evaivations perfomed to confim the adequacy of expansion anchors whose installation was not in conplete confomance to Drawing 054162 requirenents.
108. As reported in the January 27, 1984 letter, design factors of safety are almost always well above 3 for Hosgri and DDE load cases. A factor of safety of 3 is considered to be fully acceptable by the industry. The January 27, 1984 letter reported that certain hypothetical limiting conditions night result in a few anchors having a factor of safety less than 3. However, an expansion anchor randon sampling program, which was perforned subsequent to subnittal of the January 27, 1984 letter, found that out of more than 100 electrical supports, having more than 400 anchors, there was no case in which an anchor had been installed
- such that its factor of safety was below 3. These results were reported tO to the NRC Staff by a letter dated February 16, 1984. Further, contrary to the statenent made by Mr. Stokes, this encompasses all load cases including Hosgri and DDE. Expansion anchor design, including actual safety nargins and redundancy, is reasonable, conservative, and neets all licensing criteria.
XXII. It is alleged that:
Due to deficient design drawings for welding, inconsistent and incorrect assumptions were made about certain welds. (Stokes, 2/8/84, p. 6) 109. Contrary to the allegation, there were no deficiencies in the design l
drawings. The Diablo Canyon design groups use a corner radius (R) equal 1
to 2T (where T is the thickness of the tube steel) for tube steel with a
(
t l
perineter of less than 14 inches. A corner radius of 2-1/2T was used for perineters greater than 14 inches. In no case was a corner radius of 3T used. These conditions are reiterated by a March 4, 1983 neno l from Dan Curtis to Diablo Canyon Unit 1 pipe support group and a March 21, 1983 neno from Leo Mangoba to OPEG pipe support engineers.
110. Mr. Stokes alleges that Japanese tube steel with a radii of 1.5T was used at Diablo Canyon. That allegation is false. Pullnan Power Products purchase orders indicate that naterial shall be donestically manufactured as required by the contract. We have researched all structural steel nill certificates to detemine origin and have j confirmed that no Japanese tube steel has been received. We have deternined that two purchases of a small anount (3000 ft) of Canadian tube steel has been used; however, the manufacturing was per U.S.A.
l requirenents.
- [
111. Fifty pipe supports with tube steel nenbers with perimeters greater than In 14 ir.ches were chosen at randon and the corner radii were measured.
' a few cases radii insignificant 1y less than 2-1/2T were neasured. The testing described below has shown that for the radii slightly less than 2-1/2T, an effective throat of 5/16R is obtained and the design f
requirenent net. The tests also show that the 5/16R requirenent is net I when the radii are 2T.
112. AWS D1.1 Structural Welding Code Section 2.3.1.4, allows the use of an effective throat of 5/16R (where R = outside corner radius of tube steel) for single flare groove welds without perfoming a weld procedure qualification test. The 5/16 R dimension is accepted as being a 1
l
. - . - . . - , . , - - - - . s. +
conservative effective throat that can be increased if additional O verifications are nade in accordance with Section 2.3.1.4 (2) of AWS I
D1.1.
113. Test prograns have been conducted which substantiate the effective throat assunption of 5/16R as conservative. One test progran was perforned at Diablo Canyon by Pullnan ar.d a second test progran was conducted by Pullman and United Engineers at Seabrook Station.
114. The tests at the Seabrook Station were conducted using standard Pullman Welding Procedures for carbon steel naterials. The technical report describing the tests is attached as Exhibit 2. The purpose of this test program was "To verify, as a mininun, that the effective throat thickness for a flare-bevel-groove weld when filled to the solid section l
of the bar will equal 5/16R, where R is equal to the radius of the bar." Four sizes of structural tube steel were welded using 3/32" and i
Q 1/8" dianeter E7018 electrodes in the flat, vertical, and overhead welding positions.
115. The results fron the Seabrook Station tests showed that the actual effective throat equalled or exceeded 5/16R (where R is 2T for tubing with a perineter less than 14 inches and 2-1/2T for perineters greater than 14 inches) by as much as a factor of 1.0 to 1.92, with an average factor of 1.4. The nininun effective throat occurred when 3X3X1/4 tube steel was welded using a 3/32" electrode in the flat position. In that case, the effective throat equaled 5/16R.
116. Tests at Diablo Canyon were conducted using Pullnan's Diablo Canyon Project welding procedures. A brief sunnary is attached as Exhibit 3.
l s O
i The tests were performed to verify that the actual effective throat net i or exceeded the 5/16R for the worst case identified by the test progran perforned at Seabrook Station. Six tests were conducted to detemine the typical effective throats which would be achieved for flare bevel joints when welding 3X3X1/4 tube steel using 3/32" and 1/8" E7018 electrodes in the flat position.
117. All tests done at Diablo Canyon indicated that the amount of effective throat exceeds 5/16R by a factor of 1.4 to 1.7.
~ 118. In conclusion, field investigations and tests at Seabrook and Diablo Canyon denonstrate that the design requirements concerning effective throat are consistent with as-built conditions.
XXIII. It is alleged that:
weid procedures ad techaiaues r 41 d to co=a a==t far i
\ O weaknesses in design drawings. (Stokes, 2/8/84, pp. 6 and 7) 119. As discussed above, the design drawings had no deficiencies or
" weaknesses" that required welder or welding procedure conpensations.
120. All weld procedures are written and qualified to ASME Section IX and/or l AWS D.l.1. The weld procedures were not intended, and do not allow, .
welders to conpensate for " deficiencies in design drawings."
121. Weld procedures assure that a completed weld will develop the required l
strength for the type and size of the welds specified in the design drawings. For exanple, the qualification tests for a full penetration weld would ensure that the specified strength of the naterial is developed or exceeded.
O.
127. The engineer specifies a weld type and size as deternined by O conprehensive weld size calculations. The welding is controlled by procedures and is perforned in a nanner that ensures the strength of the ,
1 l
Weld as specified by the designer is obtained. l
- 123. Weld procedures are most definitely not written to allow the welder the flexibility to select weld electrode sizes to compensate for what a welder night perceive as a shortcoming in design. Contrary to the Stokes inplication that only 3/32-inch dianeter or smaller welding ,
electrodes were acceptable to compensate for design deficiencies, and ,
that 1/8" dianeter electrodes were incorrectly used, the tests referenced above have shown that the 1/8-inch diameter electrodes are acceptable.
((]) XXIV. It is alleg?d that:
Weld procedures, specifically Pullman ESD-223, did not require joints to be welded flush for flare or bevel welds. (Stokes, 2/8/84, pp. 6 and 7) 124. Pullnan Power Products Specification ESD-223 establishes the procedures for the installation, inspection and documentation of the final assembled configuration, i.e., as-building of pipe supports. The ,
current version of ESD-223 does not permit flare groove welds to be installed without the weld profile at least flush with the flat portion of structural tubing. Past revisions to ESD-223 have had provisions for i neasuring flare groove welds which were not welded flush. However, these provisions of ESD-223 were not used at Diablo Canyon because the l
Unit 1 design drawings did not permit less than a flush weld.
I
<(:)
~-
t 125. The Diablo Canyon Unit 1 pipe support design groups did not specify i
O dimensions along with the flare groove weld synbol. The flare groove l weld symbol alone requires that the flare groove weld be filled at least flush.
126. Because the flare groove weld synbols on the design drawings did not specify or permit flare groove welds as being other than flush, Construction was required to provide welds which were flush. This was verified by QC inspection. If Construction had provided flare groove welds that were not flush, Engineering would have detected and not accepted the weld during the as-built review program.
127. To verify that this was done consistently, a random sanple of flare bevel welds was inspected to deternine if they were welded flush. A total of 233 welds were examined. All were found to be welded flush, except for minor variations in five instances. Four welds were 1/16" I
Q under flush and the fifth one was 1/32" under flush over a part of its length. The effective throat on each of these five cases was, however, within the design requirements.
128. In sunmary, all flare groove welds were intended and specified to be flush. The design engineers had control over final acceeptance of the welds through the as-built approval process. Verification, through a sanple reinspection, has assured that the welds are, in fact, adequate.
XXV. It is alleged that:
The allowable angle of skewed fillet welds is unacceptable. (Stokes, 2/8/84, p. 8) lto 129. Contrary to the allegation, the angles, bevels, and weld configurations
'O specified by Pu11 nan in their procedures were qualified in accordance with ASME Section IX and/or AWS D.l.1 and were conpatible with design assunptions.
130. ESD-223 did not provide dihedral angle linitations for skewed fillet welds. Limitations for dihedral angles, where applicable, were provided in the design drawings. ESD-223 does not and need not 11 nit the application of skewed fillet welds since such limitations are a design 4 concern, not an installation concern. For buildings, there are no specific linits on the dihedral angle to which a fillet weld can be applied. The AWS Dl.1 Code limits the prequalified status and the nethod of qualification.of skewed fillets. Skewed fillets for angles less than 60 degrees are considered by AWS D1.1 as partial penetration welds for purposes of qualification. Mr. Stokes has confused the (Q ESD-223 provisions for partial penetration welds with skewed fillet welds. ESD-223 has a requirenent for measuring skewed fillet welds by I using a special gauge.
1 31 . The fact that the partial penetration weld table incluaes a 150 angle for this type of weld is only of academic note, since an angic that i
shallow was never specified by Engineering on the design drawings.
XXVI. It is alleged that:
37-1/20 groove welds were improperly used. (Stokes, 2/8/84, p. 7-8) 132. Mr. Stokes is correct in stating that the 37-1/20 weld preparation angle for groove or partial penetration welds does not satisfy the l .O
i requirenents for prequalified joints in AISC/AWS. However, these Codes j
O do not require exclusive use of these prequalified weld joint l
configurations. The codes sinply state that these prequalified joint configurations may be used without further testing. The codes also I provide that other joint configurations are allowed, but they first must be tested to demonstrate acceptability. The groove welds nade prior to June 23,1983 were qualified by testing based upon a 37-1/20 weld l preparation angle as set forth in paragraph 146 below.
XXVII. It is alleged that:
ESD-223 and welding procedures were not available to wel ders. (Stokes, 2/8/84, p. 8) 133. ESD-223 addressed installation and inspection requirenents for pipe supports. The docunent is not a welding procedure specification (WPS)
(O and there would be little reason for a welder to have a copy of ESD-223. Welders need not have copies of a WPS in their possession.
( They need only be familiar with and have access to WPSs. WPSs for pipe supports are so fundamental and basic that the qualified welders would not need copies in their possession during welding activities. As set forth in the affidavit of Richard Etzler, filed contenporaneously herewith (paragraphs 6 and 7) welder qualification, testing, and certification ensure welder knowledge of proper weld procedure.
XXVIII. It is alleged that:
The ESD-223 fillet weld table is inaccurate and does not use the same effective leg as San Francisco design engineers assumed. (Stokes, 2/8/84, p. 8) c0
d i
134. Attachment I to ESD-223 has a table which converts the design weld synbol dimensions to convenient working dimensions for construction and inspection personnel to use, because the direct neasurenent of the I design dimension required on the weld synbol is not possible for skewed weld joints. The use of a table converting design dimension to a working measurable dinension is a fairly common practice and improves quality control functions by making measurements easier and nore direct. Mr. Stokes has confused this table for partial penetration welds as being a fillet weld table.
XXIX. It is alleged that:
Inspection personnel were not qualified to the AWS Code and were not issued weld synbols. (Stokes, 2/8/84, p. 8) 135. The AWS Structural Welding Code did not, when Diablo Canyon started, and (O does not today, require AWS qualified inspectors. Inspectors need not be issued the AWS weld synbols. Knowledge of these~synbols, like nuch other material, is part of an educational, experience or training background. These symbols are connonly available in references and need not be issued to inspectors.
XXX. It is alleged that:
The Quick Fix pipe support engineers removed illegible weld synbols to improperly receive approval by QC inspectors. (Stokes, 2/8/84, p. 8) 136. The purpose of the PSDTC group was to assist in clarifying, on a l case-by-case basis, pipe support design tolerances which were not explicitly included in Pullman Power Products Specification ESD-223, uO
" Installation and Inspection of Pipe Supports." This activity was controlled by PEI Number 12. An integral part of this activity was the interpretation and clarification of weld synbols.
137. As construction workers encountered weld symbols about which they had some questions or as they encountered welds which could not be perfomed due to inaccessibility, the drawing was referred to a PSDTCG engineer for interpretation or adjustment. During that process it is very possible that PSDTC engineers may have substituted welds which provide effective throats sufficient to neet design criteria for welds which are inaccessible or inpractical. In these cases, the PSDTC engineer would elininate the old weld synbol and provide a new weld synbol. An example would be when two sides of a flange are required to be welded with fillet welds, but where only one side is inaccessible. A PSDTC engineer nay substitute a groove weld with the same or greater effective throat and the sane sectional properties as the fillet welds originally specified. The PSDTC engineer making this kind of change would not require access to the support calculations because there is no decrease in the support capacity. Such changes are connon and are documented in the appropriate PSDTC forns. A conplete calculation package, including the as-built hanger drawing is reviewed for final acceptance as set forth in paragraph 139 below.
138. No welds which are necessary for the structural integrity of a support l
have been eliminated by PSDTC engineers without one of the following alternatives being taken:
(a) Substitution of a weld which provides at least an equivalent effective throat;
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(b) Mod ication to the weld pattern to conpensate for the renoved (c) Reference being nade to the design calculations to ensure that the
- structural integrity of the support was naintained; (d) Providing other nechanical connections to achieve equivalent strength.
139. The engineering decisions of the PSDTC engineer are verified during the l as-built review process. At that time the as-built drawing would i
reflect the final weld configuration, as specified by PSDTC, and this configuration would be evaluated by another design engineer to assure the qualification calculations were compatable with the revised weld configuration.
'( XXXI. It is alleged that:
QC was not provided with proper instructions and calibrated tools to neasure radii of flare and flare bevel welds on an as-built basis. (Stokes, 2/8/84, pp. 8 and 9) 140. There are no specific Diablo Canyon or general code requirenents existing for field neasurement of the radius of outside corners for structural tubing. Dinensional and nechanical requirenents are i
controlled through purchase specifications. Pullnan Power Products purchase orders required that naterial be donestically nanufactured to ASTM A-500 specification required tolerances.
1 41. Flare welds (flare bevel and flare groove) are perfonned on tube steel conponents. Design documents indicate where such flare welds are to be installed. Pullman Power Products approved standard ESD-223 requires l
l l ._ . . .
such welds to be flush with the face of the tube steel. Pullman Power O Products Quality Control Inspectors verify acceptability by applying j this criteria. Weld gauges are issued to Pullman Power Products Quality l control Inspectors to facilitate their inspections.
142. As discussed in paragraph 127 above, a recent randon sample of flare bevel welds was reinspected to assure flush welds.
XXXII. It is alleged that:
Pullman changed its procedure to standardize weld bevel for partial penetration welds to 450 in June 1983.
However, welds prior to that date did not neet this requirement. (Stokes, 2/8/84, p. 9) 143. There was a procedure change in June 1983, by Pullman that standardized the weld bevel used for partial penetration welds on pipe support l
conponents to 450 This action, however, was not the result of any
'(O action on the part of Mr. Stokes. A large influx of QC inspectors around that time made it necessary to develop a more standardized approach to the pre-weld fit-up measurenent. Hence, the 450 angle was chosen as a standard with which most inspectors were familiar, not to provide a more acceptable nethod of welding. As described below, the 37-1/20 bevel angle has been qualified by tests and it was and still renains an acceptable bevel angle.
144. In a recent inspection, the NRC Staff questioned what bevel angle was used on carbon steel support nenbers. The review showed that current and recent practice has been to use 450 bevels. This was based upon a June 28,1983, neno to QC inspection, observations in the field, and interviews by the NRC Staff with several QC inspectors. The reference t.O .
1 l.
by Mr. Stokes to a June 23, 1983, meno by Russ Noble does not appear to O be related to the allegation as described, but refers to stainless steel weld procedures 15/16 and 129 to be used for welding butt joints in the pressure boundary of piping. The neno did not apply to partial penetration welds for pipe supports.
145. Notwithstanding the above, PGandE has reviewed welds prior to June 1983. Conversations with QC inspectors and production personnel who have been onsite from the early stages of the Project indicate that the practice was to provide a 450 bevel angle. However, the weld procedure, 7/8, which is applicable to pipe support installation, allows bevel angles of both 450 and 37-1/2 0 and therefore, one might assune there are welds with bevel angles of 37-1/20 146. To qualify the 37-1/20 bevels, Pullman perfomed tests to detemine the anount of effective throat that would be obtained using their (Q welding procedures in a tee joint, welding 3/4" plate with a 5/8" deep partial penetration weld bevel at 37-1/20 using the shielded metal are process. This joint configuration is a liniting configuration because it does not provide the accessibility of a butt joint. In this case, the design engineer would have assuned a 1/8" reduction in the 5/8" weld size which would give an effective throat of 1/2". The actual measured throat on the test weld exceeded that required by the designer (1/2").
147. In addition to the tests perfomed by Pullman, existing partial penetration welds were exanined fron previously installed naterial which had been renoved from the plant. One had a bevel angle of 37-1/2 0 (plus or ninus construction tolerances) on a 3/4" base plate to support 1 -O i
l i.
77/19SL. Its effective throat was measured in two places (5/8" and
'O 43/64"). Both these neasurenents net or exceeded the value required by design (5/8" which is 1/8" less than the depth of bevel preparation).
l l 148. In sumary, the designer specified partial penetration welds that were compatible with single bevel weld preparation angles used by
)
construction. The designer derated these partial penetration welds by Further, sectioning 1/8" to account for the lack of fusion at the root.
and neasurenent of actual test coupons of typical joints denonstrate that the procedures used by Pullman on partial penetration welds with bevel angles of 37-1/20 produced effective throat dimensions that were compatible with the designers' requirements.
l XXXIII. It is alleged that:
re==1t or ar state's inauiry. iacorrect 6=414ta9 = ave eats
/
O ^=
were changed to reflect proper building novement. (Stokes,
' 1/25/84, Tr.11-13).
149. The seismic displacement of the buildings is provided in DCM C-28 which was originally issued on October 7,1982. In this design criteria nenorandun, deflection of the base of all the structures for the Hosgri evaluation is considered as zero since, as it always has been, the seismic evaluation was based on a fixed base nodel. The fixed base nodel is, of course, an idealization of the actual case, but for all practical purposes it is a reasonable idealization for the relative deflection between adjacent buildings at Diablo Canyon. Contrary to Mr.
St.okes' allegation these deflections have never been changed as a result of Mr. Stokes' connents, or for any other reason, since they are i correct and meet all criteria.
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150. For the DE/DDE evaluation, a nodel was used for some structures which had soil or rock springs at the base. For this type of nodel, an actual deflection was determined and, for the case of the Auxiliary Building and Containment, these deflections, to the nearest 1/100 of an inch, were 0.00 and 0.05 inches, respectively. These deflections were I reported in Rev. O of DCM C-28 and in every subsequent revision. The models used for the other two structures for which DE/DDE analyses were
, perfomed, Turbine Building and Intake Structure, need fixed base nodels. These deflections were, therefore, zero. Regardless of the calculated value of deflection, from a practical point of view, no safety problem exists since the movement is extremely small for any earthquake motion.
- ] XXXIV. It is alleged that
In 1983 a management representative from San Francisco, Mr. Dan Curtis, refused to answer numerous questions and challenges from site engineers who believed that Document 049243 was not a conservative basis for the seismic redesign program. (Stokes,1/25/84 Tr.13-14) 151. Drawing 049243 contains eleven standard support details with associated l
allowable load ratings. The authorized standard supports and acceptable paranetric limits are prequalified by a worst case analysis, the calculations for which are located in San Francisco. So long as the-qualified load ratings are not exceeded, use of a prequalified support results in an acceptable design. The very nature of the worst case analysis and the establishnent of acceptable Ifmits results in a conservative but efficient nethod of qualifying small bore pipe
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' supports. While PG8E did send an engineer to the site to discuss O drawing 049243, his purpose was only to explain recent changes in the procedure and not to discuss or " defend" snall bore criteria as inferred by Mr. Stokes. (See Affidavit of Daniel J. Curtis.) The presence of such person should not be taken to mean that the project at any time considered the use of 049243 to be inappropriate or unconservative.
XXXV. It is alleged that:
Management did not freely distribute professional codes
- that supposedly paralleled conputer analyses relied on by engineers in the seismic design review. In some cases the only reference materials to guide the engineers were the conputer analyses. That is improper, as management effectively conceded in the fall of 1983 through instructions that the conputer analyses were merely a guide and not meant to replace the professional codes.
Unfortunately, the program had officially been completed when nanagenant disclosed the non-binding nature of the computer analysis. (Stokes,1/25/84, Tr.16-17,120-31) 152. This allegation is unclear and not true due to Mr. Stokes' confusion between industry codes and conputer codes (the referenced transcript i
pages do not discuss computer analysis). A proper reading of Mr.
Stokes' transcribed connents, together with his other allegations, leads one to relate this issue to his disagreenent with Diablo Canyon Project criteria (DCM M-9) for angle sectioned menbers and U-bolt load ratings.
Mr. Stokes alleges that only the AISC code specified bracing criteria for long " angles" and ITT Grine11's U-bolt load ratings should have been allowed instead of the criteria in DCM M-9. This is but one exanple of Mr. Stokes' limited understanding of why project specific requirements are used. Mr. Stokes' lack of knowledge as to the basis for these I l
project requirenents has led him to allege that industry codes and standards were not used. In actuality, industry codes do allow the use of testing or other more sophisticated methods to develop project specific standards which are then used in lieu of code specified values.
153. The transcript (Tr.121-2) indicates Mr. Stokes' confusion and disagreement with code provisions that allow for more sophisticated nethods or, in these cases, test data as discussed earlier in paragraphs 16-31 and 76-81. In fact, in the technical discussion with Dr.
Hartznan, Mr. Stokes sumarized the basic cause of his disagreenent:
i "I would like to nake a statenent: that I have never professed to be a PhD in one specific area of all the allegations I have brought up. I only profess to be a practicing engineer with reasonable knowledge of industry 4
practices, as any other engineer, and nore in sone cases. Anything that is new research, unaccepted as an industry whole, has no point being in a new plant, in ny
{. opinion." (Stokes, Tr.129-30) 154. Therefore, Mr. Stokes rejected, as not meeting code requirements, project criteria that were based on testing or more sophisticated methods. Such judgments regarding licensing criteria are beyond Mr.
Stokes' specific job and overall professional experience.
155. The technical adequacy of the project requirenents (DCM M-9) for U-bolt i
load ratings and angle section nunbers has been described previously in paragraphs 16-31 and 76 to 81.
XXXVI. It is alleged that:
The M-9 computer analysis for angles onitted the relevant p(rovision "AISC") code of the American for allowable Institute bending stress, of Steel Construction contrary to c0 l
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licensing connitnents. Management officials stopped engineers from using that section of the code, because compliance required angles to be cut out and replaced with tube steel, or at least reinforced through braces.
(Stokes,1/25/84, Tr.15-21) 156. The technical aspects of this allegation are addressed in paragraphs 76 to 81 above. The basis for the nanagenent direction for use of the Project specific criteria is discussed in paragraphs 152 to 155 above.
XXXVII. It is alleged that:
Management imposed inconsistent standards for l'
nodifications in the seismic design review: as the number of nodifications approached the limits beyond which PGandE had connitted to expand the so.nple, nanagenent refused to fix deficiencies, even if obvious and more severe than those previously corrected.
Instead, engineers conducting the first round of calculations were told to make assunptions contrary to fact, such as restraints that did not exist. (Stokes, 1/25/84, Tr. 23-24)
(O 157. This statement is inconsistent with the small bore piping sample plan and results identified in the PGandE Phase I Final Report for the Design Verification Program. The plan consisted of a connitment to review a h
specific size sample for certain design considerations and to evaluate i the results of this review. A specific acceptance criteria established l
for evaluation of the review results, such as the 5 percent linit alleged by Mr. Stokes, was never set. The details of the small bore reverification progran are set forth in the History of Onsite Engineering Affidavit filed contenporaneously herewith. (Attachnent B).
158. All nodifications found to be required during the review of the sanple were identified by cause. The cause was then related to a design l
consideration and the Generic Small Bore Program was expanded to address so s-
this consideration for all affected piping and supports. The following issues, initially a part of the sample program, were transferred to the .
I generic progran when it was detemined that the existing design did not neet all licensing requirements: .
- 1. Computer thema11y analyzed small bore piping and associated seisnic analysis,
- 2. Equipnent seismic and themal anchor novement,
- 3. Unusual concentrated mass configurations, e.g., nunerous valves or equipnent in a concentrated configuration,
- 4. Nozzle loads on equipment which were upgraded to show compliance with seismic criteria, and I
- 5. Vents and drains.
159. All piping and supports have been reviewed and are shown to be qualified for those design considerations related to the generic progran.
160. No modifications were found to be required for those design considerations addressed and qualified by the sanple progrsn; therefore, it was not necessary to set an acceptance criterion, such as the 5 percent Mr. Stokes alleged.
XXXVIII. It is alleged that:
Bechtel issued out-of-date conputer STRUDL nanuals to engineers in the seismic design review. Inexplicably, the office at Diablo Canyon was not on the route slip for updated naterials on the computer, and even after that deficiency was corrected the materials consistently were outdated. The nanual provides backup infomation to engineers who wanted to check or go beyond the program.
Engineers in the seismic design review did not have written procedures to guide their use of the STRUDL j conputer program. As Mr. Stokes explained, "All we had
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was the form handbook of a STRUDEL [ sic] progran ninus the pertinent infomation such as the model load O points." (Stokes,1/25/84, Tr. 27, 29,146-47) 1 61 . For static analysis used by pipe support designers, the Bechtel STRUDL user's nanual consists of two docunents:
- 1. MIT STRUDL II, The Structural Design Language Engineering User's Manual, Volune 1, Frane Analysis, MIT Research Report R68-91. :
- 2. STRUDL III User's Manual.
162. Bechtel's Data Processing Library issues the revisions and user infomation bulletin to the controlled copy holders. Contrary to Mr.
Stokes' allegation, three OPEG engineers had controlled copies of the STRUDL user's nanuals.
163. The first document is essentially the basic user's manual which has not been revised since its first edition in November 1968. This docunent is not a Bechtel controlled document. It was originally issued by MIT, Q
Cambridge Massachusetts. The conplete STRUDL input can be prepared from this manual.
164. The second document provides specific instructions, such as how to run STRUDL on Univac; STEEL DATA code, and other enhancements to nake it easier for the user.
165. Knowledge of the first docunent allows the user to prepare INPUT files correctly for running STRUDL. Experience in the use of the STRUDL progran and a nininun of three years of nuclear pipe support experience were requirenents for hiring OPEG job shoppers or casuals in pipe 2 support design. Therefore, it was not necessary to prepare written procedures on the use of STRUDL. In addition, because the basic STRUDL v
O
_ _ _ _ _ _ _ ~ _ _
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user's manual has not changed in 16 years, Stokes' concern about O
uncontrolled copies of the nanual is largely acadenic.
Input foms were also established for. OPEG to ensure: (1) unifomity of 166.
input, (2) consistent consideration of maximun load combinations and (3) increased efficiency by eliminating rewriting the nandatory STRUDL comands. More load cases and other connands for the analysis of the i pipe support frane could be added as required by OPEG engineers, including Mr. Stokes.
XXXIX. It is alleged that:
Engineers in the stress group relied on outdated seismic
- data that was necessary for their calculations. It took up to six months to receive updates, by which time the
! newly-arriving naterial was out-of-date. (Stokes, 1/25/84. Tr. 29) rO 157- oesisa or 5m=11 bare 59 95as reiies apaa seisaic spectra iaputs fraa ocas C17, C28, and C30 developed by the Civil discipline and seisnic anchor novements (SAM) of large bore piping to which the small bore piping connects. It is nomally desirable to delay the analysis of sna11 bore piping until all such inputs are finalized. However, the Project recognized that some schedule advantages could be gained by beginning with prelininary seismic input assumptions for the analysis of small bore piping, with final analysis being completed as final seismic input becane available. The use of this initial preliminary input data is not l
of concern since subsequent finalization of the calculation would have corrected any differences in the input infomation. The process of ensuring that the latest sef snic input was used in calculations was k.O
controlled by Piping Procedure P-27. This procedure required docunented O review of all calculations affected by C-17 C-28, and C-30, to perfom new analyses where required, and to respond, in writing, when all actions were complete. While it was recognized that response spectra and structural novenents were undergoing a complete review, controlled 1
copies at the seisnic input criteria were assigned to OPEG in early 1983. As C-17, C-28, and C-30 were finalized, the reviews required by piping procedure P-27 were perfomed thus assuring that all final input infomation was included in the calculations.
XL. It is alleged that:
Bechtel's conputer progran did not have an adequate "menory' for engineers to conduct full analyses of complex hangers. As a result, engineers had to ignore relevant factors as the worst case scenarios for force on the support frane. (Stokes,1/25/83, Tr. 27, 29, 38-39)
(Q 169, In perfoming p{pe support Work on various nuclear power plants Bechtel has never experienced a case where STRUDL nenory limitations prevented the analysis of any pipe support frane.
169. Bechtel's STRUDL conputer program allows the analysis of problems which require up to 262K nenory. Analysis of a support with fifty joints and more than 45 loading cases should not require nore than 80K nenory. In fact, a STRUDL analysis perfomed on a pipe support for another project t
had more than 200 joints and 45 load cases and only required 120K of nenory. At Diablo jobsite, calculations of 155 joints and 32 load cases and 57 joints and 49 load cases have been successfully run.
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170. ' The 80K memory limitation at Diablo Canyon is one that is imposed by the If it is desired to have results
~
type of output b'eing requested.
irriediately printed on the sane teminal that the input was prepared, l
then a limitation of 80K r:ust be imposed. However, if the engineer does not require fenediate res'ponse, a batch node of operation can be selected and full 262K memory can be used.
1 171. The problens e'necuntered by Mr. ' Stokes in perfoming computer analysis seen to stem from his Rack of knowledge of the. efficient application of the STRUDb computer program, rather than a limitation on progran nenory.
172. An onsite STPfl)L specialist was availcble for consultation who provided i
guidance,! as needed, to the pipe tsupport personnel, including Mr.
e Stokes, for the efficient application of the STRUDL.
173. Mr. Stokes also expressed concern (Tr. 87-88)'about not being allowed to perfom all aspects of the calculation process. At Diablo Canyon, two gro'ups were involved in STRUDL analyses. STRUDL input data were prepared by pipe support engineers who are skilled in the application of STRUDL for pipe support frane analysis. The second group consisted of conputer operators who did not do any engineering work. The computer
(- .,
operators are skilled in the computer operation of the STRUDL progran.
This division of effo'et has resulted in ,an efficient operation because 6 2 j
the engineers were relleved:cf non-engineering effort, such as typing in their own input files. Mr(Stok'esisapparent1kcomplainingaboutnot being allowed to per om the clerical function of typing in the STRUDL input.
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1 XLI. It is alleged that:
I 1
There was no consistent procedure or criterion to guide engineers who checked calculations in the seismic design
! review: they could check whatever they wanted through l any method. (Stokes,1/25/84 Tr. 31-32)
- 174. Mr. Stokes is simply nistaken in this allegation. Engineering Manual l Procedure 3.3, " Design Calculations", Section 4.2, identifies the requirements for checking calculations. As shown in the following
! quotation from the procedure, the procedure identifies the itens to be checked, acceptable nethods to perfom the check, and the checker's _
actions if the calculation is unacceptable:
i
" Checking of the calculations shall include:
a) Checking the basis of the design, such as the design nethod, design concept, proper use of design criteria, and assumptions.
lg
- b) Checking the design loads, forces, flows, currents, voltages, naterial properties, foundation conditions, etc.
c) Checking the results, d) Conparing the results with the drawings to assure confomance of dimensions, materials, etc. ,
Manual calculations shall be checked using an alternate calculation nethod if possible. When alternate calculations are not feasible, the calculations nay be checked by a detailed review of a copy of the originals.
This copy shall be clearly marked to indicate that it is i
the calculation check.
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i Computer calculations shall be checked for:
a) appropriateness of the progran to the design or analysis b) correctness of inputs tO
c) reasonableness and application of outputs v d) conpleteness of Computer Calculation Index Sheet infomation.
When the checker has detemined that calculations require .
correction, the calculations shall be presented to the /
originator for correction. The checker shall check the corrected calculations."
175. In addition, the Piping Group developed an implementing procedure P-6,
" Procedure for Assenbling Pipe Support Calculation Packages." This procedure includes an additional checklist of specific items to be l
included in each ' calculation.
176. Experienced engineers at Diablo Canyon are utilized in both an originating and checking function for pipe support calculations, as Mr.
Stokes states in the transcript, page 31, from the January 25, 1984, meeting with the NRC. Therefore, engineers who have the experience necessary to originate a calculation and provide the documentation package for that calculation are certainly capable of checking similar calculations by another engineer without additional detailed procedures. No additional training or instruction in how to check a calculation is required beyond the training in the Engineering Manual Procedures.
XLII. It is alleged that:
After the NRC obtained certain work packages at Mr.
Stokes' suggestion on December 8, 1983, management directed a purge of relevant files to remove any evidence of previously destroyed or censored work by engineers who failed hangers but were later overruled. (Stokes, 1/25/84, Tr. 41, 81-82)
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I 177. We categorically deny ever purging files or records. As stated before,
,O no approved calculations or approved revision to a calculation have been destroyed, altered, or purged. The Diablo Canyon Project does not retain calculations that have not been approved. This practice is standard industry practice as confimed by discussion with other individuals on Bechtel nuclear projects as well as other architect / engineering fims.
178. Mr. Stokes has grossly exaggerated the nunber of small bore piping calculations that he produced. In fact, from our records, Mr. Stokes was the originating engineer on 59 calculations and the reviewing engineer on 39 calculations. Of these 98 calculations, 56 of then related to Unit 1, including 13 involving pipe stress analysis, and 42 of them related to Unit 2. The discrepancy between the 300 he estinates (Tr. 83) can be explained rather easily when one considers that there Ig was not a requirement to produce calculations at the rate of 1.5 hangers per day. In fact, the assumed rate for work scheduling was actually 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> per support (see paragraphs 192-193). Coincidently, Mr. Stokes' time sheets for the two mc4th period from mid-March to mid-May 1983 indicate that he averaged 16.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> per support calculation.
Additionally, Mr. Stokes was assigned to the PSDTC group, where he would not produce calculations, for about four months out of his eleven month enployment history and about 1 month of this history on Unit 1 involved piping stress analyses and not hanger calculations. Therefore, one would have expected Mr. Stokes to have produced about the sane nunber of calculations that appear in our records for the 6 nonths he worked in j hanger calculations.
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.- . XLIII. It is alleged that:
When assumptions of loads were changed for preliminary calculations on pipe supports that previously had failed, typically no one redid or checked the entire calculation. This step was necessary to determine that the new conbination of variables in its entirety would support a conclusion to pass the pipe support. In Mr.
Stokes' judgnent, this allowed hangers to pass which should have failed. (Stokes,1/25/84, Tr. 50-51) 179. Design data for a pipe support calculation, such as loading information 1
and piping novenents, are supplied to the pipe support engineers by the piping stress engineers. However, these design data may have been derived fE assumptions or prelimin'ary infornation. This process is described in detail in paragraphs 102 to 104 above.
180. Once revised preliminary design data is received, all pipe support calculations are reviewed to assure qualification to current pipe loads, displacements, and acceptance criteria. These reviews cause various degrees of calculation revision. The extrenes of ',he revision are: (1) simply documenting compliance to revised load and displacenent input in cases of inputs that are less severe than those used in the previous analysis, and (2) complete recalculation, including support modification.
1 81 . In cases of partial calculation revision, the previous calculation is retained to complete the design qualification calculation.
XLIV. It is alleged that:
l Mr. Stokes was unaware of the nomenclature for
- calculation revisions. (Stokes,1/25/84, Tr. 51).
182. The small bore piping portion of the Correctivre Action Progran began in the Fall of 1982. It involved review and analysis of the installed
- O e
piping and pipe supports to show qualification or to develop O nodifications where necessary that would result in qualification under the program. l l
183. The pipe support modifications issued under the verification program !
involved either changes to existing supports or the addition of entirely new supports. In the case where modifications to existing supports were necessary, the documentation of the pre-existing support configuration was designated as Revision 0. This included all approvals granted under the jurisdiction of pre-1982 procedures.
184. The initial version of calculations completed for such supports under the reverification program was designated Revision 1 of that support calculation, with subsequent revisions numbered sequentially thereafter. In the case of the addition of an entirely new support, the initial version of the reverification program calculation was designated (Q
Revision 0, with subsequent revisions numbered sequentially thereafter.
Letter revisions of calculations were not used.
XLV. It is alleged that:
Mr. Leo Mangoba, the Bechtel official who supervised engineers in the pipe support group, approved the seismic review calculations en masse over several days without studyingandproper1Treviewingthework. Mr. Mangoba did not even get to the calculations until a few days before the end of the program. Supposedly Mr. Mangoba's approval was one of the checks and balances on the quality of the calculations, but it was pro foma.
(Stokes,1/25/84, Tr. 52.)
185. It is true that at the end of the progran Mr. Mangoba approved approximately 100 calculations in a several day period. However, to state that they were not properly reviewed is incorrect. Mr. Mangoba
-O had instructed five other senior experienced engineers to perfom a O detailed technical content review prior.to providing the final calculation package for his approval. These reviews were done in addition to the normal checking of the calculations.
1 86. Mr. Mangoba then approved the calculations as required by Engineering Procedure Manual. This final approval authority was assigned to only two individuals in OPEG pipe support group in order to provide consistancy in the final documentation package.
XLVI. It is alleged that:
Management did not have necessary documents from vendors and manufacturers to guide calculations on required supports for vendor purchases such as valves. The omission helgs to explain why engineers based their 1
analysis on past experience" at other plants brought in from previous jobs. Management at Diablo Canyon did not rO send dra $as details ad support coaditions to vaive manufacturers and other vendors for approval. The vendor's review and approval is necessary to assure that the component is being used as intended. This omission was unique in Mr. Stokes' experience in the nuclear industry. It represents more necessary infomation that was missing from the seismic design review program.
(Stokes,1/25/84, Tr. 54-55.)
187. The design of valve supports and qualification of the valves for support location and forces was not performed based on "past experience" as alleged but, instead, was based upon specific approved criteria, procedures, vendor supplied data, and review and design standards.
188. Piping qualified by computer analysis includes the modeling of each remotely operated valve. These models include the location of the ulve and operator center of gravity (C/G) and mass. The C/G location, mass, and allowable accelerations are provided by the vender and are lO o
documented in Design Criteria Documents and drawings. In a very few cases, presumably the onissions alleged, the valve supplier was no longer in business and therefore could not provide the location of the valve C/G. In these cases the valve C/G was assumed to be two-thirds ,
the distance from the valve center line to the top of the operator based upon previous experience. This instruction is contained in Piping Procedure P-ll. The calculated valve acceleration provided by the conputer analysis is compared to the vendor allowable to show qualification. If support of the valve is required to meet criteria, the analysis is reperforned with the added restraint included. The analysis results provide forces on the support and valve. These forces are then converted to equivalent valve accelerations and compared to i
supplier allowables to demonstrate qualification.
189. Piping designed by manual methodology, as directed by Design Criteria
' Menorandum M-40, required supports to be installed on all remotely operated active valves. The supports were installed in pairs: one on the pipe at the valve and one on the operator. This methodology ensured that there was no differential movement between the pipe, valve, and valve operator and assured valve qualification for both stress and operability considerations.
190. Guidance for design of valve supports was provided by design standards.
However, all valves restrained by valve supports were reviewed by either the supplier or an independent project engineering group to ensure that valve integrity, operability, and accessability for maintenance were provided. The review was directed by written procedure and the results are documented.
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- XLVII. It is alleged that:
O Management., production scneduie for tne seismic design review made it impossible for engineers to think clearly, let alone produce consistently high-quality calculations. For extended periods, they were instructed to conplete 1.5 hangers per day on a schedule of seven days and 84-120 hours per week. (Stokes,1/25/84, Tr.
62-63, 89-91).
1 91 . As Mr. Stokes himself states in the transcript (Tr. 89),1-1/2 hanger design completions per day was not a mininum standard for continued enployment. The unit rate for support design calculations used in scheduling work was assuned to be 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> per support as an average for all supports. Sone sinple supports would require less time while nore complex supports would take longer.
192. During one period from Decenber 1982 through January 1983 there were two three-week periods when abnormally high overtine was worked to support unusual schedule demands. These periods were broken up by the two week
{
Christmas holiday period when a substantially reduced level of overtime :
was worked by those engineers not on vacation. During these two periods, there were only eight instances when an individual engineer's weekly time charges exceeded 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, with the maximum being 114 hours0.00132 days <br />0.0317 hours <br />1.884921e-4 weeks <br />4.3377e-5 months <br />. With the exception of these two abnomal periods, time charges for OPEG pipe support engineers averaged approximately 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br /> per week.
I XLVIII. It has been alleged that:
In sone instances engineers approved hangers solely on the basis of conclusions in file 049243 for similar pipe supports, without any independent evalution. This was known as the " cookbook" approach. (Stokes,1/25/84, Tr.
75-76, 91.)
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193. We agree that certain small bore pipe supports were designed based O solely on the drawing 049243.
194. It is connon practice in the nuclear industry to provide conservative prequalified load rated design standards to be used in the design of small bore piping. Extensive calculations or testing results provide the necessary documentation to show qualifications of these standards to all applicable licensing criteria. The drawing 049243 describes many of these pre-qualified standard pipe supports used on Diablo Canyon.
195. It is not necessary or required for each engineer who uses drawing 049243 to review all the backup documentation to ensure that the calculations or tests do indeed meet the licensing criteria. However, these docunents are available for inspection by the NRC.
196. When a prequalified standard design was used to qualify an existing
(
support or to design a new support, all aspects of the appropriate support parameters were compared to the requirements of Drawing 049243.
If any parameters did not meet these requirements, the supports would be designed by individual analysis and fully documented.
XLIX. It is alleged that:
Early in the seisnic design review, management instructed engineers to check a blank on the fom that the calculation results would not affect the Final Safety i Analysis Report ("FSAR"), despite the engineers' protests ;
that they did not know what was in the FSAR. Eventually, i l blank foms were just xeroxed with the "X" filled in and distributed to the engineers for their calculations. The only way the engineer could ensure accuracy was by whitingout what was already there. (Stokes,1/25/84, Tr. 96-97).
197. The calculation cover sheet referenced in the allegation is the standard cover sheet required by Engineering Manual Procedure No. 3.3, Design 1
l Calculations. The sheet contains the requirement to check if the
,O calculation affects the FSAR.
198. The Diablo Canyon piping procedures, themselves, ensure that the design and analysis methodology and criteria comply with all licensing requirenents including those contained in the FSAR. Therefore, implementation of these procedures by pipe support designers assures that the requirements of the FSAR are met. This process provided the basis for supervisors' instruction to subordinates to check the "SAR change required 'No' box". Pipe support design engineers activities are directed by these written criteria and procedures, so that engineers, including Mr. Stokes, need not be familiar with the FSAR.
L. It is alleged that:
'(O t
ensineerins calculations that cailed for field modifications were altered after complaints from construction, without the knowledge or approval of the originator. Tampering with calculations in this manner was highly improper. The significance is that in an unknown number of cases, corrective action required on the basis of documented engineering analysis was informally circumvented. The basis for revising the modifications is unknown. (Stokes,1/25/84, Tr. 98A) 199. A careful reading of Mr. Stokes' transcribed remarks indicates that his conplaint involved the modification of a support sketch to resolve construction interferences. This, of course, is the process involved in or the PSDTC program for which Mr. Stokes, himself, volunteered.
200. There is nothing improper with minor modification of a support sketch by a qualified support engineer to resolve a construction problen. Such nodifications would be subsequently reviewed by other qualified engineers as part of the as-built approval process.
og l
201. It was inpractical to have each support design engineer always provide O the solution to construction problems and to review the as-built drawings to approve the changes to the specific supports that they had ;
l originated. Engineers in the PSDTC progran, including Mr. Stokes, 1 l developed solutions to construction problems and modified the design support sketch to reflect this solution. These changed as-built drawings were subsequently reviewed and approved by support design engineers.
LI. It is alleged that:
Multiple engineers independently produced preliminary calculations on the same hangers. Besides being wasteful, this practice gave management the option to throw out the calculations that failed hangers and keep those that passed. (Stokes,1/25/84, Tr.99-100. )
202. The same hanger support was not intentionally assigned to multiple Q
engineers to perform qualification calculations and, therefore, provide an option for management to accept only calculations showing qualification.
203. Sna11 bore pipe supports were assigned to design engineers by support identification nunbers. This process normally assured that each engineer was assigned a different support from that assigned to other engineers.
l 204. Occasionally several supports, each having different identification l '
numbers, are " ganged" together with interconnected structural menbers.
205. Such a " ganged" support cannot be analyzed correctly by different engineers separately analyzing each support since loads from one support i
nay be transferred to another support.
206. When the approving supervisor discovered that a " ganged" support was assigned to several engineers due to the multiple identification nunbers for individual supports, individual support calculations were superseded and the " ganged" support, with all connected individual supports, reassigned to one engineer for calculation. Therefore, while aspects of 1
the allegation are correct, the nischevious intent alleged is false.
LII. It is alleged that:
Management officials overruled engineers who attempted to calculate the effects and stresses of torsional loads, created when pipe supports were twisted to tighten then during installation. This is an obsolete technique in the nuclear industry, and according to a former engineer in the seismic design review, it is hardly ever used unless totally qualified by structural calculations.
Engineers were told not to calculate for torsion and were overruled when they did. The stated reason was that "the hanger would fail." (Stokes,1/25/84, Tr.103-04,123. )
207. Contrary to the allegation, a check for torsion in angles is required, where applicable. Piping Procedure P-6, " Procedure for Assembling Pipe Support Calculation Packages" provides standard forms to be used in the preparation of calculation packages. Attachment F to P-6 provides a checklist for STRUDL frame analysis. One of the itens requiring entry is a check for torsion. This check evaluates the shear stresses that result from torsion in the angle sections.
208. Mr. Stokes further alleges that induced warping or bending effects as a result of torsion in angle sections have not been considered. Warping is not a phenonenon that occurs in angles. The only stresses induced in nenbers where all plane sections remain plane are shear stresses.
Sections which remain plane after twisting include open sections I
comprising two thin rectangles, such as angle or tee sections. In these l
i sections the only stresses resulting from torsion are shear stresses
,O and, therefore, warping or bending effects are not considered since they do not exist.
209. Two textbooks which explain in more detail the phenomenon of torsion in ;
angles and the resulting stresses are the " Steel Designers Manual",1/
pages 105 to 116 and the Bethlehen Steel Handbook entitled " Torsion Analysis of Rolled Steel Sections =2/ page 72.
21 0. It is true that some computer programs, such as GTSTRUDL, do consider the effects of warping due to torsion. However, these programs do not, for the reasons mentioned earlier, address additional normal stresses created by warping effects of torsion on angles.
211. On the Diablo Canyon Project, the shear stresses resulting from torsion in angles are added to other shear stresses and compared to AISC allowables for shear. In the case of angles, no increase on bending iO stresses due to torsion was included, nor is it necessary for the reasons described above.
References:
- 1) Steel Designers Manual 4th Edition Granada Publishing Limited 1221 Avenue of the Americas N.Y., N.Y. 10020
- 2) Torsion Analysis of Rolled Steel Sections Bethlehen Steel Corporation I
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, LIII. It is alleged that:
Engineers onsite had to wait up to a week to obtain t
infomation on the telephone from San Francisco that would nomally be on the drawings and was necessary to draw engineering conclusions. Combined with scheduling requirenents, this system created pressure on engineers without the benefit of data on which they would nomally rely. There was no system or procedure to verify the accuracy of design infomation received on the telephone from the San Francisco offices. In the absence of any i
such procedures, the data was unverifiable despite
- engineers' doubts about its accuracy in some cases.
(Stokes,1/25/84, Tr.110-112) 212. It is possible that, during certain. periods, onsite personnel may have had a delay in obtaining infomation from San Francisco. To minimize this inefficiency, onsite engineering personnel were temporarily relocated to the home office in order to provide data to onsite engineers. This infomation was transmitted in some cases by phone in order to expedite the perfomance of prelininary calculations.
(O-Engineering Manual Procedure 6.1, Section 4.4 specifically states that all design information provided verbally must be confimed in writing.
Engineering Manual Procedure 3.3, Section 4.1.2 provides that data requiring verification at a later design stage be identified and the calculation cover sheet marked " Preliminary" until verified. This was the procedure used for such circumstances throughout the reverification program. While this practice is allowed, it was not cornonly used except for brief periods or special cases. In all cases, datc was subsequently provided by nomal document control procedures and verified prior to finalizing affected calculations.
l
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, LIV. It is alleged that:
The initial records for hanger calculations later covered l by the seismic design review are totally unprofessional
,l and unacceptable due to the inadequate underlying documentation, as well as the lack of signatures and evidence of a checker or other approval for the great najority of calculations. The records are so deficient that the seismic design review must be expanded from a sample to cover 100% of relevant hardware. Reliance on a sample assumed the existence of a comprehensive, if questionable, base of professional engineering calculations. In Mr. Stokes' professional judgnent, such l
a base did not exist. The plant cannot be licensed on the basis of a sample base of minimally-acceptable engineering calculations. (Stokes,1/25/84, Tr.113-15) 21 3. At the time that the original design of small bore piping was undertaken, the small bore pipe support design, including support
- spacing, was specified by design standards. These standards included prequalified, load rated standard support details in PGandE Drawing 049243, the calculational basis of which were prepared by PGandE's Mechanical and Nuclear Engineering Department in San Francisco. Pullman Power Products detailed and installed supports as specified by this standard.
21 4. Engineering authority was delegated to General Construction to approve minor modifications to these details where required to facilitate installation, provided that the original design intent was maintained.
In some cases, simplified calculations were performed to justify these j
deviations from the standard details. In some cases, supports were found by inspections to have been installed at variance with specified standards, and contractor discrepancy reports were written to document these problems. In order to resolve the DRs, calculations were perfomed to qualify the installed condition.
1
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i 215. In summary, every small bore support was documented b,r an individual O support drawing which had received engineering acceptance based upon the prequalified standard of Drawing 049243 or authorized deviations fron ;
049243 justified by calculations where required. The complete records l
of the drawing 049243 calculations were maintained in the San Francisco engineering offices, which would explain Mr. Stokes' lack of familiarity with them.
LY. It is alleged that:
At the time of Mr. Stokes' departure, plant operators did not have access to a centralized document center with all infomation necessary to respond to conditions in the plant. This could compromise operators' ability to make all decisions from the control room in an emergency.
(Stokes,1/25/84, Tr.115-16) 21 6. Document systems, controlled by procedures, are in place, which ensure (O that plant operators have immediate access to all drawings and docunents necessary to safely operate and naintain the plant.
21 7. The Design Control Procedure, Engineering Manual Procedure 3.60N, requires review of all safety-related design changes by the Plant Staff Review Cormittee (a plant operations cocnittee) prior to release for construction. The procedure also requires the operations organization i to be infomed upon construction completion of each design task. The operating organization has procedures which interface with 3.60N to ensure that this current infomation is distributed to all document control centers and individuals identified in their drawing distribution lists. Upon completion of construction and as-builts submitted to 9
Engineering, the permanent plant record drawings inportant to safe O operation and maintenance are revised to incorporate the changes and issued within one nonth.
21 8. The PGandE Records Managenent Systen (RMS) provides a computer-based nultiple cross-index listing of all inportant plant records. This listing provides reference to the location of records on nicrofilm.
This systen is accessible from the plant, and all microfiln required for safe operation and naintenance is available to the operating
~~
organization.
LVI. It is alleged that:
Mr. Stokes reported errors in the M-9 conputer analysis, which incorrectly instructed engineers to consider small-bore baseplates and non-computer analysed piping lines as rigid. In fact, the baseplates and lines are flexible. The assumption was inconsistent with other
.Q
! instructions to calculate displacenent for the bolts on the baseplate. (Stokes,1/25/84, Tr.138-41. )
21 9. Design Criteria Menorandun M-9, " Guidelines for Design of Class I Pipe Supports," states in paragraph 6.8.1.1; "Small bore pipe support base plates on non-computer analyzed lines may be considered rigid for purposes of pipe support evaluation."
This assumption is in agreenent with the requirenents of NRC IAE Bulletin 79-02 for pipe supports on piping systens that were qualified by conservative alternate analysis rules or " span tables."
220. In instances where baseplate flexibility could significantly affect the frequency of the pipe support, it has been considered. An example is
! the inclusion of baseplate flexibility when calculating the natural
,O
1
- frequency of a simple cantilever beam. Accordingly all simple l O cantilever beams with baseplates included consideration of base plate !
flexibility in the natural frequency calculation. In more complex structures, the nonent resistance of the frane reduces the effect that any baseplate flexibility would have. As a result, baseplate flexibility is ignored since its effect is insignificant to the overall i
support natural frequency. However the flexibility is considered in calculation of anchor bolt loads in accordance with I&E Bulletin 79-02 and Diablo Canyon licensing commitments.
221. The Bechtel procedures referenced by Mr. Stokes (Tr.141) require consideration of baseplate flexibility for calculation of natural frequency of cantilever beams. This was precisely the practice at Diablo Canyon.
'Q 222. In discussions with th'e NRC Staff (Tr. 148, 149), Mr. Stokes indicated that the STARDYNE computer code was not used for Diablo Canyon.
Instead, the progran BASEPLATE II was used when " flexible plate theory" was required. Mr. Stokes is apparently unaware that BASEPLATE II is merely a preprocessor for STARDYNE. BASEPLATE II transforms the relatively simplified input information required for baseplate analysis into more complicated STARDYNE input format. It seems ironic that an engineer who apparently probed with such attention to minute detail in
! some areas of support design was unaware of this computer progran l
application.
I l
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LVII. It is alleged that:
Similar to the experience of Mr. Stokes and others in the pipe support group, engineers in the stress trailer were transferred after challenging suspect changes -- such as eliminating eccentricities -- in the models for the seismic design review calculations. The reluctant engineers were replaced by personnel who cooperated with questionable manipulation of models. In fact, there.were considerably more personnel shifts in the stress group than the pipe support group. (Stokes,1/25/84, Tr.131) 223. As with the pipe support group, the stress group experienced reassignment of some personnel to the Unit 2 small bore effort in the Spring,1983. However, this did not involve. physical transfer of personnel since almost all stress group personnel could be located in the one trailer which they already occupied. Contrary to statements in the allegation, no attempt was made to transfer personnel in the piping
~
stress group on the basis of objections raised regarding analysis nodeling techniques. It is true that, as with pipe support analyses, a difficult or troublesome stress calculation might be reassigned to a different engineer to take advantage of greater experience or familiarity with acceptable alternate calculation techniques. We reject the implication that reassigning calculations for this purpose is inappropriate.
LVIII. It is alleged that:
i Contrary to management assertions at the December 15, 1983 meeting with NRC staff, the calculations that replaced those rejecting pipe supports were not more refined and sophisticated. In fact, the opposite was t true: less sophisticated analysis was used. The models for subsequent calculations eliminated the unique eccentricities relevant for particular pipe supports.
(Stokes,1/25/84, Tr. 85-86,152-53.)
O
224. We are aware of only two situations which, upon initial observation, might appear to support Mr. Stokes' allegation. In one case, as outlined in Mr. Stokes' meeting with NRC, hanger 100-132 was analyzed with less sophisticated nodeling techniques to denonstrate its qualification. To the best of our knowledge, including the rereview of over 100 support design calculation packages, this was a unique case.
(Also see the Affidavit of Alex Shusterman) 225. A second situation which could have led to this allegation. Mr. Stokes believed (Tr.134) that if a support component exceeded AISC criteria for bending of angles or ITT Grinnell's U-bolt load capacity, the support was not qualified, even though it would be acceptable under the less conservative Diablo Canyon Project criteria. Mr. Stokes was willing to accept only the AISC and Grinnell load ratings as (O queiiricetioa criterie.
226. In such a situation, engineering supervision would then give the calculation to an equally qualified engineer to review in accordance with project criteria, whereupon it was qualified. This sequence might lead one to believe:
(1) Since Mr. Stokes failed the calculation, it was given to another engineer to qualify.
(2) Less conservative rules were used when the support would not qualify.
While both these statements are literally true, they are the result of perfectly acceptable techniques for resolving a problem.
l c0 227. However, in a more general sense, the calculations were, in fact, based
.O on nore sophisticated methods, since the project specific criteria for !
angles and U-bolts were based upon detailed evaluations and test results.
1 (O
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i Dated: March 5, 1984 O
L &lf8:
NED C. SKEISMEI5ItK fA=Y DANIEL J. TIS
$cyL. a {
MYRpNE.LEPPKh
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kA%M ROBERT G. OMAN
.O LARRY E. PLEY '
\
a WILLIAM H. WHIIL Subscribed and sworn to before ne this 5th day //
of March,1984. % A /. erps.s GARY H. 100RE f -
&A R. Q MICHAEL J. 7ACOBSDN M cy J. Lenaster, " SEAL Notary Public in and for the City and County of San Francisco, State of California.
My comission expires April 14, 1986.
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h NA'4CY J. LEfMSTER . :h
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C;1Y AND COUNTY OF M e
WJ.I , Y.
d Sate rhANCISCo 14,1986 32 LMy Commission Espires Aptd O XX3COOOD::CCKXXXXXXXmtXXXXX f
List of Exhibits l
Exhibit 1 PGandE Menorandun dated March 22, 1982 Exhibit 2 United Engineers and Constructors, Inc. Report, Attachnent C, May 20, 1983.
Exhibit 3 United Engineers and Constructors, Inc. Report. Attachment D, May 20,1983.
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Exhibit" No. ' 1 peu persa.cxasrawv arme PACIFIC GAS AND ELECTRIC COMPANY l'
orrica or vmz caAIRM AN March 22, 1982 i
TO: PGandE OFFICERS, ENGINEERS, TECHNICIANS AND OTHERS DIRECTLY INVOLVED IN THE COMPANY'S NUCLEAR FACILITIES This letter is to reemphasize the Company's long-standing commitment to design, build, and operate safe nuclear power plants and in achieving this commitment to require all employees to practice fundamental honesty and to adhere to Nuclear Regulatory Commission ("NRC") rules and
, regulations.
l This is also to reemphasize that our communications with the NRC must be open and allow a free flow of information. We must be ever alert to any possible l misleading or ambiguous statements made either in oral or
' , (]) written communications. Any such misstatements must be corrected immediately upon discovery. Nothing less than full and open communication between the Company and the NRC can be tolerated.
In October 1975, PGandE formalized its general policy concerning employee conduct (Standard Practice 735.6-1).
The statement of policy establishes a Company philosophy regarding work conduct emphasizing that:
"It is the policy of this company that i
employees shall at all times continue to practice fundamental honesty. Employees -
shall not, nor attempt to: deceive, defraud, or mislead the Company, other employees, or those with whom the Company has business or other relationship; ... misrepresent the Company or its employees; ... withhold their best efforts to perform their work to acceptable standards; ... violate applicable laws; or conduct themselves at any time dishonestly or in a manner which would reflect discredit on the Company."
'()
This policy is particularly important to all emoloyees engaged in work concerning nuclear power.
L______. ___ - - - . .--_ _
l l
To All Addressed March 22, 1982 i ,
()
In April 1976, Mr. J. D. Worthington, and again in 1980, Mr. J. O. Schuyler, issued a memorandum to all personnel involved in the company's nuclear power work which described a program to permit such personnel to discuss their concerns regarding nuclear power. The August 1980 letter stated that
"[Our) purpose is to again reaffilm the Company's strong commitment to the protection of its employees and the general public against any unsafe situation with respect to these nuclear '
facilities and, further, to assure that you have every opportunity to communicate freely to your Company any views you might have on the safety of nuclear facilities.
"We believe that you appreciate your right and obligation to express yourselves on matters of safety and that you have the dedication and individual initiative, insofar as your t
4 (]) responsibilities are concerned, to see that our nuclear facilities are i
designed, constructed, and operated in a safe manner.
"To give you added opportunity to ask questions or to express your views on any aspect of the safety of nuclear facilities, including those outside your own sphere of responsibility, we encourage you not only to talk to your supervisor, but also, if you wish, to any one of the following people who have been designated a review team to answer questions and to evaluate the views of any employee who wishes to express any concern whatever about the safety of nuclear facilities:"
l We are proud that the application of these policies of l openness in finding and evaluating safety issues led directly to the discovery by PGandE personnel of the " mirror image" error at Diablo that otherwise might have gone undetected.
. (:) -
To All Addressed -
3- March 22, 1982 O
Recently, in February of this year, Mr. R. C.
Thornberry issued a separate memorandum to Diablo Canyon 1 Power Plant employees which reiterated the Company's policy '
concerning adherence to government rules and regulations. l i
We must strive for perfection in design, construction,
, and operation of our nuclear units. To attain this goal, it a
is necessary that we all exercise our best efforts to ,
resolve problems we encounter in our work. When problems are encountered, they must be immediately identified, clearly defined, and brought to the attention of your supervisor. This approach should facilitate the evaluation of, and formulation of timely and effective solutions to, any problem. Constructive recommendations are encouraged at all levels.
Our goal is to design, construct, and operate our nuclear facilities with full margins of safety and full compliance with NRC requirements. Strict adherence to the above policies will provide added assurance that this goal will be met.
O . w. Nh t
F. W. MIELKE,
&^
- r. B. W. SHACKELFO cc: Officers Department Heads Division Managers All Concerned Personnel l
e
M 4 Exhibit No. 2 l
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Date: May 20. 19E3 File No: 21.E,1 ,
WTTD uw.ns.v.5 & CONS *KUCTOR$ INC.
TECIDCICA1. REPORT 1 I
Date: May 20, 1983 Purpose of Baport : Qualification and verification of Flare-level Groove Walds - Square Tuba Distribution: M. P. McKenna R C591 W. J. Duffy M C589 D. C. Torsquist M C589 W. C.141thead DEC294 .
S. J. Pattisen W C262 A. Sandopadhyay R C589
- 1. W. Gregory BEC589 M. 3. Lasota M C589 .
E. E. Serg UEC196 P. E. Jathaveda:r. UEC787
- 5. C. Sethi , DEC288 3. Resu WEC589 T. M. Alman E C196 5. C. Madaras TEC5ES
- 5. 5. Caruso M C290 C. V. Neurar R C392 0 3'==**7 J. R. Slotterbach
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- 3. J. Euselton
==c2'2 UEC5E9
- 3. E. Rhoads 07U4 c. P. Ealani 09U4 E. M. Bayes WEf143 J. M. Benenati 09U9 -
- 1. E. Bryans WEC262 S. E. Guha M C2E2 M. A. Edgar R C184 J. R. Julian CEC 262
- 1. C. Seventy R C786 M. J. Essopka 07U6 G. A. Emilant UEC262 DCC Field TECit5 P. A. 34ene DEC591 DCC - PA 06:1 C. T. 11gamenti 0704 SM File UEC164
- 3. C. 34 vine ilEC262
! J. P. Cannon 1453
- 5. J. Implan 170' l . Report Prepared By: / / s T. 1. Yrcio Es.ert Approved By: \ S ' > d- - 9 l T. P. Tassa11o. Jr.
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SM: 4579A i
Purpose - Te verify. as a ad.mians:. that the effective throat thickness for a flars-bevel-groove weld eben filled to the solid section of the bar will be equal 3/16 E, whnre R is equal to the radius ef.the bar.
Matertals - Tubular steel sises 3" st 3' x %". 4" s 4" x 3/8". 6" x 6" a %" and 8" s 8" x 4" ASTM A300 tes used.
MeMh; Process - The shielded antal are usad".ag procese was used, uti-11aima ETA 3.1. E7018 alestredes with smaltipla passes.
Prahast and 1sta: pass - The miniaw!s preheat and interpass tamparature ses in accordante t-}th'4EEI/ANS D1.1. Tabla 4.2.
- f, PMures for Shielded Metal Arc - T*k welding'ses done in the vertical, overhead and flat planes scs.11 ming 3/32" and 1/8" disaster
, electrodes in nach positisc. The solding parameters were as follows: -
3/32" - DCRP. h0-120 ampa. 20-27 velts. 2 fps aim. trevel.
1/8". -DCIP' . 115 165 eraps.- Z-27 welts. 2 ipa ain. travel.
Qualification - The saapits wereisectisted for visual == amination.
The welds wre tree froc cracks end there was thorough fusion between adjacent layngs'ef weld metal and the base metals.
The welds is anneral, were visually acceptable.
Conclusion - In gemaral. 3/32" 8 alastrodes showed good penetsstion ex-eseding the assiasm thrsat thickness by approximeta17 50%
az-spt there were some problems with the 3" a 3" a k" tubes.
The small radius did not permit the depth of pometration.
. The 1/8" 9 electrodes showed excellent pesecration 'for exceeding the minians threat,thickmass for the flare-bevel-groeve welds.
It is recousanded that the Centracters be directed y) utiliae 1/8" S electrodas for the first pass to insure adhqs. ate reme-traties.
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Exhibit No. 3 lO ~
AIPullman Power Products Corporintion -c ! i f . ,
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- 1 BATE: SECDetR 9.1963 *
"v. awn is Yesms l 70: D. SOCMLL. PG&E FROM: N. KARNER. 04/1)C
SUBJECT:
RPS SEAM ATTACNNT SBD-18 AND FLARE DEVEL WELDS The NPS beam attachment 380-18. which uns in the possession of the NRC, has been esamined by N.T. and U.T. Please find copies of the results of these esaminations attached. ,
The NRC discussed with Pullaan Power Products unld penetration .
for flare bevel melds an tube steel as used at Diablo Canyon. .
i An favestigation had previously been conducted by Pullman
- Feuer Products and United Engineers and Constructors. Inc.,
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at Seabrook,5tation en tais subject. This information uns presented to the NRC at Diablo Canyon for their review.
Their review revealed that the minimum required throat was !
mest difficult to obtain on small size tube steel (3* x 3*)
when using 3/32" electrode 1a the flat position.
As a result of this determination and discussions with Mr. Sam Reynolds of the RRC. Pullaan Power Products prepared several sample unids at Diablo Canyon using 3* a 3* tube steel in the flat position with 3/32' electrode. Measurements were taken in the presence of Mr. Reynolds. The formal results of !
these sample melds are attached.
- If you have any questions. please do not hesitate to call.
l .
l sw-Earner 04/QC Manager NK:pe Attachments- (originals) cc: A. A. Ick w/attactuner.ts P. Stieger O File
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ansvLTs,er rsARE SEVEL PENETRATION TEST en Beeember S. ,1983. Pe11aan power dPredects for flare esadseted tests determine bevel pista.
the typteal penetrations utdeh will be achieveTh All welding uma performed la the flat posities with thick plate. Baselts are as follows:
3/32* and 1/9* E7010 elastredes.
=i - --i - 1/g" Elaetrede i-i U Threat (S/15 R1 3/32' Electrode
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UNITED STATES OF AMERICA O.
NUCLEAR REGULATORY CCM4ISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD
)
In the Matter of ) Docket Nos. 50-275
) 50-323 ,
PACIFIC GAS AND ELECTRIC )
COMPANY ) Design Quality Assurance
)
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) )
)
HISTORY OF ONSITE ENGINEERING SMALL BORE PIPING PROGRAM DE5GRIPiION AFFIDAVIT OF M. TRESLER, R. OMAN, AND M. LEPPKE STATE OF CALIFORNIA ) :
ss.
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CITY AND COUNTY OF SAN
)
FRANLISCO )
The above being duly sworn, depose and say:
I I, Michael R. Tresler, an Assistant to the Unit 1 Project Engineer for the Diablo Canyon Project.
I, Robert G. Oman, am an Assistant Project Engineer for the Diablo Canyon l
Project, and from August,1982 to October,1983 acted as Onsite Project Engineer at Diablo.
I, Myron E. Leppke, an Onsite Project Agineer for the Diablo Canyon Project.
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' ONSITE ENGINEERING
- 1. In the early 1970s, the design and location of small bore pipe supports was dictated by design standards produced by the PGandE Mechanical and Nuclear Engineering Department. These standards provided infomation for location and spacing of pipe supports and standard support details.
Pullman Power Products detailed and installed supports as dictated by these standards. Early in the construction process, it was found that the standard support details often required modifications to facilitate f
installation. Because the majority of modifications were minor in nature, l 1
Mechanical and Nuclear Engineering delegated design authority to Ger.eral l Construction for approval of minor modifications to these details, providing the original design intent was maintained.
Eventually
- 2. Initially, this work load required only one to two engineers. 1
'"" 'd '" '" '" '"*'" ' '"S'"**r' " '"'ad '" '"
0 effort. Engineering also delegated authority to General Construction to approve minor changes to large bore piping hangers provided the intent of the original design was maintained, and Mechanical and Nuclear Engineering personnel perfomed reviews to verify that the changes were properly implemented. The minor changes made to all of the large bore piping hangers were reviewed as part of the as-built hanger review, and compatibility with existing engineering calculations or reanalysis was .
performed. This review was dictated by the quality assurance procedures and the Engineering Manual Procedures 3.6, 3.6 ON, and 3.7.
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1
- 3. In early 1980, the i. Jtion of the large and small bore design O adjustnent activities required approximately 20 engineers. In addition, approximately 20 to 25 drafters were added to the organization to assemble final support as-built configurations. In early 1982, the decision was made to establish this group as a part of the Engineering Department.
Thus, on April 19, 1982 this organization, temed the Onsite Engineering i
Group (OSEG), was placed under the technical direction of the Mechanical i and Nuclear Enginee'ing r Department, specifically, Mr. M. R. Tresler, Diablo Canyon Piping Coordinator. Mr. M. E. Leppke was placed in charge of the group at the site. Effective April 19, 1982, OSEG began operating in accordance with the Engineering Department Manual. The change is detailed in the attached letter, (Exhibit 1). Mr. Leppke actually assumed his position in mid-March, prior to issuance of this letter.
g 4. In September 1982, Project Engineer's Instruction PElo9 (Exhibit 2) was issued, which further detailed the requirements of this group's activities. It changed supervisory responsiblity to the Project Engineer and established the Onsite Project Engineer position which was filled by Mr. R. G. Onan. It established the Onsite Project Engineering Group (OPEG) as a multi-discipline engineering group which was located at the jobsite and served as an extension of the Home Office Project Engineering Group in San Francisco. OPEG operated under the same procedures and criteria as Home Office Engineering. OPEG was comprised of .
I representatives and lead engineers from each of the major discipline groups of the Diablo Canyon Project: civil, mechanical, electrical, instrumentation, and piping, and included representatives of project j l quality assurance and quality engineering.
O 1
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- 5. One of the major tasks assigned to OPEG was design and, in the case of
' Unit 1, reverification of, small bore piping. The small bore verification effort was assigned to OPEG because (1) DCP small bore pipe design had historically been done onsite, (2) onsite work facilitated field confimation of the installed conditions of the plant, and (3) onsite work facilitated checking the feasibility of proposed modifications in such a way that the physical impact to other plant installations would be minimized.
- 6. The size of OPEG grew from approximately 20 people in August 1982 to a l
peak of approximately 270 in the Spring of 1983. At the peak, ' e group was comprised as follows:
Approximate Nunber _ Type Function
- O 6 6
Civii Ensineers Mechanical Engineers Desion Design 5 Electrical Engineers Design 2 Instricientation Engineers Design 20 A&iinistration Clerical, typing 30 Drafting Drafting Walkdown Engineers Walkdown and feasibility 60 checks in support of Home Office engineering 20 Pipe Support Engineers Units 1 and 2 pipe support design tolerance classification teans 85 Pipe Support Engineers Small bore piping support design 40 Piping Stress Engineers Small bore piping stress engineering
! 7. Design of small bore piping relies upon seismic spectra inputs developed by the Civil discipline and piping thermal modes developed by the
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Mechanical discipline, and the seismic and themal anchor movements (SAM / TAM) of large bore piping to which the small bore piping connect.
Ordinarily, small bore piping design is delayed until all these other inputs are received. The Project recognized that some schedule advantages l t
could be gained by parallel preliminary analysis of small bore piping, with final analysis after the other inputs were received.
! 8. Late in 1982, OPEG Management was planning for the staffing requirements for the entire year of 1983. It was then anticipated that an increase in personnel for the small bore piping effort would be required commencing in November 1982, building to a peak in April 1983, and subsiding to a ;
minimum work force in mid-year 1983 when the work for Unit 1 was scheduled f to be conpleted and Unit 1 personnel would be available for transfer to Unit 2. (Exhibits 3 and 4). To accomplish this temporary demand for
.O personnel, we relied upon the hiring of job shoppers and agency people like Mr. Stokes.
- 9. In approximately November of 1982, OPEG started to increase its personnel in support of the start of the significant work in the small bore reverification effort for Unit 1. Mr. Stokes was among the first pipe' support designers hired to facilitate this manpower buildup and arrived i early in of November 1982. By January of 1983, the pipe support group reached the size of abcut 35 pipe support engineers. Through February 1983, this group worked entirely on the Unit 1 small bore piping f
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reverification effort. Unit 2 small bore design activities began in Q
March, 1983, 1
- 10. In order to ensure proper management of the small bore design activities for the two Units, separate teams within OPEG were established in March, 1983. This facilitated independent management, scheduling, production, I and control of the work for the two units and facilitated coordination with the two independent Unit 1 and Unit 2 project teams in San Francisco. Separation also prevented intermixing of calculations, calculation files, and support drawings.
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- 11. Engineering personnel to staff the two separate pipe support groups cane from both newly hired individuals and from the existing OPEG personnel.
The basic consideration in establishing the makeup of the two separate teams was to provide each with an essentially equivalent mix of new
,( assignees, engineers with more project experience and supervisory At the time personnel such that each project would be supported equally.
l of the division, the four supervisors who were to be the principal leadership in the new Unit 1 and Unit 2 organizational structure held discussions to establish which of the more experienced engineers were to be assigned with the newer engineers in each of the squads of the new organization. Contrary to Mr. Stokes' claim, there was no discussion or consideration of any factor other than as discussed above in the The Unit 1 effort through this assignment of personnel to the two teams.
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55 period af ti== coatiaa d r ver$ricatiaa aad '5aa'$zatiaa af 9 9 as str===
O and pipe support calculat4ns as the necessary inp'.:t data was finalized I The Unit 2 for such items as seismic spectra, thermal modes, SAM and TAM.
effort was directed primarily to original design because design and construction of Unit 2 was not as advanced as Unit 1.
- 12. Another responsibility of OPEG, which was developed in December 1983, involved the creation of a pipe support design tolerance clarification tean (PSDTC). Engineers in this team worted directly with construction engineers and Pullman Power Products crafts in resolving construction difficulties in the installation of pipe supports, both large and small There was a team of engineers located bore, in order to minimize delay.
at each unit.
- 13. In June of 1983, Mr. Stokes volunteered for assignment to the Unit 2 pipe support design tolerance clarification team.
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- 14. In August 1983, Mr. Stokes came forward with concerns regarding welding, the spacing of expansion anchors, and the use of angle menbers in design of supports. These were submitted by Mr. Stokes in handwritten fom on In August and September, the August 12 and August 16, 1983 (Exhibit 5).
concerns that Mr. Stokes raised were reviewed and a determination wa that the concerns did not amount to discrepant situations or, in the case of the welding issue which was already under review by the Project, that there was a resolution for proper acceptance of previously installed welds h
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through an as-built procedure, (Exhibits 6, 7, 8 and 9). The Discrepancy O Reports (DRs) were assigned DR numbers and typed in October,1983.
(Exhibits 7, 8 and 9).
- 15. In carrying out previously established personnel staffing plans, a force ranking was made in May, June, and July,1983 by the support group leaders to establish the relative standing of engineers based on assessment of work perfomance. That, ranking was not intended to indicate unsatisfactory perfomance but rather, the relative standing within the i
group. When the manpower reduction was undertaken, this force ranking was used as one guideline for detemining the order of engineers to be separated. Mr. Stokes' standing in that force ranking was in the botton third of his group.
- 16. As an additional guideline for detemination of the order of separation, O =91ay at it tus was caasider d- ^ deci=$aa > = de ta reduce = apa r cost by releasing agency personnel first. This was not related to level of performance but rather to economic considerations. Agency enployees are generally more expensive than pemanent or casual employees.
Generally, a premiun is paid for agency employees largely in compensation for the uncertain nature of their employment duration. Use of agency personnel is a common industry practice to accorriodate rapid adjustnents in manpower level in response to changing schedule priorities, and the ,
i i uncertain nature of their employment is understood by those so employed to l
be one of the conditions of their assignment.
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O 17 T"' rarc' r*ductiaa af o'ca "*s'a i" a#a' af 1983 fra" ' tot ''
support group manpower level of 104 engineers. Fifteen engineers were ,
released in June and July, 9 engineers in August and September, and Mr. l Stokes was one of the 3 engineers released effective October 14 at which i time the total reduction had reached 27 pipe support engineers and the
! group manpower level was at 77 pipe support engineers.
- 18. Contrary to Mr. Stokes' allegations, his release was not precipitated by his submittal of three Discrepancy Reports but was part of the planned _
force reduction that took place in OPEG, (Exhibits 3 and 4). While we cannot state with certainty that the earlier reductions in force of June and July caused Mr. Stokes to draft his three DRs in August,1983, in attenpt to obtain job security, we are certain that Mr. Stokes was aware that a reduction in force was taking place and that there was a likelihood that the end of his employment was drawing near.
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- 19. Project engineering procedures governing the use of Discrepancy Reports (DR), as well as Nonconfomance Reports (NCR), are contained in the Engineering Manual Procedures. Training on the Engineering Manual is .
L required of all engineering personnel shortly after assignnent to the project. This training includes indoctrination in the purpose and use of DRs as well as NCRs. Project training records indicate that Mr. Stokes i
attended this training on November 8,1982, shortly after his arrival onsite. NCRs are addressed in Engineering Manual Procedure 9.1 and the DRs are addressed in Engineering Manual Procedure 10.1.
- 20. These procedures provide that any individual, who may or may not be an l
l employee of Engineering, can identify a potential discrepancy and bring l '"' "' "'" '" '"' ' " '"*" "' '"' '*"" '""**"' *"*"*
O group leader or supervisor. The supervisor is responsible for
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determining, after investigation, whether the identified item is a O nonconformance, a discrepancy, or neither, and directing that the appropriate report be prepared. During the course of the OPEG piping design effort, there were numerous instances identified by engineers which required discussion and clarification of the design basis for itt.es which were unclear to specific engineers. This is not unexpected in the nomal course of design engineering activities where solutions to engineering f
problens are developed.
Zi. The following is a brief description of the small bore piping program for Unit 1. Upon discovery of the original annulus frame diagram error in Septenber 1981, review of large and small bore piping was initiated to i
assess impact. During this review, and others promulgated by discovery of other deficiences in the annulus spectra, certain aspects of small bore design were found to be deficient. These findings were documented in a
- Q series of discrepancy reports which were reported to the NRC as Open Items. In addition, Robert L. Cloud and Associates performed an Our independent design verification which included small bore piping, findings and the Cloud findings were reviewed and that review resulted in the Corrective Action Progran identified in the first issue of the Phase I Final Report. This report was submitted to the NRC in September 1982.
- 22. The small bore piping and pipe support program consisted of two components, a Generic Program and a Sample Program.
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- 23. The Generic Program addressed all design considerations which were 0- identified to have the potentiai 1.o cause modifications due to exceedence of acceptance criteria. The design considerations were:
Piping Original Generic Program
- 1. Computer seismically analyzed small bore piping and associated thermal analysis
- 2. Valve qualification
- 3. Seismic and themal piping anchor movement (SAM /TN4)
- 4. Design class change boundaries
- 5. Hot piping designed by spacing criteria Added From Sample Program
- 1. Computer themally analyzed small bore pfping and associated seismic analysis O 2. Equipment seismic and themal anchor movement
- 3. One unique concentrated mass configuration
- 4. Nozzle loads on equipment which were upgraded to show compliance to seismic criteria
- 5. Vents and drains Pipe Supports Original Generic Program
- 1. Standard support details
- 2. Loads from seismic and themal piping anchor novenent (SAM /TN4) l O l
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. _ _ _ . _ . . _ _ = . _. _ . _ _ _ _ . _ _ _ . . _ _. _
- 3. Code boundaries O 4. Lug stress and lug local effect on pipe stress
- 24. All piping and supports were reviewed to identify that piping which included one or more of these design considerations. Generally, the piping or supports identified were reanalyzed and accepted or modified to gain acceptance. However, for a few design considerations (equipnent and piping SAM / TAM, vents and drains, and lug stress and lug local effect on pi Pe stress), a worst case analysis methodology was used to show
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qualif'ication of all installations in that design consideration category.
This approach required the analysis to either address the worst case in the plant or sequential analyses of worst cases until a level was reached which resulted in acceptance of all remaining cases.
- 25. The Generic Program caused reanalysis of approximately 28,000 ft. of pipins. The 9 1ant contains approximateiy 43.000 ft. of smaii sore
. O piping. Most of this analysis, 25,000 ft., was perfomed by computer using the Bechtel ME-101 program. The rest was analyzed by manual J
methodology using the M-40 criteria.
- 26. The small bore Sanple Program consisted of review of 5000 ft of pipe and the results allowed acceptance of 15,000 ft (including 5000 ft contained in the sample analyzed) of pipe which did not contain generic design I considerations.
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- 27. The sample program addresses all remaining design considerations and no
' o deviations from acceptance criteria were anticipated to be found in this review. This position was based on previous analysis and reviews perfomed by PGandE and the IDVP coupled with the conservatism contined in l
The following the span criteria methodology used for the initial design.
considerations were addressed:
Piping I
- 1. As-built piping accuracy
- 2. Revised seismic spectra
- 3. Concentrated masses
- 4. Effect of piping and insulation weight
- 5. Spans exceeding spacing criteria
- 6. Anchor and equipment loads
- 7. Equipment and building seismic and themal anchor movenent Q
- 8. Themal analysis
- 9. Integral valve bypass
- 10. Vents and drains Pipe Supports
- 1. As-built piping accuracy
- 2. Revised spectra
- 3. Concentrated masses
- 4. Effect of pipe insulation weight l
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- 5. Spans exceeding spacing criteria O 6. Equipnent and building anchor movement
- 7. Thermal loads
- 28. As the review of the sample progressed, certain design considerations were found to require modifications. Therefore, rather than perfoming additional sanpling, those design considerations were transferred to the Generic Program for a 100% review. These design considerations are listed in the description of the Generic Program under the heading "Added Fron
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1 Sanple Program." The cause of all modification, to piping or supports was '
O io.ntiri.o to a .ign con,io. ration to as,or, non, r.,oit. from a consideration which wa, not a part of the Generic Program.
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Dated: March 5, 1984 kwCL G MYRQR E. LEPPKE //
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R05t.M1 5. UMAN~
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Subscribed and swor to
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before me this 8 day of March, 1984.
h -- n 1h SEAL O nGy~a'.74 aster, Notary Public in and for the City and County of San Francisco, State of California.
comission expires ril 14,1986.
x.cocxxxxxxx:r ::co::cce::TcccMc" E. NANCY J LEMASTER
$ "m' '.3s f;0TARY PUBUC C.klFORNIA
' * ' ' i,3 13
- CITY Ar3D COUNTY OF N
SAN FRANOISCo fj 14,1986 2 h My Commission Expwes Apniw.,ma n -. -mn .~ maal>
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, PACtFIC ts As suw - . ....
l COPY (
Exhibit No.1
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1.21 Plant site Design of Piping and Pipe supports - nimble canyon -
april 19, 1982 MESSRS: C. E. MAIFIELD
' N. R. TEESIJE f R. D. ETZLER N. E. IEFFIE I
S e General construction Department is performing design work as assigned by letter of delegation from the Wh==ical a Nuclear Engineering Department.
To improve technical direction of design efforts, insure consistent engineering practices sad increase the scope of work which any bc 3arformed at the plant s the Mechanton1 a nuclear Engineering Department will reassasse responsibility 26, 1982. for all Design class I piping and pipe support activities effective April All design will be performed as dictated by the Engineering Department The on-site Engineering Group will be supple-Manual and implementing procedures.
mented by placing the majority of General Construction paroonnel currently assigne to piping support design work under the direct supervision of Mechanical and Wuc Engineering. Work will be assigned, directed and controlled by Mr. M. E. Imppke, i
reporting to Mr. M. R. Tresler, Diablo Canyon Piping Coordinator.
f i
General construction will continue to provide administrative control for personnel on their payroll and they will continne to provide facility support for the group.
of Mr. R. D. Etsler.
ORIGlilAL S!GNED D7. OR;;3;At Sm;iED !!
R., s. mazz 3. v. accx:a
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MELeppke/sar ocs DOBrand_
DADrand GSBates E Moore GHkster CEWelte O
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Exhibit No. 2 IN5mtDCTIM ND. 9 DIAB w CANYW PROJECT -
Page 1 cf 5 i
Revision No. O I
Effective Date: 9/17/82 i
O WSI'm PRa7ECT ENGINEERING GROUP f
APPROVED: h FRaIECT nat.
LE.T 1
[ /![2
'D E An~J6;2 dA?DE I PRQJECT JiHGlht.tx UNIT 2 V ;
1.0 PURPOSE l
h is procedure establishes the organization of the Onsite Project
- o Engineering Grou) at the Diablo Canyon jobsite and describes its
& duties, responsi)lities, and authority. ,
2 2.0 SCOPE
- Dis procedure applies to all Unit 1 and Unit 2 engineering work e performed at the jobsite by the Onsite Project &gineering Group.
T 3.0 RESPONSIBILITIES l
O 3.1 D e Onsite Project Engineering Group (OPEG) is an extension g of the home office project engineering organization. OPEG's n basic functions are to expedite resolution of insering U design problems for Construction and Startup, to expeditiously issue limited design changes to the field organization where hane office guidelines and directives permit.
3.2 Se OPEG is composed of an Onsite Project Engineer (OPE), an msite Assistant Project Dgineer (OAPE), lead discipline engineers assigned on an as-needed basis by the Project Engineer, and various engineers needed to acc eplish the assigned scope of work.
3.3 D e Onsite Project Engineer is the Project Engineer's representative in the field. He is the Onsite Project Engineering Group team leader who provides overall coordination, guidance, and a&ninistrative supervision to
. the group.
3.4 % e Onsite Assistant Project Engineer is responsible for overall coordination, guidance, and administrative supervision of the OPEG in the absence of the T E. His primary responsibility is to control the engineering work of the T EG for Unit 1. However, he can be used as needed for Unit 2 work and has SfHD Project Engineer signature authority as designated by the TE.
3.5 D e lead discipline engineers are jobsite representatives Le for the Hane Office hgineering Group Supervisors (ECSS).
Althotgh a&ninistrative direction is provided to them by the Onsite Project h gineer, technical direction and, to a certain extent, scope of work are provided by the home office EGSS.
INmtDCT1W NO. 9 DIABID CANYON PRQECT Page 2 ef 5 Revision No. 0 Effective Date: 9/17/82 Q
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Unless directed otherwise by the Ikane Office EGSS, the lead discipline engineers are responsible for determining whether a proposed t'asign change can be initiated, reviewed and approved by In the group, or if it should be forwarded to the h ee office l engineering group for resolution.
In general, all proposed design changes should be coordinated with o the SPHO EGSS to evaluate possible effects of field changes on SIMO work. Changes to systems, structures, and components important to safety (i.e., those associated with the reactor coolant pressure 1
boundary, systems required for safe shutdown, or systems rocuired
- to mitigate the consequences of postulated accidents) shoulc. be
'O discussed with SFHO EGSS to evaluate whether or not any required design ::odifications would be more expeditiously acccuplished in o the home office due to effects on design criteria, positions on v Regulatory Guides, licensing; ccumitments, andesoff-project chief involving FSAR engineer design review conar.tments. Dose interface or C revisions, procurament actions, significant which affect Design Verification shall be accaglished by Project Engineering.
3.6 Specific responsibilities of the Onsite Project Ihgineering Group include, but are not limited to, the following:
3.6.1 Assist in the evaluation of piping and pipe support modifications. D is includes walkdown to verify as-built pi ? ing and pipe support configurations and to verify t w installation feasibility of any proposed modifications.
3.6.2 Issue pipe support designs for small pipe axi spprove pipe support modifications for large or l small pipe as required.
l Requests (DCR's) 3.6.3 Review and aprove Design y(DCN s) for design changes and Design Cun;;e Notices falling within the guidelines of Section 3.5. All
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DCN's shall be issued with sketches attached, in accordance with Engineering Manual Procedure 3.6 or 3.65. Sketches shall be incorporated into drawings and issued by SFHO.
3.6.4 Resolve problems related to design modifications identified on Plant Modification Followers (PMF's) and Dm's/DCN's when determined to be within the l
! O guidelines of Section 3.5 l (' Resolve Diablo Problems (DP's) issued by 3.6.5 Construction /Startu) and issue DCN's as appropriate when determined to'se within the guidelines of Section 3.5.
9 _ . _ . - , . . . -. -
INmtDCTIN NO. 9 DIABID CANYW PRf1ECT Paga 3 cf 5 Revision No. O Effective Date: 9/17/82 O
3.6.6 For those design changes determined to be outside the guidelines of Section 3.5, review the associated DP's ,DCN's, Dm's, Action Request Transmittals l (ART's) and PMF's for caipleteness in terms of c problem definition prior to forwarding to Smo l
' Project h gineering for resolution.
O 3.6.7 Provide representation to the Systems Intere: tion Program (SIP) valkdown team to assist in the definition of potential problems as designated by -
the Home Office EGS; and assist in providing design I fixes to problems identified by DCR, ART, or PMF 0 when determined to be within the guidelines of Section 3.5.
O 3.6.8 Provide general liaison between PG&E/Bechtel )
y Engineering /Startup personnel in the field and S 50 c personnel on matters that pertain to engineering.
7 3.6.9 Represent Project Engineering at Plant Staff Review
,0 Connittee meetings, as required.
3.6.10 Provide input to Sm0 for weekly and monthly Engineering Progress Reports.
4.0 PROCEDURES 4.1 General c
4.1.1 Copies of any DCN's issued by the Onsite Project
' Engineering Group shall be forwarded to S m0 for review. In addition, any clarification of design requirements or technical direction of a significant nature to Construction or Startup shall be appropriately doctanented and distributed.
4.1.2 Sm0 Ergineering shall review each DCN for concurrence, but Construction or Startup does not require Smo concurrence prior to proceeding with the OPE approved design or design modifications.
However, Unit 1 modifications shall not be implemented stil all necessary requirements of the operating license for design changes are met. In the event that the S MO does not concur with the direction provided therein, Project Engineering O shall notify the W E immediately and resolve any s' - problems. Any design doc ments issued will be revised, reissued or cancelled as appropriate.
i
l DtMUCTIGI ND. 9 DIAnm QuWON PRCUECT Page 4 cf 5 Revision No. 0 Effective Date: 9/17/82 O l 4.2 Jobsite Initiated Design (hannes 4.2.1 Desip of piping and piping supports, and design sodifications produced by the TEG shall be in accordance N with the applicable sections of the following doc eents as murented xilow O - Project Procedures Manual, '
Diablo Canyon Project i
- Engineering Manual, Dgineering Departanent Pacific Gas &
4 Electric O - MLNE Piping Group Controlled Procedures,
) Instruction & Criteria v
C'
- Project h gineer's Instructions Manual O 4.2.2 DCR's, DCN's, N 's, and ART's requiring Engineering
.O acticn shalt be screened to the criteria of Seceton 3.5 to determine which ones could be sore expeditiously coupleted by the group in the field.
4.2.3 Ntabers for DQl's/DCR's initiated by the group shall be obtained by calling the PG&E Project Coordination Section.
4.2.4 Upon design ccupletion, the applicable docunent will be signed off as follows:
- f. mad Discipline Engineer for Discipline Group Supervisor.
- Onsite Project k gineer or Onsite Assistant Project h gineer for Project Engineer.
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4.2.5 After signing, copies of the documents including documentation of discussion with SfMO shall be forwarded to Project h ginating for action as applicable.
4.2.6 Originsi copies of Unit 1 DQt's/DCN's issued by the OPEG that fall outside the PSRC guidelines for supports and as-builts shall be forverded to the Plant Manager for acceptance and work assignment.
(O Original copies of Unit 2 DQt's/DCN's issued by the WEG shall ut sent to GC for implementation.
IN!mtDCTI(M NO. 9 DM N m Page 5 cf 5 Revision No. O z'r etiv n e : 9/17/82 O
i 4.2.7 The (FEG shall keep SMO PCS apprised at-all times of the coordination and implanentation status of any DCR's/DCN's initiated or issued by them for tracking CD pur p =.
- 4.3 Signature Authority 4.3.1 1he Onsite Project Engineer and the Onsite Assistant Project Engineer have the authorization to sign for o the Project Engineer in all matters related to i Engineering as delegated by this Instruction.
Copies of all itens signed by the OPE or OEE shall 4 be forwarded to the Project Engineer.
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- O -O exh'6't "o 3 INTEROFFICE MEMORANDUw; Oiablo CanyonProject PACIFIC GAS AND ELECritC COMPANY BECHTEL POWER CORPORATION ,
1 Ta S. Bhat Dm April 1, 1983 ,
1 F" R.G. Oman Fue No. 927 )
of Onsite Project Engineering subject Non-Manual Manpower Estimate At Jobsite Emns.on 3507 As per your request, OPEG is forwarding the estimate of non-manual to go manpower as of April 1, 1983.
O M
- R.G. Oman Onsite Project Engineer JJ/RGO/in .
Reply Requested: No cc: G.H. Moore w/a G.V. Cranston w/a l J.D. Jumper v/a l M.E. Leppke w/a J. Shryock w/a l
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Exhibit No. 4 0-INTEROFFICE MEMORANDUM Diablo CanyonProject PACIFIC GAS AND ELECTRIC CGMPANY BECHTEL POWER CORPORATION To S. Bhat om July 25, 1983 Se R. G. Oman %* 927' o' Onsite Project Engineering suwi Non-Manual Manpower Estimate ai Jobsite Eme - 3507 In response to the recent request of July 12, 1983, attached is the forecast estimate of total OPEG non-manual manpower as of July 1,19d3.
O
&C e R. G. .an Onsite Project Engineer RGO:kms Reply Requested: No .
cc: G. H. Moore w/a G. V. Cranston w/a J. W. Shryock w/a P. Snooks w/a J. Leahy w/a MMS - C IS S
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Exhibit No. 6 4
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- w%N I?!TEROFFICE MEMORA!!DUM Q
n i.. + . h- C c . . , n n. W . *m *- %@pm.s M
FACIFIC GAS AND ELECTRIC COMPANY BECHTEL PCWEF CORPORATIOiV
<a September 29, 1983 ,
Te Mike Tresler F', sa 910 r,om Leo Mangoba .
Onsite Proje:t Engineering 5 -et Discre;ancy Raports oi m Jobsite Eneace' 3067 Three discrecancy reports were proposed by one of the OPEG engineers assigned to PSDTC team.
I have reviewed their contents and conclude they are :nore like questions than discrepancies associated with the review program or other activities perfor:ned by the project.
" The following are the subjects of these write-ups:
- 1. Anchor Bolt Scacines: The write-up referenced manufacturer's recomentatidn. The original copy cf this write-up has already been given to you for review by the Civil / Structural grouo. This write-up apoears to duplicate earlier reviews by tnem.
- 2. Welds: The write-up generally duplicates the effort of John Miller, who was aiced by Tze Quan and Paul Brooks in resolving the identified issues.
- 3. "Unbraced Lencth of Anoles: OPEG has performed detailed evaluations of the succorts listed in the attachment to l
the write-up and has not found any to be discrepant.
In addition, tnis issue was a subject of the small bore review program and generally referred in the NRC SER.
These write-ups are being sent to you for your information, f
l and unless otherwise directed. OPEG will t ke no further action.
Thank you, gri.
LM:kes l Mangob Response Required: No Ne
, Sm P- NE Attacnment: Yes c:: MLe:pke w/o LShi ley w/o Schi nis wie RC::an wn f - -- . ..
l Exhibit No. 7 Procedure 10.1 Attachment A Page 1 of 1 ;
l O' -
1
- PACITIC CAS AND ELECT?.IC CO.P.O7.
C;CINEERING DEPAT.TME'.T DISC?IFANCY REPORT Control Nu=ber (3)(3) - (3)(3)(3).(3)
Diablo Canyon Unit No. 1&2
@ JECT OR PLANT (S):
Pipe Support Design Engineering DRGANIZATION ATTICIID:
G.V.Cransten Proj ect Engineer INDIVIDUAL RESPONSISLE FOR RESOLUTION:
SUEJECT (ITD!/ ACTIVITY):
F.ilti Catalog, and Phillips Catalog M-9 & ESD 223 R_:::.
_RENCES:
DISGE?ANCYNanu'actures soeci'v the rini:m.rn center to center distance to be equal to 10D where D the hole dip ~. ster. Design has used the bolt size on shells,
~
not the hole reg'd for the shell for D. This was caused due to :nissing dr.fer: stic n e- m %1. e% -,-,' a e~- chi l e e-m -= 1 Rechech anchor bolt calc. for shells, Reducing PROPOSED,AC ION:
allowables per M-9 and rechecking interaction ecuations f or :hese cases where 10D (snell hole size) exceeds that used on Dwg.
SGEDULED COMPLETION: /Oi~ 93 Initiated by: [ Date: / [3 -- g '3 Approved .by: [ k, . Date: /0!7,!7'3' Letlu M c.o.l. endu xivw ACTIONS TAKEN: 3CC A thm e H es I noe.# 0370 J 7.
O p
-' CLOSID 1M Late: A , f, 08 Approve? by: - a i Late /pddf l
Concur *Chief. h / Engineering Quality Centrol #/
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- Approved by: Date:
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C.05D A;;reved by: Date:
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- Date:
Chief, I:; neering Ouality Control
' C ly re:;; ired if i=itia:ed by IQC.
2-3/9/S1
033037 0 INTEROFFICE MEMORANDbM Diablo CanyonProject PACIFIC GAS AND ELECTRIC COMPANY BECHTEL POWER CORPORATION H.R. Tresler o,,, September 28, 1983 3
J.K. McCall ,, %.
52.3.6 Civil Engineering DR on Expansion Anchor
- Spacing, Shell Type Anchor 45/23/B34 8-1414 si Eswnsen' , , _ , , , , , , , , , , , , ,
- 1. The attached DR initiated by Charles St'okes on August 9,1983 questioned the use of spacing in accordance with DCM-M9 for shell anchors of 10 times the normal size of the bolt while manufacturers.specify 10 times hele size ..
as the minimum spacing.
- 2. Discussion of the effect of this dif ference follows:
- a. The anchor length for shell anchors (Phillips and Hilti) is in all cases less than 5d (d-nominal diameter). Appendix B of ACI 349 O
svecifies that the areas to be censidered fer reductien due te overlapping are 45* sloping cones starting at the enter of the base of the anchor. Based'on this criteria there would be no reduction required for a spacing of 10d (nominal).
- b. In 1962 Doberne and Elgenson, Counsulting Engineers of North Holl)vood, performed a series of tests on Phillips Red Head concrete anchors to determire the effects of spacing on pullout capacities. The reported results showed no decrease for 10d on " diameter anchors and 9.14d for 7/8" diameter anchors. Reduction of spacing by a factor of 2 only reduced the capacity by 20 percent. This reduction is small co= pared .
to the factor of safety used.
l
- 3. Recom:nendation 1
Accept as is the spacing requirements of DCM-9.
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- ~
J. K. McCALL
%-. Ih*
JrDicCIEEEpst ein:dnl Reply Requested: No Attachment i
j y "j ', cc: GVCranston i G10doore ann.ite
{DNLOO3]
I
, Exhibit No. 8 i
" ~
Procedure 10.1 !
At tach,=ent A Page 3 ef 1 PACITIC CAS AND ILICTRIC CO)2 ANT E;CINIIKING DEPART.E;T DISCRIPANCY REPORT Control Nur:bar
, (3) () - (~0) (3) (3)-(3)
PROJICT OR PLAST (S): Diablo Canyon Unit No.' 1 & 2 ORCAN!!ATION AFFICTID: Pipe Support Design In;;ineering_
I';DITIDUAL RESPONSIBLE FOR RISOLUTION: C.V.Cranston Prej ect Eng.
SUIJICT (ITIM/ ACTIVITY):
- I AWS, AISC, M-9 Pull =an Veld Pro,:etdures
- .:.: :..,__RIN CES :
, DISCRI"ASCY: See attached sheets (5)
!- i ,,
PRO?OSID. AC 105: Make necessary changes in Lesign Guides and ISD 2 3 and Pullcan Veld Proceedures to bring the= up to A'r.'S reccire=ents for pre-qualified partial & full oenetration welds. '
S 2IDi~ID CO:2LITION: / /
Initiated by: p Lzte: /0 [O Approved by:. [ kt _ Date: /0!7 !PI tr.
l ACTIONS TAKI5: The .ablo Canyon Proj ect has no concitment to observe weldin3 l
requirenents of AUS. The issues identified by this discrepancy report du;11cate an investi;;ation conducted by General Construction (continue on Page 3 Of
%' CLCSID .^
- .ed ~~y:
f'-D 'T Date: /0 r p3 ,
( .* 7,0 07 " h EI !O.
Chiai, In;ineerin; Oual:.ty 2ntrol //
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O DI'-E.D PFD3ID:S CDhN W. D DESIE, DRNCNG DO: ZEE 22;L"' ION, INSTALIAT. ION RC CC INSPEC.' ION.
C.C. SIDKES,10/4/83-Fewritten frcra 7/5/83 paper, i 1. FIARE BENEL At3 FLARE-V GOOVE h?DS DESIG4 Bechtel San Francisco Office-Per Dan Curtis by phone E'S table 2.3.1.4 applied to radius of tube steel ob-h frcrn table in a paper ,
entitled "A Designers Gu:Lde tn Welded Joints" written by: Mark F.icnaels
' ~ ~ ~ ~ '~
of ,ma: cert outside corner radii ' table'3.3'.'
Note: The word maxinnzn radii. This is not good engineering practice.
The conservative arproach dictates that tS. mininnzn radii be used to ensure the safety of the weld joint. It should also be noted that Park Michaels paper on the design of welded ,1cints has, to my knowledge never been approved by the Engs.neering Depar=nent and issued as a control doc-ument to eng2.neering for use on Diablo Canyon.
Bechtel Site-Per a handicok supplied by the tube manufactures M institute, all tubing manufac:dreci in the U.S.A~.~ is made or rolled with a radius of 2t to 3t for all sizes. Having assumed 2t to be the miminnrn, all calculations were made to AWS. table 2.3.1.4 using 2t = R.
Per site investigations, Jeff Van Klcrapenberg, Ken Palmer and myself disccwered sczne tub 2.ng on site (Diablo Canyon) has a minimirn radius
. of 1 t. Therefore, all welds of this type per this design grcr.:p are not conservat.tve.
Westanchouse-Through review ci drawings issued, their designs are l also in question.
- 2. PARTIAL AtO FCII, PETIRA"' ION GRO:NE KIIOS DESIGN Bechtel San Francisco-Per drawings supplied to field, very few if any are de:i h m ectly. Symicls indicate or=lete joint welds.
t.is is true for all joints requiring preparation. No angle for prep-
. aratica has been indicated and it is n.,;. cinious that the designer is aware cf the miminnzn joint reg 2irernent.s per NG 2.3.12 and Fig. 2.10.1.
Eo'ever, on joints created by natural i:Eersection of 2 marbers, it is circious that the joint in many cases is a partial oer EG and not a 50.11
.. penetration wid due to angle created by in.ersecting me-bers being to s: .111 for a full penetratien weld to be ade.
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Bechtel Sitee group has tried to ccr: ply with NG. recuiremnts in designating teth S and E per 2.1.3 and 2.10.3.1 table 2.10.3 and Fig. 2.10.1 and the included dihedral angle el+hr on preparation or by natural creation of n 'ers intersecting. . However, Pullman QC have continually refused to check weld per call out because ESD does not provide them with a procedure
- for perfo:=u.ng verification. They have recuired the (E) effective throat call out to be rem ved.
Westinghouse-See cxzments for Bechtel, San Francisco.
i
- 3. S-3ED JOINT FIIJ7 WEI.DS DESIGN Bechtel San Francisco-Per drawings supplied *e field and continued use cf fillet all around call out 2nstead of a specific call out adjusting
.ne leg size for the dihedral angle adiustment. It's obvious that the d:.nedral angle has not been considered in the joint design or if it was, it was done incorrectly.
Bechtel Site-One gren.:p has tried to be consistent in adjusting the inner and outer fillet leg size based on dihedral angle so that +h effective
-t'ioat on all sides is the same. This allcus the joint to be analyzed as though it is an equal leg fillet all around with only a leg adjustment *w ch-M 1 this when velded. Tnis is shcun on d%., so construction only
! has to make what is shcun and not interpret what is shcEn using a table which is rc: us"*11y at hand.
Westinhouse-See Deci?.el San Francisco.
t
- 1. E .AF.E BEVEL AND IIARE-V GRCCVE VEI.DS DPEING REF.ESEN17LTION Bechtel San Franciscc-No partial welds have been s'run since S(E) have been crititted. NG 2'.1.3 and 2.10.3.1 and 2.10.3 state that (S) groove weld depth and (E) effective throat shall be specified on shop or working drawings. Tne hanger drawings sent to the field are 'mh shop and working draw-i.ngs. Also, M-9 states tut only pre-qualified joints should be used on Diablo Canyon. Many inmanagee.nt cc . tend that this job is not covered by NG cois. However, M-9 states that it is governed by AISC 7th, Ed.. In AISC sec-ics on welded joints, page 4-131, paragraph 4, AISC states that small
' dreia: ions are possible per NG code and other joint fc_-.rs and welding pro-cedures may be eployed provided they are tested and cualified in ac::ordance J
4.~ w.d. NG Dl.1-72. 7nerefere, Diablo is ccrecrne? by NG D1.1-72.
- Prge. 3 CORRECT .DW(s. REPRESENTAT10N Q !
L k
scs))q s(E)k FL ARE -V FLARE BEVEL S=Raf.ius of tube E= Effective Throat per 'able 2.3.1.4 AWS Bechtel Site-One group has ccr.92ied with call outs above. Ecrever, Pulltran QC per ESD 223 have .W the rent:nral of S (E) frcn all drawings.
This was because the ESD does not provide a procedure for them to use to verify the welds al:cve. Tne symools left, after renoval of S (E) indicate
, full penetration welds even thougn, QC states this is not the case.
' Westinghouse-See Bechtel San Francisco.
- 2. PAE"IAL AND FUIL PENM"'ICN GCOVE KET.DS DREING REP. SEN'TATION Bechtel San Prarx::iscr>-No partial welds are shown since S(E) call outs have been cmitted along with preparation angle. Primary weld used is bevel.
Now indicate'd f Per AWS this si, a full penetra' don weld.
N L
l Should show SCE) -<
N With c=nsideration given to AWS section 2.3.1.3 in specifing S(E) and Bechtel Site-one grcr.:p on site has tried to ccr: ply with the acove call out per AWS 2.3.1.3. however, they haw.neet ocn.inued resistance frcrn QC in that the weld specs used for installation were written for piping per AIG O chapar 10, Fig.10.13.1.1A is 37h* differs frcn the prep angles for structural V- steel specified in Fig. 2.9.1 and 2.10.1, wt.ich usually inficate a redni. :. angle l
of ;35 l
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P:ge 4 Westinghouse-See Bechtel San Francisco Cu...-rc.s.
- 3. SWD JOINT .N7 WEI.DS DR. WING REPRESENIATION Bechtel San Francisco-Per Drawings are shcMn as #411et all around for all angle 3. Per AWS fillets are limited to a min. unum dihedral angle of 60* and a maximum external dihedral angle of 135*. All fillet made beyond these angles are considered partial pere & tion welds since AWS
- requires a reduction of e'ffective throat of 1/8" wnen angle is less than 60* and greater than 45' and h" reduction when less than 45' but greater than . 30*. 30* is the minum:n dinedral angle for structumil steel except '
tube steel which has a minimum angle of 15*. Per ESD 223 the leg is adjusted i for dihedral angle but no increase has been added to account for throat re-duction required by design resulting in adequate effective threats per AISC and AWS.
Bechtal Site-One group has cc!.sidered dihedral angle. To facilitate design, joint was sized assuming a constant effective throat size. After sizing effective throat requirements then using d2.hedral angle the 2 non 90*
1 sides were adjusted so that the installed effective throats would be correct per design. Fillets are not called out when dihedral angles are less the.n 60*.
Partial penetration welds are shat 2.
11 2) 3) INSm'" ION AND QC IN.vwacN All installation 12.s been rade per Pullran's weld procedures. Per a copy obtained frczn Pullran QC of these ywce&res. These procedures as written state that they are for the installation of pipe and pipe attachments.
No mention is made of their use in installation of pipe supm. All prep angles and joint details are written for pios, no notes or #'ications are indicated for their use in ins"1 W g pipe supcort steel. QC has been supplied ;
with these procedures and ISD 223 to inspect pipe suygL welds. They have been supplied no in#o= ration as to the correct installation of pre-qualified jcints per AISC or AKS. Furthernere, per ISD 223, they have been instructed to check scme weld joints and not others. i 1
A jeint forned by =*e-bevel or flare-v welds en all sides,such as l 2 : 6 es crossing does not require checking and per -e 223 sten the size cf a. ;
l )
flare-bevel or flare-v is sha ., the rethod of weld measurerent per ESD 223 does ret su.colv the effective threa er any di.ension stich can be used to de =~T the effective threat.
i
~
Page.5 O
Per atta=rrneitt I of ESD 223, no limitation are indicated for =M:ral or tic steel dihedral angles. NG gives the minine dihedral angle 30' for strue: ural and 15* for tube steel. Also no throat in=rease is included to ccceensate for recuired throat r e ions based on dihedral angle per AWS 2.3.1.3.
Per attae5. ent J of ESD 223, of stat use is the meast mr.nt. of S ? Per MG unless S=R. R=Padius of tube, E can not be de+a-ined through use of table
~2.3.1.4. Also upon terriew of this attachment and table of raxis:n radii of tube steel in paper by Mark Michaels, it can be shown that table J was and is based upon the wax.umm radius and not the mininun resulting in a non-cansarm.ive design.
P ACTICS TAKEN: and Engineer. The resolution of this issue was monitored by Project QA. See chron No's 023208, 024230, 031964 and 033851 plus attached QA " Work plan and log" sheets in which no discrepancy was noted.
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Exhibit No. 9-Procedure 10.1 Attachment A Page 1 of 1
?ACITIC GAS AND ELECITCC COMPAh7 E;GINEERING DEPARTMEhi DISCREPANCY REPORT
~
Control Number
~ ~
() () - (_4 6 )() G )
Diablo Canyon Unit 1&2 PROJECT OR PLAhT(S):
ipe Support Design Engineering ORGANIIATION AFTECTED:
G.V.Cranston .Proj ect Engineer INDIVIDUAL RESPONSI3LE FOR RESOLLTION:
Unbracea Length of Angles SU3 JECT (ITEM / ACTIVITY):
AISC 7th edition Sperion 1.5.1.4.6b See attached underli.ad -
ccpy of AISC.
Many angle ne=bers used to build pipe supports exceed the D* SC?I?ANC, O
i
- naximu= length f or which the allowable bending stress =ay be taken as .6(Fy)
These sembers should be sedified by boxing or bracing at PRCPOSED, ACTION:
cr:.tical compression points to comply with AISC or the allowable bending stress reduced and Interaction Equations reverified.
SCF.IDULED COMPLETION:
Initiated by: ! _ _gj ds. /
Date: /d / 8 ar/ 3_
Approved by:
c Ad.v Date: /0!7!?3
/I A list of 18 supports identified bv C. Stokes as discrerant ACTIONS TAKEN:
uere evaluated for bendine stress and all 'enad a k. .--.--. i. see.A .---
p Si?SS 3.10.1 Rev 0. This procedure =akes use of mathe=atical (see nace 2) sg CLOSED Date: /O *7-D Approved by: {h- f M Cencur*,44/ Date:/#d/ff
' Chief. Engineering Quality Control '/
1 2_
- O.. ,
2
~
' expressions providing more exact esti=ates of buckling strength see
! commoentary on paragraph H 1.5.1.4.6 in A I.Sp manual. See attached excerpts of A I.S.C code and letter P. Schurer. .
1
.1 I
i i
0
'l l
i
+
1 i
i t
I i
O. -
i_-____ __ ,_ _ . _ _ _ _ _ _ . . . _ _ _ _ . - _ , , . _ . _ , _ m -_ .,_ _ . _ _ _ . _ _ . . - . . _ 4._, . . _ _ _ ,. ... , _ _ _ _ _ - .
. Saw.wnt i:u: --- ia_.ar.g . s . ;;'
l
./'Nn :he gas r:e=i.are process s..all con. 1. .1.0.:: On the gross section of m:.311y!a:ded r:m resst::mem ers ,
.L.) : E:ss:rocer . cr Gas .Vetel.A rr Gan:r. anen Ki ? e:ceec.s C,: '
v:.uons of See:.1.17.3: EEOT or E CT tg _" ,~e.
- r crocess shais comfor= to the Specidea. (1.5-2)
- ce I~-.:. Cored-Are Wekiing. AWS A5.20 F. = E N 'Y r.f S-et. 1.17.2. 1.5.1.2.3 On the gross section of axiaHy loaded brac=g anc secondary n sna] consutute su5cient evidence of con. members, when ter exceson 120*: b- s-- - --
7 , F. (by Formula (1.51) or (1.5-2D STRESSES
- I 1.5 --
200,-
- 4. 1.6. 1.7, 1.10. 1.11 and in Part 2. au com-be so proportioned that the stress, m kips -
1 5 1*3.4 On the groes area of place g. .roer stifeners:
rd the icilowing values. except as tney are F, - 0.60F, 1.5.1.3.5 On the web of roUed shapes at the toeof the fHet: crippling.
~ see See:. 1.10.101:
F. - 0.75F,
.t pm holes:
e , - 0.60F, dKnb. 1.!.1.4 Bending A .1.5.1.4.1 Tension and compression on extreme fibers of compac: hot-e *' mum tens 5e strength of the steel. rciled or buDt.up members :except hybrid girders and members of A514 4s in eyeoars.. pin. connected piates or ouilt. s:ee! syr::me:: cal about, and loaded in. the piane of their =imor axis and meeun; the rfquirements of this section:
., = 0.45r ,
- ~ '
ts see Table 1.5.2.1.
In order to qualify under this section a member must meet the foUowing re:;ui.:=ents:
0[40F, i a. The Sanges shad be continuously connee ed to the web or webs.
fabricate: 1.. :es msy be taken as b. The width-thickness ratio of.;mstifened projec:ing elements of the 12..c tne thiccess of the web. Sre Sect. co=pression fange, as dedned in See:.1.9.1.1. sna3 not exceed n wec's. For disc.:ss:en ci bis 5 shear stress 52.2. VF,.
,se:iuns of membe s wnose wees lie m a c. The width. thickness ratio of stifened elementa of the compression Sec:.1.5.1.2. ' fange. as desned in Sect.1.9.2.1. shall not cresed 190 'V F,.
The depth. thickness ratio of the web or webs shan not exceed the 2, value
- ion ef az.: ally ioaded ce=;ressi'en members
$ d he.
e .ce ness rano of:nv unersce:i segment as '
d.t - 412 1 - 2.33 b!
F l V F, f
(1.5-41
- ~
El r t' . excen: that it neec not be less than 257 V7,.
M- Ti:e cornpres.sion :iss;c shall be supported latersuy st intervais Oc.: - 1..%1 e
ar noo
'.~;*_~..;*'
'- ~~~
.:: :: e :caed 70.05. Y T, nor
. , ga s a:r .
E.:.. ::- v:: : ;:r :e-s and members of A514 stee:. r.e2- s and gi-iers
-j . .; :rv .iee.s :es:r .ed on tae basis of composne :c::en wh:c:: meet
,, I ,f, -
e- re.=e:n nts of suu r: s; :phs a. b. c. d and e soove and are con.
mencat vasure fer var: ave tram ,f neer me *. '
- ~. r tr
- ast E' is *aKen BJ 4:Uty.
g y ; W c s -scr s M Q * ?.h r & .:E.# l . = M.- --
-- " - uc.a= ~
.ad rem a r.e-=-. .~*~w
.c c.e..:** ? 1 : :.r
....,s.
~.,
,_'.. -e,2
_m
~.-p<<~ ~
.s em ... . -
- .g.# 3
.p _
Y.e,:./ . ,e].,: %.: a 1.
,, s - '"*~
- d. -- -* -
p--
- .. 3. .
_f,;;;;r;;y- ,- y . . -
I$5Y==,- I$- . t . '>Ch* rb W *
( h h .h. & $ Y h h S $ $ b 5 $ W ?
~A
e
_ . _ _ , ._. . .- -.u.. - - 2 h. ::. .. .: . . . . , , ,
, Sowure Sinalfor Bw.:.ungs . s.2
. . . 510 x 102C.
- enec :sr % ot the negative b..en :. r
- r
- en are maxi =u=> at poista of r)} e ers, de =sr=uss positive moment
- sg:tive moments. This redact 2on I*
- 170 X 10'C. (1.5 6bt gi
.cading on castilevers. If the nega.
diy f.r::nned to the oeam or g:rtier. the , . , Or. wnen the compression fange is solid and approximat.rly rectangular in
.cring the coiu=.s for the combined cross.section and its area is not less than that of tte tension dange nst .n2 stress, f due to any con.
n:t exceed 0.15F 7, ,,,12 x 10*C. (1.5 7) d .:rdt.rs and members of A514 ateeli war , .
L5.1.4.1. except that b,,'22. exceeds .In the foregoing, m:y be designed on the basia of an ,
- distance between cross. sections braced against twist or lateral
~,, dismiacement of the compressaon Aange r- = :::ifus f gyr: tion of a sect 2on comprising the compression *:nge 013 (_:I' ':
) v'F',, (1.5 5'.
pics one.ttird of the compression web area. taken noout an sxts tn the piane of the web ton on extreme ibers of doubly- 4, = ,,rea 'of the compression Sange
=ceting the rect:irements of Sect. C. = 1.75 - 1.05 Of. 'Jfi) -- 0.3.tafi,afdr. but not more than 2.3 *.
best rioout their =inor a.zes iexcept where Af. is the smaHer and Jf, the Isrger bending moment at the i square bars; and solid rectanguiar encs of the unbmeed length, taken about the strong axis of the member and wnere Jf Jf.. the ratio of end moments, is positive
,3 ',,,=
wnen Af and Jfr have the same sign . reverse curvsture nending and negative wnen they are of opposite signs. single curvsture en on extreme ibers of box-type bencing;. When the bencing moment at any point within an
.:ng2 er web wicth thicknese ratio uneraced length as larger than that at both ends of this lench. the
- .1.5.I.4.1 but does conicrm to the va}ue of C. shah be taken as unity. C. shan & be dez2 as ission Sr.nge is braced later31]y umty in computing the value of F., and F., to be used m, Formu!s h transvene cirtance out.to-cut ;1.6.is t. See Sect.1.10 for further ti=utstion tn paate girser g
, ange stress.
'0F, _ For hybrid plate girders. F, for For .ulas '1)-!,,ai and q,,,1,.QI is the s of "szursi members not covered m.
yieic stress of the compressiona snee. Formula .1.5 7) shall not appiv to "
g.briu g'.rcert 1.4.4: .
gy' 1.3.1. 4. 6b Compression on extreme fibers of "ezurai members in. ,,,
ciudee under Sect.1.5.1.4.5 but not included in Sect.1.5.1.4.6a: -
me $bers of aerurai members in. ~
tx.:.s of symmetry in. and ioaunti in, byjd , d F. = 0.50F,
, on extre:re $I.:e!$ of Casene $* Dent I
i* :".nuten P.*.* F *.-tm hss , a.J.ve e ur p7pyjgy,i ! at sections bent about their rnajor 3xis are braced later2i}y in ess a n;:::ct vaine can Fe i :tri$et". t :he re:.cn of cornpression .. strms at intervais n,ot exceeding 76.06, s rf -
. m:! no! = ore in:n 0.60F..' / 2...A..,
- Bearing 28cn4L 's L
- 33 ' 4 L 6 so.47"
.. contset ares >
- }.{ 1.3.1.3.1 .\I*ile?. surise.-es. ine:ucing be=:.n; !!ifer.: : :r.
rear e. :r-.e:. or oorec hoses:
=i. ; in v.* ,
1.5.o.z.
F, = 0.*0F,"
10:C. , . *
. C. esn oe conservauveiy tamen an umsy. For =ailet values see Appenet: A.
Tig. AI. 3. 5104.
j .4.n. ** "sVne . nar-* m centact have c;ferent yield stressee. E, shall be the unailer vasue.
.1.4.o. .as! 1we p. igr p.ts.
/
- D.[:bl~ 'd@.nd
~~~
- .- #2'P . . 92*$*PEMf?M
- w. .
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7 q-e '"
r? P .
u.t,.5-, r.
m.e .tav. .T.. .<'.e, rc. v .
m.m 44 .. w .l.eb.y ..e w .<5 *w** . - _ _ _ -- - s w.r. ==. .* . 4m a *. .
o mycwmm % 4. . o t _ -
." - - :*a *~s~ M ;E~5 &~ S ~~ T 5 5I O'" W w x 'E ~~- "*:
.: ~*'. .. ; &~.:.~ n S
- a s..
..: -t'.M Qi%W'
- ^
~~
a-.2'E-; '
$w?*iue=xm.tf,di+}G&5'M_ w
=-
%=e **
- " % ~ gr emm. & 'm.s.- ?.*-K*j"'"k;_W P - e r e - m_ - - -mea t--
. s .-- ..
y ? m g m.; ;a=.6-,
-.~,. . . - . .me y v.t -n e . 4- ~_ m g '.
..,.m. ,_ n=~n = 1 ~ , %
e ,
m& ; - 1 = <. 1 - --
m,;;.7,43@l5 5.5M-TW_l@W~~_Q$M
- x. ~ w-_ _w . ,e m-e.---
n m_ _ e. . g. g ..
, m- - _e m ,-m - .
Lek __s_ ___iehL%@.m.-_._
~ _ - - -
i - -- -_
n#=_D Baa&diiBnwWe
-w . . . _ - - _ _ . - - __ = _ _,5 g 3 - ga _
p: . - .
I 5,.11L4 A15C Spei4emnon j4 SECTIONI.11 COMPOSITI C 1.5.1.4 Bending .
Delete suoparagraph d in iu en=rety and substitute the 1.21.2 Design Assstmptions-1.5.1.4.1 .
fotower -
2,11,;,; gg 3,3,,.__:. , of The oepth. thick.=ess rauo of tne weo or sebs shan not exceed the "For construction without tempo
" d.
=odulus of the trz.=sfor=ed co=;
vajue pven by For=uias (1.5-4ai or .1.5-4b: as applicabla. *
(referred to de botto= 1.=ge of .n.
stitste the foUow=gt d
412 1 - 2.33E,/. woes E, -f. 5 0.16
-- (1.~u a)
- VF, "For en=struccon without ta=
stress =ay be co=puted !.ro= the t for=ed secten =odulus S.,. except d 057 when p/ > 0.16 (1.5 4b)" ; shan not exceed that of ~"or=ula 'l
' YI' '
approprtate value of Sect.1.5.1."
{
t 6
1.5.1.4.2 1==ecately fotom.ng the words "of Sect. 1.5.1.4.1" add a ec- uc :.==eciately foCow.mg tne words "except that by S " delete f SECTION 1.15 CONNECTIONS tae CoC.=.a.
~
1.15.5 Restrained Members Chamge for=uia = umber "fl.5 5)" to "fl.5.5a)"
In t.he f st line of de seco=c strained".
1.5.1.4.3 Add a second paragraph as fouows:
"Dou'oiy.sv==etncal I. and H-snace me=bers bent about their =inor ax:.s .exce-t hyc :d g:rders and =e=oers of A514 steel- meeti=g the requ:re. SECTION 1.23 FABRICATION re:ts o_f meet. 1.5.1.4.1. suoparar.r.apn a. except wnere of 3, exceeds
.0. v F, out is less than 95.0 VF,, may be desi ned on the basis of an 1.03.1 S enigh tenin g M,aterial a' owao;e oe:ci=g stress Delete th:.s sub.neac and se entt the worcs " Rosed =.ater a.1". and su F, = F, 0.933 - 0.00~5 / 'I.5 5b!" reading as foUows-
\ -r N .F.J "1.03.1 Camber ng. Curving, an.
The local at,piscation of beat er 1.5.1. 4. 6 a !==ediately fobos. g de werds wer h 1.WC Wuce or ce mW. CC ace: u: =cett::: tne requ:rercent. o.<- see. - -.
- 9..*.a." of heated areas. as measured by appr for A514 steel nor 1000'F for other m 1.5.1. 4.6 b 1==ediately fo"omng he words "under Sect.1.5.1.4.5."
aed: " arc =een=g the requ:re=e=ts of Se: 1.9.1.2,". I..*3.8 Welded Construction j '" " In Table 1.':3.6. for thickmess "T
- the heading "Weldi=g Process", cha=
1 S ECTION 1.10 Pl. ATE GIRDERS AND ROLLED BEAMS ;
l
) '1.10.5S L:."e n ers .
1.10. *,.3 I: the di-d para r se .. :==e:i'steiy foUowimg the werds "noies 2..: :e suen t .st' . ceiete: "tr.e z=s..er parel di=er:.uom. o or h.
and sues:tu;e tne word.s "/. does set exceed str.e.avs; ue net;.sen exc-e:
oy Fer=ula 345r v[" 1.101."
1
She:urr.i Smi for L;4 ~.gs .1 :t T . :: . - 1.5/. ( 00.0 1.S.2 Sidesway Prevented
.~ . .:. - 1.5/, G ~.0 In frs:ces wnere laterai stability is provided by adeounte s::sch=ent tw cre: 7. - ;dc. - 1.5/. ( 20.0 diagonal bracng, shear walls, an adfaces: s: 2:: re hsnn; adequate laters.
'.' ~ . ~
stabi!.ity, or to Soor slabs or roof deck.s secured horizont. ally by wails or
? - 50.; - 1.5/. ( 40.0 bracing syste=s caraEei to the plane of the f.rame, and in trusses, tne efee::te
-r = 70 0 - 1.6/. ( 54.0 legth factor. K. for the compression members snall be taken as unity,
.r *Jo- -- . . unless analysis shows that a smaller value may be used.
- y t e same ic ces. sha3 not exceed the
~;n:.s. t!.s anear strets ailowed in Sect.
. 1.S.3 Sidesway Not Prevented .
J In frames where lateral stability is dependent upon the bending stiEness of rigidly connected beams and columns, the efective ler.gth Kl of compres.
T, '" 15.0 1 - /.A./T.)
sion members. snall be determined by a rational method and shall not be T. ~ 20.01 - /J./T.) less than the actual unbraced length. .r ut to a ..:ree: ioac :p.Gd to an of the . .
. n:ns: ;.reter.aion ;c4 c.i sne bolt. 1.S.4 Maximum Ratios The siender sess ratio. Kl r, of compression membe s sh:Il not esceec CONNECTIONS SU3 JECT TO 200.
MTION OF STRESS FATIGUE) The s,i enc,erness rat;o. K1 r, of tension memoe s other than rods.
preferably should not exceed:
For main members . 240
- ion is ceined as the d:.= age that may For brac=g and other secondary members 300 u.= 5er of St:c:::nons c: stress. Stren d these Suetustions. In :ne case of a
.e computed as tne numen=:1 sum of SECTION 1.5 WIDTH-THICKNESS RATIOS re::sive stresses or the su= of =ar.=um
-io= st a given poin:. rese.! ting f. rom 1.9.1 Un.stisened Elements Under Compression 1.9.1.1 Unstirened sprojectingi compression eiements are those has.
.r"' vention:.1 buildi:p need to be de- ing one free edge parallel to the direction of ecmore'ssion stress. The wictn i.,, 's in suca structures occur or.!y a of unstifened pistes snall be taken from the free etige to tne drs: rcw ei
- '-or stress :ue:usuens. ,,,,A he t':steners or weids; the width of legs of angies. cnannei sad zee danges, anti
,:r necaue loses .s :co infrequent to stems of tees shall be taken as the full nominal dimension; the wuith of g However. er:ne runwsys 2nd sup- f.anges of I.snape members and tees shall be tsien as one-half :ne fui; 2 ecciprnent are often subject to fatigue . nomm:I midth. The thickness of a sloping *snge snsll be rne:sureti haif.
way between a free ecte and the corresponding face of the wee.
4(.-
~
1.9.1.2 Unsnfened elements subject to axici compression or -a .
nression ciue to b.-d; r eksil he co-eid. d -. '.a;v e3.- . . wnen theE
- 2. sue ec: ts : .:irue .e::.iin 15 ced.ned U Ntn to tm0Kness is not grester than tne fodowing:
. ; , s:Q:- :ne 1:rse range .:rn::2tions t .. .
W b dmgle. anele struts:
.J coscie angse struts with separ: tors 76.0 N T'
. .. c , s n .-
.m. . . .e. .: .,.... . w. e.
5truts comorisi..c ocubie ang:es in enntae: sr.:.e or n:. tea armenn: from : . ne s cr.,.::mn.* "r .. . -
- rrenssn m e = t'ers: iamnrersmn dan:es e,f i:es;ns; ni:L..r, un
- M0\F '
, : :: :ne :: u- ur. a , ... and ic:
Ste: .s of tees ,10 s ?
- fo of sn nail" scace: compression h..e the setus: . vie: 5.to. thickness r:t:o exceed; these vslues. the
- .5.1.0.J,. tue
- enztn nu . Le tsaen .ts i:s .
. . . _ . . . . .. ..a
. ,, . .a . . .a e t .. _.. .. .. .., . .. ""*- stress snail us governe: by :ne ;:rovisior.s cf Ara.en..ix C.
-L. 's _
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- eta. -D.4*C,. - - - ~ = -
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r -
=___-_- -_ - -
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5 3
- C:~me .=w c- A!SC Sne:.4=rson
- 1. ~, 1. :.1. En tyee =::=cers are termon.Cly ve- mf.* Tne entical For=ttla (1.5-D is a c=nvenien Aterzi torsionai Duch'. Leg. for ille Compress 2cc '2=ge eDCe of Doth latera.1 benc.:.:g res1.Et*
=eXursa stress sue 'u af m "me be:= .cadec : tne ui=e cf its n sor r.25 s., ts to and about ,
Due to the ciference between da;
~
ns =mcr :au.s .= ve ontame-: us:.=g Formu;s 1. ~, . . .otn an equivalent guder, it is destrable to base the late necce ness r:w. ey tne er.:ress;o=
torsion cf the f a=ge. He=ce. use of
e=bers. Its agree =ent mts
strength ofi=termittently braced f
[' ,
15.11 5 ,
} y,F, neotts sections havt=g subst.antialre:
\r . , , , ,
in the cas.e of doubly.sv==etncal s wnere i :s tne istance between oomts of lateral support and S,.1,-asd J For so=e sections havt=g a cc cre res-+cm et tne .:jor ans scenen = oculus. =: or ar =c=est of than the tensson fange area. For=u
- nnn: == t:e . ts;=al e:: start of the beam crcsi sect:c:. It enn be its use is li=ited to secuens whos.
snow: :nat wne: d ,100 an: 4 o < '!.500 F,. the anowaoie ec=oression great as the tensson d a=ge. In pl.
i:.:ge stress menested my tre above e:;uauen will appron=. ate 0.607,. higber d/A fratio sham rolled W sha 3eycod tnis =.;t dedection rather than stress is liicely to be de des 2gn the co~nservative side. For such r en ~-* no n- For=ula (1.5-6a) and. at ::=es by :
1.~,.1.4.5 = C 1.~,.1.4.6 ~~ne sliewso;e bencin: stress for all other
= ate of buckling stremgth. %"hile strength somewnat because taey ig
'er rs; ar.empers ;s men as 0.60F,. provme:. ;ne sc=pression dange is
, the prose, dis @ty for such ses
.rsce: .: terr :t ::.;;uve;v a.ese mte-v:Is .I c., f 76.0 v F] *I O ***^****** O'#NC # e. is h.
Me=ce s cent secut ine:. =ajor aus sec navmg :n ns of sy=.::etry It sh uld be noted that r or=u.
= Ine :::ne cf lo:cm: =sv ue scepuately cracec '. ster:lly at ;; rester : ster- expressi ns it rep aces. is wntten fc
. 23 ;f tne man =um nenc:nt stress.ss reuuvec su:5aently to prevent pre- ti n is not prowped for uus :orm
=sture out k.;rg of tne ec=:ression $snge. M :nemancal exeressions wnen actual conctnons ofloac appl: .
..~ ,: int :in ewt esn=ste of the Ludline stre. - ein m sur- ---.-. <-.--
are censidered, any unconservauve D ; s- w.:nr-r-n nimu: .ne:. .e etum. r .m 3t
- ne c o essiun '--_ e b,3ED*'F =' ENC 2h " P-
- g ._
- .t tes:e. n w en .
tm ,- w v ' <
- v u e. ,. . a -
eending due to lateralloscag acuo
. . . . . . . . . - - .m. ; . . . . e _. . . .o r e. ... . n r . . % t u .. . -... ts uruenuen t upon
.. Suc: reemoe s ususily can be croot
.e.... .- . .
ns re stn;n: r**tr: mi ut *w.nts ol Late :1supDCrt .
strets wCen inst strefs is proOuced
..'...'.w...
... . . . . . ., .. ~. . , . . . . . . .n.... .. ges.. , c ... , ri,. . . . ore .. . .sn . e ...,2ne....,n g ,s os e. .m". %_nere the ; ..s:;ure moce c u..: .an L- nsvant a 11reer ce= ressio~ tnsn ter s n t, 1.;. . prw wes r -~~ u;r . 3- t: : , a. .
_ ... .t. . . " . . 2.e the D'.i=i5SiCI' Dend m g Stie55 Ccn
. .. . , e ns e . . . . u. .. .
or 1.0- , -
.s = : ~ .*
- PP. . ,. :P e ., h. .. et;.tlen C ; "*. ? Pl*".'C nesDOC. t oculas Through the intreouction of 1:
c.. r ." . Je u:se': on tne :ssu=r$ tion inst only the hending stress is per-- "'bie wee = there ts =c dren :: :nc . -aresnon snge ,
w;11 crevent the ateral c,isplacement of except wnere. in the esse of co=0m
- r. .! e
- en en! 7 s een brace; no:nts. .ne new r or .u::s 1.s. ..ca and adjustment is provided by the facte
. .c a a:er :v t,e e true. r ormu;a ,. . m two ways: Formulas ii.5.c.a s and 1.o. ...o c
. ,E nere.u ::e cariser provisions reputre: no stress reduction when Venant and warping torsion by su'r -
.*w. . e tn:n 40 rer:rcless of yiem stress value :nd then a equivalent radius of gyration. r.,,,,.
recue: ;o tne value co'.s:ned :.rc= ine psrsbolic expression. the priate expression giving the enticsl e new fer .=:4. by meressmg r.. at 1 - 0 :.rorn 0.60F, to 27, 3. ,ange of a bea=t mth that of an ax crev ne= ' connnuous s:ress reistionsha wi:h the u=br: seed length ,
I
! u nen :" - reauced fre= .ne =:nmu- pe=:ssible value of 0.60F,. l
.. .A. .nerc. . me e:raer s:npe r ormula 4 3r:clied even in the range ]
cf e:;n .;ucM:ng stress .on the as.n:=ption that For-nula (5) m u. --n :ne replacement of F r=.:ls 4 i is liberali:ed in l
. . . - -
- Column Reesien Couned Guioe
. e scc:t:en vi an r a.- "cr+ excression, si:ce this Mernoe s. Seron: am:en a;.
S.
k n s entre- " .I bui.. I:.
- ID'8" 48 *. ' M. - ' - *. . . .
~~
5' '
..- .n . n. .. .; ..c e . ., ".m :: n . e.: for W:.1 Cocapremon
- Ibic.. I;. . " *.
- g.2
Scrue:ura.! .Swel ser Bui;.Lrms . s . ;"
v For=ula (1.5@ is a convenient approximation which assumes the pres-
] .: m
- .: .: . res .. :.:nge T . ::ical .
- D' , of 'cota hte-s! Dending restance and 5t. Vesnt torsional resis .nce.
. ...... : ;;; . E . -; : . . .: ..m.t ' as to the diference between dange and web yield strengd of a hyorid
. 5. . .... r.n ecu:vsient order. it is desirable to base the lateral buckling resistanes solely on warping
- torsion of the dange. Hence, use of Formula (1.5-7)is not patted for such me=ners. Its agree =ent with more exact expressions for tse buckling
. . .i . strength of inter =it*.antly braced derural members' is closest for homoge-
'y* -;.'
neous sections having substantial resistance to St. Venant torsion. identidable in the case of doubly symmetrical sections by a relatively low d/A., ratso.
..erci nmoort an.2 c.. .', and J For some section.s having a compression dange ares distinctly smaller r..u.. =:ni - .n. =nment of A- the tension fange area. For=ula (1.5m may be unconservative; bence.
.: u: . creu .e. :: ,c. h enn be its use is limited to sections whose compression aange area is at least as .
F ine .diowau.e compresnon gre.at as the tension Sange. In plate girders, which usually have a much
. .r , m i' . arrr :n:=:te 0.60F,. higber d/A., ratio than rolled W shapes. Formula (1.5 7) may e:r grossly on r;..s ., ;R ei, :.. .e 2. ces:gn tne conservative side. For such members the larger stress per=itted by Formula (1.5-6a) and. at times by For=ula #1.5-6b). afords the better esti-mate of' buckling stra=ph. % nile these latter ion
.e oen::nc s r+n a- .r.n other -
streng;h somewhat oecause they ignore enantthe St. N.
torsional =ulas np:..stv -
of underest2 oe: :ne comareuten n.:nce is the proc.le. this n. . .. peaty for such sections is relatively small a.nd the marps
' ;- ~
\ *~
of overconservatism. therefore, is likewise Emil-
- n. nn a.-c sn :.n:. : sy= metry It should be noted that Formula il.5-7),like the more precise. compiez ersced latersil;. n: crester inter. expressions is replaces. is wnsten for the case of elastic buckling. A transi -
- cco su
- hetentzy to nrevent pre-tion is ut Wdd for this fod b h 'mh m w hse.
- ge. 51.ane=: tic:1 exnress:ons when actual conditions ofload application and variation in bending =oment
+tre - n er sucn memoe-s. wnica are considerea. any unconservative error without it must be small.
.u ". ....e.. .. .o.... .. .. .. ,.. . .,3 n;*. s.t. Singty.symmetscal, built.up. I.dape mk.s. -
sud as some e an
,neg. a: :ne:r com::re*s:en nsnge prders. often have an increased compression image ares in order to resist I c wrs.u:: . are in.> comunex :or bending due to lateralloading action in conjunction with the verticalloads.
l .e. ateur:cy :s ceueno nt upon Suct members usually can be proportioned for the full permissible bending
(%.no more 3,g
-c - s & - * =- <" ~" c r*'
inan e .c:neer ng stress when that stress is produced by the combined vertical and hon: ental loadi g. %nere de failure node of a si gly-spede.d I.s.au m su naving a Isiger ec=;ression than tension dange would be bv lateral buckling.
.-L,... . -ravme*
Ant. . 3 de a saible budh smss cu be duind h W'For=ula 1.5-6a i L(*;*- _
~
or .1.5m.
.. tne ni,e. :nostion r rmu1** Through the introduction of the modi 5er" C., some libera"-' tion in
, ---i<n "" %- n. .enesing stress is per nissible wnen taere is moment gradient over the unbraced length en. .av }arr.:s .iire ;r:aent f except woere, in the case of combined bending and azisi compression, this ne new r ormus.:. I . >.n.. anu adNsts ct is prmided by de fac or C. in Fala '1.6-in
'A t'" 3F8 Formulas ti.5-6ai and il.5-6bi may be reined to include both St.
. ;;r c-: nu nrens renue::on wnen j Ven:nt and warping torsion by suostituting a derived value for r.. This l
- . nress .: a. , .: :: tnen s eouirsient rsdius of gy stion. r...,,, can be cotained by equsting the acpro-
"
- 9. ..:.* e- a pu;.)L. ine 1
~la enete exnression giving the critical eisstic bending stress for :he comoression 1..-.,. ..,,. ,
l
.. . a. :isn;e of a besmt w7th dat of an sxially loaded column.-
o u - x - , u . - . . -i ." t v :
r -- e .:m . "*F
. : ,- - ! .. **.:ncT
. r
. .-
- C..u=n Rese:ren Counen A.ce to Desirn C .te .a :c.: Me:a: Ccrn resron i
' ' * ** * .\1er= ers. eroe IJuao:i. I 4.5. .
" Ibui Ig. 4.11 losc.. I
- m .. ::,r.2 for W:.u . . 'omatession . Ibic., .3c.
os. s4.Sc :. #4.:m. 4 m or e t..*.*:.
m
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n "'
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- - - ~
k%"f p= " W}.
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.1
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- es -
'W*--9 :* 9FQ" Q* 'd o?. =" W :- c' ~ ' 3-, e __- t
'ww .
... .rs .a ..
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A W.s,,,
- c%... -. - ~ -T+3. ;:..,.
., ~'.'.~~:7 v
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-QC ,
M *tJ4JJv
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~
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-. --macer ,
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l, uo%y ; .nm (_h }FS X ( y"o..
} a n#4 g
.- .c .* y-m!zt.:~m, ==. ~ ~ F .-. . . . = _ ~,
.-- -;-- c. ..
.e- - -. r.,?h.- -m-. .
_.,....s
. - - - wa-a - m h =- m ,u.awr--
m.wx..4=in.-&M: - s? ___*._'*~. % ~~~ ,wp 7 W, q"wp' MM d< . . "*-
"{ s_ -Nm 5. I
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wwg.o.~-Z ~ -
d< .,Q.{ -8
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"g .
__- g -% Ws-d6 d* -"iy;;e - _ . -
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,.. = 4 m-. M- - > -= *T:nbiEsikj a- -
wn?-- --_- m -
h '* 'm--
! @ i M' sM dW7W N,NEWMWA$_6 NyoC8 -
~
_ _.- sw w. _ _ qw
_F , - - ,- _ _ _ - - _. _, c m 5 - 1:3 . Commeu.v. ~ on AISO Soan.sm. con
-rer :ne esse a:. a cous:v-ey=.=ce::2 2.sc. ape :>ea=.
1.3.2.1 5 hear
, n,.....
C.O: ecuens w::ch tr'et ,.cac
,. .. a ' .
- O C' s w
23, l, are Categor :ed as "fr:Cu03.:ype" 01 no. o: su.5c:e :.!v. :.ign cH : = g fcree -
.), :.s ::s La)C
- ' e *? . , 3
- r.e =.= 0; a=s == = e : O! =erna o; : e =e=str, Yhe istier depe:d uoon eCOtaC* cf *.*
- L5** COO 200" 21 20 ho;es to tra:sfer tne los: i c= one co:
3 ~. , ',.,
.The a=ou=t of cla=p=g force ce J .,
cooling and oy A007 och.s a unpredi=
a a vent cc=olete suppare at the een:.u
- cTE eC*:ons add cetrecuC:s = ace -C.:
1.3.1.3 Searin g be ar= g.t ype. T e nigh cia.=p=; feres stre:gt.h t'olts is 5u$c:en: to prever 1.3.1.3.1 As used ::.mgnout the Spec = at:c: the te. "=il:ec sur. equal nu =ber of these bolts a e suesn
- a ce.....=;...ec.. e r .. -
g.. are =:e=dec to =cace surfaces wr.:ch s. ave .
wou;d be required to : s:s=:: a pven g+-: aeru. rate:y sawe: er --u ec .o a t.ae paa:e ey any su::ao;e =ca 2-r2 vets and A490 bolts for A5CO Grace 5 e re:::==e::e: Dea .:r nress :: :=s :s not me same :s fer :ve:.s. The Toe e5c:enev of tnreacec faner aer va ue. r :e.tenes cf :ne ne;: suess :f me ru: con r - g ::e pm connecuens is recueec s ne: :ne :..r n: e. : ev=es a safer;ar: aga=n =sta:Ory f e :.ste oeyonc ::e noie. ' betwee: :ne c==nectec carts. I::ne c -
.. . - - ... . ~, . - . e . .a y. a.r . .. g.. . . . a , v e. .. .noie.
able shear nress values are pve:: ont shear p12:e and one where ' :s Co .
1.~.2 Rivets. Solts. and Threaded Parts tha feature in :he case of A007 bo.:s.
te=d =to tne sner ;.a:e a:: :ne 2 t
...e..a.., .e .. ston gross area. a recueec ace r==ry.
As = earher ec:: ions. :e. ;ss.b:e stresses fer r. vets are ; ven i= ter ns
- h::cie to me n===a; :ress-sect:nal area cf tne r: vet oeiore c-vmg. I3** - B e arm <
- 7er ; ener eensemence = tne ;rcoernon=g of h:;n stren; n oolted con. Bearing vaiues re :rev::ie:'.. nct a ren.::s. w--e :.e nresses i:r =e ocks are pve== :e =s sppi:csb;e to their it neecs =o suen tretecuan. bu: ss :n
- -.na. vocy ares. i.e.. :ne een of me untnreacec sn:ng. However, for corneuted in accerc:nce m:r. 5ect.1.1-
.C'." sc.:s v=:n re :v: a::e = s::es u; to 4 :n. :n d::=eteri and inresded 333,.n:;ed min . vets or mm no.ti. n
- ci ner
- r.an n::n s r**0* *. =01:s. =* zhC'Y::4e tern ' 9t*ess :s 3pDU: oie ne =retence Or sbient* f f Inre ~8 :-
Tc....re. :nte =edia:e
. r n s : r: e:n : . . :,:.4 ..s -
.. : ~ a
.umu
- nas e snc.s n u=: ce :enn.e s:
"n : en ces :n: re: :: :ne roc: of ine =re t= vnen =uhi; bed by
. ne:n =:a. ;r:w :es ci :ne un:reace: ==:er.r... nas oce: found to
'. a: red wren tne oe:r:: Oressure a: .
fanecer a as =u:n as W t:=es :ne t e
- t aosezy : rec:= me :e:s:.'e nre ;- . cf irger cre:er : reaced ; arts, of the part. In th:s =resuration the s.= as =.i;nt ce used icr an nct ooits er upset rocs. setording to the usual ec=vennen. :s -
In receptue: cf :ne rotecuen aga=n noten effen in the threading. eter and th:ck=ess of :ne connected :
r.arured by :ne recu: rec =:ual dgnien=g of high nrenrih bolts, the Re- tween sineie. shear bear ng and encios-ser:n Cou:cd c= Raete: anc Scitec Structural Jc::ts nss ree:=rmended the recommenced werc:s stress is :.
- . reauvely h:gner wer,u:g stress i te=s:c: ict =;n stren;th bolta. snear bearmg s=d appro==ately ec -
A=y act:ne:a; issierer -e:s:c: rend:=g fr.= ;mn; sc:fon due to dis- stress reco==e:ced for ceter--H; re
- n::: ci :ne ::::eenon ce:si;s snould be added to :ne stress calculated '
urm.y '- ' e 2:o.;ec tens:c: := preocr.:c:=g faste..ers for an applied .
l nr e :orce, un:g me socc::ec scrang s: esses Deandi:g upon the 1.3.3 Welds i Y ..sure n= ness cf me fa.s:e=e s and the cenneet:en =ne-i:J. his p y=g .ss i: the past, tne au..c+ac,e wori 2:n:n =ay oc :erar :!e er :: =sv be a suonsntial he total te=sion
. n e iane:e-1. . .
penetration weics re the .e.e as :. c .
vices the rneenansc= proper.:es ::. :..+
a u n s .- ,. ....r.nec:e: Pa te .,.. c . ,. u . ,. . rer.scr:ws .. . __ er exceec tnose ci me wesien : sca
.4 ..mara c.~ Es.a: C:n e::;.St .:: .:.: 7 : .s=:.or s.
ect t :..ce1 r: .
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la c.".y ha. ge: .: aviavad :as 5f.5-15R wh ch . ras qualifiac by the h =e ~
cific.a .=d shecid, theraf era, he acceptabla aga'.ss: :he above :-1:er:.a.
_ _ . . . . ._.s<. . . ..a., a_, a.cye :.o
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I
O . UNITED STATES Of AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL 80ARD In the Matter of ) Docket Nos. 50-275
) 50-323 PACIFIC GAS AND ELECTRIC COMPANY )
) Design Quality Assurance (Diablo Canyon Nuclear Power )
Plant Units 1 and 2) )
)
AFFIDAVTT OF LEONARDO B. MANGOBA STATE OF CALIFORNIA )
) ss.
CITY AND COUNTY OF SAN )
- O rRANCISCO
)
I, Leonardo B. Mangoba, being duly sworn, depose and say:
- 1. I am employed by Bechtel Power Corporation (Bechtel) as Lead Pipe Support Engineer of the Onsite Project Engineering Group (OPEG) located at the Diablo Canyon nuclear power project near Avila Beach, California. In this capacity, I have had primary responsibility for managing the OPEG pipe support group since October 1982.
- 2. The OPEG pipe support group encompasses different engineering activities relating to the design verification and construction of small bore pipe support systems in Ihits 1 and 2 of the project. Charles Stokes was a
" job shopper" who worked in the OPEG pipe support group f rom November 8, 1982, to October 14, 1983. As a job shopper, Mr. Stokes was employed by lO -
.:td ll-vm "2 l M n ! C-t
"O .
neither Bechtel nor Pacific Gas and Electric Company (PGandE). Rather, he was employed by an engineering employment agency which contracted with
- PGands to provide Mr. Stokes' services on a temporary basis. Job shoppers like Mr. Stokes, whose employment agencies had contracts with PGandE, are referred to as "PGandE job shoppers' as distinguished from "Bechtel job shoppers," whose e@loyment agencies contracted with Bechtel.
l l 3. Job shoppers have been used at Diablo Canyon to provide engineering services on a temporary basis. Thus, Mr. Stokes and other job shoppers were retained in the fall and winter of 1982, and the spring of 1983, to meet a short-term need for more pipe support engineers than could be satisfied by the engineering staffs of PGandE and Bechtel. Because of the l
temporary nature of their employment, job shoppers generally conumand l{
- premium compensation.
- 4. A plan was instituted by management to reduce the number of job shoppers in OPEG (see OPEG History affidavit, Exhibits 3 and 4). Pursuant to the plan, several job shoppers were released from the project in early July 1983. Mr. Stokes was not considered for this reduction in force because
! he was at that time assigned to a particular group which was performing a high priority function.
- 5. In mid-July 1983, shortly after the aforementioned reduction in force, the cost of the job shopper staff was further reduced when 8echtel decided to terminate the services of the Bechtel job shoppers unless they elected to become " casual" employees of Bechtci. A Bechtel " casual' is an engineer l employed directly by Bechtel on a temporary basis and generally earns less l
l compensation than a Bechtel job shopper. Most of the 8echtel job shoppers O _ _. _ _ _ - - - _ - .. . . . _ _ _ _ _ - . _ _ - _ _ _ _ - . . - _ _ . - . -
O who were offered the opportunity to convert to employment as Bechtel casuals accepted the offer. Those who declined to accept casuai status were replaced by new Bechtel casual employees. Three of the Bechtel job shoppers were not offered an opportunity to convert to casual employment because I felt that the quality of their work was below what I could expect'to find in new Bechtel casual engineers who would replace them. As a PGandE job s' hopper, Mr. Stokes was not affected by Bechtel management's decision to convert or replace the 8echtel job shoppers.
- 6. Pursuant to the plan to further reduce the pipe support engineering staff in general increments,' I established estimated dates when each of the remaining engineers in the pipe support group would be released from the project. I established these individual release dates in July 1983, j ,Q shortly after the aforementioned conversion of the Bechtel job shoppers. ~
i These release dates were spread over a time frame beginning September 30, i
1983, and continuing through the fall of 1983. Generally speaking, the i
release date which I assigned to an engineer was based on three considerations: the quality of his work relative to other engineers, the expected need for his services to perform particular assignments or tasks, and the employee's status as a job shopper as opposed to a direct employee i
of PGandE or Bechtel.
- 7. idhen I assigned the release dates. I scheduled Mr. Stokes to be released on September 30, 1983. I did so because his supervisor had previously ranked him in the bottom third of the engineers in Ms group in terms of their relative performance. Three other engineers, whom I will refer to i
as Engineers A, 8, and C. had also been ranked in this bottom group.
i'O .
1
O Engineers A and 8 were 8echtel job shoppers who were terminated in July 1983 without being offered an opportunity to become Bechtel casuals l
because I felt their job performance was relatively inferior to the degree that they could be replaced by new Sechtel casual engineers. Engineer C, a P6andE job shopper who, like Mr. Stokes, was not affected by the conversion of the Bechtel job shoppers, was also scheduled to be released on September 30, 1983. The low ranking which Mr. Stokes and Engineer C l received from their supervisor was consistent with my general impression of their relative standing in the pipe support group. This low ranking, which appeared to equate Mr. Stokes and Engineer C with Engineers A and B, made them the primary candidates to be released on September 30, 1983 --
the date for the next scheduled reduction in force.
- 8. Approximately two weeks before September 30, 1983 I gave P6andE notice of
{ my intention to lay off Mr. Stokes, Engineer C, and three other P6andE job shoppers on September 30, 1983. I did not receive P6andE's approval until about October 13, 1983, when I was authorized to lay off no more than three of these five job shoppers. The three whom I selected were Mr.
l Stokes. Engineer C, and one of the three other P6andE job shoppers.
- 9. On October 14, 1983 I informed Mr. Stokes and the other two job shoppers that their services were being terminated that day.
- 10. Just before the initial reduction in force in early July 1983, there were over 75 job shoppers in the entire pipe support group. As of October 14, 1983, when Mr. Stokes was released, the number of job shoppers had been reduced to less than 15.
'O -. . -. - - . .
. O 11. I did not become aware of Mr. Stokes' three discrepancy reports until I saw them in late August 1983 as I was not at the jobsite when they were submitted in mid-August. These discrepancy reports and the claims made in the reports had no effect on my decision concerning when Mr. Stokes would be released.
- 12. The three typewritten discrepancy reports which are attached as exhibits to Mr. Stokes' affidavit of November 17, 1983, are formal versions of the discrepancy reports which he originally submitted to management in August 1983 (see OPEG History affidavit, Exhibit 5). The issues raised in the handwritten discrepancy reports had been fully investigated by the project by September 29,1983 (see OPEG History af fidavit, Exhibits 6, 7, 8, and 9). At the request of my supericr. Myron Leppke, I presented the discrepancy reports to Mr. Stokes in typewritten form hAlch Mr. Stokes
(( )
{ then signed after making a few minor corrections or changes. I never
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infa wd Mr. Stokes that his services were needed to review the resolution of the issues raised by his dist.repancy reports.
Dated: March 4, 1984 Y$ O e
/ Leonardo B. Mangoba Subscribed and sworn to before me this 4th day of March 1984.
~
2 WeW6y J. Lemaster, SEAL Notary Public in and for the
( City and County of San Francisco, ,
State of California.
My comission expire!, .
April 14, 1986 __
f'2- 22E },~hMASTER t
NOTARY PUBLIC-CAllFORNIA CITY AND COUNTY OF SAN rpANctsCO Eug;= 2M l
i
UNITED STATES OF AMERICA .
NUCLEAR REGULATORY C01911SSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of ) Docket Nos. 50-275
) 50-323 PACIFIC EAS AND ELECTRIC )
COMPANY ) (i.j,ign Quality Assurance
)
(Diablo Canyon Nuclear Powey' )
Plant. Units 1 and 2 ,, )
f 6[F.LDAVIT OF AZRIEL SHUSTERMAN STATE OF CALIFORNIA )
) ss.
. 'O citv ano couNrv or san FRANCISCO
)
i I, Azriel Shusterman, being duly sworn, depose and say:
- 1. I an employed by Bechtel Power Corporation (Bechtel) as a Pipe Support Engineer and was assigned as a Squad Leader of the Onsite Project Engineering Group (OPEG) located at the Diablo Canyon nuclear power project near Avila Beach, California. In this capacity, I supervised i
the design activities of approximately eight engineers in the OPEG pipe support group from October 1982 through March 1983.
-(:) ____ -
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! l 1
O Mr. Stokes has alleged that I intentionally authortred, approved, and 2.
" covered up" a recalculation of a hanger which had been previously disqualified. There is the clear implication that I may have done this on other occasions. (January 25 Tr., pp. '195-197)
- 3. Contrary to the assertion of Mr. Stokes, the verification of existing pipe support 100-132 was not reassigned to " cover up" a prior analysis, but solely to qualify that support, if possible, prior to requiring modification. I have never assigned calculations or recticulations for j any purpose other than' legitimate profe:isional reasons.
- 4. The analysis of small bore support 100-132 was originally assigned to 1 Mr. G. Katcher in about January 138 3,and his analysis was checked by Mr. D. N. Patel. The results of this analysis showed the support i :O basepiate to be overstressed.
- 5. In discussions between myself. Mr. Katcher and Mr. Patel, it was agreed that the completed analysis was done assuming the support baseplate to be rigid. I proposed that a more sophisticated reanalysis should be performed using the more realistic assumption that the support baseplate was flexible. This analysis method is an &ccepted industry practice and did not result in the violation of any licensing criteria. Mr. Patel agreed that this approach would be acceptable, but Mr. Katcher did not.
l I did not consider it necessary in this case to require modification of the support untti other entirely acceptable, analytical methods had been tried for demonstrating support qualification. Furthermore, I did not O
\
i o consider it prudent to be limited by Mr. Katcher's apparent lack of familiarity with another acceptable analytical method. Consequently, I assigned the reanalysis to another engineer, Mr. G. Shah, who was familiar with the proposed approach. Mr. Gautam was assigned to check the reanalysis. I then considered, and now consider, that reassignment to be entirely proper and clearly within the normally accepted prerogatives of a supervisor.
- 6. This analysis reassignment coincidentally occurred at the time of the i
division of the pipe support group into separate Unit I and Unit 2 teams located in separate trailers. At that time, Mr. Katcher was assigned to the Unit 2 team.
- 7. The reanalysis completed by Mr. Shah and checked by Mr. Gautam showed 1tia iirictia rarteata
<O ta >=aa ri t 6 a iiri a-However, the calculation contained an error which went undetected in the checking process. Had the error not been made, the support would not have qualified in the reanalysis using the method that was actually employed. The reanalysis did not consider baseplat'e flexibility as I had instructed, and was completed using a sure simplif ted model than was used in the original analysis by Mr. Katcher.
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O 8. This error was discovered during an audit of this calculation in j December, 1983. At that time the calculation was assigned to another pair of engineers, Mr. R. Amin and Mr. Singh, to reanalyze and check using an apprcpriately detailed model and considering baseplate flexibility. This calculation demonstrates acceptable qualification of the support and meets licensing criteria in all respects.
Dated: March 4, 1984
/-
ksw, AZRIEL SHUSTERMAN Subscribed and sworn to
!O before me this 4th day of March, 1984.
l
' " SEAL Raficy J. Lemaster, Notary Public in and for the City and County of San Francisco, State of California.
My commission expires April 14, 1986.
pe::.=x.:xxxx:cxecec::m: .
y NANCY J. LEf/.ASTERg N0*4RY FUI'.C-CA.tFORNIA
$ (h.
e gh,' city see couraY OF sa*: r:masco 4
M O*# _
'2 My comrnissic- Et?'ts Apnt 14,1986
'NX:ZO::C:NX:GXXXF:C:M*(MD:*CCOM O
UNITED STATES OF AMERICA O NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD l
)
In the Matter of ) Docket Nos. 50-275
) 50-323 PACIFIC GAS AND ELECTRIC )
COMPANY ) Design Quality Assurance
)
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) D J
AFFIDAVIT OF RICHARD D. ETZLER I
STATE OF CALIFORNIA )
'O ) ss.
CITY ANC COUNTY OF SAN )
FRANCISCO )
The above beir.g duly sworn, deposes and says:
I, Richard D. Etzler, an Field Construction Manager for Diablo Canyon construction activities at the jobsite. I have held this position since Septenber 1978. Prior to my duties as Project Superintendent, I was Resident Mechanical Engineer. Fron 1971 to 1977 I was a Field Engineer and group leader who reported to the Resident Mechanical Engineer.
I have carefully read the affidavits of Charles Stokes dated Novenber 17, i
1983, and February 8,1984.
l 0 ,
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It is alleged that:
QC inspectors were not consistently qualified to the AWS code, that none of them had been issued copies of that code, and that the inspectors could not read weld symbols on design drawings. (Stokes. 2/8/84, p. 8, para. 19) l
- 1. Contrary to the implication of Mr. Stokes, Pullman QC inspectors are all qualified and certified in welding procedures and symbols pursuant to Pullman Power Products Quality Control Procedures which are consistent with ANSI Standard N45.2.6. It has always been the case that, as part of this qualification and certification program, all quality control welding inspectors
> have received training in AWS symbology. Such train $ng includes mandatory reading assignments and testing in weld symbols.
- 2. The AWS symbols " code" referred to by Mr. Stokes ( AWS 2.4, Symbols for Welding and Non-Destructive Testing) is not a code O but a standard. This document is not required to be issued to Pullman QC inspectors by the NRC, PGandE specifications, or industry code requirements, and is not the sole reference on AWS welding symbols available at the Diablo Canyon site. Various -
handbooks and other standards contain such information.
Q: __ _ _ _ . _ . ._ _.__ . _ _ . ._ __
l It is alleged that:
Welders did not have Pu11 nan Power Products procedure ESD-223 or welding procedures. (Stokes, 2/8/84, p. 8, para. 17)
- 4. ESD-223 is a Pullman Power Products procedure based upon PGandE Specification 8711, design documents, codes, standards, and manufacturing reconnendations. This procedure provides detailed criteria for installation of pipe supports. Controlled copies of ESD-223 are issued to field QC and engineering personnel and are available to the craftsnen through forenen, general foremen, superintendents, field engineers, and QC 1
inspectors. Current controlled distribution of ESD-223 is over 300 copies.
- 5. For the nost part, ESD-223 is not a document utilized or needed by craftsnen, such as welders. Typically, for a pipe support installation, a
! Pullman Power Products field engineer reviews the design documents and, utilizing ESD-223, prepares a process sheet. The process sheet specifies the weld procedure to be used and notes specific requirements such as preheat, interpass temperature, and electrode type. The process sheet and 4
the design documents which are given to the welder provide all the necessary infornation to do the job correctly.
- ensures that the welder is familiar with the paraneters and essential i
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i vertables reevired by that w1 ding procedure. Once the wider has been qualified and certified to wld to a specific procedure, it is so noted on the Walder Perfoneance Qualification Card, which is in each wolder's l l
l possession. General m1 ding inforsuation is also included on the back of this card. The Weekly Wolder's Qualification status, a document summartzing the qualifications of each wlder, gives additional information about each w1 ding procedure with respect to base fasterial.
root type, weld process, and filler festal. Each welder has access to this document through his foreman.
- 7. Before a welder my work on a specific wlding assignment and prior to obtaining wld rods, the Quality Assursace representative verifies his qualification to do the wld precedure specified on the welding red requisition. Only after this verification is filler material issued to the craftsen. In addition. all wlds are inspected by Quality Control to ensure that they meet criteria specified in ESO-223.
Dated: March 6,1994 state of California county of san Luis ontspo 88-D RICHARD D. ETILER
-'f h'h-g Subscribed and swom to before me this 6th day of March 1984.
OFFICIAL SEAL l IDA DUTitA
%*w"*J*.*s, .cn butna.
1 lO
O UNITED ST m S Or Aia ICa NUCLEAR REGULATORY C0094ISS10N BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of ) Docket Nos. 50-275
) 50-323 PACIFIC GAS AND ELECTRIC CopFANY )
) (Design Quality Assurance)
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) )
g )
AFFIDAVIT OF HOWARD 8. FRIEND STATE OF CALIFORNIA )
) ss.
CITY AND COUNTY OF SAN )
FRAdCISCO )
The above being duly sworn, deposes and says:
- 1. I, Howard B. Friend, am Project Completion Manager for the Diablo Canyon Project. I am thoroughly familiar with the contract between Bechtel Power Corporation and Pacific Gas and Electric Company covering the work to be perfomed by Bechtel to complete the Diablo Canyon Plant.
- 2. During the course of his meeting with members of the NRC Staff on January 25, 1984 Hr. Charles Stokes alleged that the contract Bechtel has with PGandE is a lump sum contract (Tr. at 22). This is not true.
I The PG4E/Bechtel contract is a cost plus contract. That is, direct costs and expenses incurred by Bechtel in the perfomance of services are reimbursed at cost. The indirect costs incurred by Bechtel are recovered by //
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Q adding a pert:entage to the direct payroll costs. This is known as "an allowance for indirect costs." There is also provision for a fee.
Dated: March 5,1984 HOWARD 8. FRIEND /
Subscribed and sworn to /
before ne this 5th day of March,1984.
Wncy J. Limaster,v SEAL Notary Public in and for the City and County of San Francisco, State of California.
O % comission expires April 14,1986.
NA C T 5 h',\ NOTARY PUBLIC-CALIFORNIA E 87;
. .; , i't. cit- AND COUNTY OF p i'- %- SAN ritANCISCO tj My Commission Expires Aprd 14,1986 g
- vec F - . . W4 O
O UNITED STATES OF AERICA NUCLEAR REGULATORY CON 41SSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD i
)
In the Matter of ) Docket Nos. 50-275
) 50-323 PACIFIC GAS AND ELECTRIC )
COMPANY ) Design Quality Assurance
)
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2 l J
AFFIDAVIT OF DANIEL J. CURTIS O )
STATE OF CALIFORNIA
) ss.
CITY AND COUNTY OF SAN )
FRANCISCO )
I, Daniel J. Curtis, being duly sworn, depose and say:
- 1. I am employed by Bechtel Power Conoration (Bechtel) and am currently assigned as Piping Group Supervisor of the Onsite Project Engineering Group (OPEG) located at the Diablo Canyon nuclear power project near Avila Beach, California. From the Spring of 1982 until November,1983, I was assigned as the large bore Pipe Support Group Leader of the Diablo Canyon Project engineering team in San Francisco, California. ,
0-C'l1&W O;
O Mr. Stokes alleges that during the period of his employment at OPEG I was 2.
j sent to tKe Diablo Canyon jobsite for the purpose of explaining the basis l l
and use of PGandE drawing 049243 to the small bore pipe support group.
l
- 3. At no time was I ever requested to, nor did I ever go to, the Diablo 049243 or any other docweent of Cartyon jobsite to explain PGandE's drawing procedure to the small bore pipe support group.
- 4. I did, however, ask Mr. Charles Magee to make a trip to the jobsite in 049243. Mr. Magee was February 1983, to explain recent changes to Drawing 049243 and was, at that an agency engineer knowledgeable about Drawing time, working in my large bore pipe support group. He was not sent to explain the conservatisms in Drawing 049243 or to defcad the basis for its development as alleged by Mr. Stokes. While Mr. Magee was generally knowledgeable in the application of 049243, he was not authorized to Q provide interpretations of small bore piping design criteria or to establish small bore support design criteria or policy.
Dated March 5,1984.
DANIEL J URTIS Subscribed and sworn to before me this 5th day of March 1984.
c
- f 'EL Yancy J. Lemaster, Notary Public in and for the City and County of San Francisco, State of California.
My cornission expires
^Prii 14. i'85 O. pccc::ccm7Jo:xx/.x:::conoccoccq NANCY J. LEMASTER y i:
l5 , NOT ARY MI.?'.tC.CAUTORNIA p,
- f. g.p)-
CITY AND '/J'J:3Tf 0F i-M Q SAN NNCISCO ! x My Commission Expires Aont 14,1986 6:monocoocco:x:~m:-:scex="
- c. O UNITED STATES OF AERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD
)
In the Matter of ) Docket Nos. 50-275
) 50-323 PACIFIC GAS AND ELECTRIC )
COW ANY ) (Design Quality Assurance)
)
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) )
)
AFFIDAVIT OF J. D. SHIFFER, R. PATTERSON, J. M. GISCLON, K. C. DOSS, J. B. HOCH,- R. C. THORNBERRY, R. D. ETZLER, R. K. RHODES, AND E. M. BURNS STATE OF CALIFORNIA )
) ss.
CITY AND COUNTY OF SAN )
FRANCISCO )
The above being duly sworn, depose and say:
I, Janes D. Shiffer, am Manager of Nuclear Plant Operations in the Nuclear Power Generation Department of Pacific Gas and Electric Company.
I, Robert C. Thornberry, am the Plant Manager for Diablo Canyon Power l Plant.
I, Robert Patterson, an Assistant Plant Manager for Diablo Canyon Power Plant.
I I, John M. Giscion, an Assistant Plant Manager for Diablo Canyon Power Plant.
ein= y w] _ w .. ,
, . . _ , , , , , . , ~ , , . ,
!O I, xen C. Doss, am e member of the Dnsite Safety Review Group for the Diablo Canyon Power Plant and from September,1977 to April,1982 was a Senior Instrument and Controls Supervisor of the Instrument and Controls (180)
Department for Diablo Canyon Power Plant.
! I, John B. Hoch, am Project Manager for the Diablo Canyon Project.
I, Richard b. Ettler, am Project Superintendent for the Diablo Canyon Project.
I, R. Keith Rhodes, am Technical Services Supervisor for the Diablo
! Canyon Project.
I, Edward M. Burns, an a Lead Licensing Engineer for Westinghouse i
Electric Corporation.
- 1. The purpose of this affidavit is to respond to the affidavit of John Cooper dated January 23, 1984, and filed with the Joint Intervenors' O Motion dated reeruary i4, i984. Mr. Cooper s aiie ations can se aroadir grouped into three categories: (1) that the design of the Residual Heat Removal (RHR) System is deficient; (2) that there was an inadequate 1
management response when Mr. Cooper expressed his concerns regarding alleged safety considerations; and (3) that Pacific Gas and Electric Company (PGandE) took unwarranted retaliatory actions against Mr. Cooper f as a result of his raising safety concerns,
- 2. This affidavit will discuss Mr. Cooper's allegations based upon our personal knowledge of the issues and events as they transpired.
Statements of our background and experience are attached.
l O-
O I. Residual Heat Removal System Design
Mr. Cooper was employed by PGandE as a construction inspector and an operations department instrument and controls' naiatenance technician between March 1976 and November 1979, and again as a construction field engineer between April 1981 and March 1982. The design issues raised by Mr. Cooper have all been considered and responded to by PGandE nanagement, as well as by the NRC in their subsequent evaluation and detemination of design adequacy.
- 4. The NRC has thoroughly reviewed and evaluated the design of the RHR O
' system at Diablo Canyon. As documented in the Safety Evaluation Report (SER) for Diablo Canyon, NUREG-0675, October 1974, the NRC has found that the design of the RHR system met all of their safety requirements.
Further, in later supplements to the SER, the NRC specifically reviewed the single RHR suction line from the RCS hot leg design in Supplement 7, dated May 1978 (Exhibit 1) and the RHR interlock design for RHR overpressure protection in Supplement 8, dated November 1978 (Exhibit 2) l and found these designs acceptable. Further, allegations made by Mr.
Cooper subsequent to his employment at PGandE have been extensively considered and resolved by the NRC in Supplement 21, dated December i
1983. The NRC's response to Mr. Cooper's allegations is documented at pages 2-85 through 2-113 of SER Supplement 21 (Exhibit 3).
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O 5. The primary function of the RHR systen is to renove residual and sensible heat from the core and reduce the temperature of the Reactor Coolant System (RCS) during the second phase of plant cocidown. During the first phase of plant cooldown, the cenperature of the RCS is reduced by transferring heat from the RCS to the steam systen using the steam generators and Auxiliary Feedwater System (AFWS). To accomplish the second phase of plant cooldown, the RHR syst'en is aligned to take suction from one RCS loop hot leg and pump the reactor coolant through ,
the RHR heat exchangers back to the cold legs of the RCS loops. A schenatic drawing of the RHR systen is attached (Exhibit 4).
- 6. The RHR systen is also utilized as part of the Safety Injection Systen (SIS) and Containment Spray (CS) systen. The SIS provides energency
' core cooling in the unlikely event of a break in either the RCS or stean systen. If required to operate as part of the SIS, the RHR pumps, along with the centrifugal charging punps and safety injection pumps, function initially to inject borated water from the Refueling Water Storage Tank (RWST) into the RCS. When this injection is complete, the RHR systen is aligned to deliver water from the containment sump through the RHR heat exchangers and back to the RCS for long-tem decay heat renoval. The RHR systen also functions as the water source for the containment spray l
systen du. ring post-LOCA recirculation. Water from the containment sunp is delivered to the containment spray rings by the RNR punps, as well as to the suction of the SIS punps.
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lO 7. Mr. Cooper's affidavit focuses on the availability of the RHR systen to renove residual heat from the core. For heat renoval during the second phase of a nomal cooldown, the redundant pump / heat exchanger trains of 5
the RHR system both take suction from tne RCS Loop 4 hot leg via a l single suction line. Two motor-operated valves (8701 and 8702) are located in series in this line to isolate tne RHR system fron the higiner l
pressure of the RCS when the RHR systen is not operating in the decay
'~
heat renoval node.
His allegation is incorrect. While later Westinghouse designs have employed two separate suction lines, Diablo Canyon and most of the other operating Westinghouse plants enploy the single suction line design.
These other operating plants include North Anna 1 and 2, Beaver Valley 1, Zion 1 and 2, D. C. Cook 1 and 2 Salen 1 and 2, Surry 1 and 2, Sequoyah 1 and 2, and Trojan. The acceptability of the single RHR suction line design with two . isolation valves has been documented in the NRC Staff's SER, Supplenent 7, dated May 1978, page 3-3 (Exhibit 1) l and Supplement 21, dated December 1983, page 2-95 (Exhibit 3).
l . m >
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l 9. The Diabl,o Canyon design fully neets the Diablo Canyon licensing i _
criteria. General Design Criterion (GDC).34,10 CFR Part 50, Appendix A, post-dates the Diablo Canyon design and is not part of the Diablo Canyon licensing basis. However, as stated at page 3.1A-ll of I the FSAR, the Diablo Canyon design does, in fact, comply with the . intent l
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f lO r
of GDC-34. The decay heat renoval safety function is provided by the diverse, safety-related AFWS (together with the stean generators and the atnospheric dunp valves).
- 10. The AFWS is specifically designed to renove decay heat under nomal circunstances as well as under circunstances such as loss of offsite power where the reactor coolant pumps are not operating. The capability ,
of the AFWS to perfom this function has been denonstrated at other i plants and will be demonstrated in the special low power test progran scheduled for Diablo Canyon.
containment and where the core is damaged. Mr. Cooper is wrong in several respects.
Cooper, a small break LOCA with insufficient loss of reactor coolant to
' go into the recirculation node, would not involve any core danage and would not require that the RHR systen perfom its SIS function. If the operators should choose to use the RHR systen in its nomal cooldown node (which they would not have to do), and it is unavailable because of a closed valve on the single suction line, the containment should still i
remain accessible and the operators could manually open the valve.
- 13. For small break LOCAs where there is a large anount of water released to containment, the operators could initiate the recirculation mode, just I ,
l l
O as in a large break accident, by injecting water fron the RWST into the J
RCS and then switching over to recirculation when there was sufficient water in the containment sump. In the recirculation node, the RHR systen is aligned into redundant flow paths fron the containnent sunp.
Thus, the decay heat renoval function of the Diablo Canyon design satisfies all single failure requirenents.
14 Mr. Cooper's other nain technical concern, which is repeated throughout his affidavit, seems to be that the design of the RHR systen and related circuitry is such that the valves are subject to being closed by unanticipated electrical signals. This has happened at Diablo Canyon on two occasions in the past during naintenance and testing activities.
Since then, the problem has been solved by revising plant procedures to require power to be renoved from valves 8701 and 8702 at all times other than when the valves are being operated. This practice, which has been approved by the NRC and included in the plant's Technical Specifications (Exhibit 5), precludes the possibility of valve closure by unanticipated signals.
- 15. Mr. Cooper's proposal is to avoid unanticipated signals by redesigning the power sources for the valve relays so that they bypass the Solid State Protection System (SSPS). The PGandE resolution is based upon l
evaluations by Westinghouse, the systen designer, which show there are reasons consistent with Westinghouse design standards for redundant protection channels to have all such signals, including the RHR valve relay signals, channeled through the SSPS. By modifying plant procedures to preclude the potential for the unanticipated signals, both Q
O the Westinghouse design standards and Mr. Cooper's concerns are directly and effectively accomodated. In arriving at this resolution, all of Mr. Cooper's concerns on this issue, as described in his affidavit and as raised by hin in his May 19, 1981 menorandun to Mr. Giscion (Cooper Affidavit at 9), his October 10, 1981 Design Change Request (Cooper Affidavit at 11,12), and his discussions and correspondence with the NRC (Cooper Affidavit at 9,11,12) were specifically and appropriately addressed. Mr. Cooper's affidavit is dedicated to. pronoting his own solution, but gives no reasons whatsoever why the NRC-approved resolution is inadequate. See SSER 21, Decenber 1983, at pages 2-85 through 2-113, and NRC Inspection Reports 50-275/82-26, 50-323/82-13 and 50-275/82-42 (Cooper Affidavit, Exhibits 10b and 10c).
Mr. Cooper (Cooper Affidavit at 10 and 13) expressed concern over the Q. 16.
lack of control roon annunciation to alert the operator of loss of RHR fl ow. In fact, PGandE has connitted to install an RHR low flow alam.
In addition, there are several other means for the operator to detemine that loss of RHR flow has occurred due to the closure of an isolation valve. Flow through the RHR systen is indicated in the control roon on flow instrunents in the RHR/ SIS return lines. The RHR suction valve positions are indicated in the main control roon by red / green status lights near the valve control switches. The suction valves are included on a monitor light box such that the monitor light is on when the valves are open.
- 17. In sunnary, the design of the Diablo Canyon RHR systen is a standard Westinghouse design employed in most of its operating plants and is fully in accord with the Diablo Canyon licensing basis requirenents. l l
O Plant procedures have been inplenented to preclude the possibility of unanticipated valve operation in the RHR systen in accordance with the j requirements of the operating license Technical Specifications. ,
4 1
l O
t i
O-
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1 I
l t _ _ _ _ _ _. _ _ , _ _ _ - . . . _ . . , _ . . _ . _ _ , . , . _ . , _ - , , _ . . . _ , _ , . . . . ~ ~ . - _ . _ -
l
' O II. Management Responsiveness
! 19. PGandE has had a longstanding policy of encouraging employees to identify items or areas of concern and bring them to the attention of management. Management, in turn, is responsible for assuring that these concerns are appropriately resolved in a timely manner consistent with their significance, regulatory requirenents, and other ongoing work activi ties.
- 20. This policy has been embodied in various mar.agement directives which have been issued over the years (See Exhibits 6-11). To assure that this policy was effectively implemented on a day-to-day ') asis, fomal systems were developed to document and track problems. These systens are described in written procedures developed by each department. For example, Nuclear Plant Operations (NPO) has issued Nuclear Plant
! Administrative Procedure C-12: " Identification and Resolution of i
Problems and Nonconfomances." Similarly, General Construction (GC) has issued G.C.Q.A. Program Procedure GCP-12.1: " Documenting Discrepancies and Assessing Reportability" (Exhibits 12 and 13).
- 21. Under the NPO system, each identified problem involving equipment,
( design, materials, procedures, etc., is first documented on a Nuclear Plant Problem Report (NPPR). Any individual who identifies a problem may initiate a NPPR, regardless of the organzation with which he is affiliated. Employees are encouraged, and for many types of situations O. - are required, to initiate NPPRs, even for what may be considered tc.be s >
c I p
l 4
l l
O trivial matters. After the NPPR is generated, it is routed to plant supervision for approval and then to plant management for review. Plant management review serves several purposes: (1) to assess potential significance and reportability, (2) to assure that the NPPR is routed to the proper group for resolution, (3) to propose and/or concur in a j
resolution, and (4) to establish a priority for resolution.
- 22. Any item which is judged to be potentially reportable or which meets significance criteria established by the PGandE Quality Assurance Department is elevated to the level of a potential nonconfomance, and a Nonconfomance P.eport (NCR) is initiated. The NCR is used by all departments, and all potential nonconfomances are fomally dispositioned by an ad hoc cornittee which includes representatives fron Quality Assurance as well as other affected departments. Each NPPR and O
s NCR is tracked by computer until ultimately resolved.
- 23. Under these policies, an enomous number of such documents are initiated, all of which must be carefully controlled to assure appropriate disposition. In the years 1978 through 1983, approximately 20600 NPPRs were generated at Diablo Canyon. Of these NPPRs, approximately 3700 were initiated by the Instrument and Controls Maintenance Department, in direct contradiction to Mr. Cooper's assertion that there was a policy within this group to discourage the writing of NPPRs.
- 24. With the large numbers of NPPRs which have been generated, a l
correspondingly heavy burden is placed on plant management for timely j
and responsible disposition. Accordingly, some prioritization of these 1
tasks is required. Inevitably, some items will remain open longer than
' O' others, depending on the prioritization process. However, the plant
- - - - - , - --m m. - pv-y ,
O staff at Diablo Conyon continues to work aggressively to close out these items, and they will be closed out on a schedule cor.sistent with need, l significance, and regulatory requirements. For example, as of January 31, 1984, approximately 18500 of the aforementioned NPPRs have been dispositioned. All NPPRs required to be closed out prior to fuel loading were closed out. Similarly, all NPPRs required to be closed out prior to cemencement of heatup were dispositioned prior to comencing heatup. This process will continue for initial criticality and ultimately for full power operation.
- 25. A similar situation exists regarding the disposition of Design Change Requests (DCRs). Since the design verification program comenced in late 1981, approximately 3500 DCRs have been initiated. The sheer number of DCRs has required prioritization . As a result, certain items remain open longer than others. At this time, approximately 3360 have been dispositioned by Engineering. Again, these items are closed out on a schedule consistent with need, significance, and regulatory requi rements.
- 26. As discussed below, in each of the instances cited by Mr. Cooper in his 4
affidavit, particularly those related to his concerns over the RHR system, the issues he raised by memorandum, NPPR, DCR or other means were responded to by management and properly dispositioned prior to the time when such dispositions were needed.
.?7 Mr. Cooper (Cooper Affidavit at 9) states that he wrote a memorandum to Mr. Giscion on May 19,1981, " explaining that valves 8701 and 8702 would fail closed when SSPS [ solid state protection system] output fuses were removed and that emergency procedure OP-8 ' Control Room Inaccessibility'
'l l
4 O -s in error.- Contrary to the impiication in the affidavit, Mr.
Cooper's concern was addressed on July 30, 1981, through the issuance of NPPR DCO-81-TI-P0237.
- 28. During the interval between May 19 and July 30, 1981, Mr. Giscion had been engaged in discussions with the NRC staff regarding the Technical Specifications, including the manner in which valves 8701 and 8702 would be operated in their role for low-temperature RCS overpressure protection. The staff agreed in July that power should be removed from
! valves 8701 and 8702 during operating modes 4, 5, and 6 when the reciprocating charging pump was in operation. The HPPR issued on July 30,1981 thus encompassed and responded to the topics raised by Mr.
Cooper's memorandum and established that the emergency operating procedure would be revised to require that the circuit breakers on h valves 8701 and 8702 be opened after the valves were opened.
- 29. Mr. Cooper suggests (Cooper Affidavit at 9) that he was reprimanded for sending the memorandum to Mr. Giscion, and warned against going to the NRC. In fact, Mr. Cooper had ignorJd the established procedures by sending a memorandum directly to a departnent head in an entirely different organization. He should have initiated a NPPR or a Minor Variation Report, or initiated a Design Change Request (DCR), any of which would have officially entered his concern into fomalized tracking systems. In this case, there was no reason for Mr. Cooper's failure to follow established procedures, and he was requested by supervision to follow thest procedures in the future. This was not only reasonable, but necessary. The established reporting systems would break down and concerns would not he systematically and adequately addressed if
. . _ ~ .
i employees bypassed the established procedures and sent then by memorandum directly to managenent, particularly in another organization.
- 30. Mr. Cooper was not warned against going to the NRC or threatened with He was the loss of his job as he alleges (Cooper Affidavit at 9).
' requested by supervision, Mr. P. Gilbreath and Mr. R. D. Etzler, to attempt to resolve problems within his organization and within PGandE prior to reporting directly to the NRC. At no time was he told, either directly or by implication, that he would lose his job or be subject to disciplinary action. In fact, Mr. Cooper connunicated with the NRC on several occasions after the alleged threat (Cooper Affidavit at 9,11, 12-13) with neither loss of job nor reprinand or criticism fron management.
- 31. Mr. Cooper initiated a DCR on October 10,1981 (Cooper Affidavit at 11, 16-17). In that DCR he requested that the valve interlocks be renoved, since the renoval of power from the valves would make the interlocks no longer usable. While he complains that his DCR "has been sitting in some engineer's in-basket since 1981, unreviewed and unresolved," (Cooper Affidavit at 16), he had previously noted (Cooper Affidavit at 11) that the Onsite Safety Review Group (OSRG) had reviewed and recornended rejection of the requested design change on the grounds that it would increase the probability of overpressurization of the RHR systen. .The DCR, by now a decidedly low priority iten, was officially dispositioned by Engineering, which rejected the requested design change in 1983. Procedures were correctly followed, and dispesition was appropriately made prior to the tine when the RHR systen would be called r
- into service to remove decay heat.
O-l l
. . - - _. . _ . - - - - = . - _ _ . _ - _ -.
Mr. Cooper states (Cooper Affidavit at 13) that his notification of an Q 32.
error in the Plant Manual "had not been corrected eight months after the original notification." This involved a correction to Volume 16 of the Plant Manual identified by Mr. Cooper in April 1981. The correction j involved a Hi-Lo Level alam on the reactor coolant pump lube oil system which was incorrectly identified as a Lo Level alam. A long tem mvision of Volme 16 was in progress at that time. Since the plant was l
not then in operation, plant management decided to incorporate the correction into the long tem revision. The milestone requirement
~
established for completing this revision was prior to plant startup. In late 1982, in order to address Mr. Cooper's concerns, and at the suggestion of the NRC, a plant engineer, Mr. R. L. Fisher, issued a mvision (on-the-spot change) to the existing version of Volume 16.
This revision was ultimately superseded by the complete revision of Volume 16 issued in September 1983, prior to Unit 1 fuel load in November 1983.
- 33. Mr. Fisher also discussed the suggested correction with Mr. Cooper in June 1981 and advised him in writing via a memorandun dated June 9, 1981, that appropriate changes had been nade in the draft copy of Volume
- 16. In a response to Mr. Cooper's memorandum to Mr. Thornberry, dated December 19, 1981, Mr. Fisher again doceented the status of this matter in a memorandum dated January 11, 1982. The January 11 memorandun l
contained a marked-up copy of the Volume 16 pages showing Mr. Cooper I that his concern was being addressed.
- 34. Mr. Cooper states (Cooper Affidavit at 13) that NPPRs he had written three years previously had not b w n resolved. The two NPPRs referred to l
O-J -- -. -- . _
i by Mr. Cooper involved a change to a PGandE drawing to correct an air supply pressure for a control component, and a change to two vendor These NPPRs
' drawings to show the correct instrunent channel numbers.
As were closed out in September 1983 and September 1982, respectively.
discussed previously, problems are corrected and NPPRs closed consistent with the time at which equipment is required for plant operation. Both problem reports would have been addressed more rapidly if required sooner by plant operational or maintenance needs.
l 35. Mr. Cooper alleges without explanation "that the FSAR description of this system [ presumably the RHR system] was incorrect, and PGandE refused to change it" (Cooper Affidavit at 13, 14-15). To the contrary, PGandE has been and is engaged in a comprehensive revision of the FSAR as required by 10 CFR 50.71(c). In accordance with an extension granted i
O by the NRC, the revision is due on September 22, 1984. The revision will include the updated description of the RHR system.
- 36. Mr. Cooper (Cooper Affidavit at 23) alleges an " unwritten policy" to attempt to cover up plant deficiencies. There is no policy, written or unwritten, to cover up any deficiencies identified in plant design, construction, or operation. In fact, as discussed above, PGandE's policy is just the opposite. Indeed, the most significant example which l
refutes this claim is PGandE's response to the discovery of the diagram design error in September 1981. At that time, PGandE reported it to the NRC and voluntarily stopped preparation for fuel Icading at Diablo -
Canyon Unit 1.
l
(LERs). During 1983, 6409 NPPRs,113 NCRs, and 36 LERs were issued at Diablo Canyon. This level of documentation shows that PGandE is actively resolving problems, not covering then up.
- 38. It is the policy of the nanagement at Diablo Canyor that all NPPRs, and all such other connunications involving safety concerns, be addressed l
and resolved. The resolution of such problen reports is performed on a ll schedule consistent with when the problen itself nust be resolved. In l
other words, if there is a problen report which affects the safe operation of a particular system, it will be resolved prior to declaring the systen " operable." If a stated problem is not worthy of action, it is so stated in the answer / resolution section of the problen report before the report is teminated or conpleted. PGandE has identified i five nenoranda which Mr. Cooper generated in the spring of 1979. All of the concerns and all of the points raised in these nenoranda have been O addressed. All of the problen reports which Mr. Cooper specifically referred to in his affidavit have sinflarly been addressed and closed.
- 39. Mr. Cooper alleges (Cooper Affidavit at 24) that "problen reports" are destroyed by nanagenent. Management at Diablo Canyon has never deliberately destroyed or voided a NPPR in order to cover up or dismiss a plant safety issue. Any employee found deliberately destroying a problen report for such a reason would be subject to disciplinary action.
- 40. Any person who believes he has identified a plant problen can write a l
NPPR to document the problen. When a NPPR is generated, it initially is ,
reviewed by first level supervision to verify that a legitimate problen exists. If supervision detemines that the problen is not valid, the NPPR can be voided. Real problems identified with safety-related
4 If O Procedures or safety-reiated eaufPuent are not voided by maaa9emeat.
j a NPPR is voided, an explanation is provided to the person who initiated the NPPR. If the employee is dissatisfied with the explanation or resolution and continues to believe a valid problem exists, he can Pursue the matter through successive levels of management via the "Open Door Policy." Once entered into the tracking system, if a NPPR is voided by supervision, it remains a pemanent plant record.
- 41. In December 1983, completely independent of Mr. Cooper's allegation, the Diablo Canyon Quality Control Department conducted a management review of the NPPR logging and tracking system. The status of the approximately 20600 NPPRs generated during the period from 1978 through 1983 was examined in the management review. Prior to the advent of the current conputerized logging system, NPPRs were tracked with a h handwritten log maintained by the department. When the current computerized system was established, the records for previous years were entered into the system. A search of the computerized records showed that 100 NPPR numbers in the ISC Department were indicated to be mi ssing. Based on a review of the handwritten NPPR log and other records, it was detemined that nine were missing due to data entry l
errors.,11 were shown to have been dispositioned,19 are still open and will be dispositioned as scheduled, and 61 were voided by the supervisor l
l in charge. The voided NPPRs are being researched to ensure that the perceived problem has been adequately addressed. Following this process, they will be officially closed out. At the present time, approximately 16 of the voided NPPRs have been closed and we anticipate completion by March 31, 1984. Thus, there are no identified instances of destruction of problem reports or improper voiding of NPPRs.
, . - - -_ - _ - . . . _ - -,o._ # . .-,.-
O 42. Since the is78-79 time frame referred to by Mr. Cooper, the system for :
tracking NPPRs has been significantly upgraded and improved. Under
, current procedures, after approval by the supervisor, a number is assigned to the NPPR and a brief description is entered into the Records Management System (RMS). The status of the NPPR is tracked using this The computer-based system until the NPPR is completed and closed.
current procedures are intended to ensure that a NPPR cannot be lost or remain unprocessed, although occasionally, due to clerical errors in processing the large number of NPPRs, some may become temporarily lost, h all such cases an investigation is conducted to locate the NPPR and resolve the issue.
- 43. Mr. Cooper alleges (Cooper Affidavit at 24) that he was " reprimanded" for correcting an error. The accusation takes on a sonewhat different hue when examined in its proper context. . Administrative controls are in place to assure proper implementation, documentation, and resolution of problems. Notwithstanding Mr. Cooper's statement that all that was involved was "a simple wiring change," the error " corrected" by Mr.
Cooper constituted an unauthorized design change, contrary to the requirements for approval and control of all design changes. The procedures for resolving the error were underway, through connunications with liestinghouse, when Mr. Cooper made the unauthorized change.
Unauthorized changes of t'he type made by Mr. Cooper constitute serious l
violations of procedures which are deserving of reprimand.
[ 44. Mr. Cooper alleges (Cooper Affidavit at 25) that maintenance technicians 1
' were " routinely denied access to necessary infomation," and that "the problem of inadequate reference materials was not corrected." The
(
O statements are not true. Contrary to Mr. Cooper's allegations, required documentation was always available to the technicians, and significant improvements were made in the degree of availablility and accessibility
! of the documents.
- 45. The documentation in question consisted of design infomation desired by the instrumentation and control technicians to perform their maintenance activities. At no time were the technicians " denied access" to documents. The required documentation was always available, although the limited number of controlled copies sometimes made access inconvenient. When management was requested, in a letter dated April 3, 1979 from the technicians, to improve document accessibility, timely I
action was taken to distribute and make available more copies of such high-use documents. It should be noted that a " satellite" file, or extension of the plant master file, had been established in the
~
instrument shop since about 1975. Also in 1983 an RMS data teminal was located in the I&C Maintenance shop area to provide more convenient access to this system.
- 46. Several technicians also had concerns about the adequacy of existing plant procedures. Mr. Giscion responded properly and reasonably by l
requesting the technicians to document any inadequacies that were l
believed to exist. This documentation came in the fom of four memoranda and associated NPPRs from Mr. Cooper. All of the items raised in the memoranda were addressed and responded to.
Nuclear Notepad, and NOMIS (Nuclear Operations and Maintenance Infomation Service) is regularly disseminated to personnel at DCPP.
l This program was established in 1981 as part of our response to
! requirements contained in NUREG-0737, " Clarification of TMI Action Plan Requirements."
l 48. And finally, Mr. Cooper's allegation (Cooper Affidavit at 26) that i management " destroyed" individual files is actually a criticism of i
PGandE for proper records control. In 1982, a continuing program was in progress to assure that current infomation was available to all personnel working on safety-related equipment. In many areas, management found it necessary to remove unofficial documentation from working files to assure that the documentation in use reflected the actual configuration of the plant, i.e., if a design change had been made to a cin:uit, it might not have been translated to a technician's notes and unofficial drawings. Using this incorrect, unofficial documentation could have resulted in a maintenance error.
- 49. Mr. Cooper (Cooper Affidavit at 10, 11-12) alleges that a violation of .
internal control procedures occurred in the disposition of the NPPR initiated to document the spurious closure of valve 8701 on Septenber 29, 1 981. The concern was that the NPPR had been finally dispositioned without initial plant management review. The allegation is correct, insofar a~s a portion of the NPPR fom had not been signed as required.
However, the reviews required by the initiation of the NPPR were properly, adequately, and timely perfomed.
- 50. The NPPR fom has three principal sections. The top section is where the problem is reported. The middle section is for an initial Ov management review, where the responsible department head looks at the l
+- -
O cirtuostances and decides whether the p mbiem is a potentiai nonconfomance or is poteatially reportable, establishes a priority for resolution, and identifies any further instructions he feels to be i
appropriate. The bottom section indicates the final resolution and sign-off. ,
- 51. Review of this NPPR indicates that the bottom section was signed, l
- indicating final resolution, but that the middle section was not signed as required by Administrative Procedure C-12 to indicate the initial review, Mr. Sexton, the responsible manager, cannot recall the exact circumstances surrounding the NPPR. He notes, however, that because the resolution (issuance of a procedure change) and an initial punp test and inspection were completed imediately after the problem occurred, and prior to his seeing the NPPR, the initial review was superfluous, h Ordinarily, resolution of a problem is not accomplished so quickly, and the two-step process indicated on the NPPR fom is gemane.
- 52. Plant Management responded to the September 29, 1981 valve closure in an appropriate nanner. The Shift Foreman's log for that date clearly indicates that the Operations Department, in conjunction with the Maintenance Department, inspected the pump after the valve closure and
- the Operations Department successfully completed Surveillance Test Procedure P-38 to detemine operability of the pump. Further, a procedure change was initiated by the Operating Department on September 29, 1981 to Operating Procedure B-2 to help prevent a recurrence of the event. Additional followup action took place on this event. The GSRG initiated a second NPPR (NPPR DC-1-81-NO-P0010) on November.13,1981, calling for a perfomance test (P-3A) to be run on the pump and the ieted.
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- 53. Mr. Cooper suggests throughout his affidavit (Cooper Affidavit at 1, 9, 12, 19, 24-27) that he has suffered retaliation from management because of his efforts to identify safety concerns and bring them to management's attention. There is no policy, written or unwritten, Pemitting or encouraging management to take retaliatory action by any I
means against employees who raise safety concerns. In fact, the policy is just the opposite; it encourages employees to identify safety concerns and specifically reassures them that such action will not be hamful to them.
- 54. This policy is strongly stated in Mr. Thornberry's menorandum on the "Open Door Policy" (Exhibit 9):
O voicias coaceras r 9 rai 9 91 at f tv or operatiaa 411 not be documented in the employee's personnel record and i WT1 never be used for any type of disciplinary action.
! I give my fim guarantee of that.
- 55. We have carefully reviewed the circumstances described in Mr. Cooper's affidavit. The incidents cited by Mr. Cooper involved no elements of retaliation or punishment for attempting to bring real or perceived l
concerns to the attention of management or the NRC, and we find no evidence that he was disciplined or retaliated against in any way for expressing his concerns. This will become clear in the following discussions of his various allegations.
- 56. Mr. Cooper alleges (Cooper Affidavit at 9) that he was threatened with loss of his job if he spoke to the NRC. This is false. See paragraph l 30 above.
.M 1 _ , _ .
O 57. Mr. Cooper's allegation (Cooper Affidavit at 24) that he was reprimanded for correcting an error is true. As discussed in paragraph 43 above, i
Mr. Cooper's actions constituted an unapproved design change in violation of plant procedures, and was deserving of reprimand.
- 58. Mr. Cooper (Cooper Affidavit at 25) alleges that retaliation was taken i
against whistleblowers by assigning them to the "more distasteful jobs" or giving them poor perfomance reviews. The only examples he cites, however, are his own. He states that when he found errors in safety-related procedures, he was " isolated" in the coffee room and He also given the " distasteful" job of rewriting the procedures.
alleges that in 1979 he was given a poor perfomance review.
- 59. With respect to rewriting procedures, he was the obvious and nost logical candidate for the job, since he himself had identified the l needed revisions. It comes as a surprise to learn that he considered it a distasteful job. As Mr. Cooper well knows, there are literally thousands of procedures required for the operation of a nuclear power plant. . Prior to operation, the initial preparation of these procedures and subsequent revisions after initial use is a monumental effort which requires the combined efforts of virtually every engineer, foreman, and technician in the plant. To the extent that procedure preparation is
" distasteful," it is a distaste to be shared by all.
- 60. Since the shop area lacked office space for such work, temporary tables
' were often set up in the " coffee room" of the administration building.
i This room had ready access to the central file system located in the l
administration building. No restrictions were placed upon Mr. Cooper's Q
)
O freedom to go where he pieased, and he was no more isoiated from the crew than any other person, including the plant's ISC Supervisor, whose work location was in the administration building.
- 61. Mr. Cooper's allegation regarding his perfomance review is equally erroneous. PGandE supervisors are trained and instructed to give perfomance reviews which accurately reflect the employee's true perfomance, including strengths and weaknesses and areas needing improvement. Perfomance reviews are reviewed by at least one higher -
level of management to insure that the evaluation is fair to the employee. There is no policy at Diablo Canyon to use the perfomance review for retaliation or any other unjustified reason.
- 62. Moreover, the perfomance review quoted by Mr. Cooper in his affidavit was given in 1979, well prior to his becoming, in his words, a i
"whi stleblower."
- 63. PGandE did not consider the perfomance review to be a particularly poor review. This is substantiated by the decision to re-employ Mr. Cooper in April 1981 after he had resigned in November 1979. No reputable company intending to retaliate against an employee by using a perfomance review would rehire the employee 1-1/2 years later. One can only speculate why Mr. Cooper wanted to return to work for PGandE in c
l 1981 with such ill feelings toward management which he apparently developed in 1978-1979.
- 64. Mr. Cooper alleges (Cooper Affidavit at 10) that he was being " punished" However, when his security clearance was interrupted on August 6,1981.
Not as Mr. Cooper goes on to note, it was reinstated the very next day.
only can this action not in any way be construed as harrassing, or even O
__~1 _ _ -. . _ _ , _
_ - _ _ _ . _ _ =__. ___ - __ - __- - - .- _ . _-. -
! significantly inconvenient, no suggestion was ever made to Mr. Cooper that the one-day interruption of his security clearance was in any way retributive.
! 65. Security reconis that old are routinely removed from the files, so we have no records to check on the incident. Mr. Cooper's supervisors in General Construction have no knowledge of a harrassing or disciplinary i
Mr. Thornberry certainly
" hold" being put on his security clearance.
e did not put a " hold" on Mr. Cooper's security clearance. If Mr.
Cooper's security clearance was, in fact, temporarily interrupted, there For could be several purely administrative reasons for this to occur.
example, if an employee's status level expires in the security computer, It the employee cannot enter Unit 1 until the status level is updated.
is not an unconnon occurrence for employees (or even entire departments) to have their status level expire because of various paperwork It problems. Also, security status must be reinstated every 30 days.
is not uncommon to miss a particular individual's 30-day update. Such occurrences nonnally take less than a day to correct.
I 66. Mr. Cooper alleges (Cooper Affidavit at 27) that PGandE's " Behavioral Reliability Program" will be employed to " weed out whistleblowers" and inflict punishment by " involuntary psychiatric examination and loss of The j ob." This charge is astonishing, and totally without foundation.
behavioril observation program in effect at Diablo Canyon is a necessary component of PGandE's required and approved Physical Security Plan and has been developed in accordance with NUREG 2076, " Behavioral Reliability Program for the Nuclear Industry"; Appendix B of NUREG 0768,
" People Related Problems Affecting Security in the Licensed Nuclear Om ,
I l - -
radustry > a4 aast/^as st ad ra 3.3-i982 security for nuciear power O
Plants." As part of the NRC-required security program, Diablo Canyon employees are administered the Minnesota Multiphase Personality Inventory (WI).
- 67. Mr. Cooper has a history of vociferous opposition to the administration of this program, and notwithstanding his acceptance of employment with PGandE, he has declared to both the press and his employer that he would not work for a company that " spies on its own employees" (Exhibit 14).
That, of course, is not quite how the program works. A detemination that a person exhibits unreliable or untrustworthy behavior which could jeopardize the safety of the general public, the plant staff or the plant itself can be made only after a recornendation by a licensed clinical psychologist or psychiatrist not in PGandE's employment. Thus, there is no way to use the behavioral observation program to " weed out O whistleblowers", inflict punishment, or spy on employees.
, 68. Mr. Cooper states (Cooper Affidavit at 12) that he was transferred to a
" remote, snowbound site" as punishment for " speaking out on safety concerns at Diablo Canyon." The facts indicate otherwise.
- 69. Mr. Cooper was notified by supervision, Mr. R. X. Rhodes, on February 24, 1982, that he should report to the Helms Project in the Sierra Nevada mountains, on March 1,1982. His assigned task was to make instrument take-offs in order for General Office Engineering to provide set-point data back to the Helms Project. Mr. Rhodes explained that this was a temporary assignment to which Mr. Cooper was assigned due to I
his experience and familiarity with instrument systems. His work was completed on approximately Man:h 9,1982. Mr. Cooper was requested to O_
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j.
1
! renain at Helms through the end of the week (March 12, 1982) to answer any questions concerning the take-offs in the absence of the forenan in charge of the instrument group. Mr. Cooper was also told that he would be sent back to Diablo Canyon early in the week of March 15, 1982.
- 70. Instead of complying with his assignment instructions, Mr. Cooper left l the jobsite on March 9,1982, without prior notice or explanation to anyone. When contacted by supervision (Mr. Rhodes) on Monday, March 15, 1982, and asked to explain his actions, Mr. Cooper stated that he had resigned'as of March 9,1982. He told Mr. Rhodes he was going to quit
! anyway because of his dispute with PGandE over the requirement i. hat he take the MFI.
- 71. It is standard PGandE policy and practice within General Construction to make work assignnents to employees on relatively short notice for various durations due to the nature of construction and startup work.
l Mr. Cooper was definitely not assigned to the Helns Project as l
punishnent.
//
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i Q
~ _ . ~~ _ - . .
O Dated March 5,1984. ,
2 JAMES D. SHIFF g ( J0tWB.]CH
_ _M e - Mh- _%
EDWAKU M. BURN 5 g/ JOHN M. GI5CLON V 0-4
~ KEN C. DOSS Subscribed and sworn to before me this 5th day of March 1984.
g r 4
VM7 SEAL Rancy J. Le6 aster, h Notary Public in and for the e- City and County of San Francisco, O. State of California.
My corviission expires April 14,1986 rx:x:ccono:we=ccm::cceeer N NANCY J. LEMASTER NOTARY PUC'J0GLIFORNIA
[. h t l..., ' . , CITY A'*O CC'.J!!TY OF SAN MAra0iSCO
, My Commission Emp.'es April 14.1986
. r4We":OccCre:.tCO:MTcCCCCCCCM l
1 O
JAMES D. SHIFFER O
JOHN M. GISCLON KEN C. DOSS JOHN B. HOCH EDWARD M. BURNS Subscribed and sworn to before me this 5th day of March 1984.
Nancy J. Lemaster, Notary Public in and for the City and County of San Francisco, State of California.
O's My comission expires April 14, 1986 i aN _
~
6 ' ROBERT PATTER 50N ROBERT C. THORNBERRY RICHARDD.ETZLFy
$W R. KEITH RH06ES h&
Subscribed and sworn to before me this 5th day of March 1984.
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b[l6- i b1771 N Y SPRO L notaer eveuc cAUFORNIA
, WendySproul,g W Luis OSISPO COUm A Notary Public in and for the <
, U City and County of San Luis Obispo, I" M"M"_ !
State of California.
My comission expires June 30, 1986
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List of Exhibits Exhibit 1 NRC Safety Evaluation Report, Supplement 7, pages 3-3 to 3-4, May 1978 Exhibit 2 NRC Safety Evaluation Report, Supplement 8, page 7-1, November 1978 Exhibit 3 NRC Safety Evaluation Report, Supplement 21, pages 2-85 to 2-113, Decenber 1983 Exhibit 4 Schematic Diagram of Residual Heat Renoval Systen Exhibit 5 Technical Specifications, Section 3.4.9.3 Exhibit 6 Letter from J. D. Worthington, S bject: PGandE Policy on Identifying and Reporting Safety Concerns, April 29, 1976.
Exhibit 7 Menorandum from R .D. Etzler,
Subject:
Personnel Particpation via the Enployee Suggestion Plan, July 25,1980.
j Exhibit 8 Letter fom J. O. Schuyler,
Subject:
Reaffimation of PGandE
' Policy on identifying and Reporting Safety Concerns, August 7, 1980.
Exhibit 9 Letter from R. C. Thornberry,
Subject:
Open Door Policy and PGandE Quality Hotline, February 5,1982.
Exhibit 10 Letter from F.W.f"-1ke, Jr. and 8. W. Sbckelford,
Subject:
Reaffimation of PGandE Policy on Identifying and Reporting Safety Concerns, March 22, 1982.
! Exhibit 11 PGandE Hotline Notice l
J Exhibit 12 Nuclear Plant Administrative Procedure C-12 " Identification and resolution of Problems and Nonconfomances", Revision 6 June 20, l
1983.
Exhibit 13 General Construction Quality Assurance Progran Procedure GCP-12.1
" Documenting Discrepancies and Assessing Reportability,"
Q Revision 3, June 30,1983.
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Exhibit 14 Article from SAN LUIS OBISP0 TELEGRAM-TRIBUNE dated January 23, 1982.
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EXHIBIT 1 - SSER 7, page 3-3, May 1978 O .)
- memel instrumentation is evettable ta esintain surveillance of important primary and necendary systes parameters such as pressure, temperature, and meter levels.
. The applicans has presented an analysis of the primary system ester vol se sarink-f- age due te coeldsun and has dotarained that the shrinkage is sufficient to accom-I 8
- aseate chemical and veime control systas fget required for teration and cooling
- of me pressurizer. , Therefore, the normal letdeun system f a not ree.f red ta i - achieve long-tare coelfag with residual heet removal systas. Staffarly, the applicant Aes demonstrated that dogastffcatten of the reacter coolant is not necessary because the potential nyeregon esacantration in the solution is less 1Aan the sataratten value at cold shutdeue sendittens.
All of the operetse actions noseed la perform plast coeldsun (encept for periodic survef11ance of the toren concentratten) con to accomplished free the control rees ass efag ne single failure. The applicant has demonstrated that redundant patas or systams are aveff able ta perfere the essential functions using guelffied equip-sent in the event of a single fa11ers. In same fastances this mould require .
sperstar actfen outside of the conteel rees to activata the redundant pe u.
trita regard ta the residual heet removal systas, the auctfen Ifas fs a sfagle line with tue isolatten velves in series. 'A failure that prevents eponing these valves O ,
- usu14 provost activetien of the residual heat removal system for long-tare cold
,h , savtdene heet reseval. An electrical failure could readily to corrected by sens-ally actuating the valve (s), ide mere concerned atest a possible anchenical failure of one of these valves. The applicant andressed SAfs guestian in a letter dated January 2g, Ig73,' asintaining that the probatt1fty of a valve disc senareting free l
the valve stas is law e'eeagh that it need not to considered in the single failure study.
Sased on our review of this aetter we have concluded that this design festare is acceptatie for tas reasons states below. If anchanical valve failures of the type that preclude opening the velves are considered ta to rendes events, taen the
. pestability of such failures occurring et the same time as a severe earthquaka does appe' e r ta be quite low. gn the etaer hand, such failures any reasoneely be considered to be related to the earmouake. Ida de act believe that the pree-anf11ty of an eartaquete causing the failures has been quantiffed, llouever, an earthquate sneuld not affect the valves since they are designed to withstand the Idesgel event. Accordingly, altasugh the ceastned prohott11ty level has not been quantified, it is unifkaly that a severe earthqueta will occur fa coesinetfen with mechanical failures that preclude eponing one of the tus residual heat removal j system suction valves. Fureerence, if a severe eartaqueta should secur in com-tinett u with eschanical failures that preclude opening one of the reslauel heat removal section velves, feny tare heet removal con te accesplisted indefinitaly
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witn the steam generators and the auf1tary feedmeter systes rather taan the residust heat removal systas. As a result, we consider tas likalthe:J of this 2,
failure in causination with its potential consequences to to acceptasie.
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Sinca, without offsita power, the reacter coolant peups could not be run ts provide reacter coolant systes einfng, the appIfcant has committed ta perferning a natural circulatten test ta demonstrate adeouate beren afzing and accostaale . t seeldsen contitfens using natural circulation. These tests will be performed !
during start s when core heat is avaffatte.
The systans and equipment needed ta perfers these functions will be quelfffed for the Nesgrf event and have been included in the seismic reevaluatten program wnich fs discussed in other sectfens of tafs swelement.
The following f tmas are still outstanding in this portfen of our review:
(1) A rovfew of the system functfen fadication avaffable to the operator in the i
control room fn connection with perferetag the shutdeen. (Sectfen 7.5 et this soplament)
(2) A review of the sefsmic qualf fication of the rar wetar storage rese-vefrs, faciuding:
(a) A review of the potential for safches causfag a sfgnificant less of meter from the row meter starege reservefrs. (Sectien 2.4 ef tais -
O -e ene (b) A review of the potential for slopes sliding fata the raw water storage reservefre. (Section 2.3.3 of tafs septement)
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(c) A review of the propertfee of the rock underlying the rer meter starage reservefrs. (Sectfan LS.3 of this supplement)
(d) A review of the previsions ta ensure tAat the reservefr seuld not drafn through connected pfpfag that is amt emelffled. (Section 10.5 of this supplement) .
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We have reviewed the capabfifty to cool the plant te cold sautesen conditfens and provice long-tare coeltag. The app 1fcant has demonstrated that sufficient systems are avafleele for residual heat removal with or witaeut offsite power and assuming a single failure in accordance with Criterien 34 of the General Design Critaria.
Steilarly, these systems will be qualfffed for operation fa the event of the Nesgrf event fa accordance with Criterien 2 of the General Sesign Crftaria.
Accordingly, subject ta satisfactory rosesution of the setstanding questions described aheve, we find these provistens acceptatie.
We will provide sur eva'luetten of the outstanding f tens discussed above in a future supplement to the Safety fvaluation Raeort.
EXHIBIT 2 - SSER 8, page 7-1, November 1978 7.0..fMSTRUMENTATION AND CSTROL. . ..
7.2 Reactor Trio Svstaa In Supplement Neber 7 to t.% $afety Evaluation Aaport we fcund ce basic seisaic scram systas procesad by the apo11 cant acceptaale.
!fewever, we reeufred further infomation free De applicant regarding .%w L%
systas would satisfy our requirements for separation, isolation quality, testacility, and qualification for Class 1E circuits.
, TheasolicanthasprovidedadditionalinformationonthissubjectinAsencuent67 ta tne Final Safety Analysis Report. Based on our review of the aeditional infomation, we have concluced that the salsaic scras systas is of a sfailar sesign and meets the same criteria as the reactor protection system and is, therefan, accootaole.
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f W consider tais matter resolves, d
- 7. 4 Systems tecuiete for Safe Shuteown In Susolement Naser 7 we stated that we would reevire further Information aneut tne insication availasle to the control reos operator in c=nnec*. ion vita perforsing a Sautaoun after a Mesgri event. ide have iew coesleted our review of this asttar anc it is resolves as discussed in Section 3.2.1 of uis suoclament.
- 7. 5 4H4 Ove-seestue. 7 . tact ion inteetoems
- n :ne Safety Evaivation Report .e aesc-ites ue inter octs for :ne notar acerste salees on 24 esioual nea: -emo..I sys.as section ! ~nes (Vaives 3701 sno '70".
- he interlocas acerate on diverse 3rfaciples ta srevent scening us saives nen reactar :solant systas pressure saceecs aneut a*! pounos per scuare f aca ano :s automatically : lose the valves aneg reactar coolant systes pressure sacaecs aceu SCO souncs per souare inen. The interlocu are provicec to prevent oversressuri:a-tien sf tne residua; neat reenval systes .nen reac.:r :solant systas =ressure is sign, pr. mary curing coeration. se founo mese inter'ocas ac:sstasia.
i In Jur "f at protect on review i .as costulateo cat d'*e samage ta electrica*
as;es ::w;4 cause seu valves is seen. To :ar ect 2 :s ce 3:: " : ant *as : ::cse:
- escving :o er ' es :ne es*,es notar :oerst:rs my cen W sa.ca : : wit :ressa*S.
w i'nce :n's .'!' 3reven: ce sostdatec " e :amage * :a :cen'.g =cta ,a!,es -e
EXHIBIT 3 - SSER 21 Peges 2-85 to 2-113 Dacember 1983 Task: Allegation No. 37 Q ATS No.: RV 83A41 8N No.: 83-169 (10/20/83) 1
-l Characterization The solid state protection system (SSPS) relays that initiate closure of RHR letdown isolation valves 8701 and 8702 perform no safety function, reduce the reliability of the RHR system, and cause a potential for RHR pump damage.
Therefore, these relays should be removed.
Implied Significance to Plant Design, Construction, or Operation The RHR letdom line contains two isolation valves (8701 and 8702) in series that are normally closed during power operation. These valves are opened when entering Mode 4 (hot shutdown) to allow the RHR pumps to take suction from the reactor coolant system (RCS) to the RHR heat .exchangers for decay heat removal.
O 8eth vaives 8701 and 8702 are interiocked so that they iii automaticaiiy ciose to isolate the RHR system from the RCS if RCS pressure increases to a pre-de-tennined setpoint. This automatic isolation function (performed by the West-inghouse designed SSPS) is provided to protect the low pressure RHR system piping,from higher RCS pressures. Isolation is accomplished using a " fail safe" design (i.e., on a loss of SSPS power, valves 8701 and/or 8702 will automatically close). The concern here is that a loss of SSPS power will cause an unwanted (spurious) isolation of the RHR letdown line causing event-ual RHR pump damage assuming no operator action.
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Assessment of Safety Significance Isolation of the low pressure RHR system from the high pressure RCS must be provided to prevent RHR system overpressurization that could potentially re.
sult in a loss of coolant accident (LOCA) outside containment. Therefore, RHR letdom line isolation is a safety function. The SSPS, including relays, which performs this function is safety related and designed to Class IE re-qui rements. Both valves 8701 and 8702 are provided with this automatic clo-sure interlock on increasing RCS pressure so that a single failure will not prevent RHR letdo m line isolation. Therefore, the relays used to initiate closure of these valves are essential and should not be removed.
O Oiverse iadicatioas aad aiarms ar arovided ia the coatroi room (inciudias a RHR system low flow alarm to be install,ed during the first refueling outage) to allow the operator (s) to assess RHR system status and to alert them to potential system degradation. Technical Specification surveillance require-ments at Diablo Canyon include periodic verification of RHR system flowrate when using the RHR letdo m line. In addition, diverse means of decay heat removal (i.e., reactor coolant loops) can be readily made available should the RHR letdom line be' inadvertently / spuriously isolated.
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i Based on the above, the staff concludes that the existing SSPS design regard-ing RHR letdown line isolation is acceptable.
e Staff Position This allegation does not involve considerations that question plant readiness for power ascension testing or full power operation.
Action Required None.
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.t Task: Allegation or Concern No. 38 ;
2 O ATS No. RV83A47 BN No.83-169 (10/20/83) t Characterization PG&E is ignoring evidence that the spurious closure of a motor operated valve is not " impossible."
Implied Significance to plant Design. Construction or Operation The allegation suggests the licensee has not satisfactorily analyzed operational data.
Assessment of Safety Significance The alleger has described operating events at the Diablo Canyon facility and other Westinghouse facilities during which motor operated valves in the residual heat removal (RHR) system have, upon spurious initiation of their automatic closure circuitry, moved from the normally open position (for RHR operation) to the closed position, these presenting the potential for damage to RHR pumps.
The staff has examined in denth the licensee's actions in response to an event invnivino the sourious initiation of RHR motor operated valve closure as well as the concerns exoresi.ed hv the allenar reaarding the potential for such event, and concluded that timely evaluation and corrective measures were taken to preclude O- 2.
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repetition of such conditions. (See Allegation or Concern Nos.: 42 & 44).
Staff Position The staff's position regarding the interlock cricurity which causes automatic clost of the RHR isolation valves is duscussed in Allegation or Concern No. 45. It does appear that the licensee is giving proper attention to the spurious closure of the valves in question.
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Task: Allegatten #39 ATS No.: RV 83A47 BN No: 83-169 (10/20/83)
O, Characterization There is no control room annunciation provided to alert the operator (s) when the RHR letdown line has been isolated Juring Modes 4, 5, and 6 (hot shutdown, cold shutdom, and refueling respectively).
Implied Significance to Plant Design, Construction, or Operation During modes 4, 5. and 6 the residual heat removal (RHR) system is aligned in the shutdo m cooling mode by taking suction from reactor coolant system (RCS) loop 4 through the RHR letdown line to the RHR pumps. The RHR pumps direct flow through the RHR heat exchangers for decay heat removal via the component cooling water (CQl) system, and then back to the RCS cold legs. There are two f solation valves (8701 and 3702) in series located in the RHR letdow line.
O, if one of these vaives shouid inadvertentiy ciose RHR ,um , suction wouid be lost. The, concerns here are loss of decay heat removal capability and poten-tial damage to the RHR pumps. It has been estimated that pump damage could occur as soon as 10 to 15 minutes following a spurious isolation of the RHR letdom line.
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Assessment of Safety Significance For those modes of operation where RHR shutdown cooling is used, only one RHR train oj; one filled reactor coolant loop is necessary to provide sufficient decay heat removal capability. The Diablo Canyon Technical Specifications require either two RHR trains be operable and/or two filled reactor coolant loops be available in order to allow for single failures. If both RNR trains are being used and the RHR letdown line becomes isolated, the operator (s) would have sufficient time to fill at. least one coolant loop (assuming no loops are filled) for decay heat renoval. Control room indications of loss of decay heat removal include RCS temperature RHR system flow, and RHR pump discharge pressure. With less than the required number of reactor coolant O,
loops and/or RHR trains operable, the Technical Specifications require in-mediate corrective actions to return the required loop / train to operable sta-tus as soon as possible.
Indication provided ir. the control room of RHR letdown line isolation in-cludes position indication for valves 8701 and 8702 (red and green position status lights next to the valve control switthes on the main control board) as well as RHR system flow, pressure, and pump status information. Although l
these features do provide a capability to assess RHR system status, the staff has recognized the need for installation of a RHR low flow alarm. Accordingly, b
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the licensee is required to install a RHR low flow alarm during the first re-fueling. This requirement is documented in Supplement No.13 of NUREG-0675
" Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2." The staff has concluded that the existing con.
trol room indications and procedures are sufficient to assure adequate decay heat renoval in the interim.
Staff Position This allegation does not involve considerations that question plant readiness for power ascension testing or full power operation.
Action Required None.
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Reactor Systems Branch O Task:
Allegation #40 ATS No.: RV83A 47 BN No.: 83-169(10/20/83)
Characterization The question raised was with regard to whether or not the single RHR pump suction line from the RCS hot leg meets safety related standards.
The newer PWRs are designed with redundant RHR pump suction lines from the RCS hot legs.
Implied Significance.to Plant Design, Construction or Operation The RHR suction line from the RCS hot leg in Diablo Canyon contains two t
Q isolation valves (8701 and 8702) in series that are normally closed during power operation. When the RHR system is operated as a part of the ECCS, the RHR pump suctions are aligned with either the RWST or the
' containment emergency sumps. The RHR suction line from the RCS hot leg is only used during modes 4 (hot shutdown while RCS temperature is less I
than323*F),5(coldshutdown)and6(refueling). A postulated failure l
t of either isolation valve (8701 or 8702) in the RHR suction line to open during plant shutdown could prevent the plant from reaching a cold shutdown condition.
Assessment of Safety Significance .
In the Diablo Canyon SER Supplement No. 7, the staff states that the single RHR suction line from the RCS hot leg was acceptable. The staff O- coaciu=iaa > 6 =.e aa ta <= iia ias:
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I (1) Th3 Diablo Cany:;n design has a safety related Auxiliary Feedwater System (AFWS). The condensate storage tank is the primary source of AFW with about an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> water supply. In order to ensure the I
capability to remove heat via the steam generetors for extended periods, provisions have been made to connect the raw water reservior to the suction line or the AFW pump. This will provide enough AFW to allow an additional 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of steam generator operation for both units.
(2) The licensee has indicated that the combination of a mechanical failure of the itHR isolation valves and an earthquake results in a risk of about 10% of the core melt risk from all causes calculated in the Reactor Safety Study.
Branch Technical Position RS8 5-1 was not approved at time SSER No. 7 for Diablo Canyon was issued. In accordance with the implementation O schedule of BTP RS8 5-1, the Diablo Canyon Units are considered class 2 Table 1 of plants which are not required to fully implement this BTP.
BTP RS8 5-1 shows what is necessary to be implemented for class 2 plants. A single RHR suction line from the RCS' hot leg is considered acceptable for a class 2 plant as long as a single failure could be corrected by manual actions inside or outside of containment, or the plant could be returned to hot standby until manual actions (or repairs) areaccomplished.(page5.4.7-16of5,RP5.4.7). Also, BTP RS8 5-1 for class 2 plants requires that the RHR isolation valves have independent, diverse interlocks to protect against one or both valves being open during an RCS pressure increase above the design pressure of the RHR D- 2-94
system. There was no assessment of the degree of complianco of tha ,
Diablo Canyon design against BTP RSB 5-1 documented in any staff SSER.
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Based on the above facts, the staff evaluation of the subject allegation is as follows:.
The RHR suction line from the RCS hot leg is not required for ECCS functionability. .The RHR pumps take sucticn from RWST or containment emergency sumps, and serve the ECCS function during a LOCA. The suction line from RCS hot leg is used only for modes 4 ( 323*F),5and6. GDC 34 of Appendix A to 10CFR 50 requires that the decay heat removal safety function should be accomplished assuming a single failure. THe Diablo Canyon design complies with this requirement by having a RHR system plus a safety related AFWs (with steam generators and atmospheric steam dump valves). Based on the above, we conclude that the Diablo Canyon design meets GDC 34 and the intent of BTP RSB 5-1. The current RHR design is adequate for safe operation at Diablo Canyon, 1
The staff is currently conducting a reevaluation of the adequacy of the decay heat removal system design of all LWRs. This work is being performed as an Unresolved Safety Issue (TAP-A-45), and the Task Action Plan is projected to be complete within one year. Diablo Canyon, will be subject to any new requirements that may result from the work of TAP A-45.
Staff Position .
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This al' legation does r.ot involve considerations that question plant I
readiness fo'r power ascension testing or full power operation.
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Action Required None l
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Task: Allegation or Concern Ns. 41 ATS No.: RV83A47 BN No.: 83-169 (10/20/83) i O Cnaracterizat4en The power source of certain relays is not shown on certain drawings and this caused an operational problem, the failure (closure of RHR isolation valves).
Implied Significant to Plant Design, Construction or Operation Sufficient infomation may not be readily available to plant operators or maintenance personnel regarding the effects of deenergizing certain portions of plant safety related systems causing unexpected plant behavior which, in turn, can be of safety concern.
Assessment of Safety Significance Preliminary examination by the staff of the drawings and circuit schematics of concern to the alleger revealed that a detailed review of several drawings, circuit diagrams, and logic diagrams is necessary to fully comprehend the effect of the removal of power to the SSPS output relays. This removal of power can O
cause this RHR hot ie, suction vaives to ciose, resoiting in ,otencia, dama,e to safety-related RHR pumps, and a condition which may not be detectable by operators in the control room.
The alleger's specific concern is that removal of power to a portion of the SSPS on September 29, 1981 did result in unexpected closure of the RHR isolation valves with an RHR pumo running. (See Allegation or Concern No. 44).
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i Examination of facility records and discussions with licensee personnel know-legable of the circumstances of the event of September 29, 1981 revealed the following. In preparation for " trouble-shouting" the cause of apparent power supply difficulties in a portion of the SSPS, a "... Clearance Request and Job Assignment Sheet" was prossed and approved, as required by plant administrative procedures, to authorize such activity. Subsewent disablement of the power supply (removal of a fuse) caused automatic closure of the RHR isolation valves !
thus interrupting RHR system flow. Initiation of the closure of the RHR valves had not been anticipated by either the operation supervisor or maintenance personnel involved in the activities Operations personnel did respond to the unpected closure of the RHR isolation valves in a resonably timely manner such that the RHR pump continued to operate without flow for approximately five minutes. The pump substained no detectable damage in this instance.
O It was also revealed in discussions with licensee personnel that a simplified sketch of the RHR initiation circurtry has been constructed to clarify inter-actions between various components previously shown only en individual plant drawings and circuit diagrams. The construction of this simplified sketch has resulted in a much improved understanding of the cricuitry by the plant's maintenance as well as operations personnel.
Staff Position Activities involving maintenance or texting of systems associated with the nuclear plant should be planned in advance sufficiently to anticipate the reasonse of such systems when these activities are undertaken. Adequata preplanning measures in this regard appear not to have been taken by the licensee in this instances. However, measures have been taken by the licensee to preclude a repitition of the specific occurrence in this instance.
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No further specific action'is required. The staff will focus attention in this inspection program to the preplanning and procedural precauticas established by the licensee in carrying out maintenance and testing activities of a similar nature in the future.
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Task: Allegation or Concern No. 42 O
ATS No.: RV83A47 4 BN No.: 83-169 (10/20/83) d Characterization Licensee management was unresponsive to recommendations to prevent spurious closure of the isolation valves on the residual heat removal (RHR) system.
, Closure of the valves disables operation of the RHR system for decay heat j[ removal.
Implied Stanificance to Desian, Construction or Operation O .
4 iack of aanropriate respease bv the iicensee. couid indicate aa undesirabie 4 level of management sensitivity toward employee concerns and recommendations aimed at improving operation of the reactor facility.
Assessment of Safety Stanificance Facility records were examined, discussions were held with facility personnel, and observations were made by the staff.
Periodic discussions were also held j
l with the alleger. Since the alleger's concerns had been examined by Region V l
inspectors previously, reports of prior inspections were reviewed and discussions were held with Region V inspectors relating thereto. In addition to the specific concern (or allegation) characterized above, other concerns of 4
the alleger, as discussed below, were also examined.
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The alleger had documented concerns regarding spurious closure of the RHR isolation valves because of certain steps in an emergency operating procedure
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related to safe shutdown from outside the control room. The licensee's response consisted of the initiation of a nuclear plant problem report, and investigation of the alleger's concern. The licensee's resolution to the concern was to revise the emergency procedure.
A design change request (DCR) authored by the alleger addressed the alleger's more general concern of potential for inadvertant closure of the RHR isolation valves.
A revision to the DCR was subsequently initiated by the alleger providing the Licensee Event Reports (LERs) of other facilities relating to instances of RHR system disablement due to spurious closure of the isolation valves similar to those which were the subject of the alleger's concern.
(]) The alleger's preliminary evaluation of the DCR determined that the requested change involved an unreviewed safety question requiring prior NRC approval in accordance with 10 CFR 50.59. The OCR is still under consideration by the licensee's engineering department, the plant operating department and Westinghouse.
Preliminary discussions have been held between the licensee, Westinghouse and the NRC staff relating to an informal proposal by the licensee (supported by Westinghouse) to remove the RHR interlock circuitry from the Diablo Canyon facility. The proposals and actions required to resolve this DCR are still open.
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l The staff determined that other procedural changes have been made by the O " " '" '" ' " ' "d* '" "' '"' """ ' " ' " "
spurious actuation of the interlock circuitry.
l The staff also reviewed a concern documented by the alleger in a meno in April 1981 to plant engineering regarding reactor coolant pump bearing oil level annunciators. In postulating a tube failure in the lube oil heat exchanger, the view was expressed that an incorrect alarm response procedure may lead the operator to take improper action. Written acknowledgement of the alleger's concern was provided by a plant engineer in June 1981, indicating that the procedures manual was being revised to resolve the concern. The alleger
, observed approximately eight months later, that no change to the Plant Procedures Manual had been made. The alleger documented this observation by an additional memo. .The same plant engineer who had previously responded to the alleger responded to this memo. The engineer explained that the Plant Manual had been O the subject of an extensive revision effort for the past year and all changes resulting from this effort were to be incorporated into the Manual "... in one major revision" which would be published "... definitely prior to low power physics testing." A major revision, which included the alleger's initial conuent, was subsequently made to the Manual in September 1983.
During the intervening period between the time of the alleger's second memo and implementation of the major revision to the Manual NRC resident inspectors pursued the alleger's concern with licensee personnel. In response, the licensee implementing a temporary change to the specific procedure of concen. This l
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O, temporary change was acconiplished by the issuance of a Procedure ON-THE-SPOT Change in early 1983.
Staff Pesition A period of approximately 2 years appears to be excessive in attempting to resolve the RHR concerns of the alleger. The issue is not yet fully resolved.
However, unusual circumstances did exist in that resolution of the alleger's concerns regarding the RHR system and his specific recommendation to remove the.
interlock circuitry involve substantial safety analyses by the licensee, as
, well as NRC staff review and approval. In the interim, procedural changes had been implemented by the licensee which had substantially addressed the concern of the a11eger. A similar period, approximately 21/2 years, to formally address the alleger's concern regarding the accuracy of an annunciator response O <s ,,ocedure aiso, under no mal circumstances, a,, ears excessive. In this instance, however, the unusual circumstances of a ma,ior revision to the procedures manual was in progress.
It is the judgment of the staff that there is not a prevailing attitude by i
licensee management which in itself discourages employees from expressing concerns or making recomunendations for improvement in facility operations.
Action Required The Region V staff will give particular attention in its ongoing routine inspection program to evaluate the performance of licensee management in this area.
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Task: Allegation or Concern No. 43
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l ATS No. RV83A47 BN No.83-169 (10/20/83)
' Characterization The loss of the residual heat removal-(RHR) system on 9/29/81 due to unplanned closure of the RHR isolation valves was an event which should have been reported to the NRC in accordance with 10 CFR 50.72. The licensee's failure to make such a report was in violation of NRC regulations.
Implied Significance to Design. Construction, or Operation The failure of the licensee to report this occurrence, would indicate a deficiency in the licensee's management control systems to provide adequate review and reporting of events to the NRC.
Assessment of Safety Significance The circumstances associated with the event were examined by review of facility records and discussions with licensee personnel.
The loss of residual heat removal capacity during a time when significant fission product decay heat is present in the core would have safety signifi-cance. In this particular instance, fuel had not been loaded into the Diablo Canyon Unit 1. Therefore, no fission product decay heat was present and loss of RHR capability had no actual safety significance.
O.
. 2-104
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/
The intent of then applicable provision 10 CFR 50.72 of the NRC regulations /
O was to insure that holders of operating licenses for power reactors report promptly by telephone to the NRC Operations Center significant events such as those which involve intitiation of the licensee's emergency plan; the nuclear reactor not be in a controlled or expected condition; fatality or serious injury or radioactive contamination of personnel; or acts which seriously threaten the safety of the reactor or site personnel.
The event in question was reviewed by the staff and it was detennined that this event is not required to be reported in-accordance with 10 CFR 50.72. Licensee representatives did indicate that an infonnational report of the event was to be made in writing to the NRC.
Staff Position The staff concludes that the event did not meet the reporting requirement of 10 0 CFR 50.72. .
Action Required None l
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CL 2-105
Task: Allegation or Concern No. 44 O ATS No. RV83A47 I BN No.83-169(10/20/83) l Characterization The licensee failed to properly process a Nuclear Plant Problem Report.
Implied Significance to Design. Construction, or Operation The allegation, could indicate a weakness in the implementation of the licensee's Quality Assurance Program for Operations.
Assessment of Safety Significance O
The Nuclear Plant Problem Report (NPPR) is.the document used at Diablo Canyon to record events such as significant equipment failures and operational problems.
The NPPR form becomes the record of the identification of a problem, its evalua-~
tion, and the action taken to correct and prevent recurrence.
On September 29, 1981, inadvertent closure of the residual heat removal (RHR) system isolation valves occurred while the RHR pump No. 1-1 was running. The alleger's concerns are that the NPPR which was initiated following this event was not processed properly in that it was, " signed off as complete without any plant management review... classified as 'non-reportable' and without any follow-up action such as an RHR pump inspection or investigation into the cause of the event."
O 2-106
l The processing of the NPPR was assessed through an examination of facility records; discussions with facility personnel (including all those persons whose O identity was provided by the alleger) and the alleger; and observations by the inspectors.
l The NPPR record in question was examined, Itwaswrittenon9/21/81andclosed on 10/5/81.
The resolution of the three issues are as follows: !
Signed-off without any plant management review l The inspector detemined that licensee management, including the. plant super-intendent and operations supervisor, were involved in the review and evaluation O of the NPPR.
! The alleger's concern included the fact that when he examined the NPPR (after if had been completed) there was no signature to indicate the results of management's evaluation of cause and corrective action (s) taken. The alleger had called this discrepancy to the attention of a QC superytsor, who obtained the proper signature on the NPPR. When the NRC inspector examined the NPPR record (in December 1983) the Operation Supervisor's signature was found on the document.
It was undated. In discussions with the NRC inspector, the Operations Supervisor stated he may have signed the NPPR after it had been closed, but he could not accurately recall the circumstances.
- NPPR classified as "rion-reportable" O .
2-107
The inspector verified that the NPPR was in fact classified as "non-reportable" j O by iicensee monasement. The ciassification is coasidered appropriate by the staff and is addressed in Allegation or Concern No. 43. 1 l
- M followup action was taken into the cause of the event l
.. )
The NPPR in'di ted that a revision.to operating procedures was necessary to
/ prevent recurrence of the event,'and that such a revision had been implemented.
Facility records indicate that the NPPR relating to the event was the subject of review by the On-Site Safety Rev'iew Group (OSRG) on two occassions--
/
October 19, 1981 and November 24,1981. - On October 29, 1981 the OSRG observed that the operating procedures had'been' changed, and that a proposed ch,ange to
~
remove the RHR isolation valve initiating circuitry had been proposed. The latter, it was determined, was a Design Change Request (DCR) which had been O ._
initiated by the alleger (see Task Allegation or Concern No. 42). The OSRG determined that it would review the even't further during a subsequent meeting.
- On November 24, 1981 the OSRC directed,that an operational test of the RHR pump be conducted, and that the DCR not be ~ approved since it would provide less protei: tion for RHR over pressurizatioit than presently existed.
I
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Staff Position '
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- *v .
s -
The NPPR was properly processed and subsequently reviewed by the OSRG.
1 Action Required .
None. y' 1 l
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2-108.
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0 Reactor Systems Branch TASK: Allegation #45 ATS NO.: RV 83A47 BN NO.: 83-169 (10/20/83)
Characterization:
Section 5.5. of the Diablo Canyon FSAR describes the autoclosure inter-lock for the RHR Suction line isolation valves (8701 and 8702).
Section 3.4.9.3.a of the Diablo Canyon Technical Specifications requires I
power to be removed from these isolation valve operators during modes 4 (Hot shutdown when RCS cold leg temperature is less than 323*F), 5 (cold shutdown) and 6 (refueling). This requirement defeats the function of autoclosure interlock for the valves.
Implied Significance to Plant Design, Construction or Operation As the result of Technical Specification Section 3.4.9.3.a. the iso-lation valves (8701 and 8702) will be left in an open position with O
power removed during low pressure / temperature operation of the plant.
The automatic closure interlock to these isolation valves causes them to lose their design function. This will result in a situation in which there is in sufficient isolation capability feature to prevent an intersystem LOCA between the high pressure RCS and the low pressure RHR system.
j Assessment of Safety Significance i
l Section 5.5 of the Diablo Canyon FSAR states that during low pres-i sure/ temperature operation, the isolation valves (8701 and 8702) between the RCS and the suction of the RHR pumps are interlocked with a pressure signal to automatically close the valves whenever the RCS pressure increase above approximately 600 psig. Section 3.4.9.3.a of the Diablo Canyon Technical Specification requires the RHR system isolation valves 2 109
?' ~ ._. - . - -
_ - _ . _ _ _ - _ . -- . _ . _ - - - = .-- -- .
- O l (8701 and 8702) to be open with
- power removed from the valve operators while the positive displacement charging pump is in operation. The applicability of the T.S. is during mode 4 when the temperature of any RCS cold leg is less than or equal to 323'F, mode 5, or mode 6 with the reactor vessel head on this Technica1' Specification requirement defeats the automatic closure interlock function as designed.
Power removal from valves 8701 and 8702 while the RHR system is operat-ing was required by the staff as the result of a meeting with the licensees on RCS low temperature overpressure protection (1. TOP) and RHR pump protection concerns. Since the Diablo Canyon design has only one RHR suction line from the RCS, a spurious automatic closure of the
~
iso 1ation valve would result in loss of RHR pump suction f1'ow and would result in a RCS pressurization as a result of the loss of letdown flow.
However, there was no documentation (SS'ER, letter or meeting minutes) of the staff's basis for requiring power removal from those isolation valves during modes 4, 5 and 6.
IntheDiabloCanyonSERSupplementNo.13.section6.3.(ECCS), dated April 2,1981, the staff concluded that the licensee should be required to provide an alam to alert the operator to a degradation in ECCS (during long tem recirculation). A low flow alarm was stated to be an acceptable method to satisfy this concern and the staff indicated that -
an alam should be installed at the first refueling' outage. Until then, procedures and dedicated operators were tc be implemented during long
- O. tem recircuistion to -nage and -nitor ECCS perf-ance. There was no documentation to indicate that the licensee comitted to this 2- 11n
O' )
staff position, nor was this staff position included in the Diablo Canyon low power license. SRP 5.4.7 (BTP RSB 5-1) requires an autoclosure interlock on the RHR suction line isolation valves. Without power to the valve operators, the autoclosure function is defeated.
Based on the above facts, the staff evaluation of the subject allegation is as follows: '
. Without power to the isolation valve operators, the plant design does not confom to BTP RS8 5-1 Position 8.1.C. for the requirement of autoclosure interlock. By having power available to the isolation valves during shutdowns ensures an event V (intersystem LOCA) will not O *'"'"""'***'''''"'"'''***"*'""'
during a return to power.
With power on the isolation valves, a spurious closure of the isolation valves would result in a loss of suction flow to the RHR pumps. Howev-er, the low flow alarm discussed in SSER No.13 would enable rapid operator detection and mitigation. The licensee has infomally indicat-ed that a minimum of 10 minutes without adequate suction pressure would be available without pump damage. Also, there are numerous indications available to alert the operator to improper RHR valve alignment ( A list is provided in staff evaluations to allegation No. 37 and 39). -
Staff Position To implement the staff position stated in SSER No. 13, the installation of a low flow alarm for RHR pump protection is being considered as a 2-111
1 O license condition in the Diablo fanyon full power license $
Additionally, it is the staff position that power be available to the l RHR MOVs when in a shutdown condition. However, there is a question as to when these requirements should be implemented. If the low flow
! alam were not installed until the first refueling outage, r'einstating power to the RHR MOVs in the meantime would result in the autoclosure interlock being enabled to provide protection against intersystem LOCA However, the chances of spurious autoclosure and consequent loss of RHR suction pressure (without the low flow alarm) and of an overpressure event would he increased. If power restoration to the RHR MOVs were not implemented until the low flow alars is installed at the l first refueling outage, the chance of loss of RHR suction in the interim is reduced but there is a possibility of an intersystem LOCA. To -
detemine which option results in the safest operation of the plant, the staff considered the following:
- 1. During the first cycle of operation, plants operate more frequently on the RHR system as a result of maintenance, testing and training
, requirements for a new plant. Thus, the period of vulnerability to l
a spurious RHR suction MOV closure may be greater than in subse-quent cycles.
MOVs left open. The operator would have to shut at least one MOV to continue the plant startup.
O ,
first MOV can fail in either of two ways by either the "open pemissive" interlock failing along with the operator reinstating power to the valve, (it is required to be de-energized) then attempting to open the valve. The second mode of failure would be for the valve to rupture in such a way that flow between the two ,
systems occurred. Both of these failure modes are judged to have an extremely low probability. However, the consequences of an intersystem LOCA could be severe.
4 There have been many occasions .of spurious RHR suction valve closures on operating plants. Tnis has resulted in not only a loss of decay heat removal, but also an overpressure event due to the loss of the letdown flowpath.
O <
ACTION REQUIRED -
Based on the above factors, the staff believes the best course of action is to continue the current technical specification for power to be removed from the RHR MOVs during Modes 4, 5 and 6 until the low flow alam is installed. However, the staff position that would permit the licensee to wait until the first refueling outage before installing the low flow alam was taken over two years ago. Staff will puruse with the licensee a comitment to a schedule for accomplishing this Installation at the earliest possible time. In the interim, until the low flow alarm is installed, the staff believes that strict administrative controls should be developed and implemented to ensure O that Movs 87o1 and 87oz are ciosed with power remove. dur4ng piant startups when RCS pressure is above the RHR design pressure.
2 113
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O O I
l TO TO l CONTAINMENT CHARGING SPRAY PUMPS d i d '
TO RCS COLD LEGS - ,
IE2 T '
RHR HEAT
(( C AINMENT EXCHANGER 1 '
p
' ' ' FROM RCS LOOP 4 H ESC )( M
- 1&2 7 E
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M FROM RWST TO RCS COLD LEGS :
3&4 M , i -
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" CONTAINMENT i RHR HEAT SUMP EXCHANGER 2 KEY:
l HE Pr8 1 r 1 r O LLY '
$t TO CONTAINMENT TO SAFETY INJECTION (AT POWER)
MOTOR-OPERATED SPRAY PUMPS .'
j VALVE l NORMALLY OPEN (AT POWER) ,
- MOTOROPERATED VALVE i AIR-OPERATED VALVE i I I SCHEMATIC DIAGRAM OF j M CHECK VALVE RESIDUAL HEAT REMOVAL SYSTEM
--__ __ _ -- .. - ~ _ - __.____ _ - - - - - . ...-- -- - _ - - - - - - _ _ _ _ .
1 l i l
- EXHIBIT 5 - TECHNICAL SPECIFICATIONS, SECTION 3.4.9.3 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 The following overpressure protection systems shall be OPERA 8LE
l a. RHR system isolation valves 8701 and 8702 open with power removed i
from the valve operators when the positive displacement charging pump is in operation, and
- b. Two power operated relief valves (PORVs) with a lift setting of less than or equal to 450 psig, or
- c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.07 square inches.
l APPLICA8ILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 323*F, MODE 5 and MDDE 6 with the reactor vessel head on.
j ACTION:
' O isoiation vaives or secure the positive dispiace ent chareiac .. '
b.
\
With one PORV inoperable, restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through a 2.07 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- c. With both PORVs inoperable, depressurize and went the RCS through a 2.07 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- d. In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs er vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- e. The provisions of Specification 3.0.4 are not applicable.
O DIA8LO CANYON - UNIT 1 3/4 4-32
"'^c' O " c '^"' 5'5""
SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
- a. Performance of a CHANNEL FUNCTIONAL TEST on the PORY actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPE *A8LE.
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.
- c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORY is being used for overpressure protection.
- d. Testing pursuant to Specification 4.0.5.
- e. Verifying the RHR isolation valves 8701 and 8702 are opened with power removed from the valve operators when the positive displacement charging pump is in operation at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- when the vent (s) is being used for overpressure protection.
0(
"Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
l l
DIABLO CANYON - UNIT 1 3/4 4-33
d EXHIBIT 6
. ;, + . .
'. PGWE O ' - - ' ' ~ ~ " -
IO',*mE SENIOR VICE PRESIDENT x ,
l Rs Lstets er j sums, Nuclear Facilities , ,
I April' 29, 1976 s , ,
To: PGandE ENGINEERS, TECHNICIANS AND OTHERS DIRECTLY INVOLVED IN 'tWE, COMPANY 'S NUCLEAR PACILITIES , -
..f This letter is written to the members of Pacific Gas and Electric Company who are, or have beu, directly involved in the design, construction, and operation of our nuclear ,
facilities. Its purpose is to reaffirm the Company's strong l oosmitment to the protection of its employees and the general public against any unsafe situation with respect to these facilities and, further, to assure that you have every oppor-i tunity to communicate freely to your company any views you might have on the safety of nuclear facilities. ,
We believe that you appreciate your right and obliga-O ( --
tion to express yourselves on matters of safety and that you have the dedication and individual initiative, insofar as your l
- responsibilities are concerned, to see that our nuclear facilitie I are designed, constructed, and operated in a safe manner. l Asnucleardevelopmenthasacceleratedthe$ehasbeen an intensifying discussion on nuclear issues. To give you added opportunity to ask questions or to express your views on any aspect of the safety of nuclear facilities, including those outside your own sphere of responsibility, we encourage you not only to talk to your supervisor, but also, if you wish, to any one of the following people who have recently been designated a review team to answer questions and to evaluate the views of any employee who wishes to express any concern whatever about the safety of nuclear facilities: ,
Company Telephon Wallace 3. Allen Director, Environmental Quality Department 1675 Clinton P. Ashworth Supervising Mechanical 3305 Engineer, Mechanical &
Nuclear Engineering Dept.
Alfred W. Medcalf Steam Ginneration Engineer, 1292 0 - == a a r=**== a P*- l
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Page' 'hto - $
ADril 29, 1976 . -
a
..i-t '-
The individuals on this review team have broad nucisar knowledge, and they have full latitude for drawing W the expertise of others and stimulating action if it is necessary.
You may contact any one of them and arrange for a discussion with him alone or he will arrange for a joint meeting with other review team members. .
J. D. WORTHINGTCH ', ,.,',
JDW:BJD ' *'
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cc: Officers ~
Department Reads '
.' ,' ?
Division Managers
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, EXHIBIT 7 FACIFIC GAS AND ELECTRIC COMPANY STATION CONSTRUCTION DEPARTMENT DIABLO CANYON PROJECT O
July 25,1930 NDiogANDUM 70: DIABLO CANYON GENERAL CONSTRUCTION FIRSONNEL RE: Personnel Participation
. To provide you with additional opportunity to participate in the Project, a new avenue of connuaication is now open. There are three locations where you can suhait written questions and/or suggestions. Boses have been located la the following places: The Project office asil room, j General Construction tool room and the Welding Crew change room.
You are all invitsd to ask any questions and to offer any suggestions you may have regarding the F oject. This includes anything from project administration, policies and rules to technical concerns such as plant 4
design, construction and operation.
O I se interested in your suggestions for improving the efficiency of our work at Diable including safety, quality, productivity and cost control.
I an equally interested in answering any questions you any have now or later about all aspects of working on this Project or for PG&E in general.
Your questions and suggestions will be responded to as fast as possible to you directly or on the bulletin boards if no asas is given. If your suggestions appear to deserve consideration and/or award, via the formal Employee suggestion Plan you will be advised.
I as looking forward to your participation, questions and suggestions.
l R. D. Et:1er Project Superintendent l
t O.
i
e es esie a v. m EXHIBIT 8
.,PGwE -
, prget tortaa-coespaNY USES ,
VICE FRESIDENT
{ SEs",., NUCLEAR POWER GENERATION
( Pn.E No .
Rd Lavetet eer aussect Nuclear Facilities August 7, 1980 4
TO: PCandE ENGINEERS. TECHNICIANS AND OTHERS DIRECTLY INVOLVED IN THE COMPANY'S NUCLEAR TACILITIES -
This letter is written to the seabers of the Pacific Gas and Electri Cotspany who are, or have been, directly involved in the design, construc-tion, and operation of our nuclear facilities. Its purpose is to again reaffirm the Company's strong comitment to the protection of its employees and the general public against any unsafe situation with respect to these facilities and, further, to assure that you have every opportunity to coussunicate freely to your Company any views you might have on the safety of nuclear facilities.
We believe that you appreciate your right and obligation to r < =7 =4 == = 7 = a O t
>r 7 tv = =t r and individual initiative, insofar as your responsibilities are concerned,
a 4 at= =1
- to see that our nuclear facilities are designed, constructed, and operated in a safe manner.
Since the Three Mile Island accident, there has been an intensi-fying discussion on nuclear issues. To give you added opportunity to ask questions or to express your views on any aspect of the safety of nuclear facilities, including those outside your own sphere of responsibility, we encourage you not only to talk to your supervisor, but also, if you wish, to any one of the following people who have been designated a review team to answer questions and to evaluate the views of any employee who wishes to express any concern whatever about the safety of nuclear facilities:
Company Telephone Wallace 3. Allen Director, Environmental 1675 l
Quality Department Clinton F. Ashworth Supervising Mechedcol Engineer 3305 Mechanical & Nuclear Engineering Department Alfred W. Madcalf Sr. Nuclear Generation Engineer 1292 Nuclear Plant Operations O - Department
I, .
Page 2 August 7, 1930 The individuals on this review team have broad nuclear knowledge, and they have full latitude for drawing upon the expertise of others and stimulating action if it is necessary. You may contact any one of them and arrange for a discussion with him alone or he vill arrange for a joint meeting with other review team members.
J. O. SCHUYLER JOS(3096):ar cc: Officers Department Beads Division Managers O
O.
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4 . . v .mi NBn 9
, PGwE ma maa-cou-v vess O o -==ai
-- NUCLEAR PLANT OPERATIONS Diablo pnyon Power Plant
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REISSUED APRIL 14, 1983 Rt LETTER OF s m ac, Adherence to NRC Regulations '
Open Door Policy REVISED FEBRUARY 2, 1984 M" PGandE Quality Hotline February 5, 1982 TO ALL EMPLOYEES:
Since we have so many new employees at Diablo Canyon, it is a good time to reiterate PGandE's policy concerning adherence to governmental rules and regulations. All employees are urged to follow approved plant procedures and plant technical specifications. In doing so, the plant staff will meet all governmental rules and regulations, including those issued by the Nuclear Regulatory Comission (NRC).
Should any employee feel that such rules or regulations are not being followed, he should pursue the matter with his supervisor.
If the matter cannot be resolved through his supervisor, he should O pursue it via the "Open Door Policy", which has been previously stated and is repeated on the following page.
~
Should any employre feel that NRC rules or regulations are not being folicwed and plant management is not correcting the situation, NRC regulations allow him the right to discuss the problem with NRC inspectors (see NRC-3 notice posted on the bulletin board). Two NRC inspectors, Marvin Mendonca and Mark Padovan, are located in the office trailer adjacent to the Administra: ion Building (plant extensien 2439). Should an employee prefer not to talk to the onsite fiRC inspectors, the Region V NRC office address and tele-phone number is:
U.S. Nuclear Regulatory Comission Region V 1450 Maria Lane Walnut Creek, CA 94596-5368 (415) 943-3700 This memorandum does not authorize any employee to leave his assigned job or interrupt his work assignment to meet with NRC inspectors during work hours. It is preferred that, if necessary, employees contact NRC inspectors during nonwork hours. However, if a meeting must take place during work hours, an employee may request permission from his super-visor. Any employee who requests time to meet with NRC inspectors will be granted such time off at a scheduled, convenient time.
h.
l I
i TO ALL EMPLOYEES February 5. 1982 O
I OPEN 000R POLICY I would like to take this opportunity to reiterate our policy concerning the employee's right and obligation to inform manage-ment of safety concerns. Any employee who feels he has identified a concern or problem affecting plant safety or reliability should discuss this matter with his innediate supervisor. Following such 1
' a discussion, if the employee feels that his concern is not being adequately addressed, he should pursue the matter with progressively higher levels of supervision, up to and including the Plant Manager. '
While it is desirable to keep minor complaints from reaching this level of management I would like all employees to feel free to approach the Plant Superintendent and the Plant Manager with his concern if he feels it is necessary.
Voicing concerns regarding plant safety or operation will not be documented in the employee's personnel record and will never be used for any type of disciplinary action. I give my firm guarantee of that. ,
This policy also applies to any situation in which the employee feels he is being required to do any work which would jeopardize his personal safety.
O PGandE QUALITY HOTLINE I wish to call attention to signs which have been posted around the plant and project concerning the " Quality Hotline". This is a PGandE hotline and NP0 employees ar'e urged to use it if they deem it necessary. The hotline number is plant extension 3567 (541-7567 if calling from a public or outside telephone).
l i
j NO[
R. C. THORNBERRY r ll RCT(3350):ws xc RDEtzler LERosetta JDShiffer l
O.
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EXHIBIT 10 pea nwen .comerawv oens i 1
PACIFIC GAS AND EI.ECTRIC COMPANY O OFFICE OF TRE CNAIRMAN March 22, 1982
! TO: PGandE OFFICERS, ENGINEERS, TECHNICIANS AND OTHERS DIRECTLY INVOLVED IN THE COMPANY'S NUCLEAR FACILITIES 1
i -
This letter is to reemphasize the company's long-standing commitment ,to design, build, and operate saf a nuclear power plants and in achieving this commitment to l require all employees to practice fundamental honesty and to adhere to Nuclear Regulatory Commission ("NRC") rules and regulations.
This is.also to reemphasize that our communications .
l with the NRC must be open and allow a free flow of t
( information. We must be ever alert to any possible misleading or ambiguous statements made either in oral or O wfitten communications. Any such misstatements must be l corrected immediately upon discovery. Nothing less than full and open communication between the Compa'ny and the NRC i
can be tolerated.
l l In October 1977, PGandZ formalized its general policy j concerning employee conduct (Standard Practice 735.6-1).
The statement of policy establishes a company philosophy
- regarding work conduct emphasizing that: ,
"It is the policy of this company that employees shall at all times continue to practice fundamental honesty. Employees shall not, nor attempt to: deceive, defraud, or mislead the Company, other l employees, or those with whom the l Company has business or other l relationship; ... misrepresent the l Company or its empicyees: ... withhold their best efforts to perform their work
~
to acceptable standards; ... violate applicable laws; or conduct themselves at any time dishonestly or in a manner which would reflect discredit on the O Company.-
This policy is particularly important~to.all employees engaged in work concerning nuclear power.
O
__ _ _____ _ _ _ - - - - - . - . _ . , . . - . . . , . . . _ _ . . . . _ - - . . _ _ . - _ - , - . . . ..,.,m,,,. . , . _ -r ,e,.y .,e.,
l To All Addroosed Mcrch 22, 1982
()
In April 1976, Mr. J. D. Worthington, and again in 1980, Mr. J. O. Schuyler, issued a memorandum to all personnel involved in the company's nuclear power work which described a program to permit such personnel to discuss their concerns regarding nuclear power. The August 1980 letter stated that:
"[our] purpose is to again reaffirm the Company's strong commitment to the protection of its employees and the general public against any unsafe situation with respect to these nuclear facilities and, further, to assure that you hWve every opportunity to comununicate freely to your Company any views you might have on the safety of nuclear facilities.
"We believe that you appreciate your right and obligation to express yourselves on matters of safety and that you have the dedication and individual initiative, insofar as your responsibilities are concerned, to see
(]) that our nuclear facilities are' .-
designed, constructed, and operated in a safe manner.
"To give you added opportunity to ask questions or to express your views on any aspect of the safety of nuclear facilities, including those outside your -
own sphere of responsibility, we encourage you not only to talk to your supervisor, but also, if you wish, to any one of the following people who have been designated a review team to answer questions and to evaluate the views of any employee who wishes to express any concern whatever about the safety of
. nuclear facilities:"
We are proud that the application of these policies'of openness in finding and evaluating safety. issues led directly to the discovery by PGandE personnel of the " mirror image" error at Diable that otherwise might have gone undetected.
l I
To All Addressed March 22, 1982 O '
Recently, in February of this year, Mr. R. C.
l Thornberry issued a separate memorandum to Diablo Canyon j Power Plant employees which reiterated the Company's policy i
concerning adherence to government rules and regulations.
We must. strive for perfection in design, construction, ;
and operation of our nuclear units. To attain this goal, it is necessary that we all exercise our best efforts to
- resolve problems we encounter in our work. When problems are encountered, they must be immediately identified, clearly defined, and brought to the attention of your supervisor. This approach should facilitate the evaluation of, and formulation of timely and effective solutions to, any problem. Constructive recommendations are encouraged at all levels.
Our goal is to design, construct, and operate our nuclear facilities with full margins of safety and full compliance with NRC requirements. Strict adherence to the
) , above policies will provide added assurance that this goal .
will be met.
O ;f3 g-F. W. MIELKE, r. , B. W. SHACXZLFO cc: Officers Department Heads -
Division Managers All Concerned Personnel e
O l.
O O O
- 1avc a GUAUTY CONCEHN? HOTUNE.
Checklist for using if you have a the HOTUNE quality concerrt When calling the HOTUNE, please be prepared to provide the following
- Information:
Elf the answering service recorder is $MM~
- [,"jef,"ur
, call TELL YOUR SUPERVISOR.
EIGive as complete a description of the quality concern as you can, for l Instance: No Results . . .
l sea.= m TELL YOUR MANAGEMENT 1 -
- c ,s a u
! s a u.=in e,
Still no resulta - 0
- p 5 i
om u USE THE HOTLINE 5
--e l a., ,s u e ,
se.msn an.
OR STOP BY THE HOTLINE OFFICE 3 r
if possible, Identify individuals who may s be able to provide additional or PG&E EXTENSION corroborative information.
l 58 Give other methods by which the concern was expressed previously.
tas.nes.:sesp i.e 3567 v s .e I =*.r .e in. ra j E pFTIONAL identify yourself and a PUBLIC PHONE
- method by which we can reach you i a) to obtain niore information or $4j - 7$$[
i b) to provide results of the invest!0ation
! to you.
j Alternately,if you choose not to identify 2
yourself, you may phone the HOTUNE i again in several weeks during business hours and inpire as to the status of the quality concern.
i l NOTE: Every effort will be made to protect the identity of those callers h
} who identify themselves.
I e
i O t
EXHIBIT 12 *
. , 69 018 (7/01) ;
"
- g% t(21z
. HDGaurI]!!; Pacific Saa sad Bestric Company ,[," l
~
""'*^a ^a' aa^aa= a^ts a/2s783 O DEPARMENT PAGE*1 0F 42 TITLE VED y l "
IDENTIFICATION AND RESOLUTION 0F PROBLEMS AND MONCONFORMANCES (,[h MANAGER n==u===========:mummuune 1
- 1. SCOPE l This procedure describes the Nuclear Plant Problem Report (NPPR) system, the General Office Problem Report (GOPR) System and the Nonconformance Report (NCR) system utilized for the identification of equipment, material, procedural, or other problems and nonconforwences which occur at or affect a nuclear facility. The requirements for identification, documentation, routing, evaluation and detemination of cause, resolution, and corrective action are also discussed as they relate to problems and Nonconformances.
In this procedure the major factors involved in identifying, resolving and reporting problems are discussed in general tems, and detailed instructions and suggestions for completing and handling the forms are provided.
Q II. DISCUSSION The NPPR, 60PR, and NCR systems are used to ensure that conditions adverse to plant safety, such as failures, an1 functions, deficiencies, deviations, defective unterial and equipment, abnormal occurrences, discrepancies in work activities, and nonconformances are promptly identified, evaluated and corrected. The systens also ensure that the appropriate levels of management are notified and that independent ,
reviews are conducted as required. -
! These systems also satisfy several other objectives, including:
l A. Document significar.t problems involving only one department.
B. Document problems which are not necessarily directly involved with operating plant equipment, such as procedural problems and spare parts problems.
C. Document the determination of the cause and the specification of the corrective action (steps to prevent recurrence) as part of the review and resolution of major problems.
D. Provide a reminder that many major problems require reporting to the IstC and provide the required formal procedure specified in 10CFR21 to assure that such reporting is accomplished within established time limits.
O
=
NRiOI 1 l
l l
i
, , C -028 (7/h ) -
l WUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE
' REVIsloh W112 Q TITLE IDENTIFICATION AND RESOLUTION OF PROBLEMS AND NONCONFORMANCES DATE 6/15/83 PAGE 2 0F 42 E. Provide formal documentation of the final resolution of the Technical Review Group for most major problems.
F. Provide for initiation of reporting to the Nuclear Plant Reliability DataSystem(NPRDS).
The NPPR and GOPR systems also serve as a written method for requesting intradepartmental work on any item, whether it is safety related or not, thus insuring that:
A. There is no misunderstanding as to precisely what is being requested.
- 8. Scheduling and planning of work is facilitated.
C. Affected departments can keep track of whether the work has or has not been done.
D. A record of the work performed is provided.
E. A mechanism is provided for keeping management informed of the plant status and department activities.
O ~
F. Inforention is provided on coepleted work.
The NPPR. 60PR. and NCR systems are iglemented respectively by the use of Forn 61-4516 the " Nuclear Plant Prtblem Report". Form 69-025 the
" General Office Problem Report", and Forn 76-286 the "Nonconformance Report." The NPPR and GDPR forms describe the problem, document i
preliminary decisions which are made regarding its significance, serve as
' a work request, and document the basic resolution of the problem. The ICR form documents the findings of the Technical Review Group (s) and provides for follow-up on items which are judged to be nonconformances.
II
I. PROCEDURE
A. Nuclear Plant Problem Reports (NPPR)
- 1. General Requirements for Initiation of Nuclear Plant Problem Reports. Appendix C.
- a. Each plant and/or department may specify criteria for the l
initiation of Problem Reparts subject to the following minimum requirements:
1
- 1) Ary malfunction of permanently installed plant systems or equipment important to safety..
- 2) Any item which must be reported as a Nonconformance O- (see Section C of this procedure), because NPPRs initiate NCRs.
NPG01 2
- - .. . T_. . - - - _ - - - -
- . O-022 (7/h )
NUMBER PFAP C-12 WUCLEAR PLANT OPERATIONS DEPARTMENT W-112 REVIS10 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
- TITLE PROBLEMS AND NONCONFORMANCES PAGE 3 0F 42
- 3) Any incident involving equipment defects,'
malfunctions, personnel overexposum, effluent miease in excess of limits, or other similar
. occurrences which represents a violation of established procedures and/or may require a special report to the NRC (see NPAP C-11/G0AP W-111).
- b. When a plant is under construction or is undergoing major modifications, such that the plant staff is interfacing extensivelywithoutsidegroups(engineering, construction, consultants, contractors), it is permissible for the various parties involved to adopt mutually agreeable alternative problem reporting methods which meet or exceed the requimments of this procedure. Such alternative methods shall be described in appropriate Temporary Procedures, or in the QA/QC manual, for the ,
outside group.
l c. Arty NPO individual who identifies a plant problem may l initiate a NPPR, but as a minimum shall notify supervision i
(preferablyhisimmediatesupervisor). The supervisor
(, notified is responsible for notifying the supervisor in O \ charge of the work, if other than himself, and for
- assuring that a NPPR is initiated if mquired.
NPD-General Office initiated NPPRs shall be coordinated with and transmitted to the Plant Superintendent for further processing.
- 2. Guidelines for Conducting Required Work l
- a. Occasionally, plant management has to correct an item immediately without waiting for submittal and approval of l a NPPR. An example of this type of a situation would be finding a protection channel out of calibration in the unsafe direction, which is a reportable occurrence.
Obviously, the instrument should be immediately recalibrated if possible. The NPPR would still be prepared for the purpose of informin!1 plant management and others of what had transpired, and gnving them a chance to approve the accomplished work and decide whether additional corrective action is appropriate. Although the work may be couplete, the problem is not officially msolved until all sign offs are complete.
- b. Except as prohibited by 2.c. below, the decision as to whether an activity should be performed pending submittal I
of a NPPR is left to the supervisor in charge of the work.
O =- r* aica =a 11 aat 6 eaad ct d Priar *= r e $Pt =<
plant annagement approval include:
NPM1 3
~
1 l
6 % 318 (7/i.1 )
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER RN G.12
- 2 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS!0 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 l TITI.E PROBLEMS AND NONCONFORMANCES PAGE 4 0F 42
- 1) Activities which may involve an unrevieweid safety question as defined in 10CFR50.59, or a change in the Technical Specifications or Final Safety Analysis
, Report. An activity is considered to involve an unreviewed safety question if:
a) the probability of occurrence or the consequences of an accident or an1 function of equipment important to safety previously evaluated in the safety analysis report may be increased; or b) a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or c) if the margin of safety defined in the basis for any technical specification is reduced.
- 2) Activities which involve a design change.
O 3) *) cias aaacaara'=ian =P r
- rts aad ==teri=1= ia service without the appropriate reviews.
- 3. Design and/or Pmcedure Changes If the work involved on a NPPR results in a design and/or a procedure change, the administrative controls described in Nuclear Plant Administrative Procedures C-1 and E-4 shall be followed.
B. General Office Problem Reports (GDPR)
- 1. General Requirements for Initiation of General Office Problem Reports, Appendix D.
- a. Each group within Nuclear Plant Operations may specify i
criteria for the initiation of Problem Reports subject to the following minimum requimments:
- 1) Any General Office deviation and/or deficiency in work activities from approved procedures or from department implementing instructions.
- 2) Ariy item which is discovered which may require a specialreporttotheNRC(seeNPAPC-11/GOAPW-111).
' 3) Any item which must be reported as a,Nonconformance (seeSectionCofthisprocedure),becauseG0 prs initiate NCRs.
NPG01 4 1---___ _ - - - - _ -- _
l .
- (s028 (7/h) tfUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER fPAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE 112 REVIS10 O
10minCATiON AND RESOLUTION Or a^75 5/25/83 TITLE PROBLEMS AND NONCONFORMANCES 42 PAGE 5 _ 0F i
i
- b. When a plant is under construction or is undehoing major sodifications, such that the General Office staff is !
interfacing extensively with outside groups (engineering,
, construction, consultants, contractors) it is pemissible for the various parties involved to adopt mutually i agreeable alternative problem reporting methods which meet or exceed the requirements of this procedure. Such alternative methods shall be described in appropriate Temporary Procedures, or in the QA/QC manual for the outside group.
- c. Any NPO individual who identifies a General Office problem i may initiate a GOPR but as a minima shall notify supervision (preferably his immediate supervisor). The i
supervisor notified is responsible for notifying the supervisor in charge of the work. if other than himself, and for seeing that a G0PR is initiated. NPD-plant initiated 60 prs shall be transmitted to NPO-QC for further j processing. NPO plant initiated NPPRs requiring General 4
Office action shall be transmitted to the responsible
., Supervising Engineer for further action.
- 2. Guidelines for Reviewing G0 prs Supervisors are responsible for reviewing and conducting additional investigations as necessary to make one of the following determinations:
I
- a. The problem could be a potential Nanconformance as defined in Section C. unless a Technical Review Group has almady detemined otherwise. If it is a potential Nonconformance. the msponsible supervisor shall process the matter acccrding to Section C of this Procedure.
- b. The problem meets the requirements for a Problem Report as defined herein, in which case the supervisor shall process the problem in accordance with this Procedure.
- c. The problem is neither of the above. The supervisor shall dismiss the matter or handle it through normal work practices. No additional documentation is requimd. For l example, spelling errors, and other clerical errors in documents which clearly do not affect the obvious intent of the document should not be considered discrepancies; but they should obviously be corrected.
O-l NPAn1 8i
_-.J .- - _ - . . . . - - - - _ - -
. (. 018 (7/61) esuCLEAR PLANT OPERATIONS DEPARTMENT MER 12 NUCLEAR PUWT ADMINISTRATIVE PROCEDURE 2 REVISION 6 O rarxtiricatio" aan arso'urro" or "^ ''
TITLE PROBLEMS AND N0NCONFORMANCES PAGE 6 0F 42 1
1
- 3. Design and/or Procedure Changes l
If the work involved on a GDPR results in a design and/or l procedure change, the administrative controls described in the '
appropriate General Office Administrative Procedure shall be followed.
C. Nanconformances(NCR)
- 1. The Campany's Quality Assurance Program (QAP 10.1) requires that "Nonconformances" be reported in writing and resolved in a fomal manner with appropriate management review. The administrative steps involved in the identification and msolution of Nanconformances Appendix E, are summarized below,
- s. Responsibility for Classifying Problems as Nanconformances ,
The Manager Nuclear Plant Operations has delegated the msponsibilities for determining whether a problem identified in a GDPR is a Nonconfomance to individual General Office Swervisoring and Senior Engineers. The Plant Manager 1 and plant department heads have been delegated the responsibility for detemining whether a problem identified in a NPPR is a Nonconformance. In addition, other individuals may be delegated this responsibility, provided this delegation is in writing.
When such an item is identified it is the responsibility of that individual to initiate an NCR.
i b. Criteria for Classifying Problems as Nonconformances
- 1) Definition A Monconformance, as defined in Quality Assurance i
Procedure (QAP) 10.1, is a discrepancy or departure from requirements in purchase specifications, drawings, approved practices, established Quality Assurance policies or procedures, or NRC regulations which mquires resolution and may require measures to be taken to prevent its recurrence. Nanconformances consist of conditions adverse to quality which, if left uncorrected, could have resulted in any of the following:
l 1 For those plants without a Plant Manager, all responsibilities assigned to this position are transferred to the' Plant Superintendent.
NPG01 6
,,-.-.%.,. _ , _.-.- p- ._ e+ y _,,__g%
_ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ .__________~. . _ _ _ ._. . _ __ _ _ _ . _ _ _ _ _ _-.
. . . O-018 (7/h )
NUMSER WAP C-12 NUCLEAR PLANT OPERATIONS DEPARTn4ENT O ""c^"^"'^""'"'5'"^""""' "* M *~"'
DATE 6/15/83 IDENTIFICATION AND RESOLUTION OF TITLE PROBLEMS AIO NONCONFORMANCES PAGE 7 0F 42 .
I a) Degradation or loss of the integrity of the tsactor coolant pressure boundary; b) Reduction or loss of the capability to shutdown
- the reactor and maintain it in a safe condition, including the compromise of design objectives during construction and/or modification activities; or c) Lack of effective control over items or activities (including quality program implementation) that could reduce the capability l to prevent or mitigate the consequences of
. accidents that may result in potential off-site exposures comparable to the guidelines set forth in 10CFR100, " Reactor Site Criteria."
- 2) Examples of Nonconformances Appendix A is a compilation of examples to aid in determining if a problem or potential problem is a i
' Q Nonconformance.
- c. Role of Technical Review Group in Resolution of Nonconformances In accordance with Quality Assurance Procedure 10.1, once a problem has been classified as a Nonconformance by one of the individuals assigned such authority per paragraph C.I.a. above, a Technical Review Group must be convened within 30 Calendar days to review the Nanconformance, l
< determine its cause, approve the proposed resolution and establish the corrective action to prevent recurrence.
Appendix B discusses the makeup and functioning of the Technical Review Group.
- d. Role of Plant Staff Review Committee in Resolution of Nonconformances
- 1) The Plant Staff Review Cannittee shall review, at least monthly, the status of all NCRs that were issued on-site or which affect on-site safety-related items or activities. In most cases, this review will l take place after the Nanconformance has been I resolved. When a Nonconformance Report involves a system design change it shall be reviewed b the PSRC prior to reliance upon the affected system (y) s for a 0 -
fety-rel ted f action. (a for to Nucle r ri at Administrative Procedure C-1).
NPG01 7
i
. sN03 8 (7/i,1)
NUCLEAR PLANT OPERAMONS DEPARTMENT NUMBER WAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10f' W-n2 O
IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 l TITLE PROBLEMS AIG NONCONFORMANCES PAGE 8 0F 42 The Secretary of the PSRC should prepare a short summary of each NCR which was resolved during the month and include it in the announcement package for
, each regularly scheduled monthly meeting. The NCR shall be reviewed as part of the PSRC's normal review of operating, maintenance, and surveillance testing experience.
- 2) The Technical repmsentative on the Technical Review Group is responsible for assuring that NCRs am mferred to the PSRC(secretary) prior to implementation of the resolution where such referral is required or desirable.
- 2. Determination of Reportability When a problem arises. 'it is extremely important that
, knowledgeable individuals promptly evaluate the problem and
! determine whether or not it is reportable to the NRC or other I
government agencies; and, if it is, enure that such reports are made in a timely manner. The initial evaluation is O performed by the appropriate department head or other assigned individual and if the problem is considered to be reportable per the guidelines of NPAP C-11/GGAP W-111 then that individual or department head is responsible for convening a Technical Review Group meeting. if time permits. It is emphasized that in some cases the initial verbal report may have to be made so quickly (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) that it may not t
l be practical to convene a TRG. In these cases, the department ;
head should promptly notify the Plant Manager and/or Manager of Nuclear Plant Operations, who will then make the verbal notification. The function of the TRG in such cases is to myiew the occurrence and assume that followup written reports are made, According to Standard Practice 420.3-1. each site shall establish a Site Review Group for the purpose of investigating potentially reportable items. This Site Review Group shall be i the Plant Staff Review Committee or a designated subcosmittee l
thereof. Since most reportable items also involve a Nonconformance it is most convenient to consider the Technical '
Review Group which investigates the Nonconformance a subcommittee of the PSRC for purposes of making the initial determination of its reportability. Therefore, the ass 6ership and functioning of the Site Review Group shall be as described for the Technical Review Gmup in Part C.1.c above and in Appendix 8.
NPA01 8
~ _ . , . _ - . - . . . ,- . . - -
e -
c~ola(741f -
- NUMBER WAP C-12 eluCLEAR PLANT OPERATIONS DEPARTMENT 112 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10 l DATE 6/15/83 l . - IDENTIFICATION AND RESOLUTION OF TITLE PROBLEMS' AND NONCONFORMANCES PAGE 9 0F 42
- 3. Audits
- a. Wanconfomances .
The Quality Assurance Department, as stated in QAP 10.1, shall maintain a listing of all nonconfomances, issue
> periodic reports of NCR status, and auditing the nonconformance report system annually for trends or other
- indications that practices should be improved. The Quality Assurance Department is responsibb for issuing a i report of such annual audits which sets forth the findings
> . and recommendations.
- b. Reportablii Items '
The General Office Nuclear Plant Review and Audit Committee (GONPRAC) is responsible for myiewing the implementation of Standard Practica 420.3-1 at least annually. -
D. - General Requirements 0 1. Problem Report Resolutien Process and Instructions for Completing and Handling NPPRs, GDPRs, and NCRs
~
The NPPR, 60PR, and NCR forms have been developed so that they follow the general processes for handling problems - from discovery through resolution and signoff. Appendices C, D, and E to this procedure provide general instructions for completing these forms. Appendices D and E also give general instructions
, for routing of GDPRs and NCRs. Because of the differences in l the organizations of the individual plants, each plant shall prepare a supplementary procedure specifying how NPPRs are handled at the plant. Also, the suggested general routing instructions for GOPRs and NCRs may be altered if desired by preparation of appropriate supplementary erocedures.
- 2. Maintenance of Records
- a. Records which are related to an actual, substantial safety hazard shall be retained for the life of the affected basic components or the affected licensed facilities.
These records include:
- 1) Records of the activities and meetings of review groups.
- 2) NRC notifications.
0 -
- 3) Documented reviews and evaluations.
/
hprn1' O
,. w l
. . . _ _ _ .- . . = .__ - _ _ _ _ _ _ _ _ _
, , 0 -028 (7/61)
NUCLEAR PLANT OPERATIONS DEPARTt4ENT NUMBER WAP G-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS! 6 O IDENTIrICATION AND RE50turION Or nate 25/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 10 or 42
- b. Problem Reports which involve equipment important to safety shall be retained for five years.
- c. Problem Reports which do not involve equipment important to safety should be retained for one year.
l Y. REFERENCES A. ANSI /ANS 3.2: " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants."
- 8. PGandE Standard Practice 420.3-1: " Defects and Noncompliance:
Reporting of to the Nuclear Regulatory Casuiission."
C. Quality Assuran'ce Procedure 10.1: "Nonconformances and Corrective Actions."
! D. Quality Assurance Procedure 10.2:. " Verification of Reporting of Defects and Noncompliances."
l E. Title 10. Code of Federal Regulations, Part 21: " Reporting of
@ Defects and Noncompliances.'
F. Title 10. Code of Federal Regulations. Part 50.72: " Notification of Significant Events."
t
- 6. NRC Regulatory Guide 1.16. Revision 4. August 1975: " Reporting of Operating Information - Appendix A Technical Specifications."
l H. NRC Regulatory Guide 10.1, Revision 4. October 1981: " Compilation of Reporting Requirements for Persons Subject to NRC Regulations."
I. NUREG-0302, Revision.1:-'Raserks;Prosented1Q0estion#Antwers '
Discus' sed) At Public RegionalllleetfiIgfic Discuss Regulations (10CFR21) for Reporting of Defects and Noncompliance."
J. NPAP C-11/GOAP W-111. "Nonroutine Notification and Reporting to the Nuclear Regulatory Commission (NRC) and other governmental agencies."
K. Nuclear Plant Adninistrative Procedure C-1, " Design Changes."
VI. APPENDICES A. Appendix A - Examples to Aid in Determining if a Problem or Potential Problem is a Nonconformance.
O s- ^PP adix = - a t P ad r=actiaa =<
Restoration of Nonconformances.
T cha5= 1 a vi ar==P ia th-NPG01 10
. s'A 028 (7/61) i -
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER fPAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10gGAP W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
- TITLE PROBLEMS AND NONCONFORMANCES PAGE 11 0F 42 C. Appendix C - Description of Nuclear Plant Problem Report Resolution Process and Instructions for Completing and Handling Form 61-4516.
, D. Appendix D - Description of General dffice Pmblem Report Resolution }
Process and Instruction for Completing and Handling Fom 69-025.
E. Appendix E - Description of Nmconforiaance Report Resolution Process and Instruction for Completing and Handling Fom 76-286.
VII. ATTACHMENTS A. Form 61-4516: " Nuclear Plant Problem Report."
- 8. Form 69-025: " General Office Problem Report."
C. Fom 76-286: "Nonconfomance Report" i
i oC l
I 9
NPG01 11
i , 0-028 (7/61) l C UCLEAR PLANT OPERATIONS DEPARTMENT NUMSER IFAP C-12 i
NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS100[0A W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
. O TITtE enostEns ANo NaNCoNroaaANCEs ,A.a 12 or 42 ,
l' l
I l
APPENDIX A -
Exasoles to Aid in Detemining if a Problem or Potential Problem is a Monconfomance Nonconfomances represent problems or potential problems in items or activities important to safety as defined in this procedure. Nonconformances may include physical defects, test failures. incorrect or inadequate documentation, or deviations from prescribed procedures (i.e. inspection, test or processing).
The following are saaie examples to illustrate when an item should or should not be considered as a Nonconfomance. These examples should not be -
considered as all inclusi,ve:
A. Procedural Problems
- 1. While performing a procedure, a person identifies some changes which l will make it more workable. A temporary change is issued pending a
, revision. No Nonconformance is involved. Such changes would be l considered as part of the normal review process. In fact, this item need not even be reported on a Problem Report inasmuch as the Q temporary change sheet provides a written reminder to plant management that the procedure requires revision.
~
- 2. While perfoming a calibrat1on check on a reactor trip circuit it is noted that due to an error in the procedure, the instrument has had an incorrect setpoint for several months. A definita Nonconformance is involved which is indicative of a weakness in the procedure review process and represents a " lack of effective control" over activities important to safety.
l
- 3. Repeated failure to follow approved procedures or to provide l required documentation after a discrepancy or problem has been identified and reported qualifies as a Nonconformance because it represents a " lack of effective control".
B. Maintenance Problems
- 1. Normal repairs involving expected deterioration or waar need not be considered as Nonconformances but such repairs shall be documented in maintenance records.
- 2. During routine testing or preventive maintenance on a piece of equipment important to safety which was operable at the time it was cleared, component parts are found to be deteriorated but still operable, and are replaced. No Nonconformance is involved unless the problem is an unexpected one which indicates a basic design deficiency or unexpected wear rather than the nomal wear and tear.
l C Basically in this exas.ple the testing and preventive maintenance NPG01 12
. C -ola (T/h)
NUARER NPAP C-12 NUCLEAR PLANT OPERATIONS DEPARTMENT NUCLEAR PLANT ADMINISTRATIVE PROCEDURE
-112 REVIS10N IDENTIFICATION AND RESOLUTION OF DATE 6/15/83
- TITLE PROSLEMS AND NONCONFORMANCES PAGE 13 0F 42 APPENDIX A (Continued) program has done its job - i.e., discovered a deficiency before it
, produces a failure. If the parts had deteriorated to a point that the equipment was inoperable, a Nonconforiaance would exist because a mduction in safeguards capability was involved (see paragraph C.1.b.1)b).
, 3. A piece of safeguards equipment fails in service, but is reported and repaimd within the allowable license interval. A Nonconfomance is involved if the equipment would not have done its job, because the problem may be indicative of a deficient testing or preventive maintenance program (i.e. lack of effective control as per C.I.b.1)c).of this procedure), and a reduction in safeguards capability (C.1.b.1)b) of this procedure).
C. Test Equipment Problems
, 1. A gauge is marked as requiring calibration inmediately prior to each use, and on one occasion is found to be out of calibration. No Nonconformance is involved because the calibration program was
, O <~
(, adeauase to identifs the erobien in time to avoid usino an instrument which is out of calibration and there was no lack of effective control.
- 2. Test equipment (including transfer standards) which has been used for quantitative work on equipment important to safety is found out of calibration to the extent that subsement results and/or calibrations are questionable. A nonconformance is involved which indicates lack of effective control and which may mquire shortening the frequency or reevaluating the accuracy specifications. There are many instruments which are calibrated routinely which are not so used.
D. Material Problems
- 1. A spara part for a system important to safety is received and is rejected during receipt inspection. No Nonconfomance is involved -
the inspection program has done its job. The problem would be documented on the standard receipt inspection form and the defective item should normally be returned to the supplier as part of the procurement process.
A " defect" of this nature, if it could create a substantial safety hazard if used, may be reportable to the NRC under the provisions of 10CFR21 (see NPAP C-11/G0AP W-111) although most spare parts are "consercial grade" itans which are exempt from this regulation i n -
unless they have been " dedicated" for use as a " basic component."
l U ,
However, as long as the part is returned to the supplier, the NPG01 13
~
csola M/E1)
NUCLEAR PLANT OPERATIONS DEPARTMENT MER NUCLEAR PLANT ADMINISTRATIVE PROCEDURE 2 REVIS100f'6
' O TITLE raratrricaTroa aao arso'urtaa or PROBLEMS AND NONCONFORMANCES PAGE 14 0F 42 l
, APPENDIX A (Continued) supplier is responsible for making this report if one is required.
, On the other hand. if the defective item is not returned to the supplier then an NCR is initiated (i.e. it is repaired on site for subsequent use), and the reporting burden shifts to the purchaser (Company).
If the received item is obviously defective or incorrect and is so identified by the materials facility prior to conduct of a QC receipt inspection, the item may be returned as part of the procurement process without filling out either an inspection report i
or a problem report (i.e. the item is, for all practical purposes, not considered to have been received for possible use). However, a defective item must still be returned or discarded in order to eliminate the requirement for filing a report under 10CFR21.
- 2. A spare part is received without adequate documentation, and it is necessary to use the part anyway. A Problem Report should be filed and the item classed as a Nanconformance since this condition creates a deviation from the purchase specifications and it requires review and approval.
- 3. A spare part is received but prior to receiving inspection adequate documentation or some other part of the order is found missing, and the part is not needed to be used soon. The part may be tagged and placed in the QC hold area until the missing items are received. No Nonconformance is involved, nor should a Pmblem Report be filed.
Obtaining the missing items is a part of the normal procurement process. If the part was subjected to the receipt inspection and approved and later found to be defective because something was '
missing, a Nonconformance would be involved.
- 4. A part degrades during storage to a point where it would no longer perfors its function, and it is discovered during the established
' system for inspecting parts in storage prior to the time it is installed. A Problem Report must be filed (see Nuclear Plant Administrative Procedure D-503). The problem may or may not be a Monconformance, depending upon the circumstances. If there is reasonable assurance that the defect would have been discovered prior to use by either the materials facility or the group which installs the part, then there has been no loss of control and no Nonconformance is involved. On the other hand. if it was only .
i through good fortune that the defective part was not placed in
' service si.e. effective control was lost de facto), or if the deficiency appears to be indicative of a substantial breakdown in established quality control procedures for conduct of the materials facility operations, then a Nonconformance is involved.
O NPG01 14
. O-03 8 (7/h )
. NUCLEAR PLANT OPERATIONS DEPARTMENT MER 2 2
NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION 6 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 l
. TITLE PROBLEMS AND NONCONFORMANCES PAGE 15 0F 42 l
APPENDIX A (Continued)
E. Design Probless ,
, . A discrepancy or departure from requirements in design documents or activities shall be identified and documented as a Nonconfomance if:
- 1. It was identified after the design docisment had been approved and it was issued for use as a basis for further design; or
- 2. There was a basic design error meeting the definition of a Nonconformance (not a design change) in a design document which was approved and issued for use in construction or modification of existing facilities; or J
- 3. There was a failure to conform to the approved procedural and quality program requirements which were committed to in the Safety Analysis Report, Construction Permit, requirements issued by the Nuclear Regulatory Commission subsequent to the Construction Pemit, or Operating License.
F. Miscellaneous A- . _1.. _ Discrepancies or problems of a relatively insignificant nature but U {- which, due to their repetition, require action by management should be considered as Nanconformances.
- 2. Repeated failure to correct by the mutually agree-upon commitment date those discrepancies or departures which were identified in audits, surveillance reports, or inspections, when such a delay is determined to have a significant 1spect on quality, should be considered as a Nonconformance.
- 3. If a deficiency in an item is noted at an operating plant during an inspection performed to identify such deficiencies so as to allow correction prior to relying on the item to perform its safety function, a Nonconformance Report may not be required.
- 4. If a deficiency is noted during a construction process and checking and correcting is part of the routine normal course of work prior to sign-off and acceptance, it need not be reported as a Nonconformance.
- 5. When administrative, test, maintenance, or other such procedures identify items which must be classified as Nonconformances, these procedures govern.
- 6. In most, but not all, cases items which require special reporting to the NRC will involve a Nonconformanco. Appendix 1 to NPAP C-11/GOAP W-111 provides a sunnary of all the generic items which require a special report to the NRC, and identifies those which should be documented on a Nonconformance Report.
O .
NPG01 15
, , 0 038 (7/61)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER WAP C-12 REVISIO N5 W'II2 O NUCLEAR PLANT ADMINISTRATIVE PROCEDURE ME 6/15/83 IDENTIFICATION AND RESOLIJTION OF TITLE PROBLEMS AND NONCONFORMANCES PAGE 16 0F 42 APPENDIX 8 Makeup and Function of a Technical Review Group in the Resolution on Nonconformances.
)
This Appendix describes the makeup and functioning of a Technical Review Group in the resolution of Nonconformances.
A. Manbership
- 1. The Technical Review Group shall, as a minimum consist of the i following personnel:
- a. Technical Representative (Chairman)
- 1) The technical representative shall be or be selected by the department head of the affected department (i.e.
l normally the individual who made the determination that the problem was a Nonconformance) or either the Manager.
Nuclear Plant Operations, or the Plant Manager /
X U Superintendent. The technical representative shall be a supervisor responsible for or affected by the item or activity. The department head may choose to appoint more than one technical representative to the group, in which case he will select one to serve as chairman.
- b. Quality Assurance Engineer or authorized delegate.
! In the event that the authorized delegate (s)isQuality not onsite, Assurance Engineer the review group may or his meet without him and obtain his concurrence orally. The oral approval shall be noted in writing in the review group's written decision.
- c. Quality Control Representative.
- 2. Others should be included in the Technical Review Group to assure representation by all affected departments (i.e., those affected by the NCR and those who may be affected or participate in determining the cause, resolution or corrective action.)
! a. A representative of NP0 shall be included on each TRG involving a nuclear plant.
1 O
i MPG 01 16
, , s' *.-018 ( 7 / 61 )
. NUCLEAR PLANT OPERATIONS DEPARTMENT NUMSER A l NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION 6 DATE 6/15/83 O ,
TITLE IDENTIFICATION AND RESOLUTION OF PROBLEMS AND NONCONFORMANCES PAGE 17 0F 42 APPENDIX 8 (Continued) -
- b. For the duration of the Diablo Canyon Project, a Project representative should be included on the TRG if the item involves the project or project schedule.
B. Conduct of Meetings
- 1. The time and location of the meetings shall be determined by the Chairman. Meetings should be timely, consistent with the significance of the problem and the availability of personnel, but need not be held immediately upon identification of 3 Nonconformance. For example, the Chairman should prepare a meeting I agenda or may ghoose to investigate the problem prior to convening a meeting. However, the Chairman is responsible for assuring that meetings are held in time to assure that Reportable Occurrences are indeed reported within the required time period (see Appendix 1 in NPAPC-11/60APW-111).
- 2. A written record shall be kept of the review group's findings. Form 76-286 is used for this purpose, although supplemental sheets may be attached as required.
O L- C. Responsibilities and Authorities
- 1. The technical review group is responsible for:
- a. Evaluation of the Monconformance.
- b. Determining its reportability (see NPAP C-11/G0AP W-111).
l
' NOTE: An item may already have been reported prior to the time a TNG is convened in which case the TRG is concerned only with followup reporting.
- c. Determining the probable cause, if practical.
- d. Determining and approving the resolution of the Nonconformance and any appropriate corrective action to prevent recurrence.
- e. Establishing a schedule for implementation of the resolution and corrective action.
l
- f. Verifying the implementation of the resolution and corrective action.
o-NPG01 17
l 6 % o3 8 (7/61 )
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER WAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION GOA W112 IDENTIFICATION AND RESOLUTION OF WE 6/15/83 TITLE PROBLEMS AND N0NCONFORMANCES PAGE 18 0F 42
~
APPE@ik ii (Continued)
- 2. The technical representative (s) on the group is responsible for:
, a. Detemining the technical aspects of the resolution and l corrective action. l 1
I
- b. Detemining whether any emergency measures am required to l achieve safe operating conditions prior to final resolution. l
- c. Referring items to the Plant Staff Review Committee when required or as desired.
- 3. The Quality Assurance representative on the review group is A responsible for:
- a. Ascertaining the acceptability of proposed dispositions and corrective actions with respect to the established Quality Assurance Program. The Quality Assurance Department does not
- have the responsibility to determine that the proposed
, dispositions and corrective actions are technically correct.
- b. Verifying that disposition actions have been completed and signing the Nonconfomance Report to so indicata.
- 4. The Quality Control representative is responsible for:
- a. Assisting the technical repmsentative in researching problems and developing an appropriate resolution and corrective action.
l b. Acting as interface between a site Technical Review Group and the PSRC when items are referred to the latter.
- c. Serving as secretary for the review group meetings.
- d. Keeping a log of all outstanding Nonconfomances. Separate logs will be maintained for General Office / Plant initiated NCRs.
- e. Conducting inspections as may be mquired when nonconfoming items are reworked.
- 5. All decisions of the Technical Review Group must be unanimous. If the members cannot agree, the matter shall be referred to the Manager, NPO, Plant Manager / Superintendent and the Manager, Quality Assurance.
O 6. In the event of an emergency, the Plant Manager / Superintendent or his authorized delegate may take actions necessary to restore a safe operating condition regardless of unanimity.
NPF,01 1A
' . C,-032 (7/61)
NUMSER WAP C-12 NUCLEAR PLANT OPERATIONS DEPARTMENT W W-112 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10N IDENTIFICATION AND RESOLUTION OF DATE 6/15/ 3 TITLE PROBLEMS AND NONCONFOP".ANCES PAGE 19 0F 42 APPENDIX 5 (Continued)
D. Dispositions The Technical Review Group has several options available for detemining an acceptable resolution, including:
- 1. Accept As-Is
- a. Procedures, specifications, drawings, activities, or items may be accepted "as-is". This resolution requires that technica1 evaluations be perfomed, documented, and approved to assure that there will be no adverse effect upon the safety, operability, or maintainability of the items or of the
, component or system in which it is installed. Examples of circumstances which may lead to accepting "as-is" include specifications which originally were excessively stringent, modifications to related items and activities which render the "Nonconfomances" acceptable; or the initial identification of the Nonconformance was in error and the requirements are being met without any change being necessary.
- b. When a supplier 1 proposes that PGandE accept a nonconfoming' ites, the PGandE specifying organization shall require the supplier to identify the nature and extent of the Monconfomance and the reason for proposing its acceptance, use or installation.
- 2. Rework
- a. Itams or activities may be reworked or repaired to confom with original design and/or specification requirements if practices and procedures are used which were approved in advance of the work being done.
- b. Items which have been reworked or repaired shall be inspected to the original requirements, or to criteria established by the Technical Review Group if it considers the original requirements to be no longer applicable,
- c. If the Nanconfomance involves ASE Code materials or items and the disposition is to rework or repair, then the Authorized Inspector shall be promptly notified and his concurrence obtained.
2The term " supplier" includes any individual or organization who can furnish items or services. Included are manufacturers, contractors, subcontractors, O' ,
distributors, service shops, and consultants.
I 1
j , MB (7/h) tiUCLEAR PLANT OPERATIONS DEPARTMENT MER 2 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE 12 REVISION 6 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 20 0F 42 APPENDIX 8 (Continued)
- 3. Reject Items may be rejected. Items that are rejected shall be clearly marked and/or identified to prevent their inadvertent use.
- 4. Repair. Revise or Modify
- a. Itaas or activities may be repaired, modified or revised so that they are acceptable for use, although not always conforming totally to original requirements.
- b. When a modification involves a design change, the requirements of the applicable General Office / Nuclear Plant Adninistrative Procedures shall apply.
- c. When a modification involves revising approved procedures or development of new procedures, the requirements of the applicable General Office / Nuclear Plant Aeninistrative Procedures shall apply.
- d. Items which have been repaired, revised or modified shall be inspected to the original requirements, or to criteria established by the Technical Review Group if it considers the original requirements to be no longer applicable.
The Technical Review Smup may determine that all or part of the issues raised on a Nonconformance Report should be reworded.
processed separately, or submitted to a different Technical Review Group. In such instances, the disposition of these issues may be handled by issuing additional nonconformance reports, as applicable.
E. Resolution of Nonconformance by Suppliers i
If a supplier has an approved quality assurance program for handling nonconforming items, the PGandE specifying organization may delegate the authority to the supplier to perform a technical evaluation of those nonconforming items which are under the direct control of the supplier.
t PGandE retains the responsibility to assure that the resolution of Nonconformances is acceptable and satisfactory. The organization I delegating the authority shall establish procedures and controls and I monitor the supplier's activity to assure that nonconformances are handled correctly and affectively. When these conditions are met. PGandE l 1s not required to issue a separate Nonconformance Report fbr each nonconformance report that a supplier may issue. Whers such authority is delegated specifica11 in the contract or purchase order, the supplier is O " "" ' "'"' '":
l l
NPG01 20
_ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ m _ - --r- . , - ,--e.e - w. ,- -
w., -,-m._. , , , . . , - . , _ _ _ ,,. .-,..-- . -
,y,_,, . , , , , , . . _. ,.-*,_-__--w.y,,,._pe
t' .-0M (7/h )
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER WAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE G W-112 REVIS10N 0 -
. TITLE ioENrinCATroN ANo RESoturioN oP na's 5/25/83 PROBLEMS AND NONCONFORMANCES PAGE 21 0F 42 APPENDIX g (Continued)
- 1. All procedural actions comply with the requirements or
, specifications approved by PGandE;
- 2. Personnel performing the evaluation are qualified. PGandE reserves the right to review the qualifications of contractor personnel !
performing such evaluations.
h OC l
O ;
NPG01 21
i c 018 (7/ Ell NM ET RNED E NUMBER REVISION PFAP C-12 III O IoENTirICATION ANo RE50tuTION Or aaTE 25'83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 22 0F 42
! APPENDIX C -
INSTRUCTIONS FOR COMPLETING NUCLEAR PLANT PROBLEM REPORT (FORM 61-l i This Appendix describes the various sections of the Nuclear Plant Problem l Report and the type of information to be included on the form. Procedures which specify who completes, myiews and approves the various portions of the form; and how forms are distributed and tracked shall be developed by each plant as a supplement to this procedure. Quality Control Departments at each plant should provide an annual trend analysis of NPPRs which may require further action by MPO Management.
A. INITIATING DEPARTMENT
- 1. Identification '
The person who identifies a problem shall report it promptly to his supervisor. The supervisor is responsible for assuring that a Nuclear Plant Problem Report (Fres 61-4516) is initiated (except that the person who identifies the problem may choose to defer X filing a Problem Report until after conferring with his supervisor).
U If a NPPR was written and a problem does not exist, the supervisor umst sign the form and state why the problem is not valid. The form must then be routed to management for their concurrence and normal .
l processing to close the item.
i Upon notification of a potential problem, the supervisor of the individual who identified the problem will decide whether or not the problem is valid. If not, the process stops at this point. If valid, the INITIATING DEPARTMENT portion of the form shall be completed and signed off.
- 2. Instructions
- a. IDENTIFICATION This is a four part twelve character description code that is to uniquely identify each Problem Report.
- 1) The first three characters identify the applicable site / plants and the unit; authorized codes are:
DCD Diablo Canyon. General (affecting all units)
DCI Diablo Canyon. Unit 1 DC2 Diablo Canyon. Unit 2 HBO Humboldt Bay. General (affecting all units) 21 Humboldt Bay. Unit 1 0
- , , - - - - - - - cy e,m, - ,. - , - e w-e y e--r+-.+ y- --
- ( -OH (7/h1
- NUMBER W AP C-12 NUCLEAR PLANT OPERATIONS DEPARTMENT NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10h W-112 DATE 6/15/83
- IDENTIFICATION AND RESOLUTION OF TITLE PROBLEMS AND NONCONFORMANCES PAGE 23 0F 42 APPENDIX C (Con'.inued)
HB2 Humboldt Bay, Unit 2 1 , 23 Haboldt Bay, Unit 3 ,
HB4 Humboldt Bay, MEPP SPO Stanislaus Nuclear Project, General (affecting all units)
NGO Nuclear Plant, General
- 2) The second two characters are the last two digits of the year. 81 is for the year 1981.
- 3) The third pair of characters identify the department which initieted the report.
Plant Staff (Nuclear Plant Doerations) PG Plant Staff Review Committee PS Operations OP Instrument and Centrols TI Chemistry and Radiation Protection TC O ('
- ri at snaia rs w Mechanical Maintenance M Electrical Maintenance EM Technical Support ST Security SE Quality Control QC i Materials Facility MF Office (Personnel and Geheral Services) 0F Training TR Document Control DC Suggestion System SS ISI and E E Services IS Bionssay (DER) BE Quality Assurance QA On-site Safety Review Grouo SR Station Construction SC Resident Electrical RE Resident Mechanical m Resident Civil RC Resident Startup RS i
Although not normally found on Nuclear Plant Problem Reports, O ,
the following are found on Nonconfomance Reports issued by other Company departments.
WDCf11 9'J
sN ole (7/h )
eeuCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISI0pA 6
O TITLE IDENTIFICATION AND RESOLUTION OF PROBLEMS AND NONCONFORMANCES DATE PAGE 6/15/83 24 0F 42 APPENDIX C (Continued)
Encineerine. General EN Electrical EE Mechanical and Nuclear ME Civil CE Engineering Quality Control EQ Engineering Services ES Design Drafting DD General Construction GC Materials - ML Encineerino Research ER Nondestructive Testing NT Standards Laboratory SL Sitina SI O aac' r 'a s n r t$an as Nuclear Plant Operations (General Office) NO Muclear Projects NP Quality Assurance QA Meteorolony Office MD
- 4) The fourth part of the identification is a four digit unique number which is assigned by the organization which initiates the report. These numbers will start at P0001 each January first and increase sequentially through the i
year.
As an example the fourteenth NPPR identified by the Diablo Canyon operators on Unit 1 in 1977 would be identified as:
DC1 77 SP P(D14 The prefix P is used to distinguish Problem Reports from Monconformance Reports (Forn 76-2,5) which use a similar numbering system except that they use the prefix N.
- b. ITEM OR ACTIVITY Describes the general system or activity involved (e.g.,
O reactor coolant pump, containment leak rate test).
l l
! nenni 24 l
- - . - - * . - - - - - -,e, ,-g .- , ,.e ---,m- ,---+----e ,-e-ge-e.-i-- w- y -- p
C -028 (7/EI)
NUMBER WAP C-12 NUCLEAR PLANT C?ERATIONS DEPARTa4ENT GOA W-112 Q NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS10N DATE 6/15/83 IDENTIFICATION AND RESOLUTION OF
, , TITLE PROBLEMS AND NONCONFORMANCES PAGE 25 0F 42 APPENDIX C (Continued)
- c. PROBLEM
- 1) Describe the discrepancy or departure in sufficient detail i to illustrate the problem. ;
- 2) A suggested resolution may be made. This can serve two purposes:
a) In many cases, the person reporting a problem has a good idea for what should be done and his suggestion could be helpful in determining the disposition. l
.This is optional, and trivial comments ( repair it")
are not appropriate or necessary.
b) The second use for suggested resolution is to indicate work which has already been done for those i cases where work has preceded the issuance of the '
form. Although the work may be complete it still Q only has the status of a " suggested resolution" until the appropriate approvals have been received.
- d. STATUS The use of this section of the form is to inform those people who will review the form, and to whom it will be transmitted for resolution, what the status of the problem is.
In the first portion of this section, the originator should indicate the status of the suggested resolution. if any. That is, has it already been implemented and the purpose of the report is simply to indicate what has been done. is the work in progress, or has the suggested resolution not yet been started?
In the remaining portions of this section more detailed information on the status of the item can be given. As a minimum, the appropriate portion (s) should be filled out if the s resolution has not yet been determined, or if it has not yet been started or is not complete.
- 1) For plant equipment problems, indicate the operational and/or clearance status; 1.e.. in service, cleared and tagged to Shift Foreman, partial clearance (electrical, mechanical),notcleared. provisions are also provided for indicating whether the item is tagged with a Maintenance Work T orInformationTag;thetypeof
- O -
cie r ace r-appropriate)auir41circi.PiantordisPatcher.s
- whether a SWP is required; and any other special conditions or cautions that are appropriate.
- -. _ \
NPG01 25 I
0%D28 h/i1)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAP C-12 I
NUCLEAR PLANT ADM1N1STRATiVE PROCEDURE REV1510,f>0A W-112
- O T1TLE 10ENTiriCAT10N ANo RE50 tut 10N Or
=^'s PAGE 5/25/83 0F PROBLEMS AND NONCONFORMANCES 26 42 APPEND 1X C (Continued) ;
- 2) For material problems, such as nonconforming spare parts.
. the originator should indicate the location of the material and whether it is tag ed (see Nuclear Plant Administrative Procedure D-500 .
, 3) For procedural or administrative problems, the originator
! should indicate whether the item is complete, continuing.
l or halted pending resolution. Pmvision is also included for amplifying remarks as appropriate.
- 4) Finally, a remarks section is provided where explanatory notes' can be entered or types of problems can be discussed which are not adequately covered by the previous sections,
- e. PR10R1TY This section is so the originating group can provide their I
assessment of the priority which should be placed on the item.
l This is for information purposes only, and does not represent a i
Q binding order on the department which must perform the work.
The final priority is determined by plant management.
"N/A" should be used in the case where scheduling is not a factor, such as when the work is already accomplished and the form is being submitted for reporting purposes only.
. "Begin ASAP" should be assigned to work which is necessary to
' restore plant equipment for which allowable repair times have been specified in the Limiting Conditions for Operation in the Technical Specifications. This priority should also be assigned to work which is intended to correct a problem which is actually causing a Unit curtailment or which results in a significant decrease in Unit reliability.
"Begin when schedule permits" priority should be assigned whenever the previous priority considerations are not a factor.
This priority should be specified whenever possible because it provides the affected department maximum flexibility in scheduling its work.
" Unit outage" priority should be used when radiation levels, considerations of unit availability and reliability, or other I factors dictate that a Unit outage is required.
"$TS/LC0" if an item is a Technical Specification requirement and involves an LCO, this fact should also be so indicated.
! O l
1_ - _ _
6 -.a - e 4 m . a_. - e. +A _ _ -- --- -- - - - - - - -
l
-l . s'.'.-02 B (7/61 )
l NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REV1810g0AP W-112 6
IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 27 0F 42 APPENDIX C (Continued)
- f. INITIATING DEPARTMENT SIGN OFF The initiating department should sign off in accordance with plant procedures.
B. INITIAL MANAGEMENT REVIEW j
- 1. Problem reports shall be distributed to an appropriate level of !
plant management for an initial review of the item to detemine if '
it is important to safety, environmental quality related, a nonconformance, or is potStially reportable. The review should also consider the approprit Umess of proposed esolutions and use of procedures.
- 2. ITEM / ACTIVITY IS l
- a. If the item is Q-Listed, it should be so checked.
- b. .If an item is Safety Related, it should be so checked. The
{, criteria for identifying safety related items is given in Nuclear Plant Administrative Procedure D-1.
- c. If an item is a potential Nanconformance, it should be so noted and an NCR initiated. Criteria for identifying Nonconfomances are given in Appendix A of this procedure.
- d. If an LER is mquired, it should be so noted. Guidance for submission of LERs is contained in the Tech. Specs., Reg. Guide 1.16 and NPAP C-11/GOAP C-111.
- e. If a report to the NPRDS is required it should be noted.
- f. If the item needs to be piaced on the outage schedule it should be noted.
l 3. RESOLUTION
- a. If a resolution has been proposed and the reviewer I concurs, then this should be checked.
- b. If an activity has been halted (or proposed corrective action l t
has not been started) the reviewer can indicate whether work say or may not proceed by circling the appropriate statement.
O
. i 8966G. --
- ..m e:nu _
l AfUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS100p0A W-112 IDENTIFICATION AND RESOLUTION OF 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 28 0F 42 l
APPENDIX C (Continued)
- c. If the reviewer chooses to modify a pmposed resolution
, or to offer an alternative resolution, he may so indicate under i "Other". i
- d. In some cases the reviewer may be unable to cousnent in the preliminary myiew regarding the resolution, in which case N/A can be checked. This often occurs on equipment problems when the appropriate resolution is unknown prior to the time the msponsible maintenance group takes the equipment apart and diagnoses the problem.
- e. If the pmblem is well understood and other departments will be involved in addressing the problems the routing of the NPPR should be so noted.
- 4. RESPONSIBLE DEPARTMENT This is the department to whom the Problem Report is being transmitted for action and resolution. The department checked here m presents the opinion of the reviewer as to who should handle the problem, and the purpose of this section is mostly for assisting the clerical staff - i.e., into whose mail box should the form be placed. No comunitment is intended that the checked department must actuall head (s)y resolve that thedepartment another problem, ifislater moredetemined by the department appropriate.
The blanks may be used to specify a sequential routing order if the problem involves several departments.
Final priority, if other than that already specified, can be assigned.
l S. PROCEDURE REQUIREMENTS The reviewer may choose to specify the requirements for use of specific procedures. The use of procedures for various , jobs is discussed in the Adninistrative Pmcedures applicable to the type of work being performed.
If procedural requirements are not specificall "To be Determined by Responsible Supervisor". The y known, lattercheck simplyN/A or means that the foreman or other first line supervisor will make this determination, based upon his assessment of the nature and complexity of the task.
1 a
( .-02 5 (T/il)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER l@AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE 112 REVISIO IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 29 0F 42 APPENDIX C (Continued)
- 6. POST-MAINTENANCE TEST On equipment problems where operability must be proven or equipment needs to be tested to diagnose the problem, the reviewer should check that a test is requimd.
- 7. SIGN OFF -
The person (s) who completes the management review should sign off.
Although there are two sign off blanks, this does not mean that two signatums are required - only one is. However, if two (or more) people myiew the fom, they should both sign.
- 8. Convening Technical Review Gmup l If the department head identified the problem as either a nonconfomance or potentially reportable, he will be responsible for assuring initiation of the NCR and convening the Technical Review
{_.,. Group in accordance with Appendices B and D of this procedure.
C. IMPLEMENTATION OF RESOLUTION
- 1. The supervisor (s) who receives the Pmblem Report will perfom the resolution (assuming, of course, that there is no restriction placed on him by the management reviewer. He should also complete the Nuclear Plant Problem Report form.
- 2. SUPV: ASSN. TO The supervisor, and if appropriate, the person doing the work, and the inspector (if there is one) should be identified here.
- 3. RESULTS OF INVESTIGATION CAUSE/ ACTION TAKEN The results of investigation by the department doing the work and any action taken should be discussed in detail here. Efforts should be made to detemine the cause of the failure. If a separate report is warranted, it should be referenced to and included as an attachment to the Problem Report fom. If after working on the problem, the Supervisor later determines that the problem should be resolved by another department, it should be passed on to them for resolution.
O :
GeMMM9 AM
t' 602 5 (7/61)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVIS!g[ W-112 O 10 m !FICATiON . 0 RES0 tut!0N 0, n*TE s/1s/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 30 0F 42 APPENDIX C (Continued) i
. Once the work is couplete and post-maintenance testing is required, the supervisor in charge sends the fom to the Shift Foreman or an appropriate planner / coordinator who then would detemine and arrange the required test (s).
The supervisor and person who perfomed the test should be
_ identified in the designated block.
- 4. REPORT CLEAR When the job is complete, the tests evaluated and the system again operable. the supervisor in charge should sign off as " reporting clear" and date the fom. If a piece of equipment is involved and he reports clear to some other supervisor, he can so indicate or l mark M/A in the space marked "To".
D. FINAL PLANT MANAGEMENT REVIEW. SIGN OFF AND DISTRIBUTION O After the work is complete. the fom should be routed for final -
management review and signoff in accordance with plant procedures.
O enan, en
. t'%032k7/i1)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER WAP C-2 112 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIO IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 T!TLE PRD8LEMS AND NONCONFORMANCES PAGE 31 0F 42 APPENDIX D DESCRIPTION OF GENERAL OFFICE PROBLEM REPORT RESOLUTION PROCESS AND Ib5TRUCTIONS FOR COMPLETING AND HANDLING FORM 69-025.
This Appendix describes the general process to be followed from the time a problem is identified until the problem is resolved. GO-Quality Control should provide an annual trend analysis of initiated GDPRs which may mquire further action by NPO Management. -
A. INITIATING DEPARTENT
- 1. Identification The person who identifies a problem shall report it promptly to his supervisor (typically a senior engineer). The supervisor or the identifying person is responsible for assuring that a General Office Problem Report (Forn 69-025) is initiated (except that the person who identifies the problem may choose to defer filing a General Office Problem Report until after conferring with his supervisor).
If a 60PR was written and a problem does not exist, the supervisor must sign off on the form and state why the problem is not valid.
O (' The form must then be routed to management, as instructed in L paragraph D, for their concurrence and novinal processing to close the item.
- 2. Upon notification of a potential problem, the supervisor of the individual who identifies the problem will decide whether or not the problem is valid. If not, the process stops at this point. If valid, the person who identified the problem or the supervisor thould complete the INITIATING DEPARTMENT portion of the form, i
l
- 3. Instructions
! a. IDENTIFICATION This is a four part twelve character description code that is to uniquely identify each Problem Report.
- 1) The first three characters identify the applicable site / plants and the unit; authorized codes are:
DC0 Diablo Canyon, General (affecting all units)
DC1 Diablo Canyon, Unit 1 DC2 Diablo Canyon, Unit 2 l HBO Husboldt Bay, General (affecting all units) l El Humboldt Bay, Unit 1 H82 Humboldt Bay, Unit 2
- 23 Humboldt Bay, Unit 3 O- '
SPD Stanislaus Nuclear Project General (affecting all units)
NGO Nuclear Plant, General (includes General Office)
. sheet 99
s' s 038 (7/il)
CUCLEAR PLANT OPERATIONS DEPARTMENT MUMBER IFAP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIONGOA W-112 DATE 6/15/83 O TITLE IDENTIFICATION AND RESOLUTION OF PROBLEMS AND NONCONFORMANCES PAGE 32 0F 42 APPENDIX D (Continued)
- 2) The second two characters are the last two digits of the
, year. 81 is for the year 1981.
- 3) The third pair of characters identify the department which initiated the report.
General Office Staff NO (Nuclear Plant Operations)
Training TG Department Administration DA Nuclear Safety and Engineering MS Personnel and Environmental Safety PE Although not normally found on General Office Problem Reports issued by the General Office, the following are found on Nanconformance Reports issued by other Capany departments.
Engineering. General EN O Electrical EE Mechanical and Nuclear ME '
Civil CE Engineering Quality Control EQ Engineering Services ES Design Drafting DD General Construction GC Materials ML Engineerina Research ER Nondestructive Testing NT Standards Laboratory SL l
Sitina SI Nuclear Power Generation NG Plant Staff (Nuclear Plant Operations) PG Operations OP Instrument and Centrols TI Chemistry and Radiation Protection TC Nuclear Engineers TN Mechanical Maintenance 191
_ . - _ _ _ _ . , , - - - - - - - - - - - - - - ' ' ~ ^ ' ~ ' ' ' '
, C%313 (T/E1)
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12
. NUCLEAR PLANT ADMINISTRATIVE PROCEoURE REVIS10 W112 0 -
TITLE inENTiriCAT10N ANo RE50turiON Or PROBLEMS AND NONCONFORMANCES
- ^ 25'8 PAGE 33 0F 42 APPENDIX D (Continued)
Electrical Maintenance EM
, Security SE Quality Control QC Materials racility MF Office / Record 0F Bioassay (DER) BE Quality Assurance QA On-site Safety Review Grouo SR Station Construction SC Resident Electrical RE Resident Mechanical RM Resident Civil RC Resident Startup RS O Nuclear Projects Quality Assurance NP QA Meteorolony Office MD
- 3) The fourth part of the identification is a four digit unique number which is assigned by the organization which initiates the report. These numbers will start at G0001 each January first and increase sequentially through the year. Quality Control controls the numbering sequence.
As an example the fourteenth GOPR identified by the General Office QC Engineer 1982 would be identified as:
NGO 82 DA G0014 The prefix 6 is used to distinguish Problem Reports from Nonconformance Reports (Forn 76-286) which use a similar numbering system except that they use the prefix N.
- b. PROBLEM
/
Describe the discrepancy or departure in sufficient detail to illustrate the problem. Indicating whether the item / work is complete, continuing or halted pending resolution.
O_ .
Nor,01 '4'4
l l
C -M 8 U/h ) '
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIV
F. PROCEDURE
REVIS10NGOAP W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 34 0F 42 APPENDIX D (Continued)
- c. A Recomunended Solution may be made. This can serve two
, purposes:
- 1) In many cases, the person reporting a problem has a good idea for what should be done and his suggestion could be helpful in determining the disposition. This is optional, ;
and trivial comments (" correct it") are not appropriate or i necessary.
- 2) The second use for Recomunended Solution is to indicate work which has already been done for those cases where work has preceded the issuance of the fom. Although the work may be complete, it still only has the status of a :
"Reccennended Solution" until the approval of the Technical Review Group has been received.
- d. INITIATING DEPARTENT SIGN OFF 4
' The person who reports the problem and/or originates the form should sign off. In addition, his supervisor should sign off g- and date the fom, indicating that he is in agreement.
- e. DISTRIBUTION Following the completion of the INITIATING DEPARTMENT, the form should'be distributed as follows:
- 1) Original to the supervising / senior engineer of the department which will handle the resolution in accordance l with General Office procedures.
- 2) Infomation copy to Quality Control Engineer, where it will be entered into NPO's tracking system.
B. INITIAL MANAGEMENT REVIEW l
- 1. Upon receipt of a Problem Report, the appropriate supervising / senior engineer is responsible for making an initial review of the item to detemine if it is nuclear-related, environmental quality related, a nonconformance, or is potentially reportable. He should also determine the priority level and indicate his views regarding the appropriateness of proposed resolutions and use of procedures, and should document these decisions on the Problem Report form.
- a. RESOLUTION
- 1) If a resolution has been proposed and the supervisor concurs, then this should be stated.
- 2) If the supervisor chooses to modify a proposed resolution or to offer an alternative resolution, he may so indicate.
IfDCA1 1A
W
, s'102 8 D/h )
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE W112 REVIS10 DATE 6/15/83 IDENTIFICATION AND RESOLUTION OF TITLE PROBLEMS AND NONCONFORMANCES PAGE 35 0F 42 APPENDIX D (Continued)
- b. CAUSE OF PROBLEM Define how and why the problem occurred,
- c. CORRECTIVE ACTION TO PREVENT RECURRENCE List any measures taken to preclude repetition of significant or recurring discrepancies, departures, or conditions adverse to quality from reoccurring again.
- d. IMPLEMENTATION ASSN. TO The supervisor, as well as the person doing the work (if other than supervisor himself), should be identified here. The responsible department represents the opinion of the originator i and his supervisor as to who should handle the problem.
- e. SIGN OFF -
O b' The Person (s) who comp i etes the Initiai Mana.ement Review should. sign off. However, if two (or more} people review the form, they may all sign.
- f. Distribution Following the initial review by management, the form should be distributed as follows:
- 1) Original Transmit the original to the person who has been assigned the work.
- 2) Copy Transmit to Quality Control Engineer. This will serve to inform him of problem status. He should also maintain an active file of outstanding problem reports and , assure that resolutions are timely. If the problem has been identified as a potential NCR, the QC Engineer should send an infonnation copy with an assigned NCR number to the QA department.
O NPG01 35
Cso]B(7MI)
WUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REV!SI(( -112 O TITI.E 10tNTInCATION AND REs0tuTION Or =^T5 5n5/83 PROBLEMS AND NONCONFORMANCES PAGE 36 Or 42 APPENDIX D (Continued)
- 2. Convening Technical Review Group If the supervisor identified the problem as either a Nonconformance or potentially mportable, he will be responsible for convening the Technical Review Group in accordance with Appendix 8 of this procedure.
C. IMPLEMENTATION OF RESOLUTION The person (s) who receives the Problem Report will carry out the appropriate resolution (assuming, of course, that there is no restriction placed on him by the supervisor).
D. FINAL MANAGEMENT REVIEW
- 1. The results of investigation by the departamnt doing the work and any action taken should be discussed in detail here. Efforts should be made to determine the cause of the failure. If a separate report is warranted, it can simply be referenced to as an attachment to the O Problem Report form. If after working on the problem, the supervisor later detemines that the problem should be resolved by t
another department, he should state his findings on tne form and return the GOPR to his supervising engineer. The supervising engineer signs off the FINAL MANAGEMENT REVIEW block before he passes on the form to the other department.
- 2. Sign Off and Distrib:Jtion .
- a. The supervising engineer reviews the form and signs off.
l b. After sign off, the supervising engineer should pass the l
original and a copy to the QC Engineer for processing.
E. QUALITY CONTROL The QC Engineer should complete and distribute the copies as follows:
- 1. Complete verification.
- 2. Copy and any attachments will be sent to the supervisor who origireted the Problem Report for his information.
l 3. Route copy to the Manager, Nuclear Plant Operations for his information.
- 4. Make additional copies of the report and distribute as necessary.
O-
- 5. File the original in the Central File Room and RMS.
MDCn1 '4K
f
. c.-02 8 (7/h )
WUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER IFAP C-12 NUCLEAR PLANT ADHINISTRATIVE PROCEDURE REVIS!0gDA W-112 DATE 6/15/83 IDENTIFICATION AND RESOLUTION OF l
. TITLE PROBLEMS AND NONCONFORMANCES PAGE 37 0F 42 APPENDIX E DESCRIPTION 0F NONChFORMANCE REPORT RESOLUTION PROCESS AND GPIPLETING AND HANDLING FORM 76-286. l This Appendix gives instructions for completing Nonconformance Report (NCR)
Fom 76-286. This form is used by Nuclear Plant Operations to identify and process nonconformances as required by Quality Assurance Procedure 10.1.
4 Forn 76-286 "Nonconformance Report", is used as the basic " minutes" of the Technical Review Group meeting when a problem has been classified as either a ,
nonconformance or potentially reportable. It is also used to document reporting and the verification of the completion of the resolution.
Completion of the form is the responsibility of the Chairman of the review group. ,
A. Item 1. IDENTIFICATION This is a four-part, eleven-character description code that is to uniquely identify each NCR.
0 (~ 1. The first three characters identify the applicable site / plants and b the unit; authorized codes are:
DC0 Diablo Canyon, General (affecting all units) j DC1 Diablo Canyon, Unit 1 DC2 Diablo Canyon, Unit 2 NB3 Humboldt Bay, Unit 3 SPD Stanislaus Nuclear Project, General (affecting all units)
- NGO Nuclear Plant, General
- 2. ' The second two characters are the last tuo digits of the year.
81 is for year 1981.
- 3. The third pair of characters identify the PGandE department most appropriate to provide the chairman of the Technical Review Group.
GENERAL 0FFICE Engineerina. General .
EN Civil Engineering CE Design Drafting DD Electrical Engineering EE
. Engineering Quality Control EQ Engineering Services Es Mechanical and Nuclear Engineering ME.
...... r-=
I
.%02 E M/i1i NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER W AP C-12 NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIONGOAP W-112 IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 TITLE PROBLEMS AND NONCONFORMANCES PAGE 38 0F 42 APPENDIX E (Continued)
Encineerina Research , ER General Construction GC Materials ML Neteorolony Office M0 Nuclear Power Generation NG Nuclear Plant Operations NO Nuclear Projects MP Quality Assurance QA Sitina SI SITE Bionssay (Encineerino Research O ~
. Plar.t Staff. General (Nuclear Plant Operations) PG 4
Chemistry and Radiation Protection TC Electrical Maintenance EM Instrument and Controls TI Materials Facility MF Mechanical Maintenance lei Nuclear Engineers TN Operations OP Quality Control QC Security SE Technical Support ST Training TR ISI and NDE Services IS Station Construction. General SC Resident Civil RC Resident Electrical RE Resident Mechanical m Resident Startup R$
l
- 4. The fourth part of the identification is the letter N followed by a i
t three-digit, unique number which is assigned by the department identified in c. above. These numbers will start at 001 each i
n U January first and increase sequentially through the year. Each department which may issue a noaconfomance report shall establish '
controls for assigning sequential numbers for Item 1 for each l calendar year. '
l woem so
._ _ - _-_=-_ _ - __ __ . - _ -. - -_ _ _ _ _ _ _ _ _ _
( %328 iilil)
Mutt 3ER IFAP C-12 NUCLEAR PLANT OPERATIONS DEPARTMENT NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISIc h W-112 DATE 6/15/83 IDENTIFICATION AND RESOLUTION OF TITLE PR08LEMS AND NONCONFORMANCES PAGE 39 0F 42 l
APPENDIX E (Continued)
As an example, the fourteenth NCR assigned to the Diablo Canyon l i
, Plant Staff for a nonconfomance on Unit 1 in 1980 would be identified as DC1-80-PG-N014 The prefix N is used to distinguish Nonconformance Reports from Plant Problem Reports and General Office Problem Reports which use a similar numbering system except they use the prefix P and G.
respectively.
B. Items 2 througi 5 shall be completed by the person issuing the report.
- 1. Item 2. IMM OR ACTIVITY
- Describes t'+ general system or activity) involved (e.g., reactor coolant pues containment leak rate test .
Iter 3.
O -, 2. REFERENCES / REQUIREMENTS This identifies the basic reference document which contains the requirement (e.g., drawing number, specification number. Technical Specification reference) which was not met. Other pertinent documents, audit reports, inspections, de p rtmental problem reports, etc.. may also be referenced here.
- 3. Item 4. DESCRIPTION Describe the discrepancy or departure in sufficient detail to illustrate the problem.
- 4. Item 5. ORIGINATED l The issuer lists his department, the date, and signs the fors.
Distribution shall be made in accortlance with Item 15 - ,
DISTRIBUTION.
C. The Technical Review Group must first determine whether the problem is actually a Nonconfomance meeting the definition in Quality Assurance Procedure 10.1. If it is, the group completes Items 6 thregh 12. If it is determined not to be a Nonconformance. it must be so stated in'the disposition and shall be referred to the responsible supervisor.
- 1. Item 6. CAUSE OF NONCONFORMANCE O A brief expianation of
- ca- of a =confo-nee mi, be
- stated in this section. In some cases the explanation may involve a study which can be referenced. If the cause cannot be specifically 1 determined, the most probable cause(s) shall be described and !
entered in this section.
1
. .e m . l
(M B (7/h )
NUCLEAR PLANT OPERATIONS DEPARTn4ENT MUMBER IFAP C-12 g NUCLEAR PLANT ADMINISTRATIVE PROCEDURE REVISION GOA W-112 l
IDENTIFICATION AND RESOLUTION OF DATE 6/15/83 i TITLE PROBLEMS AND NONCONFORMANCES 42 PAGE..m ..40 0F ,
APPENDIX E (Continued) i
- 2. Item 7. RESOLUTION 1
i
' In this section, the Technical Review Group specifies the I disposition of the Nonconfomance. l The resolution shall be described as one of the following four options: accept as is, rework, repair or modify, or reject, with l specific amplifying infomation. When procedures, practices, or '
administrative controls are involved in a Nonconformance, the most likely resolution will be " modify." If an NCR aust be revised, a !
new Nonconformance Report shall be written. The new Nonconformance )
Report shall reference the one it supersedes under " Reference 1 Requirements" and an entry shall be made in the " Additional Remarks" I section of the superseded NCR referencing the number of the new NCR.
- 3. Item 8. CORRECTIVE ACTION TO PREVENT RECURRENCE Whenever possible, the steps necessary to prevent recurrence of the O aaacaafar==ac =* 11 6. *==cr16 d- >< *a. aaac afar ==ac. $$ =
rendom incident, an isolated case, or for some other reason it is determined that no corrective action is merited, then the basis for
, this determination shall be stated.
- 4. Item 9. SCHEDULED CG4PLETION The proposed schedule for completion of the resolution and corrective action shall be entered.
- 5. Item 10. REPORTABILITY A E NOTIFICATION
- a. The determination of whether or not an item is reportable, whether or not an item is a Sestantial Safety Hazard, the bases for these determinations, and the applicable vrting
' requirements by which it was reviewed shall be stated; for example, 10CFR20, 10CFR21, 10CFR50.36, 10CFR50.55. Technical Specification Requirement 6.9.1.12.a. etc. If the review group cannot unanimously determine reportability, this section shall state the matter was referred to the next level review group for determining reportability; for example, Site Review Group and/or General Office Review Group. (If there are any, identify those reporting requirements under which the item was l
reviewed for reportability, even if it turns out that the review trcoup does not feel the item is in fact reportable, t
This wW assist the General Office Review Group if they have to reconsider a controversial item.)
Guidance on determining reportability is contained in NPAP C-11/GOAP W-111.
_ - ,au e m-m - - = ' ' ' ' ' '
- l I
i l ,
- N 038 (7/ill MUMSER OFAP C-12
) NUCLEAR PLANT OPERATIONS DEPARTb4ENT l
. O '
auc'saa e'^at ^oatatSTRATivt eaoctouar =EvisiOM "-222 DATE 6/15/83 ;!
IDENTIFICATION AND RESOLUTION OF
- TITLE PROBLEMS AND NONCONFORMANCES PAGE 41 0F 42 I i
~ ' '
APPENDIX E (Continued)
- b. The type of report (s) required by the NRC or other agency and their timing should be entered. For the initial report the 4
reporting method (telephone, written etc.) should also be ,
indicated. (Follow-up reports are always in writing.) l Provision is also included for entering the infomation regarding the actual reports when they are a6de. This infomation should be provided to the department holding the original NCR by the department making the actual report.
- c. The Plant Manspr and/or Plant Superintendent and the Manager, i Nuclear Plant Operations, must be promptly informed of all !
potentially reportable items. Likewise, the Manager. Quality l Assurance, must be promptly notified if a substantial Safety Hazard may be invJ1ved.
- d. In many cases. items which are reportable to the NRC are also reportable to other agencies, such as the Regional Water O'( Quality Control Board on NPDES 1 permit violations and the Department of Transportation (DOT) on radioactive material L shipment anomalies. For completeness, a place is provided where these other agencies can be identified.
- 6. Item 11. ADDITIONAL REMARKS Provision is included for entering addit lonal comments and remarks that are appropriate.
- 7. Item 12. REVIEW GROUP APPROVAL
- a. Each mem6er of the Technical Review Group shall sign arid date i the report. All review group decisions must be unanimous. The Quality Assurance representative shall sign approving the quality aspects of the cause and disposition for conformance to the quality assurance program. The quality Assurance representative does not have the responsibility to detemine that the proposed dispositions and corrective actions are technically correct.
- b. All nonconformances pertaining to an operating plant must be reviewed by the PSRC. The date of the PSRC meeting at which the report was reviewed should be entered here.
D. Item 13istobecompletedbytheorganization(s)implementingthe I
Q ,
resolution and corrective action.
l -
IMPDES - National Pollutant Discharge Elimination System
_ ._ .y - _ , , _ _ , . - , _
y .-, _m_ ___ _ _ _- _,,_______,_____m.
i l
, . O-03 8 (7/h )
NUCLEAR PLANT OPERATIONS DEPARTMENT NUMBER OFAP C-12 O auc't^a ^at anat"257a^T2vt raoctouar aavis2a'/' ^! "-22" DATE
! IDENTIFICATION AND RESOLUTION OF 6/15/83 TITLE PROBLEMS A W NONCONFORMANCES PAGE 42, . 0F 42 <
i
, l APPENDIX E (Continued) ;
Item 13. IMPLEMENTATION j
- a. sign and date when the' resolution is complete l
- b. sign and date (if twquired) when resolution has been inspected l
- c. sign and date when the corrective action is complete
. d. sign and date (if required) when corrective action has been l inspected E. Item 14 is to be completed by the Quality Assurance Department when all the preceding items have been coupleted.
- 1. Item 14 VERIF,1 CATION I
The method of verification is dependent upor tha type and extent of the resolution and corrective action and may include such activities as document reviews and inspections of physical work. The Manager.
Quality Assurance, shall designate appropriately qualified individuals to verify and sign-off Nanconformance Reports. idhen Item 14 is correctly signed and dated, the report is considered closed.
- 2. Item 15. DISTRIBUTION 1
The distribution blocks at the bottom of the form shall be checked when the form is initially issued by the issuing department. If the issuing department is not the responsible department, the original of the Nonconformance Report must be transmitted to the responsible department for resolution. The Quality Assurance Department and the appropriate Plant Manager and/or Plant Superintendent shall always be checked. If Engineering Department is involved, the appropriate discipline chief shall be written in the " Engineering" block, and Nuclear Projects and Nuclear Plant Operations shall be written in the " Nuclear Power Generation" block. Other departments may be checked as deemed appropriate. Informational copies shall be so marked or identified so that they cannot be confused with official working copies. After verification. completed copies shall be )
i distributed to the appropriate departments by the Quality Assurance ,
Department. The Quality Assurance Department is responsible for i maintaining the file of closed Nonconformance Reports. The originating department is responsible for maintaining file copies of the Nonconformance Report forms which they initiate.
~
NOTE: The previous discussion applies to the distribution of i
completed NCRs. In addition to the above, a preliminary copy O shall be telecopied to the Manager. NPO when the NCR is first initiated. The Manager. NPO should send an infomation copy to the Project Completion Manager.
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65 -02E NUCLEAR PLANT OPERATIONS G/2C/82 Rev I GENERAL OFF2CE Pit 0BLEM REPORT O $1TE(5) YEAR INT DEPT NLD3ER
- 4. ICENTIFICATION l ! I I I l l III l l I l l
TO BE COMPLETED BY INITIATING DEPARTMENT
- PROBLEM 1
ECOM4t!CED SOLUTION i
SIIiNATURE DATE
- 2. IMITIAL MANAGDENT REVIEW NUCLEAR-ELATED _YES _ NO NONCONFORMANCE _YES _ NO POTENTIALLY EPORTABLE _YES _I PRIORITY BEGIN ASAP PROBLEM SCHEDLLS TO BE COMPLETED WITHIN 1 M0h*
PROCLD4 SCHEDULED TO BE CGtPLETED IN 2-6 M*THS OTHER(SPECIFY)
MSOLUTICH CAUSE OF PRUBLEM O
( CORRECTIVE ACTION TO PEVEh1 ItECURRENCE IMPLEMENTATION ASSIGNED 70:
SIGNATURE DATE
- 3. 10 BE C9tPLETED BY DEPARTMENT IMPLEMENTING RESOLUTION ACTION TAKEN ___, _
SIGNATURE DATE
- 4. FINAL MANAGEMDIT REVIEW ACTIONS TAKEN TO RESOLVE AND PREVENT RECURRENCE HAVE BEEN REVIEWED AND ARE ACCEPTABLE.
CONMENTS:
SIGNATURE DATE
- 5. TO BE CGtPLETED BY 00ALITY CONTROL ThE RESOLUTION AND CORRECTIVE ACTION ARE CG4PLETE DATE VEi:IFICATION:
DISTKIBUTION DECISION SCHEMATIC FOR NAISLING NONCONFORMANCES
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EXHIBIT 13 1
- PACIFIC G45 AND ELECTRIC COMPANY GENERAL CONSTRUCTION O Quat!Tv ASSuR4NCE ,ROGR4x Procedure GCP-12.1 Rev. 3 Documenting Discrepancies and Change Notice No. 10 TITLE
- Assessing Reportability Page 1 of 11 APPROVED: Date: b% "A A 14 '
8"A T1ce President - General Construction (/
i 1.0 SCOPE i
1.1 This procedure establishes the method for documenting discrepan-cies and assessing reportability.
1.2 A discrepancy is any departure from the requirements of specifi-cations, drawings, procedures, codes, or other applicable docu-ments. .
2.0 RESPONSIBILITY 2.1 The Senior Site Representative is responsible for implementing Q( this procedure.
2.2 Quality Control is responsible for maintaining a record of and monitoring the status of discrepancy reports.
2.3 In all areas of this procedure, supervisory personnel who have been assigned responsibility for a task have the authority to delegate performance of that task to subordinates. This dele-gation shall be by written job description or by specific written assignment.
3.0 APPLICATION 3.1 This procedure applies when a discrepancy is identified in an ites, work activity or documentation which is important to safety or environmental quality, is reportable, or requires quality assurance; or when a supplier or contractor violates P G and E quality assurance requirements. This does not include insignifi-cant departures which can la corrected in the nomai course of work, unless the Technical 3 u.i'ications of a plant with an oper-
- ating license have been vio).M ,
4.0 PROCEDURE 4.1 Discrepancies shall be documented on either a Nonconformance Report (Quality Assurance Fom 76-286), a Minor Variation Report (General Construction Fom 776-101), a Nuclear Plant Problem l O Re, ort (Nucitar riaat oRerations ro- 51-4 sis). or a contractor's discrepancy report.
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Procedure GCP-12.1 Rev.
Page 2 of 11 O 4.1.1 A nonconfomance is a departure from requirements in pur-chase specifications, drawings, codes, standards, proce-dures or Nuclear Regulatory Conunission regulations which is reportable or which, if left uncorrected, could result in any of the following:
4.1.1.1 Degradation or loss of integrity of the reactor coolant pressure boundary; or 4.1.1.2 Reduction or loss of the capability to shut down the reactor and maintain it in a safe condition, including the compromise of design objectives during construction or modification activities; or 4.1.1.3 Lack of effective control over items or activi-ties (including quality assurance program imple- ,
mentation) that could reduce the capability to '
prevent or mitigate the consequences of accidents that may result in potential off-site exposures comparable to the guidelines set forth in Title
- 10. Code of Federal Regulations. Part 100. "Re-actor Site Criteria".
4.1.2 The following examples may also be considered nonconfomances:
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4.1.2.1 Repeated failure to follow approved procedures or to provide required documentation after the dis-crepancy has been identified and reported.
4.1.2.2 Discrepancies of a relatively insignificant nature but which, due to their repetition, require ac-
- tion by management.
4.1.2.3 Repeated failure to correct by the mutually agreed upon commiitment date those discrepancies identi-fied in audits, surveillance reports, or inspec-tions, when such a delay is determined to have a significant effect on quality.
4.1.3 A minor variation is a discrepancy which is not reportable and is not a nonconfomance, but which nevertheless is a departure from specific requirements.
4.1.4 A Nuclear Plant Problem Report shall be initiated for problems identified by General Construction which requi e resolution by Nuclear Plant Operations. These problems may be of a material nature regarding equipment. systems, components, and structures or they may address procedural or administrative deficiencies which affect the coordina-tion of work efforts between General Construction and Nuclear Plant Operations. A Nuclear Plant Problem Report O.
Procedure GCP-12.1 Rev Page 3 of 11 O may also be initiated for problems at a licensed plant or unit which have been identified by General Construc-tion involving a question of acceptance or operability of systems, equipment, components or structures.
4.1.4.1 Site procedures shall be developed in accordance with Procedure GCP-5.1, " Approval and Control of Field Procedures", to provide instructions for issuance, approval and control of Nuclear Plant Problem Reports.
4.1.5 A contractor's discrepancy report is required when a supp-lier or contractor having an approved quality assurance program violates P G and E or self-imposed quality assu-rance requirements. The discrepancy shall be documented in accordance with the problem mporting requirements of the supplier's or contractor's program, and shall be con-trolled in accordance with paragraph 4.4 of this procedure.
4.1.6 In some instances, items or activities which appear to be discrepant should not automatically be considered as non-confoming but should be handled by established procedures and practices. The following examples are given for cla-rification:
4.1.6.1 A deficiency is noted during a construction pro-0 ( cess, and caec* ins and correctin are eart of the routine nomal course of work prior to sign-off and acceptance.
4.1.6.2 Normal repairs involving expected deterioration or wear; however, such repairs shall be documen-ted.
4.1.6.3 If, prior to the completion of a receipt inspec-tion of items that require quality assurance, it is observed that the received items are damaged or do not conform to the purchase specification, such items may be rejected and returned to the supplier.
4.1.7 Discrepancies which may be reportable shall be promptly i reported to the supervisor responsible for the item or activity. Discrepancies which affect operating plant equipment shall be immediately reported to the Shift Fore-man by the individual identifying the discrepancy. With-in one working day from the time that a discrepancy is identified, a discrepancy report shall be initiated and submitted to the supervisor responsible for the item or activity. The supervisor shall insnediately review the report for potential reportability to the Nuclear Regu-O
i Procedure GCP-12.1 Rev l
Page a of II :
1 O latory Commission, and to detemine classification of the discrepancy as a minor variation nr a nonconfomance. If '
the discrepancy is determined to be potentially report-able, a Noncenformance Report shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with paragraph 4.2. Additional cri-teria and guidance for assistance in detemining reporta-bility and classification of the discrepancy are included in the attached Nuclear Plant Administrative Procedure C-11.
4.1.8 A discrepancy which is not reportable and is not a noncon-
' fomance shall be processed on a Minor Variation Report in accordance with paragraph 4.3 of this procedure.
4.1.9 A discrepancy which is determined to be a nonconfomance or potentially reportable shall be processed as a Noncon-fomance Report in accordance with paragraph 4.2 of this procedure.
4.1.10 The Senior Site Representative or his delegate (Assistant Project Superintendent or Lead Startup Engineer) shall be promptly notified of all nonconfomances that are de-temined to be potentially reportable. If the nonconfor-mance affects a plant or unit holding a construction pemit, the Senior Site Representative or his delegate shall notify O ( the Project Manager. If the nonconfomance or potentially reportable item affects a plant or unit with an operating license, the Senior Site Representative or his delegate shall promptly notify the Plant Manager. Plant Superin-tendent or Power Plant Engineer.
4.1.11 No work shall be performed to correct a discrepancy until a discrepancy report is approved, except during emergency situations. Emergency situations are defined as those situations in which a hazard to life or property exists.
In these cases it is permissible to perfcm the work, then initiate the discrepancy report as soon as possible.
The discrepancy report must state the reason that work was performed prior to approval.
, 4.1.12 Minor Variation Reports and contractor discrepancy reports ,
shall be approved by the Senior Site Representative. Non- !
conformance Reports shall be approved by a Technical Re- l view Group. The Technical Review Group shall, as a mini-mum, consist of Quality Control. Quality Assurance, and the Senior Site Representative or supervisor responsible for or affected by the item or activity. If the noncon-formance affects a plant or unit with an operating lic-ense, the chairman of the Technical Review Group shall be a technical representative from Nuclear Plant Operations.
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Procedure GCP-12.1 Re Page 5 of u O 4.1.13 The disposition for Minor Variation Reports shall be es-tablished within 30 calendar days of initiation, with the exception that the disposition of Minor Variation
- Reports initiated as a result of Quality Assurance Depart-ment Open Item Reports shall be established within 15 working days of the date the Open Item Report is issued.
The Technical Review Group shall convene within 5 calendar days of the date a potentially reportable nonconformance is identified, and within 30 calendar days for all other nonconformances, to detemine the cause of the nonconfor-mance, establish the resolution and the corrective action to prevent recurrence, and assess reportability.
I 4.1.14 Quality control shall maintain logs of Nonconformance Reports and Minor Variation Reports on the Nonconfomance Report Log (General Construction Forn 77G-54) and the Minor Variation Report Log (General Construction Form 776-55). The logs shall contain the following infomation:
4.1.14.1 P G and E discrepancy report number 4.1.14.2 Contractor's discrepancy report number (if applicable) 4.1.14.3 Unit (Minor Variation Report Log)
Q 4.1.14.4 Specification 4.1.14.5 Originator 4.1.14.6 Date originated 4.1.14.7 Brief description of discrepant or rejected item 4.1.14.8 Date forwarded to Quality Assurance (Nonconfomance Report Log) 4.1.14.9 Date closed (Minor Variation Report Log) 4.2 Nonconformance Reports shall be completed as follows:
4.2.1 A four part, eleven-character description code shall be used to uniquely identify each Nonconfomance Report.
4 4.2.1.1 The first part shall identify the applicable plant / site:
DC0 Diablo Canyon. General (affecting all units)
DC1 Diablo Canyon Unit 1 DC2 Diablo Canyon. Unit 2 O- Ns3 N e idt say. Unit a l SP0 Stanislaus Nuclear Project. General (affect-l ing all units)
NGO Nuclear Plant General
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Procedure GCP-12.1 Rev.
Page 6 of 11 4.2.1.2 The second part contains the last two digits of the year.
4.2.1.3 The third part identifies the department or organization most appropriate to resolve the nonconfomance:
GENERAL OFFICE Encineerina General EN Civil Engineering CE Design Drafting DD Electrical Engineering EE Engineering Quality Control EQ Engineering Services ES Mechanical and Nuclear Engineering ME l
Enaineerino Research ER General Construction GC l
Materials ML O k.
= teor a - Ouice MO Nuclear Power Generation General NG Nuclear Plant Operations NO Nuclear Projects NP Quality Assurance QA Sitina SI SITE Sioassay (Engineerina Research) BE Plant Staff General (Nuclear Plant Operations) PG Chemistry and Radiation Protection TC j Electrical Maintenance EM Instrument and Controls TI Materials Facility MF Mecham .7.1 Maintenance MM Nuclear Engineers TN Operations OP Quality Control QC Security SE O- Technical Support ST l
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Procedure GCP-12.1 Rev.
Page 7 of 13 O Station Construction General SC
! Resident Civil RC Resident Electrical RE Resident Mechanical RM Resident Startup RS 4.2.1.4 The fourth part consists of the letter "N".
followed by a three digit number which is ob-tained from Quality Control. These numbers shall start at 001 each January first and in-crease sequentially through the year.
4.2.1.5 These codes will be expanded as codes are assigned to new projects.
4.2.2 The general system or activity involved shall be described.
4.2.3 The basic documents which contain the requirements which were not met. and any other appropriate documents shall be identified.
4.2.4 The discrepancy shall be described in sufficient detail to identify the problem.
( 4.2.5 The originator shall identify his department. date and sign the fom. and forward it to Quality Control.
4.2.6 Quality Control shall distribute information copies to the Manager. Nuclear Plant Operations. Quality Assurance, the Plant Manager and/or Plant Superintendent. Nuclear Projects and the appropriate Engineering discipline chief if Engi-neering is involved. and other departments as detemined by the Senior Site Representative. When Station Construc-tion is identified as the department responsible for the item or activity. Quality Control shall retain the origi-nel report until the Technical Review Group is convened.
If a department other than Station Construction is respon-sible for the item or activity, the original report shall be forwarded to that department and a copy shall be re-tained by Quality Control.
4.2.7 The Technical Review Group shall convene to detemine whether the discrepancy is a nonconfomance. If it is not a nonconformance, the basis for the detemination shall be entered under "Resolut.on" and a Minor Variation Report shall be initiated. If.it is a nonconformance.
the group shall detemine the disposition, evaluate, and approve the report as follows:
4.2.7.1 A brief explanation of the cause of the noncon-
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l l Procedure GCP-12.1 Rey, l Page 8 of 11 O fomance shall be stated. In some cases the explanation may involve a study which can be referenced. If the cause cannot be specifically detemined, the most probable cause(s) shall be described in this section.
4.2.7.2 The resolution of the nonconformance shall be established. The resolution may be described as one of the following four options: accept as is, rework, repair or modify. or reject, with specific amplifying information. A resolution which accepts a discrepant condition "as is" shall include a basis for the acceptance.
4.2.7.3 The corrective action necessary to prevent re-currence of the nonconformance shall be described.
If it is detemined that no corrective action is required, then the basis for this decision shall be stated.
4.2.7.4 The proposed schedule for completion of the resolution and corrective action shall be entered.
4.2.7.5 The nonconfomance shall be reviewed for repor- .
tability under Title 10. Code of Federal Regu-1ations. Parts 21 and 50.55(e). In addition.
O( if the nonconfomance affects a plant or unit with an operating license, the nonconformance shall be reviewed for reportability in accordance with the requirements of the attached Nuclear Plant Administrative Procedure C-11. The basis for reporting or not reporting shall be entered.
If reportable. the applicable time requirement for reporting, the method of reporting and the time of the report shall be entered. Infomation regarding the actual report and follow-up report shall be recorded as it is received frge the reporting department.
4.2.7.6 All decisions of the Technical Review Group must be unanimous. If unanimous decisions cannot be reached, the matter shall be referred to the Manager. Quality Assurance Department.
4.2.7.7 The Technical Review Group chaiman shall notify the Plant Manager and/or Plant Superintendent and the Senior Site Representative or n'is dele-gate of any items or activities that are deter-mined to be r3 portable.
4.2.7.8 The Senior $1te Representative is msponsi-ble for notifying the Project Manager and .
O ii aar art t a a r* at a e r a< aii
, Procedure GCP-17.1 Rev.
Page g of 31 reportable items or activities affecting a plant or unit holding a construction pemit. Nuclear Plant Operations is responsible for notifying the Nuclear Regulatory Comission of all report-able items or activities affecting a plant or unit with an operating license.
4.2.7.9 The chaiman and members of the Technical Review .
group shall sign the report approving the tech- l nical content of the disposition. The Quality ;
Assurance representative shall sign the fom i verifying that the disposition is acceptable with respect to the Quality Assurance Program.
After all the required signatures of the Tech-nical Review Group are obtained, the report shall be filed by Quality Control until the resolution and corrective action have been completed.
4.2.8 Upon completion of the resolution, the " Resolution Completed" section shall be signed and dated by the implementing
, organization. I 4.2.9 When inspection is required to verify that the completed resolution is acceptable, the individual perfoming the inspection shall sign and date below " Resolution Completed".
( 4.2.10 Upon completion of corrective action, the innlementing organization shall sign and date the " Corrective Action Completed" section.
4.2.11 When the corrective action requires inspection to verify implementation, ths individual perfoming the inspection shall sign and date below " Corrective Action Completed".
4.2.12 All clarifying or verifying documentation shall be complete and attached or referenced. The report shall then be for-warded to the Quality Assurance Department for verification and distribution. .
4.2.13 If a Nonconfomance Report must be revised, a new Noncon- ,
formance Report shall be written. l 4.2.13.1 The new Nonconfomance Report shall reference the f one it supersedes under " Reference Requirements".
4.2.13.2 An entry shall be made in the " Additional Remarks" section of the superseded Nonconformance Report referencing the number of the new Nonconformance Report.
4.2.13.3 The superseded Nonconfomance Report shall then be processed as a completed Nonconformance Report.
O 4.2.13.4 The new Nonconfomance Report shall be processed in accordance with paragraph 4.2.7 of this proce-dure. l l
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- Procedure GCP-12.1 Rev Page 10 of 3 4.3 Minor Variation Reports shall be completed as follows:
4.3.1 The initiator shall complete the heading to include the location, unit number, specification number. page number.
responsible organization, contractor's discrepancy report number (if applicable), and P G and E report number (ob-tained from Quality Control). If a contractor is to re-i ceive a copy. the contractor's name shall be entered on the form in the space provided.
4.3.2 The initiator shall complete the " Description of Discre-pancy" section with a description of the discrepant item, explanation of the discrepancy, signature and date.
4.3.3 The " Disposition" section shall be completed with a pro-posed disposition. A disposition which accepts a discre-pant condition "as is" shall include a basis for the accep-tance. The report shall then be submitted to the Senior Site Representative for approval. If desired or required, additional concurrence with the disposition shall be ob-tained from the ASME Authorized Inspector. Assigned Engineer, supplier's representative, etc.
4.3.4 The Senior Site Representative and Quality Control shall review the report for potential reportability and for Q classification of the discrepancy. If the discrepancy
( is determined to be reportable or a nonconformance, the Minor Variation Report shall be superseded by a Nonconfor-mance Report. The Monconformance Report shall be processed in accordance with paragraph 4.2 of this procedure.
4.3.5 Quality Control shall maintain the original report until the disposition has been accomplished.
4.3.6 Quality Control shall be notified when all requirements of the disposition have been verified as being satisfac-torjly accomplished. The " Disposition Accomplished" sec-tion shall be completed, signed and dated by the verifying individual. Minor Variation Reports issued by Quality Control shall be verified by Quality Control. If applicable, the Contractor's completed discrepancy report shall be attached. Any other verifying or clarifying documents shall be referenced or attached. All attachments shall be identified in the space provided at the bottom of the re-port.
4.3.7 The report shall then be forwarded to Quality Control for review, to assure that the " Disposition Accomplished" add-resses all the requirements of the disposition and that l the attachments listed are attached and complete. The original report shall be filed by Quality Control.
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- Procedure GCP-12,1 Rey, 1 i Page 11 of 11 4.3.8 If a Minor Variation Report must be revised, a new report shall be written. The original report number shall be retained and the revision number noted on the report and
! in the Minor Variation Report Log. A revision to a Minor Variation Report shall receive the same approval. m vfew and distribution as the original.
4.4 Contractor's discrepancy reports shall be submitted to the Senior <
Site Representative for approval of the problem description and concurrence with, or detemination of, the disposition. The Senior Site Representative and Quality Control shall review the report for potential reportability and for classification of the discrepancy. If the discrepancy is detemined to be reportable or a nonconformance, a P G and E Nonconfomance Report shall be initiated in accordance with paragraph 4.2 of this procedure.
4.4.1 A Minor Variation Report may be initiated to provide add-itional control of discrepancies resulting from contractor violations.
4.4.2 3tte procedures shall be developed in accordance with t' Procedure GCP-5.1 to control the processing of contrac-tors' discrepancy reports.
O. ( 5.0 DOCUMENTATION 5.1 All documentation required by this procedure, with the exception of completed Nonconfomance Reports, shall be maintained by Quality Control for inclusion in the Records Management System. Completed Nonconformance Reports shall be maintained by Quality Assurance.
6.0 REFERENCES
6.1 Title 10. Code of Federal Regulations. Part 50. Appendix 8
" Quality Assurance Criteria for Nuclear Power Plants". Criteria XV and XVI. i l 6.2 American National Standards Institute M45.2. " Quality Assurance l Program Requirements for Nuclear Power Plants". Sections 16 and 17.
7.0 ATTACHMENTS 7.1 Quality Assurance Fom 76-286. Nonconfomance Report 7.2 General Construction Form 776-101 Minor Variation Report 7.3 General Construction Fom 776-54. Nonconfomance Report Log 7.4 General Construction Fom 77G-55. Minor Variation Report Log 7.5 Nuclear Plant Administrative Procedure NPAP C-11. "Non-Routine Notification and Reporting to the Nuclear Regulatory Consiission (NRC) and Other Governmental Agencies"
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requires an 1Mahle om. Sumimme G. Brews. edelly about to tests
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ares censuretiva years to tabs the - '"_ tussere tant east them their jobs, test. A peepened federal negolatism results from out to enid that W pGa5 doesnt -
foretag meelear power plaats te these weis are Sie he1 get as his one eens give meh tests la espected to he "Furthelsgesat test resuks een as new NBC regulations are ap, adapted seen by the Nuclear Regis. he firighesudug to senaams who le proved. These regelstlema essM heery omendemon. est smanise vem the muomessF' of require meeleer plant essere to in an apart te pry spea the fDs psycheisskal tariniashey, Bewun adualuisier tbs tests and alms re.
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l ta radiattes er tazae chaudeals pesimed perm af him test resutaa. Two ansesdr told w tems's hems '
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