IR 05000275/1978007
| ML20148G587 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/26/1978 |
| From: | Book H, North H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20148G500 | List: |
| References | |
| FOIA-84-21 50-275-78-07, 50-275-78-7, NUDOCS 8801260549 | |
| Download: ML20148G587 (230) | |
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U. S. ItJCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT
REGION V
Report No. 50-275/78-07 Docket No. 50-275 License No. CPPR-39 Safeguards Group Licensee: Pacific Gas and Electric Company 77 Beale Street San Francisco, California 94106 Facility Name: Diablo Canyon Unit 1 Inspection at: San Luis Obispo County
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Inspection Conducted februar,y 27 - March 3, 1978 Inspectors: / h M!7A H.S.'Sorth,RadiationSpecialist -
/ Date Signed Date Signed Approved By* #10 d u ctz -
8!d4/74' H. E. Book, Chief, Fuel Facility and ./ Date' 51gned Materials Safety Branch
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Sumary:
' Inspection on February 27 - March 3,1978 (Report No. 50-275/78-07)
Areas Inspected: Emergency planning - communications; radiation protection,. organization, training; radioactive waste, measure-ent QC, monitoring system calibration, procedures and preoperational testing. The inspection involved 15 inspector hours on-site by one tac inspector.
Results: No items of noncompliance or deviations were identified.
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CARDE84-21 PDR IE:VForm219(2)
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DETAILS 1 1. Persons Contacted Principal Licensee Employeq
*R. Ramsay, Plant Superintendent *J. Shiffer, Power Plant Engineer *J. Boots, Senior Chemical Radiation Protection Engineer W. O'Hara, Chemical and Radiation Protection Engineer K. Doss, Division Instrumentation Supervisor M. Stevens, Instrument Maintenance Foreman *M. Leppke, QA Supervisor *J. Diamonon, QC Supervisor C. Schulze, Shift Foreman, Training Coordinator W. Eldridge, Assistant Training Coordinator T. Martin, Shift Foreman, Instructor *M. Norem, Resident Startup Engineer, G. C.
- W. Coley, Startup Engineer, G.C.
B. Sovard, Startup Engineer, G.C.
K. Rhodes, Instrumentation Supervisor, G.C.
R. Breed, QC Engineer NUCON J. Hunt, Director, Field Testing C. Graves, Manufacturer's Representative CIT Nuclear J. Laudon. Project Engineer
* Denotes those present at the exit interview.
2. Emergency Planning-Communications Surveillance Test Procedure I-29, Emergency Signals and Communications Systems Functional Test, approved 8/02/77, was examined. The proce-dure provides for tests of the following signals and communications systems on a scheduled basis: fire signals, site emergency and contain-ment evacuation, UHF radio net, emergency (red) telephone and paging receivers. The inspector witnessed a test of the fire signal and site emergency and containment evacuation signals.
l No items of noncompliance or deviations were identified.
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3. Radiation Protection - Organization Two of the Radiation and Process Monitoring (RPM) technicians of the original staff of eight have transferred to other activities within the company. Of the remaining six RPMs, one has not ecm-plated training. The selection of replacement RPMs is in progress.
A proposal to increase the RPM staff to a total cf eleven is under consideration.
No items of noncompliance or deviations were identified.
4. Radiation Protecton - Training Twelve hours of radiation protection training for approximately 30 individuals from mechanical, electrical, technical and operational groups are planned for mid March. Operators preparing for hot licensing are completing a review of the training manual in radia-tion protection. Training of the one untrained RPM technician is being delayed until the replacement selection process is complete to permit more efficient use of available training personnel.
No items of noncompliance or deviations were identified.
5. Radioactive Waste - Measurement OC Draft procedure A-4, Chemistry Laboratory Calibration Schedule, was examined. The procedure was complete except for certain data sheets. The procedure specifies chemistry laboratory and counting room instrument service, check and calibration requirements and frequencies. The instruments included in this procedure are as follows: Analytical Balance Gas Partitioner UV-Visible Spectrophotometer Atomic Absorption Spectrophotometer Proportional Counters Liquid Scintillation Spectrometer Single Channel Analyzer Mini-Scaler Multi-Channel Analyzer I
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The plant laboratory has been design'ated as an "Approved Laboratory" by the State of California for certain chemical analyses. The approval was granted by the State in support of the NPDES permit waste discharge sampling program. California, as the enforcement agency for the EPA's National Environmental Policy Act, is requir-ing that all "Approved Laboratories" establish a quality assurance ) program in accordance with EPA's requirements. The QA program docu-mentation is to include organization, records, sampling, procedures, measurement and analysis, instruments, Quality Control procedures and , statistics. At the time of the inspection, the QA program had not been prepared.
No items of noncompliance or deviations were identified.
6. Preoperational Testing Test data from complet(d preoperational tests were examined to assure that the test results were within the previously established acceptance criteria and, as applicable, deviations from acceptance criteria were properly identified and disposed in accordance with the applicant's administrative procedures. Records of the following completed tests which had been approved and accepted by operations were examined.
Test No. Test Description Loop Test 19-7, Rev. 2 Waste filter. 0-3, and Liquid Radwaste Flow to Discharge Tunnel (verification of proper flow indication and recording and valve position).
24.1, Addendum 2 Gaseous Radwaste System Cleaning Procedure - Perform Leak Test 24.2, Addendum 1 Preoperational Test of Gaseous Radwaste System Test Procedure Examined 23.11 Silver Zeolite Adsorber and HEPA Filter Testing Tests Observed I 23.11 Silver Zeolite and HEPA Filter Testing - Containment Hydrogen Purge Syster.. (Observed perform-I - arte of 0^D test on HEPA filters.)
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i FSAR Section 6.2.5, Combustible Gas ' Control in Containment, ; Amendment 41, dated February 1976, specifies that the Containment Hydrogen Purge System is provided with HEPA filters and charcoal absorbers. The inspector was informed that the installed system was equipped with a Silver Zeolite absorber in place of the char-coal absorber. Region V contacted the Licensing Project Manager, NRR, concerning this apparent discrepancy fran the FSAR. The Project Manager confirmed that NRR and the applicant were aware of this apparent discrepancy and that the applicant planned to request amendment of the FSAR.
Startup Problem Report No. 609 on Preop Test 38.4, Radiation Monitoring System, notes that radiation monitor 1-R-22 (also applies Gas to Unit scale 2, 2-R-22g, range of 10 cpm (Decay Tank Discharge Monitor, has a fullFSAR The applicant has calculagd that the expected count rate on this instrument will be 3 x 10 cpm. The applicant reported that the instrument manufacturer has available a desensitization kit which will be obtained and installed. , 7. Process and Effluent Monitoring System The applicant's plans for verifying the manufacturer's calibration of the process and effluent monitors were discussed. The vendors proto-type calibration of monitors was previously examined and* reported, IE Inspection Report No. 50-275/76-03. It was identified that a misunderstanding existed as to the requirments for calibration of the process and effluent monitoring system. The applicant was in-formed that the following inplace calibration or verification of calibration would be acceptable: Inplace calibration: a. Instrument response is determined at a minimum of three (3) points over the instrument range (e.g.,1/4,1/2 and 3/4 of full scale).
b. An initial relationship between concentration of the materials used to calibrate the instrument and the instrument response is established (e.g., counts per second/uti/ml Cs-137).
- c. The calibration is perfonned in accordance with a written
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9rocedure approved by appropriate plant supervision.
i It is acceptable for the initial inplace calibration to be performed using only two (2) specific radionuclides, providing they are of sufficiently different energies to provide an indic?tien of the aaa e v res :nse of the detector (e.g., Cs-137 and Co-60).
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As an alternative to the above procedure, an initial calibration perfonned by the instrument supplier, is acceptable providing
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the following conditions are met: l
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_ a. Calibration certificates or detector response "curves" are available for the specific detectors installed in the systen, b. The applicant has identifiable reference sources supplied by the vendor which are used to functionally check and relate the response of the installed monitors to the vendors calibration.
c. The functional checks with reference sources determine the response over the instrument range at a minimum of three (3) points (e.g., 1/4, 1/2 and 3/4 of full scale).
d. The functional checks have been performed on the installed system using a written procedure reviewed and approved by appropriate plant supervision.. It is acceptable for the initial calibration by the instrument s supplier to be performed using only two specific radionuclides providing they are of sufficiently different energies to check the energy response of the system. It is acceptable for functional checks to be done using reference sources of only one specific radionuclide.
No itens of noncompliance or deviations were identified.
8. 1.aboratory Chemical Drain Tank Failure At the time of the inspection, the applicant was in the process of replacing the originaily installed type 304 stainless steel Chenical Drain Tank with one fabricated from type 316 stainless. The original tank had failed at the head weld due to corrosion caused by chemical laboratory wastes. The failure occurred over an extended period and did not result in the release of any chemical or radioactive wastes, i No itens of noncompliance or deviations were identified.
l 9. Exit Interview l
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The inspector met with applicant representatives (denoted in , Paragraph 1) at the conclusion of the inspection on March 3,1978.
' The applicant was informed that no items of noncompliance or devia-tions had been identified. The inspector summarized the purpose
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and scope of the inspection and the findings. In addition, the misunderstanding with respect to the calibration of the process and effluent monitors was identifed and the applicant vis provided with the information detailed in Paragraph 7. The applicant was also informed of the apparent discrepancy between the the FSAR descrip-tion and the absorbers installed in the Containment Hydrogen Purge System (Paragraph 6).
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.g 3 Task: Allegation or Concern No. 102
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Characterization: l PGaE references unissued drawings in Design Change Notices (DCN). This information was provided by Dr. Meyers and supported by a copy of an November 30, 1983 H. P. Foley Inter-Of fice Conusunication.
. Implied Significance to Plant Construction, or Operation The inspector considers that the allegation implies a failure of the PG&E document control system to issue new drawings as controlled drawings.
Assessment of Safety Significance
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The staff examined the referenced a Foley inter-office communication (IOC) and applicable DCNs, discussed the reason for the communication with the author, and discussed with PG&E document control personnel the apparent failure to issue the IOC referenced drawings to the H. P. Foley controlled files. PG&E records were reviewed by the staff to determine document status of the identified DCN's and drawings.
. The referenced inter-office communication, dated November 30,'1983, identified two DCNs (Nos. DCl-EE-1243R3, dated October 8, 1982 and DCl-EE-1056R8, dated November 8, 1982) which had a total of 14 new
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drawings attached to the DCNs and stated that these new drawings had not been transmitted to H. P. Foley, by P,G4E, for inclusion in the controlled stick files as controlled diagrams of connections.. The IOC
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went on to state that on August 26, 1983, a new DCN (DCl-SE-1331) was
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approved and listed, on the list of affected documents, four of the new drawings issued with the above DCNs. The author pointed out that these four drawings had not been issued by PG&E as diagrams of connection, even though they had been issued as attachments to DCNs about one year earlier. Therefore, the author observed that this practice makes it more difficult for engineering to know which drawing was valid and . pointed out this would increase the probability.for mistakes.
Therefore, the author stated, in the IOC, that Foley document control was not handling che new drawings as controlled documents. This practice by PG&E to list affected drawings on DCN's, which are not in controlled distribution, is considered by the author to create problems of accessibility for field production personnel and makes it difficult for onsite engineering to validate drawings as up-to-date.
. The inspector discussed the referenced IOC with the author and established that the author's concern was that if a future DCNs came to Foley affecting these new drawings, document control would be unable to locate these drawings unless someone remembered these drawings were provided as an attachment to a particular DCN. The engineer (author) stated that he merely wanted t; have these drawings issued to Foley as controlled drawings to ease drawing location in the event that future
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DCNs referenced these drawirgs.
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The inspector exaniaed the PG&E records to determine the status of the referenced DCNs and drawings. Figure 1 1,ists the applicable DCNs were l examined by the staff, for date of DCN issurance, the description of
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the change, and whether DCN is nuclear related and/or safety-related.
. A complete review was made to cross-check the drawing revisions issued by different DCNs, to the drawing revisions listed as the latest approved for construction, in the cotcporate document control and the applicable revision in onsite drawing log.
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DCN Description of Channe Nuclear Re1 Safety Re1
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DCl-EE-1056R8 Upgrade the incore Yes Yes date issued: thermocouple system and 10/21/82 provide isolation between the incore thermocouples and the P-250 computer-per DCN-
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DCl-EM-1111R3. Note: This
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DCN supersedes DCN DCl-EE-1056 R5 DCl-EE-1243R4 Provide design to modify No Yes date issued: existing inputs and to add new-12/21/82 inputs to the emergency response facility data system per j DCN:DCl-EJ-1320 R1
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, . , DCl-EE-1243R3 Schematics and connection No Yes date issued: diagrams for the emergency 9/30/82 response facility data system.
This system is required by the NRC and is a new addition to the new.
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PG4E onsite document control does not maintain drawing revision status of previously issued DCNe containing a pa,rticular drawing issued by the later DCN. .
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(1) The staff concludes that drawings contained in a particular DCN are not distributed to the contractor's document control organizations as controlled documents. Corporate document control fails to issue (to the field) the latest revisions of drawings contained in these DCNs or to update DCN contained drawings to the latest revision.
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(2) It would appear that work performed by the contractor to a DCN, may not be performed in accordance with the latest approved construction drawing revision.
Staff Position The inadequate implementation of DCN drawing document controls may have affected safety-related work performed by contractors. These deficiencies are considered to be an apparent violation of 10 CFR 50, Appendix B, Criteria VI.
NRC Action Followup of licensee response during conduct of normal inspection
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program by regional staff.
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Task: Allegation or Concern No. 102
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ATS No. - - - - - - - - - BN No. ------- .
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Characterization: PG&E references unissued drawings in Design Change Notices (DCN). This information was pr:cided by. h. ":pn-we+ supported by a copy of an November 30, 1983 H. P. Foley Inter-Of fice Consnunication.
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Implied Significance to Plant Construction, or Operation The inspector considers that the allegation implies a failure of the PG&E document control system to issue new drawings as controlled drawings.
Assessment of Safety Significance
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The staff examined the referenced Foley inter-office communication (IOC) and applicable DCNs, discussed the reason for the communication with the author, and discussed with PG&E document control personnel the apparent failure to issue the IOC referenced drawings to the H. P. Foley controlled files. PG&E records were reviewed by the staff to determine doc ment status of the identified DCN's and drt. wings. vb d
. $4d }ls([b d es udev t]sna<$4&51k a%TM. .
The referenced inter-office communication, dated November 30,'1983, identified two DCNs (Nes. DCl-EE-1243R3, dated October 8, 1982;and DCl-EE-1056R8, dated November 8, 1982) which had a total of 14 new fc/
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drawings attached to the DCNs and stated that these new drawings had not been transmitted to H. P. Foley, by P.G&E, for inclusion in the controlled stick files as contro'lled diagrams of connections.* The IOC
. went on to state that on August 26, 1983, a new DCN (DCl-SE-1331) was .
approved and listed, on the list of affected documents, four of the new drawings issued with the above DCNs. The author pointed out that these four drawings had not been issued by PG&E as diagrams of connection, even though they had been issued as attachments to DCNs about one year SO C.
earlier. Therefore, the author observed that this practice makes it more difficult for engineering to know which drawing was valid and O ' pointed out p this would increase the probability' for mistakes.
TCicrwh thor s d that Foley document control
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was not handling the new drawings as controlled documents ,.This Md[ 4*f'** practice @ to list affected drawings on DCN's, which are not in
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controlled distribution, is considered by the author to create problems of accessibility for field production personnel and makes it difficult for nsite engineering to validate drawings as u o date, A tlc # W' nyr mw 4Q - -~~p &hid 2NIaYk m c. h UJ 2hl 4$ri$Wf/fA Y Y The inspector discussed the referenced IOC with the author and established that the author's concern was that if a future DCNs came to Foley affecting these new drawings, document control would be unable to locate these drawings unless someone remembered the e drtwi j M W4s4+4 s were- d 4 Awl W "# [o[ht-provided as an attachment to a particular DCN The engineer (author) 64*%* stated that he merely wanted to have these drawings issued to Foley as controlled drawings to ease drawing location in the event tha future DCNs referenced these drawings.
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The inspector examined the PG&E records to determine the status of he-tdGL referenced DCNs and drawings. Figure 1 1,ists the applicable DCNs were examinedbythestaff,ISIdateofDCNissurance,thedescriptionof
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the change, and whether DCN is nuclear related and/or safety-related.
. A complete review was made to cross-check the drawing revisions issued by different DCNs, to the drawing revisions listed as the latest approved for construction, in the corporate document control and the applicable revision in onsits drawing log.
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DCN Description of Change Nuclear Rel Safety Rel
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DCl-EE-1056R8 Upgrade the incore Yes Yes date issued: thermocouple system and 10/21/82 provide isolation between the incore thermocouples and the P-250 computer per DCN-
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DC1-EM-1111R3. Note: This . DCN supersedes DCN DCl-EE-1056 R5 DCl-EE-1243R4 Provide design to modify No Yes date issued: existing inputs and to add new-12/21/82 inputs to the emergency response facility data system per DCN:DCl-EJ-1320 R1 DC1-EE-1243R3 Schematics and connection No Yes date issued: diagrams for the emergency 9/30/82 response facility data system.
This system is required by the NRC and is a new addition
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PG&Eonsitedocumentcontr$1doesnotmaintaindrawingrevisionstatus
W Y._ __<_,- is,u,j DCNs,____ Y...a_... W
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m . ty u % d a l n n & h w & k u ).
- ' (1) The staff concludes that drawings contained in a particular DCN are not distributed to the contractor's document control organizations as controlled documents. Corporate document control fails to issue (to the field) the latest revisions of drawings contained in these DCNs or to update DCN contained drawings to th'e latest revision. .
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(2) It would appear that work performed by the contractor to a DCN, may not be performed in accordance with the latest approved construction drawing revision.
Staff Position The inadequate implementation of DCN dtawing document controls may have
' af fected safety-related work performed by contractors. These
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deficiencies are considered to be an apparent violation of 10 CFR 50, , Appendix B, Criteria VI.
NRC Action Followup of licensee response during conduct of normal inspection ., program by regional staff. \
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I 7 (Continuedfrom Page A-1) -beated and cooled byliquidirlidelactuaDy two errors that got multi-
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, -r f five o enemeers and found that four_ -eznands and contracts. *
pbed through repeated application.
V V S.- wormetrlor montasf coe case
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- Bechtel's Howard B. Friend, com . He said the. third error was msignif.
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,- . . - . - . :e more taan a yeaC- ore mz pietion manager ior the'Diablo E m _ m&E ceg-.u.-g _and plant, responded, "There was a; icantly Shipley smaH;. t Yp.
said tiie ipes- with the nQ"Ip l Q y 10 q l VV IvI vI Iv iv w ty ass 7 arts. very consoons attemptwhen a new .sapports bout close together camed - l -g e d mas came a to besure maC.3 h ahatJanpnatures below la ' i By Carl M .
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Dy, otrol in generic trainlag," he his checker (inspector) was an,e5 degraes; Fab,2.0,-where thermal ' Smr.it.e.r ._ n .. -m .- . . _penesced,tramed person. - - e - p==iaa would not be a problem.
~ Naclear Regulatory Cor4 ten .
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Ya E d 6 6ms h, bdore any M%. M to Se Wpt. NRC ah'*2aIe are perturned b.y invesu a-i 18 iP ping design:"One esasists or analysis .by anyone, either -enseer Mark:Rartzman objected.
"20. errors;;ese consists of"30 ;traised or untrained.ds made affP :."But thEre's no, guarantee.that the,,.*
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tors, findings of problems h red - - . sign work.at t Diable Car.yc -mdN eterine li sg-4 RXwhd h aMMW-wMW WLE-
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a hw MMi's which eon-y. mad-two, R's aW.-WQQhhneh'*M ~~ ~~
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-thmg was-(donel as wevocid ir.e e l lwreeg and,leformation..And by it's supertiser.who Z,@emble W Q_4Eis espagoced asid gg,7_,'fEBeIhYef?s there Is'. alai so; guarantee that we
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- lised to have seen it," tne NECs l . - Yin salt!, bewever, that the design . dent havera ,,,Mm that says the Tom Bishop said Tuesday. - fg, . errors he found had gone through ' engtseef shall_ met:do, ridicidons g -g
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NRC afficials are concerned that a*! three levels - design, checking thags.7gg- ~*-, . ,
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design enpace we mt nam , , ;; . by the engiacers to rectify ,,",gd"I g ied p
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representat' ire George i
.. _PD*'I7l""d ens were placed a h to .
PG&E, aided by Bechtel Power J-tsether supports that they cosid inter- Addent Plant Design Chief lar . day en the NRC:cencerns, saying l Corp u,p a Wear rmw Mfere with each others' functions. ry Shiol y said that toe groups of 3 PG&E cfficials wanted to wait !!!! l' e A l
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C er r e 'I . ubl:ms ecc'd1pecur when tne ppe ar.3 7v uma tired Ur-Ym -us-~thcNRCretteirwn templete. Sq _
-.after PG&E.Jeported that some
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} ;-v==tigata==-of same:sf the design ' , ; .. work last Noeembee4fter.recemng ; 7 anegmeinsa of problems. 2: .. . - , . . Ameeg.these were charges by , l Charlea_ "
Stokes, a fogmer PG&E
, * . .- Lrelais==d be was laid j ll haffggetsber-is; retribution for... -
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munc . H -frm=enthned .--
' !: the insjera. a==nerns: al a".Dec'. 15 i i Tmieetingiwiik'.PG&E-officials, a . , ' .. transcript?sLwhich.was receved . !
I Tasaday by the T '. _ Thbaser-
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3 .". * Yin said be andst4d the records
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,- - - - - ' - ' (Continuedfrom Page A-1) - beated and cooled by liquid iriide',, actuaD'y two errors that got mrJti-
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-r f five e emeineers and foisd that four_ - ernands and contracts.
,5=_orsettor mentas ~Leoe case m 'phed threagh repeated application.
Bechtel's Howard B. Friend, com . He said the. third error,was msignif-pore than a year ~ 3rfore' 6etnz pletion manager for the'Diable icantly enan . f,, . .
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l ]Q]- {lg vlOq r, in W&E engmeerung and plant, R;--td, "There was a : Stupley said tiie~ pipes with the
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very conscues attempt 1 rhea a sewi . supports hullt close together camed W assuranggarecedures._ k of mas came en the job to be sure that..hgaids.aU~.,as belo _
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==Hy ther9 was a lac atM
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- -g%;emie training F i Staff Wr. ,_ iter- ' - - -;- - " he,_penesced,traised his Mw (inspector) was sa'JNE erpaneian person.
degreet ahrenha&where thermal would not be a problem.
_ ~ ! ' Neclear Regulatory Cere'4en
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E h. h any calcidatut . W to tim transcript, NRC '
; arnmle are perturned by inves:1:2-la ppeng desage:"One consista or analysis .by anyone, either]-empneer: Mark:Rartz I
ters' findings of p obleins h re:~- : .tNermsLene consists M88028I'* ethegene is,of'M,3W aratraised#tande,affi-L"But there's so. guarantee.that in
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sign work.at the Diablo Car.ycr. I asseC4 Rwcheded Wa
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muelear power plant bonie Monnabwhich con-y,.and% rs apprwed;Wm by inherethelinewasedegrees."4,h% '
.' "R_does not appear that evers- j
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e , t -thing was-(doneras we war.ld r.2 e 1
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I , lized to br.ve seen it." tae .*ir.Cs i
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own people, appened gg.-- Yin aald, however, that the desaga_ -des't havera preIcedure that r.ays the Tom Bishop said Tuesday.
NRC afficials are concerned that l.-
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errors he found had game through " engineeEshall metido ridiculous
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* T8'. bsaM meme ym,pe andW aR three levels - desaga,mPrewatative checknag~ #
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design engmeers were not trarred review-and resahed la physi - - properly and made errors. ' d,T y timwere engmeers to W Gewge
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placed so close to calmnearembs.w of the pl.ast:-- . e Sarhema dechmed te comment to-
- PM aided by Bechtel Power -arV"--- .
' Corp cand=ried a two-year review Ww that they could inter- NM Plant Design N h day en the NRCconcerns, saymg
~ of the plast's earthgeake safety pere with each obers' funcuons. ry Sini y said Bat toe greps O PG&E deials wanted to wac S
*
design.Ibe NRC ordered the review _j. ems cddoccunrben tne ne anM - -'or-YNuw NR mNine.
-after PG&E; reported that some
- ===nie denga pleas were reversed.
s:Ite .Insached as in - ; ,
{ -vestigation of aseme:af. the design !
'
= l work last Novennber after.recemng _
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muer=H=== af pranle=* 4 T. . . . - -
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l' ' Z Among these were charges by - i
. Charles Stakes,~a former PG&E ,
i .:e .4; elai ==I be.was laid ,
,
) I haffggeleber.is;retribetion for ' I i f:liling separtd eHamigiTd*Arma -ies i..~: . I ,
^ " '- '
3:.~;; " , _ - r henthned .- , Otbninajormama ras:st a.Dec.15 -i T_aneeting'with1PG&E-officials, a . , *
..transcnpt.LeLwhich.was receved .
Turaday by the Tedegram.7hboner
. . . * Tan said Caedstad the records h M y g See @ C,Page M ,
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balone DlIblo' 0 : Alliance Project Office l DearDiabloCanyonWorkdra I Many of you have previously received Abalone Alliance leaflets addressing our concerns regarding the Diablo Canyon Nuclear. Power Plant. We are not alone in our opposition to the plant. A recent poll has shown that 62% of local residents do not want Diablo to operate. Also, some of you have called the Diablo Project Office to eLpresa your misgivings regarding the plant's design and construction.
Since the IRC has allowed M&E to load fuel, many of us will now be joining together to express our opposition to tne planned operation of the plant. @ is effort is not a new one - the Abalone Alliance has been opposing Diablo since 1977. It has been a long and difficult struggle, and it is not over yet. We are determined to do all we can to protect the lives and safety of the people of our community.
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he next accion, the People's Emergency Response Plan, will begin on Friday, January 13, 1984. This action is designed so that everyone who wi may partici-pate, as 1 cog as they agree to the Abalene Alliance Nonviolence . This accion will include walks, vigils, leafletting and civil disobedience. All of these activi-ties will create an on-going presence at. Diablo and in our community. . '
. . r, ., .. ,
w , Individuals who decide to participate in civil disobedience must attend a non-violent preparation and briefing prior to their participating in the action. The Peoples's Emergency Response Plan will be a peaceful statement of our commitment and our beliefs. As you encounter those of us who are participants, we would ask that you understand and respect this.
,..n.
cxxr.mitment., . .- , If you have any questicns regarding the action, or want to share information with us, please call the Diablo Project Office, at 543-7518.
THE CEREACH CDILECTIVE . I W !OWIOLDXE OCDE.
1. Our attitude will be one or openness, friendliness, and respect towards all people we encounter. -
.
2. We will use no violence, verbal or * physical, tcvard any person. ' - - 3. We will not damage any property.
4. We will not bring or use any drugs or ' g alechol other than for medical purposes.
5. We will not run.
6. We will carry no weapons. ! 452 Hovera $Peet . Sa%ues Ctrsoo CA 93401 e EC5 5434614 J
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. C0NSTRUCTIDH $TATUS REPQRT PAGE NO: 2-036 DIABLO CANTON 1 N U C L'E A R POWER PLANTS STATUS AS OF: 06-30-82 PL AN T CNARACTERIST I C 5: KEY PERS0NNEL
'
DOCKET NO: 05000275 CORPORATE CONTACT: PHILIP A. CRANE, JR., ASSISTANT l CP NO. 1 DATE ISSUED: CPPR-39 / 04-23-68 GENERAL COUNSEL CP EXPIRATION DATE: 04-30-79 CORPORATE ADORESS: 77 BEALE STREET A P P L I ' C" PACIFIC CAS 1 ELECTRIC SAN FRAf3CISCO, CALIFORNIA 94106 PLAHI L U:: A T ION: 12 MI WSW OF SAN LUIS OBISPD IE REGIDH: 5 CORPORATE PHONE NO: (415) 788-4291 AE: PACIFIC CAS 8 ELECTRIC tt SSS : WESTINGil0USE CONSTRUCTOR: PACIFIC CAS 8 ELECTRIC NRC LPM: B. BUCKLEY REACTOR TYPE: PWR IE PROJECT INSPECTOR: DER (l:NE) : 1084 IE RESIDENT IllSP EC T OR : M. J . BAGAGLIO I fl 5 P ECT ION 5 T A T U 5: l FUEL L0AD DATE C lt A N G E S l TIME PERIOD AS REPORTED BY UT I L I T Y: Ay , 19,2 - Jun. 198_2 REPORT NEW REASON FOR CHANGE REPORT BEG DATE END DATE DATE 1QiSQUE RLP,pjTJD BY tlTILITY EUtpf3 0F INSPm pF INSP. . 1 04-00-74 01-00-75 SLOW DOWil IN WORK RESULTING FROM WILDCAT 82-12 01-12-82 03-04-82 STRIKE BY ELECTRICIANS WHEN SECOND SHIFT 82-13 04-19-82 04-23-82 WAS ADDED It4 STEAD OF OVERTIME. INExPERI-82-14 04-0'-82 05-01-82 ENCED NEW NIPES IN ADDITION TO REDUCED 82-15 05-24-82 05-28-82 ELECTRICIAN WORK FORCE CAUSED THE DELAY.
82-16 05-10-82 05-14-82 09-00-74 04-00-75 A STRIKE BY PIPEFITTERS IS STILL IN EFFECT
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82-17 05-85-82 05-15-82 AT THE SITE. IHE LICENSEE ESTIt1ATES AT 82-18 04-26-82 04-30-82 LEAST A 48 DAY FUEL LOAD SLIP AT THIS 82-11 05-02-82 06-04-82 TIME BUT WILL FVALUATE THE EFFECT ON THE 82-20 05-06-82 05-28-82 DVERALL SCHEDULE AT COMPLETION OF STRIKE.
82-28 06-06-82 07-03-82 10-00-74 10-00-75 LICENSEE'S EVALUATION OF STRIKE EFFECTS.
82-22 05-26-82 07-05-82 08-00-73 02-00-76 RETUBING OF MAIN CONDENSER WITH TITANaUM TUBES.
Inspection Reports are available at local PDRs. 10-00-76 03-00-77 DECISION TO PARTIALLY REBLADE PORTIONS ' OF THE LOW P2 ESSURE TURBINES.
01-07-77 05-00-77 REBLADING OF Tile LOW PRESSURE TURBINES.
04-04-78 08-00-78 REVIEu 0F PREOPERATIONAL TEST PROGRAM AND COMPLETI0tl AND APPROVAL OF SEISMIC MODIFICATION PROGRAM.
02-00-79 ES tit 1A T ED Dr.TE. HRC CONTROLLING.
08-00-80 01-00-81 ESTIt1ATED DATE. NRC CONTROLLING.
09-00-81 03-00-81 fl0 REASON KNOW?l.
12-00-88 03-03-82 ESTI!1ATED DATE.
03-00-82 N/S L 0 CAL PUBL IC D0CUMENT R 0 0 N: SAN LUIS UDISPD COUtiTY FREE LIBRARY 858 rlaRRO S1REET SAN LUIS OBISPD. CALIFORNIA 9340(.
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) . - . _ _ . - - . _ . . _ . . - - - . . - - - - - - .N '* .
. C0N5TRUCT ION 5TATUS REPORT PAGE NO: 2-037 DI Ant 0 Cath 0H I NUCL EAR POWER PLANTS STATUS AS OF: 06-34-82 C0N 5 I RUCT ION 5TA T U 5:
.
ArrtICA;11'S CURREt:T EST!!14TED FUEL LOADING DATE: N/S Ari l lC Ata t 's PREVIOUS ES tit 1AT ED F UEL L O ADitJG D?.1Et APPLICANTS CONSTEUCTIDH COMPLETION ESTIMATE AS 03-00-82 DEFINED Ill CP: (EdRLIESTs trrtICANI'S ESilt:AIED FUEL LCADING DAIE AT CP ISSUANCE: 04-00-72 (LATEST) Arr L IC AllT * S CURRL14I ESII*1ATED CD:'N RCIAL CPERA110t! DATE: N/S I CUR R E!I T ESTIMAT ED PERCElli CONSTRUCTION COMPLEIE: 99% CC;:::::4TS RfC^.R3*t:G CUOREllT C0tiSTRUCTI0l4 ST ATUS: FUEL LO ADING L ICENSE SUSPENDED BY THE NRC. NO UPDATE RECEIVED F00 TiiE SECOND QUARTER OF 1932.
ESTIMATED ACTUAL ESTIf1ATED ACTUAL ESTIMATED NOBILIZE Al;D PREPARE SITE: START SJARI_ .Cor1Pt ET iq!i SprjP_1 EIJ.DH X..CGMPLETE 06-14-68 12-00-69 300% l' TACE STRUCIURAL C0ttCRETE: 12-00-69 100% I:iSI AL L REACTOR PRESSUF:F VESSEL: 02-00-73 03-00-73 800% IllSlit L LARCE EURE PROCESS PIPE:
.
It33 f ALL L ARCE EORE PIPE I!A!4GERS, RESTRAlti!S APID SNUBBERS: Ifis: ALL CIA L L FORE PIPE: INSIALL CADLE TRAY: IfiS T AL L EXPCSED METAL CONDUIT: It:S T AL L POMER C0!4 TROL. I!STRUMENTATION AttD SECURITY CABLE: INSTALL EL EC1RICAL TERMIt:ATIONS: C 0t:DUC T RF ACIOR COLD II(CROSI ATIC TEST: 06-00-75 08-00-75 100% CritDUCT I!OT f tt:ICTIct;AL TEST: 11-00-75 02-00-76 100% C03: DUCT PREOP At3D ACCEPTANCE TESTS NECESSARY FOR FUEL LOAD:
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d . C0NSTRUCTION S T=TUS REPORT PAGE NO: 2-038 DIABLO CANYON 2' NUCLEAR POWER PLANTS STATUS AS OF 06-30-82 PLANT CNARACTER ISTICSt KEY P E R S O N N E L: DOCKET NO: 05000323 CORPORATE CONTACT: PHILIP A. CRANE. JR. ASSISTANT CP NO. 1 DATE ISSUED: CPPR-69 / 12-09-70 GEt!ERAL COUNSEL CP EXPIRATION DATE: 10-31-79 CORPORATE ADDRESS: 77 BEALE STREET APPLICANT: PACIFIC CAS S ELECTRIC SAN FRANCISCO, CALIFORNIA 94106 PLANT LOCATION: 12 MI WSW OF SAN LUIS OBISPO IE REGI0ti: S CORPORATE PHONE NO: (415) 781-42tl AE PACIFIC CAS 8 ELECTRIC NSSS: WESTINGil0USE CONSTRUCTOR: PACIFIC GAS & ELECTRIC NRC LPM: B. BUCKLEY REACTOR TYPE: PWR IE PROJECT INSPECTOR: DER (MWE): 1:06 IE RESIDElli If1 SPEC TOR : M. J. BAGAGLIO IN5PECTION 5 T A T U S: FUEL L0AD DATE CNANGES TIME PERIOD A5 REPORTED BY UTILITY Aor. 1982 - Jun. 1982 REPORT NEW REASON FOR CilANGE REPORT BEG DATE END DATE PATE $CHFDM11 RFPORTEn BY UTILITY NUMBER OF INSP. OF INSP.
09-00-74 02-00-76 A STRIKE SY THE PIPEFITTERS IS STILL IN 82-07 01-12-82 03-04-82 EFFECT AT THE SITE. THE LICEllSEE 82-08 04-19-82 04-23-82 ESTIMATES AT LEAST A 48 DAY FUEL LOAD SLIP 82-09 05-10-82 05-14-82 AT THIS TIME BUT tilLL EVALUATE Tile EFFECT 82-10 05-06-82 05-28-82 ON Tile OVERALL SCilEDULE AT Tile Cot:PLETION 82-11 05-26-82 07-05-82 CF Tile STRIKE.
10-00-74 07-00-76 STRIKE.
Inspection Reports are available at local PDRs. 08-31-75 02-00-17 1 RETUBillG OF MAIN CONDENSER WITH TIT Atesti:1 TUBES (2-3 MotiTils).
2. ItiSI AL L ATI0t3 0F ADDITIO!!AL MAIN STREAM RUPTURE RESTRAINTS (4 MONINS).
10-15-76 08-00-77 ELECTRICAL INST ALL ATION PROBLEt:5.
01-07-77 12-00-77 A GEttERAL REEVALUATION OF THE PP.0 JECT SCitEDUL E.
01-04-78 01-00-78 NO REASON GIVEN.
02-07-78 03-00-79 GENERAL REEVALUATION OF PROJECT S CliE DU L E . 04-14-78 02-00-80 GENERAL REEVALUATION OF PROJECT SCliEDULE.
02-00-50 03-00-80 NO REASON REPORTED.
05-00-80 03-00-81 NO REASON PROVIDED.
09-00-81 07-00-82 NO REASON PROVIDED.
06-00-82 N/S L0 CAL PUBL IC D0CUMENT R00M SAN LUIS OBISPO COUNTY FREE LIBRARY 888 MORRO SIREET SAN LUIS ODISPO, CALIFORNIA 93406 _ _ _ _ _ ___
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o.
. C0N5TRUCTION STATUS REPORT PAGE ND: 2-039 DIABLO CANYON 2 NUCLEAR POWER PLANTS STATUS AS OF 66-30-82 CONSTRUCT ION 5 T A T U S: APPLICANT'S CURRENT ESTIMATED FUEL LOADING DATE: H/S I APPLICANT'S PREVIOG", ESTIMATED FUEL LOADING DATE: 03-00-81 APPLICANTS CONSTRUCTIDII COMPLETION ESTIMATE AS APPLICANT'S ESTIMATED FUEL LUADING DATE AT CP ISSUANCE: 11-00-74 DEFINED IN CP: (Fr.RLIEST)
(tATEST)
APPLICANT'S CURRENI ESTIMATED COMt:ERCIAL OPERATION DATE: N/S CURRENT ESTIMATED PERCENT CONSTRUCTION COMPLETE: 98% COMMENTS REGARDING CURRENT CONSTRUCTION STATUS: NO UPDATE RECEIVED FOR THE SECOND QUARTER OF 1982.
. ESTIMATED ACTUAL ESTIMATED ACTUAL ESTIMATED l MOBILIZE AND PREPARE SITE: 5 TART iTAEL $DElflJpff $0fEL ET1911 7, _f 05PMF i 0 6- M- 68 05-00-73 800% PLACE STRUCTURAL CONCRETE: 05-00-71 100% INSTALL REACTOR PRES'SURE VESSEL: INSTALL LARGE BORE PROCESS PIPE: INSTALL LA"GE BORE PIPE HANGERS, RESTRAINTS AND SNU3BERS INSTALL SMALL BCRE PIPE: INSTALL CABL E TRAY t I:4 STALL EXPOSED METAL CONDUIT INSTALL POWER. CON!ROL. INSTRUMEHi? TION AND SECURITY CABLE:
INSTALL ELECTRICr.L TERMINATIONS: 1 CONDUCT REACTOR COLD HYDROSTATIC TES7: CONDUCT HOT Ftr4CTIONAL TEST: CONDUCT PREOP AND ACCEPTANCE TESTS NECESSARY FOR FUEL LOAD:
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REGION V -- R1 \
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.U S HUCLEAR REGULATORY COMMISSION REGIONAL ADMINISTRATOR, REGION Y g ~
MO 1450 MARIA LANE, SUITE 210 HALNUT' CREEK, CALIFORNIA 94596
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COMMERCIAL NUMBER * _ 415-943-3700_ F T S NUMBER 463-3700
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TELEPHONE ATTENDANT DUTY HOURS ARE 7:30 - 4:15 , _ P* 80 PHONE
. . om ;
$$ -~'-
. o8 DIVISION OF RESOURCE MANAGEMENT & ADMINISTRATION RICHARD GILBERT, DIRECTOR........................... 46-33710 1j 4+
ADMINISTRATIVE SERVICES BRANCH IDA ALEXANDER, CHIEF.......................... 46-33793 PROGRAM SUPPORT BRANCH KATHLEEN NAMILL, CHIEF........................ 46-33743 * .
,
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ENFORCEMENT OFFICER "
.' O ALLEN JOHNS 0N..................................... 46-33739 ;e
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DIV 0F RESIDENT, REACTOR PROJ & ENG PROGRAMS *
,[
JESSE CRENS, DIRECTOR............................... 46-33735 ;g REACTOR PROJECTS BRANCH 4 1 E DANIEL STERHBERG, wa CHIEF....................... 46-33737 og r
- < -e REACTOR PROJECTS BRANCH 4 2 --l ** "
TOM BISHOP, CHIEF............................. 46-33761 g EN DIV 0F RADIOLOGICAL SAFETY & SAFEGUARDS PROGRAMS EG
~w ROSS A. SCARAND, DIRECTOR........................... 46-33717 N"
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RADIOLOGICAL SAFETY BRANCH d5 HERBERT BOOK, CHIEF........................... 46-33755 5I OQ SAFEGUARDS & EMERGENCY PREPAREDNESS BRANCH E-LEROY NORDERHAUG, CHIEF....................... 46-33775 E$
. "a STA1E LIAISON OFFICER 2m DEAN EUNIHIR0..................................... 46-33714 *5 au PUBLIC AFFAIRS OFFICER E$
JAMES G. HANCHETT................................. 46-33712 .[ m
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(2 /S) 3 37- 5~o00 F T.S 128-/000 (_
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I. ONSITE PROJECT ENGINEERING GROUP (OPEG) VERIFICATION ACTIVITITES . A. Training - R. Oman B. Document Control - R. Oman C. Technical Issues - L. Shipley 11. EXPANSION ANCHORS - D. Rockwell/K. Buchert 111. DISCUSSION
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Of f icial Attendance Record for
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J.ngine : Department Quality Program Training ^ 4 , DATE-
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DATE REQUEST DURMiON COMPLETED REC ID ATTENDEE SUPERVISOR DISC. COURSE CONTENT REQUESTED BY RECEIVED (HOURS) MO DAY YR I182 RAO, P. V. OMAN RC M&NE SFPD QA Indoctrination 1 5-4 83 1243 KATEL, BilAILAL R. M&NE SFPD QA Indoctrination 1- 5 5 83 675 RATH, B. K. CE EMP 3.3: Design Calcs CRANSTON CV 1 3 21 83 874 H ATil, 11 . K. CE' SFPD QA Indoctrination CRANSTON CV 1 5 12 83 383 RAVINDRANATil, R J OMAN RG M&NE Engineering Manual Survey OMAN RG 2/18/83 4 2 18 83 1193 RAVINDRANATil, R J OMAN RG M&NE SFPD QA Indoctrination 1 5 4 83 Ii14 RAY, A. CE SFPD QA Indoctrination 1 5 2 83 384 REARDON, EUGENE A. OMAN RC CE Engineering Manual Survey OMAN RG 2/8/83 4 2 18 83 1169 REARDON, EUGENE A. OMAN RG CE SFPD QA Indoctrination 1 54 8 'i 1492 REARDON, EUGENE A. OMAN RC CE EMP 3.6 8 3.60N 1.5 6 24 83 385 REAS , .J . 11. OMAN RG M&NE Engineering Manual Survey 3/14/83 1208
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R.EAS, J.--II .
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OMAN RC M&NE SFPD QA Indoctrination OMAN RC 3
Q5 483 83 386 RECINELLA, E. J. OMAN RC M&NE Engineering Manual Survey OMAN RC 2/8/83 4 2 17 8L 1206 RECINELLA, E. J. OMAN RG M&NE SFPD QA Indoctrination 1 5 4-- 81 1I15 REDDY , K. II. CE SFPD QA Indoctrination 1 5 2 8 'l 1400 REGALADO, DOLORES M&NE SFPD QA Indoctrination 1 6 9 83 1360 REGALADO, J. T. M&NE SFPD QA Indoctrination 1 6 7 83 1661 REGANIT, CECILIO McCALL JK CE Engineering Manual Survey MERCADO PV 4 9 21 83 387 REMOTICUE, J. R. M&NE Engineering Manual Survey I.EONG JJ 1/18/83 4 3 2 83 770 REYNOLDS, R. W. LAI M MSNE Engineering Manual Survey LAI M - 4/25/83 4 5 12 83 1103 REYSNER, P. I. CE SFPD QA Indoctrination 1 5 2 83 388 REZNIKOV, S. N. CE Engineering Manual Survey LEONG JJ 4 4 7 83 1557 RICllARD, CONNIE E. McCALL JK CE Engineering Manual Survey MERCADO P 8/12/83 4 8 31 83 740 RICllTER, L. A. M&NE Engineering Manual Survey OMAN RG 4/19/83 3 4 19 83 1530 ROBERSON, BRUCE F. McCALL JK CE Engineering Manual Survey LEONG JJ 4 7 26 83 389 ROBl.ES, D. McCALL J CE Engineering Manual Survey LEONG JJ 2/24/83 4 3 11 83 1344 ROCilA, OSCAR CE SFPD QA Indoctrination 1 6 6 83 1558 ROCERS,ilERBERT L. McCALL JK CE Engineering Manual Survey LEONG JJ 4 7 29 83 390 HollA, A. Sl'All N CE Engineering Manual Survey LEONG JJ 2/24/83 -
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391 ROURK, R. J. CONNELL EC M&NE Engineering Manual Survey CONNELL E 1982 4 4 29 83 1559 ROWE, SCOTT E. McCAI.L JK CE Engineering Manual Survey LEONG JJ 4 8 17 8i 393 ROY, S. S. CE Engineering Manual Survey LEONG JJ 2/17/83 4 3 11 81 1116 ROY, S. S. CE SFPD QA Indoctrination 1 5 2 BT
* Ita s been scheduled in the Module No. listed 1st - Attended 1st session only ** II.is been rescheduled in the Module No. listed 2nd - Attended 2nd session only- *** To be scheduled o Attended module number listed
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Page 1 of 25 Pacific Gas and Electrical Company * Engineering Quality Control Dept.
Prepared by M. M. Sweeny 0FFICAL ATTENDANCE RECORD FOR TRAINING CONDUCTED BY ENGINEERINC QUALITY CONTROL IN Tile YEARS 1981 AND 1982 ._ FOR MECIIANICAL AND NUCLEAR ENGINEERING AS OF DECEMBER 9,1982 Date l Attendee Course Content Conducted by Duration Completed Remarks
! ! ABACAN. T. Engineering Manual Survey EQC-MSweeny 4 hrs 7/22/82 ! ABREU, K. E. Engineering Manual Survey EQC-MSweeny 6 hrs 3/18/82 j ABI-SAMRA, J. Engineering Manual Survey EQC-MSweeny XXXII/XXXV/XXXIX No longer with project pe- , .
J. Leong's memo of 11/12/2 i AGOOT, M. Engineering Manual Survey EQC-MSweeny XXXI/XXXVI/XXXX Sent letter to Supervisor
. (M. Tresler) dated 11/15/!:
l - On site per J. Leong by phone on 11/22/82 AIIMAD, M. I. Engineering Manual Survey EQC-MSweeny 4 hrs 11/19/82
@fKNDY,-Erdfp Engineering Manual Survey EQC-MSweeny 4 hrs 7/22/82 AESAN0 7 S*.B'. Engineering Manual Survey EQC-MSweeny 4 hrs 7/22/82 ALIDO, E. Engineering Manual Survey EQC-MSweeny 4 hrs 10/14/82 ALSilIllRAIDA, S. A. Engineering Manual Survey EQC-MSweeny 4 hrs 6/17/82 ANSARI, F. Engineering Manual Survey EQC-MSweeny 4 hrs 10/14/82 ANTAKI, C. Engineering Manual Survey EQC-MSweeny 3 hrs 4/20/82 ; ANTE, J. A. Engineering Manual Survey EQC-JKilian 6 hrs 12/18/81 Review EMP 3.6 ON Rev 0: Operating Nuclear PP Design Changes EQC-MSweeny I hr 7/27/82 ANTIOCIIOS, P. C. Engineering Manual Survey EQC-MSweeny 6 hrs 3/3/82 Review EMP 3.6 ON Rev 0: Operating Nuclear PP Design Changes EQC-MSweeny I hr 7/9/82 DC Project Engineers' Instructions f's 1 thru 8 EQC-EKahler 1 hr 9/28/82 ARAM, S. Engineering Manual Survey EQC-MSweeny 1st XXXIX - for 2nd session -
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No longer with project pe ERKahler's memo of 12/8/8 ARIEY, A. F. Engineering Manual Survey EQC-MSweeny 6 hrs 3/10/82
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* llave scheduled to attend classes. Ist - attended 1st session only.
- To he scheduled. 2nd - attended 2nd session only.
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Pacific Ca.t and Electrical Company f .,\ Pcg2 10of25[ -
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Engineer 1 / hality Control Dept. ~ frepared t. . M. Swerny 0FFICAL ATTENDANCE RECORD
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FOR TRAINING CONDUCTED BY ENGINEERING QUALITY CONTROL , IN THE YEARS 1981 AND 1982 FOR HECilANICAL AND NUCLEAR ENGINEERING AS OF DECEMBER 9,1982 Attendee Date Course Content Conducted by Du ra tion Completed IIUANG, N. K. Remarks Engineering Manual Survey EQC-MSweeny XXXV/XXXIX IIUANC, S. No longer with project per HURT, J. Engineering Manual Survey ERKahler's memo of 12/8/82 Engineering Manual Survey EQC-MSweeny 4 hrs 12/8/82 , 11UT11. C. A. EQC-MSweeny 4 hrs 5/19/82 i IlWANG, J. II. Engineering Manual Survey EQC-MSweeny Engineering Manual Survey 4 hrs 8/5/82 IRONMONCER, R. S. EQC-MSweeny 3 hrs JACOB, C. Engineering Manual Survey 4/20/82 Engineering Manual Survey EQC-MSweeny 4 hrs 6/17/82
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EQC-MSweeny XXXIII/XXXVI/XXXX No longer with project per JACQUEL_ OADELRAB, E. 11.
L_______ Egg.ineerlIu; Manual Surven__ ERKahler's memo of 12/8/82 Engineering !!anual Survey EQC-MSweeny 2 hrs 6/9/82 EQC-MSweeny _ 4 hrs DC Project Engineers' Instructions 10/27/82 JAllANCIR, N.
- 's 1 thru 8 EQC-EKahler
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JAIN, M. Engineering Manual Survey I hr 9/28/82 Engineering Manual Survey EQC-MSweeny 4 hrs 9/23/82 JAMES, J. Z. EQC-MSweeny 4 hrs JANC, 11. Engineering Manual Survey 10/27/82 EQC-MSweeny 6 hrs Review EMP 3.6 ON Rev 0: Operating 3/18/82 JAYARAMA, M. P. Nuclear PP Design Changes EQC-MSweeny I hr 7/9/82 IEAN, J. S. Engineering Manual Survey Engineering Manual Survey EQC-MSweeny 4 hrs 10/14/82 IEUNC, G. P. EQC-MSweeny 4 hrs Engineering Manual Survey 9/23/82 EQC-JKilian 6 hrs 12/18/81 Review EMP 3.6 ON Rev 0: Operating IILWAN, E. Y. Nuclear PP Design Changes EQC-MSweeny Engineering Manual Survey I hr 7/9/82 '0AQUIN, Y. D. EQC-MSweeny 4 hrs ONES, J. E. Engineering Manual Survey 7/22/82 Engineering Manual Survey EQC-MSweeny 4 hrs 5/19/82 e OSE, V. A. EQC-MSweeny I Engineering Manual Survey 4. hrs 10/27/82 EQC-MSweeny 3 hrs 11/8/82 *
* !! ave scheduled to attend classes. Ist * To be scheduled. - attended 1st session only.-
2nd - attended 2nd session only.
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Quality Assurance Records H. P. Foley QA Manual Section XVII, Rev. 9 dated August 15, 1983,
"provides the methods to assure sufficient documentation is prepared to ' provide objective evidence of the quality of items and the activities performed." Paragraphs 3.2 and 3.3 of Section XVII tre as follows:
3.2 Procedures shall specify responsibilities for transmittal, distribution, review, maintenance, and turnover or retention.
' 3.3 Records shall be identified, legible, and protected against damage, deterioration, or loss.
Contrary to the H. P. Foley QA Manual Section XVII, the following conditions exist. x._ 1. Contrary to paragraph 3.2, a procedure does not exist which provides instruction or guidance for the review, maintenance, and turnover or retention of quality records.
I 2. Contrary to paragraph 3.3, it appears that quality records have I been found in the H. P. Foley attic which is not a controlled storage location.
3. Contrary to paragraph 3.3, the H. P. Foley records storage voult does not provide protection against internal fires.
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( 4. Contrary to paragraph 3.3, records taken out of the records =, storage facility are kept in areas which preclude protection from damage, deterioration, or loss. These records have been observed
,- to be kept overnight in the comumn work areas and not protected from unauthorized personnel or work area hazards (water, loss, fire, etc.).
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' Task: Allegation or Concern No. 97 (c
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'ATS No. RV83A063 BN No, N/A:
Characterization SitedesignengineerhavenotbeenrequiredtoEorkusingcontrolleddocuments, resulting in~the use of different design assumptions among other problems.
Implied Significance to Plant Design, Construction, or Operation See Task Allegation or Concern 79
.
f Assessment of Safety Significance See Task Allegation or Concern 79 Staff Position See Task Allegation or Concern 79 ' Action Required See Task Allegation or Concern 79
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Task: Allegation or Concern No. 79 ( ATS No.: RV83A063 BN No.: N/A
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Characterization
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Site design engineers were not required to work using controlled documents resulting in different calculation bases, load rating, and allowables applied to their work, a Implied Significance to Design, Construction, or Operation Engineers are calculating stresses in piping in a variety of ways. Without i uniform design bases, formulations, and acceptance criteria, the adequacy of plant system safety cannot be verified and assured.
. Assessment of Safety Significance The staff reviewed engineering manuals, directives, and procedures located at onsite engineering offices to assess the degree of standardization, currency and availability of design documents. Six design engineers performing on-site design activities were interviewed as part of the review. The staff identified three instances of out-of-date engineering documents and several cases of the availability of technical articles and data not related to the design of Diablo Canyon. l
1 It was determined that there was only one set of controlled procedures ; maintained in the stress analysis group. Within this set, six procedures were _ __ _. ___
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selected for examination. Three of the six were not current, or should have e been deleted and replaced.
There was evidence of inadequacies in document control such as inconsistencies in procedure lists maintained by different supervisors in design groups and confusion about who has responsibility for maintenance of procedures and
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s drawings.
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Seven The staff reviewed the calculations fo jdifferent designs, most of which involved many revisions to the original calculations. The staff identified apparent errors in design calculations which were not detected by licensee design verificati Three e lculations (MP-397 MP-465 and MP-691) contai ed e ors but re iewtrs id not pr vide adve se comm ts. Ap arent trors were e dis vered by t e staf in calcul tions MP- 2, str s calcu tion .1 501, (. - and s ss calcula ton No. -314.
J ( Field inspection of large and small bare piping supports indicated that the snubber operability may be affected by the installation of rigid restraints in close proximity to the snubbers. The staff also has concerns about the physical configuration of and interaction between rigid restraints and anchors.
In these cases, the possibility exists that snubbers and rigid restraints may not be able to accept the intended des'gn loads.
Based on the above concerns the staff questioned the effectiveness of the licensee's quality assurance audits. Three audits of onsite engineering were conducted within the past year and should have identified significant problems, but did not.
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Several of the above areas appear to be in noncompliance with NRC requirements.
(w The staff orally presented the preliminary findings to the licensee in a management meeting on December 8,,1983. The ifcensee stated that some adminis-trative controls may have been lacking, but that the final design is adequate and free of any significant design errors. Rationale for specific designs, such as the assumptions used, were discussed. There is some disagreement as to the acceptability of certain design assumptions, such as the affect of pre-qualified support members, between the staff and licensee engineers. The resolution of the acceptability of design assumptions will determine if the apparent errors detected by the staff are significant.
Staff Position
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The staff concludes that the administrative controls imposed on the engineering activities are inadequate in some cases, and ineffectively implemented in others. Design verification and document control are inadequate. Quality assurance audits are not effective in identifying problem areas.
Action Required bNS The staff mus.t perform additional review of the licensee's snubber / restraint
. design and design assumptions. Region V followup on noncompliance response.
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Task: Allegation #55 ATS. No.: RV-83A50 BN No.: 83-171(10/27/23) Characterization Bechtel has purposely approved analysis of small bore pipe supports that have failed by altering current documentation that shows failure of piping systems and pipe supports. , Implied Significance to plant Design, Construction or Operation Small bore piping is part of various systems required to monitor the status,of the nuclear steam supply system and of systems reauired for component cooling and safe plant shutdown. Failure of such piping or its supports could preclude accurate monitoring of NSSS status.
Assessment of Safety Significance
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The NRR has received from the anonymous alleger a. set of sample problems ( where the alterations are alleged to have occured. The details are, however, unclear and are currently under review. A NRR representative reviewed a number cf small bore design packages on site in conjunction with IE representatives. It was detennined that in several cases incomplete documentation transfer existed for the source of loads for which the supports were designed. The significance of this deficiency is uncertain as insufficient detail is available as to its extent to reach a conclus' ion. However, if inadequate documentation exists, or is totally missing it indicates that a re-review of all small bore piping may again be needed before Step 2, criticality and low power testing.
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Staff Positian . The apparent or potential lack of proper documentation involving the small bore pipe supports is considered a deficiency supporting the allegation. .
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Action Required NRR will review a representative sample of DCP small bore pipe support
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design packages to assess the current quality of these designs. This sample will be provided by IE, Region V. The schedule for completion of this effort will be provided on receipt of these packages.
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( Task: Allegation or Concern No. 87 ATS No. 83A063 BN No. N/A
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i Characterization
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On site management destroyed those calculations showing supports will fail under design conditions, and assigned new staffers to re-do the calculations to show these supports were adequate. The calculation logs did not refer to the original packages showing support failures.
Implied Sionificance to Desion, Construction, or Operation
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l' ( Management pressure to compromise system design safety margin. Falsification
,
of records to cover up substandard design conditions.
Assessment of Safety Sionificance The alleger was interviewed by the staff at the site on December 7, 1983 and additional information was obtained. The staff retrieved the original design calculation logs and design calculation packages from the licensee records vault. The records and logs are being reviewed and calculations evaluated by NRR and regional staff. An interface with O! has been established.
This task will be coordinated with the findings of Task Allegation or Concern Nos. 79, 82, 84, 88, and 95.
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g ff Position The staff has not progressed to the point where a position can be taken.
. Action Required
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Continue review of records, logs, and calculations.
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( Task: Allegation or Concern No. 82 ATS No. RV83A063 BN No. N/A -
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Characterization There was minimal training for onsite p'ipe support engineers.
Implied Sionificance to Desian, Construction, or Operation Without adequate indoctrination and training, the engineers may not effectively perform their assignments.
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Assessment of Safety Sionificance
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This issue was addressed by examination of training requirements, implementa-tion records, interview of engineers, and review of engineers work products.
The staff interviewed five onsite design engineers selected from the personnel roster. In addition, managers / supervisors of the various design groups were interviewed. There were no written job descriptions for any of the pipe stress and support group leaders, lead engineers, and engineers.
The staff found that other than general site QA and technical training that were provided for the new employees, no project group specific program was in Further, the
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place in either pipe stress, or pipe support engineering group.
general QA and technical trainings received by the staffers had not been timely
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and consistent. The bases for this determination are:
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, t A (Support) 10/82 02/18/83 5/5/83 8 (support) 04/83 ,
07/15/83 5/4/83 C (Support) 09/83 record shown <- - no longer w/ project D (Stress) 05/81 06/9/83 none E (Stress) 02/83 04/19/83 05/04/83 The staff reviewed several design calculations which are identified in Task Allegation or Concern No. 79. Among the calculations reviewed, it was identi-fied that design verifications / checking were not adequate to catch calcula-tional errors which is an indication of lack of procedural knowl[ege and training in project controls.
. i ' In conjunction with task findings discussed in Task Allegation or Concern No.
79, the lack of personnel training was considered to be one of the elements within the program breakdown in site design control measures.
Staff Position The staff concludes that training in project design standards and procedures has not been timely or consistent and may not have been adequate in some cases.
Action Required Further actions related to this issue will be handled in conjunction with issues 79,8f,and8h.
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Task: Allegation or Concern No. 84 -
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ATS No. RV83A063 . Btr No. H/A
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Characterization
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Lack of sianag'ernent responsiveness to resolve the alleger's concern that he was not provided with controlled design procedures.
Implie'd Sionificance to Desion, Construction, or Operatio~ n . t .
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The use of controlled design documents insures all works are performed to '
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. Assessment o'f Safety Significance -
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The staff interviewed the alleger onsite on December 7 1983 to cla ify his concernsinth/isarea. The a11eger referred to a memorandum writt rn by line p
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management to upper management relative to his concern about a lack of con- ' trolled design procedures.
. The staff interviewed projcct team general construction personnel in relation to the memo purportedly written by supervis' ion. " In discussion with the pipe
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support group leader, on 12/6/83, he denied that he had written a memorandum to
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Messrs. R. Oman and M. Leppke in December 1982, the On-site Project and Deputy Engineers, relative to the lack of controlled design procedures' to t'e used in the pipe support group, in support of Mr. Stokes' concern.
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Mr. l.eo Mangoba agreed that he was aware of the subject concerns raised by a ' number of his staff, and had taken actions to obtain additional controlled . design procedures.
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In view of the task findings that were discussed in Task Allegation or Concern _ Nos. 79 and 82 that a large number of out-of-date procedures and drawi.1gs were fcund, tne deficient document control system; the lack of training for the ' cersonnel relative to the use of up-to-date orocedures, design revisio1, the
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management response to timely correct the problem and to prevent recur ence
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appears to be inadequate. The spirit of the allegation was substantia".ed. .
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Staff Position . . -
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The staf f concludes that site management must improve its sensivity in addr'ess-ing safety concerns and improve communication with the workers.
. . e Action Required
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. I - Task: Allegation #85 ATS No.: BN No.: Characterization
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U-Bolt Design inadequate.
!molied Significance to Plant Design, Construction or Operation U-Bolts act as load-carrying members of small bore pipe supports. As such they are useo for supporting safety related piping which is required for plant safe shutdown.
Asset ment of Safety Significance An NRR representative has interviewed the alleger (C. Stokes) at the site on December 7, 1983. He stated that the installation of the
,.- U-Bolts was poor, the manufacturer's (ITT Grinell) load rating was exceeded by approximately a factor of-four, ard that the DCP ihteraction of tension and side loads was less conservative than the manufacturers.
Pe also provided a DCP document which specifies the design load ratings and some DCP experimental data suppoYting these load ratings.
Staff Position The staff has made a preliminary assessment and has concluded, based on actual observation of a sample of U-Bolt supports, that the U-Bolts appear to be installed in accordance with current industry practice.
There may be some merit to the other parts of the allecation, but there is no safety significance until 57 power is exceeded due to negligible fission product inventory.
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Action Required The staff will assess the documents provided by the alleger as to technical adequacy.
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_ Task: Allegation #86 ATS. No.: BN No.: Characterization
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"Cede break" design.
Implied Significance to Pla'nt Design, Construction or Operation A "code break" occurs within a piping system where Design Class I (seism'ic) and Design Class II (non-seismic) piping meet. The bcundary is defined by a valve and certain support requirenents, such as an anchor on the Class II side of the valve. Thgse support requirements may induce stresses due to thermal constraint in safety related portiens I
( of piping.
Assessment of Safety Significance An !;RR representative interviewed the alleger at the site en December 7, 1923. The alleger stated that a "code break" deficiency had existed but that he wa.s unaware if and when it had been resolved.
. Staff Position The "code break" deficiency was identified as e generic issue by the l ICVP, and was addrersed by the DCP during the reevaluation of the smell bcre piping. The IDVP verified, on a sample basis, that the DCP resolved this deficiency satisfactorily. The staff, therefore, finds j no safety concern for this allegation. There is no impact on either low power testing or full power licensing.
Action None.
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, . Task: Allegation or Concern No. 88 ATS No. RV83A063 BN No. N/A .
Characterization
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There had been ways to accept on-site design supports that were determined to be incapable of meeting the' loading conditions.
' Implied Significance to Design, Construction, or Operation Management practice to compromise system design safety margin by juggling calculations and designs to accept supports that had been rejected by calcula-( tions performed by the originial reviewers could result in structures unable to perform their intended function. j
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Assessment of Safety Significance The staff met with the alleger on site on December 7, 1983. Clarification and additional information concerning specific areas of his affidavit were ob- l tained. The broad characterization of his concern has the following detail elements / ways the design group may compensate for unacceptable calculations: a. Revising pipe code break locations in order to reduce the number of safety related supports, and omitting many of those that failed in the review program.
r 1/ b. Assuming gaps that did not exist and vice versa.
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c. Assuming joint release for rigid connections, but made no attempt to If f remove the welds, O
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d. Performing calculations to determine maximum support load carrying capacity. The results were .then sent to the stress group for line model change to meet piping stress allowables.
1/e. Adding new supports within six inches of the unacceptable supports, the new supports consisted of inaccurate assumption of restraint gaps. The new supports did not have control or document numbers.
The staff has obtained the records, calculation logs and design calculations necessary to examine the above concerns. The issues will be reviewed jointly by NRR and regional staff and coordinated with Task Allegation or Concern Nos.
79,82,84,8f,and95.
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Staff Position The staff has not progressed to the point where a position can be taken.
Action Required Continue review.
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DATE ^ * TE ~ . SUPERVISOR REQUEST DURATION - .ETED Q ATTEND. DISC COURSE CONTENT REQllESTED BY RECEIVED (IlOURS) hs, sAY YR RDIAR 34 Sill AO, S. TRESLER MR M&NE Engineering Manual Survey LEONG JJ 4/8/83 -
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?3 SilICKER, B. S. McCALL J CE JJLeos
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Engineering Manual Survey LEONG JJ 3/28/83 4 4 7 83 oX OMAN RG ff&NE Engineering Manual Survey 26 SilIEll, A. M. S. OMAN RG OMAN RG 3/14/83 3.5 3 14 83 o IX (6 SIIIII, MING M&NE SFPD QA Indoctrination 1 5 5 83 o Modi CE D1P 3.3: Design Calcs CRANSTON GV
!5 S11111, MING CE 1 1-7 83 o Modi '9 Sillit, MING Engineering Manual Survey LEONG JJ 1/18/83 4 3 2 83 oI 17 SIILAFMAN, I.
CE SFID QA Indoctrination CRANSTON GV 1 5 12 83 o Modi BRADY E CE Engineering Manual Survey IS SilLAFMAN, R. BRADY E LEONG JJ 4/28/83 4 4 29 83 o XIV.
M&NE Engineering Manual Survey LEONG JJ 4/8/83 4 5 12 83 * XIV; O SitMYCELSKAYA, A. M&NE EMP 3.3: Design Calcs CRANSTON GV 3 21 83 oL 7 _ SIIO1KHET. M._ OMAN RG M&NE Engineering Manual Survey
>0 SII0 l KIIFT,_11.- . OMAN RC OMAN RG 2/8/83 4 2 18 83 o IV 2 SIIOOK, RICIIARD F.
M&NE SFPD QA Indoctrination 1 5 __5 81 o Modi M&NE Engineering Manual Survey OMAN RG 4/19/83 3 4 '7 SilOOK, RICHARD F. 19 83 o X111 8 S110RTT, C. M.
MSNE SFPD QA Indoctrination 1 5 5 83 o Modi L2 Sil0RTT, C.11.
ND QE Orientation & Training IIARDIE DB 1.5 1 17 83 o Modi ND DIP 3.11 Computer Programs 5 SIIORTT, C. M. ZELNIK CJ ND
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'D SilPADARUK, V. M&NE E!!P 3.3: Design Calcs 8 31 83 o XXII 2 SilDRER , P. S. CRANSTON GV 1 1 14 83 o Modu b
M&NE SFPD QA Indoctrination 5 5 83 o Modu SilUSTERMAN, A. 1 M&NE SFPD QA Indoctrination 3 SilUSTERMAN, A. MSNE DIP 3.6 & 3.60N 1 5 5 83 o Modu O SliUY, ADAM CE 1.5 6 24 83 o Modu SilDY, ADAM SFPD QA Indoctrination CRANSTON GV 1 5 12 83 o Modu 3 CE EMP 3.3: Design Calcs () SilVEYD, L. DD Engineering Manual Survey CRANSTON GV 1 6 14 83 o Modu g S I AllAAN, BISTOK COSTEROUSE DL 3/4/83 3 4 14 83 o XI CE SFPD QA Indoctrination b SICAD, F. 1 5 2 83 o Modu F SICAD, F.
MSNE SFPD QA Indoctrinattou 1 5 11 83 o Modu M&NE DIP 3.3: Design Cales CRANSTON GV SICINSKI, E. E. TRESLER MR MANE SFPD QA Indoctrination 1 6 14 83 o Modu SICINSKI, E. E. TRESLER !!R H&NE Engineering Manual Survey 1 5 12 83 o Modu LEONG JJ 5/24/83 4 6 8 SIDDIQUE, M. OMAN RG M&NE Engineering Manual Sutvey 83 o XVI SIDDIQllE, M. OMAN RG OMAN RG 2/8/83 4 2 18 83 o IV MANE SFPD QA Indoctrination 1 5 5 83 o Modu .. SIEKIERSKI, T.
SIMilAL, S. C.
MANE SFPD QA Indoctrination 1 6 7 83 o Modu MINK 0WSKI 11 CE Engineering Manual Survey SIMilAL, S. C. LEONG JJ 2/17/83 4 4 29 83 * VI; . MINK 0WSKI 11 CE SFPD QA Indoctrination SINGil, B. 1 5 2 83 o Modu CE D1P 3.3: Design Calcs CRANSTON GV SIMGil, DICK J.
M&NE SFPD QA IndoctrinaH on 1 1 7 83 o Modu 1 5 12 83 o Modu llan been scheduled in the Module No. 1isted 1st - Attended 1st session only
- llas been rescheduled in the Module No. listed 2nd - Attended 2nd session only To be scheduled o Attended module number listed
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DATE DATE REQUEST DURATION COMPLETED REC ID ATTENDEE SUPERVISOR DISC COURSE CONTENT REQUESTED BY RECEIVED (HOURS) HO DAY YR RI 103 DIDESCU, D. C. MERCADO V . MANE Engineering Manual Survey LEONG JJ 2/9/83 4
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1439 DIDESCU, D. C. MERCADO V M&NE EMP 3.3: Design Calcs 1172 DINALANTA, M. M.
CRANSTON GV 6/8/83 1 6 14 83 o M& N'E SFPD QA Indoctrination 1 'S 4 .83 o 104 DITO, E. EQC 1983 Iludget Acctg RAI.STON CE DJ0KIC, DANILO 2 1 6 83 1521 McCALL JK - CE Engineering Manoal Survey LEONG JJ 4 9 20 83 * le 106 DOBYNS, E. R. M&NE EMP 3.3: Design Calcs 105 DOBYNS, E. R.
CRANSTON CV 1 1 14'83 o M&NE Engineering Manual Survey LEONG JJ 1/18/83 4 3 2 83 o 107 DOKLADAL, -J. - OMAN RC M&NE Engineering Manual Survey 564 DOLIN, J. A.
OMAN RG 2/18/83 4 2 18 83 o DD Engineering Manual Survey COSTEROUSE DL 3/4/83 3 * dc
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108 DOSill, 'K. D. CE Engineering Manual Survey 979 DOS 111, K. D.
LEONG JJ 2/17/83 4 3 11 83 o CE SFPD QA Indoctrination 1 4 29 83 o 109 D0YLE, J . J . MERCADO V M&NE Engineering Manual Survey LEONG JJ 2/24/83 * rt at 677 D0YLE, J. J. MERCADO V M&NE EMP 3.3: Design Calcs CRANSTON CV 110 DREISBACII, L.
1 3 21 83 o M&NE Engineering Manual Survey OMAN RG 2/8/83 * tr me til DUI!EL, T. M. OMAN RC M&NE Engineering Manual Survey 112 DUC110N , V. A.
OMAN RG 3/14/83. 3.5 3 14 83 o CE Engineering Manual Survey LEONG JJ 2/17/83 * ui 695 DUC110N, V. A. CE 4/ EMP 3.3: Design Calcs CRANSTON GV 1 3 21 83 o 113 DUK110VNY , T. EE Engineering Manual Survey LEONG JJ ' 2/9/83 *
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5 12 83 o MERCADO V M&NE Engineering Manual Survey LEONG JJ 2/24/83 * ui 676 DUONG, Y. P. MERCADO V 7/ MANE EMP 3.3: Design Calcs CRANSTON GV 1 3 21 83 .o 899 DUONG, Y. P.
M&NE SFPD QA Indoctrination CRANSTON GV 1201 1 5 10 83 o DURANI. MANM0llAN M&NE SFPD QA Indoctrination 1505 1 5 4 83 o DURANI . MANM011AN M&NE Engineering Manual Survey _ SPEASE G- '6/28/83 3 7 15 83 o o li.is been scheduled in the !!odule No. listed 1st - Attended ist session only oc lias been rescheduled in the Module No. listed _ 2nd - Attended 2nd session only 000 To be scheduled o Attended module number listed '
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Tas k: Allegation or Concern No. 78 ATS No.: RV-83A-063 BN No.: N/A .
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Characterization Drain line support bracket bolted to the' floor with only one anchor bolt in Unit 2 as evidenced by photograph supplied by the alleger.
Implied Safety Significance to Design, Construction, or Operation The safety significance of this installation is minimal since drain lines feeding into the floor drain system, downstream of drain line' isolation valves
{ are not nuclear safety-related or quality class 1 systems.
. Assessment of Safety Sionificance The staff could not locate the subject installation in Unit 2. Drain lines downstream of isolation valves are not classifed as nuclear safety-related and are not quality class 1 installations. Therefore, they are not inspected by quality control for conformance to any specific requirements.
ThelicenseedesignprovidesaclasslgsupportwhichisinspectedbyQC, j immediately downstream of the class 1/ class 2 code boundary transition. In the case of drain lines, this is the drain line isolation valve. All portions of the drain line downstream of the code boundary isolation are classified by the licensee as quality class 2 and are not within the ASME code jurisdication.
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' Discussions with the licensee's Engineering Supervisor on-site revealed that (. PG&E'is conducting extensive reexaminations of the design and installation of U small bore pipe supports installed in Unit 2. (This action was previoulsy completed for Unit 1.) The licensee's new criteria requires an upgraded code boundary support evaluation. This reexamination is complete for class 1 supports and code boundary supports. The licensee estimates that 75% of the remaining support installations examinati.on are complete and have been reworked to more conservative criter'ia'. Thus, it is highly pr:,1,able that the instal-lation photographed has been reworked. New supports installed to this criteria are fu'll seismic class 1 supports and, as such, are inspected by QC for com-pliance with specification requirements.
Staff Position
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The staff considers that the licensee's action completed to date, and which will_be completed, provides assurance that the pictured installation, and similarly installed supports prior to this reexamination effort, will be upgraded to the new, more conservative criteria.
Action Required Review licensee action as part of continuing inspection of Unit 2.
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Task: Allegation or Concern No. 89
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s, ATS No. RV83A063 BN No. NfA
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Characterization Improper resolution of piping interferences.
Implied Significance to Plant Design, Construction, or Operation
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Piping interferences or inadequate piping support could result in piping systems being overstressed during operational or design loading conditions.
. Assessment of Safety Significance The staff reviewed the disposition related to the allegation and determined that the matter was acceptably resolved. The staff inspected areas of the con-tainment and auxiliary building looking fo'r cases where pipes were resting on conduit supports. The staff did not observe any cases. Since no specific cases were cited in the affidavit, this concern relates to the more general concerns on design control on piping and supports being addressed by Tasks related to allegations 79, 82, 84, 87, 88 and 95.
Staff Position , I
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This concern should be covered by the resolution of Tasks 78, 82, 84, 87, 88 and 95.
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. . . ,- Action Required (
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PG would prepare "as constructed" field routing drawings and
, "as built" pipe restraint drawings which would then be reviewed s .
by S&L. These design provisions and control measures are con-
. trary to the present licensee QA program philosophy which requires: (1) installation and inspection to be in accordance with the reviewed and approved up-to-date design drawings, (2) utilization of the Field Change Request (FCR) system to minimize the risk of drastic alterations or modifications after system component installation, and (3) timely identification of any nonconformances, and implementation of swift and effective '
- corrective or preventative measures.
M ! (iii) S&L Specification F/L-2739, Paragraph 301.11, "Installation of 2" and Under Piping" states:
"For two inch and under piping, Sargent & Lundy drawing numbers M-2535A through M-2616C, released under an alpha revision except an alpha revision released for ' Record Revision Only': (1) "All dimensions and configurations are conceptual to provide the basic routing of each specific 2 inch and under piping system. Should conflicts or interferences occur, the piping may be rerouted using care and judgment, so as to provide sufficient clearance around electrical switchgear, instrument panels and other equipment, structural features, etc., to facilitate good routing .
practice, (i.e., valve accessibility, interference elimina-k tion, maintenance accessibility, the ability to' properly support, etc.); and to provide a minimum of 3 inch clearance
'
from other process piping and instrument sensing lines. The criginal dimensioned routing and configuration shall be followed throughout the p!peline except in the area required to clear the conflict or interference."
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In review of the PG QCPs and cps, the inspector determined there was
~
a lack of specific quantitative field design, installation criteria, and inspection criteria to provide small bore Class 2 and 3 piping with sufficient clearance or separation from electrical switchgear, instrument panels, etc. , as delineated in the S&L specification.
b. Review of PG Calculations The inspector reviewed the following PG small bore Class B, C, and H process and instrumentation piping with maximum operation temperature
~
of 150 F:
(i) Lines ICCE3AA " and ICCE3BA ", a Class H instrumentation line connecting to ICCA2A-3" at F1. El. 394'-7".
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Task: Allegation or Concern 95 k-ATS No. 83A063 BN No. N/A
.
Characterization dngles of pipe support member are out of specification. Unbraced angel steel members within a support framework exceeded AISC bending stress allowables, particulary those supports where a bundle of small bore pipes were attached.
Implied Significance to Design, Construction, or Operation
$Y The angled could buckle under excessive loading, creating lar,ge system deforma-tion and could result in piping overstress.
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Assessment of Safety Significance The staff interviewed the alleger on-site on December 7, 1983. During a site tour, the alleger identified one specific support installed at auxiliary feed-
water pump (AFP) No. 11, steam supply trap drain line No. 447, located at ele-vation 100' in the auxiliary building.
The staff plans to inspect additional supports, evaluate the design ciiteria related to small bore pipe supports, and review calculations related to this type of support. This will be reviewed in conjunction with the licensee's small bore pipe support program. l
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_
. . -. . , * . , . ' . Staff Position \.
The staff has not come to any conclusion on this concern. .
.
Action Required NRRandtheregionalstaffwillreviewtheitemsdiscussedabove.
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6,,,,,,,,, g,,r, ,,, y, , ,,,, E60S or E70S electrodes used in the gas metal.are process shall con-1.5.1.3.2 On the gross section of axially loaded compression members norm to the Specification for Mald Skel Electrodes for Gas hictal-Arc Welding, AWS AS.18. latest edition, or the provisions of Sect.1.17.0; EGOT or E70T when K!/r enceeds C : . g cicctro.les used in the flux cured.are process shall conform to the Specifica- 12r'F p* , toon for Mdd Steel Electrodes for Flux-Cured-Arc Welding, AWS AS.20, 23tK!jr)' _ late 3t edition, or the provisions of Sect.1.17.3.
Manufacturer's certification shall cona titute sullicient evidence of con- 1.5.1.3.3 On the groes section of axially loaded bracing and secondary formity with the specifications. memlsen, when l/r exceeds 12e
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F,.. F. (by Formula (1.5-1) or (1.5-2)) SIrr!ON 1.5 ALLOWAllLE STRESSES * I (1.5-3) 1.t.
.-- Except as provided in Sects. 1.6.1.7,1.10,1.1I and in l' art 2, all cons- l 200r I ponents of the structure shall be so proportioned that the stress, in kip" 1.5.1.3.1 On the gross area i.f plate girefer stiffeners: per square inch. shall not exceed the following values, except as they are ; sounded off in Appendix A. F. - 0.00F, 1.5.1 Structural Steel 1.5.1.3.5 On the web of rolled a hapes at the tocof the fillet (crippling,
, see Sect.1.10.103:
1.5.1.1 'l,ension On the net sweetion, except at pin holes: F. - 0.75F, F, - 0.60F, 1.5.1.4 Itending hot not more than 0.5 times the minimuna tensile strength of the steel. I I.5.1.4.1 Tension ami compsesion on extreme libern of c.,mpact hot-On ahe net section at pin holes in cychars, pin-connected plates or built- Hed or built.up members tescept hybrid girders amt members of A514 up members: , steel) symis.ctrical uhout, and tomled in, the plane of their minor axis and F, .- 0.4 5 F, meeting the requircanents of thia'section: For tension on threaded parts see Table 1.5.2.1. M - 0.66F, I'0 ' ' ***' in ortler to qualify under this section a member mu t meet the following requirements: On the groe.s section: F, - 0.40F' i a. The flaniges shall lee continuously connected to the web or webs.
.The gross section of rolh.d and fabricated shapes may be taken as b. The wi.hh. thickness ratio of unstilfened projecting elements of the t he pro.hn i of the overall elepth asul the thi. kneas of the web. See Scot. compremion flange, as defined in Sect. l.9.1.1, studi not exceed i.in for reduction required for thin webs. For discussion of high shear siress 5.'.4 \/ F, wit hin houmfaries of rigid connections of members whose webs lie in a c. The width. thickness ratio of otiffened elements of the omnpression romanon plane, a,ec Conunentary Sect.1.5.1.2.)
He a damed in Sect.1.9.2.1, shall not exceed 190/MF,. 1.5.1.3 Consprenion
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, Structural Stal for Buildings .D i
1.5.1.4.4. liou-type memliers are torsionally very stiff.* The critical tiexural stress due to lateral-torsional leuckling, for the compression flange Formula (1.5-7) is a convenient approximation which assumes the pres-ence of both *.ateral hesuling reai tance azul St. Venant torsional resistance.
~ of e box-type beam loaded in the plane of its minor axis so as to liesul about ite ni.sjor axis, can be obtained using Formula (1.5-1) with an equivalent Du, to the differtuce hetwe, a llange azul weh yicht atrength of a by brid slenderness ratio, by the expression girder, it is desirabic to base Im lateral buckling resistance solely on warping torsion of the flunge. Ilence, nw of Formula t 1.5-7) is not permittet! for wish g $' g.' members. Its agreement with nuere exact expressions for the buckl.mr a
(T--* * * dil, strength of intermittently braced flexural snembers* is closest for homog+n .
neous sections having subst a ntial resist an< e to St. Venant tormion, identifiabe where I is the distance between p.>m.ts of lateral support and S,.1, umi J in the case of doubly.synunetrical .wetions by a relatively low d/Af ratio.
are, respectively, the major axis section modulus, minor axis moinent of For e,oine sections having a comprewion llange area distinctly smalur suertia urut the torsiumd constant of the beam cross-section. It can be than the tens! .n flange area, Formula (1.5-7) may be unconservative; Len;e, almwn th.at, when d < 10b and I/b < 2,500/F,, the allowable compression _ its use is lin.ited to sectionis whose compression fl.ange area is at least as
!!ange stress mdicated by the above equation will approximate 0.60F.- great as the tenaion flange. In plate girders, whi,;h usually have a much lleyom! tlas limit deflection rather than streas is likely to be the design higher d/Af ratio Ihan rolled W shapes Formula (1.5-7) may err grossly on C'ilC'5""- the con,ervative side. For such members the larger stress permitted by 1.5.1.4.5 and 1.5.1.4.6 The allowable hending stress for all other Formula i1.5-6a) and, at times by Formula (1.5-6h), affords the 1 etter esti-mate of buckling strength. While these latter formulas underestimate this flexural members is given as 0.60F,, provided the compression flangt is strength somewhat because they ignore the St. Venant torsional rigidity of braced laterally at relatively close intervals (l/b f i 76.0/ VF,). the prorile, this rigidity for 1.uch sections is relatively small and the margin Members l>ent alasut their major axis and having an axis of symmetry of overconservatism, therefore, is likewise sm.dl.
in the plane of loading may be adequately braced laterally at greater inter- it should lie noted Ihat Formula t 1.5 7), like the more precise, complex vals if the maximum hending stress is reduced sufficiently to prevent pre- exprmions it replaces, in written for the ca.e of clastic huckling. A transi-ma t ure buckhng of the compression fla nge.
Mathematical exprenions & ion is not provided for this si,rmula in the inelastic stress range because, utfordmg an uact estimate of the buckling strength of such memliers, which when m a ual corulitions of load application an.I variation in bending moment i
,
take into account their torsional rigidity about their longitudinal amirs (St. are consi.lered, any unconservative error without it must. lie small.
Venant torsion) as well as the bending stillness of their compression flange Singly-symmetrical, built-up,1. shape membern, such as some crane i hetween points of lateral supgwart (warping torsion), are too complex for girders, often have an increa3 eel compression it.ange area in order to resist general design ottice use. Furthermore, their accuruey is dependent uguin I'emln g due to I teral lo.nling action in conjunction with the vertical loads. i the validity of assumptions regarding restraint at points of lateral support l Such mendscrs usually can lut proportioned for the full permissible leending i and conditions of loading which, at liest, can he no more than engineering , st rm wtsen that stress is pr wluced by the combined vertical and horizontal * ju lgments.
1" "ling. Where the failure smale of a singly-symmetrical 1-shape meinher The combination of Formulas (1.5-Ga > ur i 1.5-61,) and (1.5-7) provides hdvi"g a tergcr compression than tension flange would be by lateral buckling, a reasonable design criterion in more convenient form. the tu rmissible hending stress can he obtained by using Formula (1.5-6a) As in Formula (4) of the 1963 edition of the Specification, Formulas or (1.5-7;. a l.5-6a) >srul (1.5-66) are based on the assumption that only the bending Through the intrmluction of the mo.lisicr" C., seme liberalization in stiriness of the comprewion flange will revent the lateral displ.acement of strew is perminihte when there i moment gr.ulient user the unbraced length that element between bracing points. The new Formulas (1.5-Ga s and except where, in the case of cmunined hending and axial compression, ihis a 1.5-6h; differ from the earlier Formula t t i is two w.sys: edjustment is proviileal by the fat tor C in Formula i t.6-la).
1. Whereas the earlier provisions required no = tress reduction when Formulas ll.5-6al asul e 1.5-6h) may be resined to include both St.
I r was less than 40 (regardless of yicht stress values and then a . Ven.n.t aml warping torsion by subotituting a derived value ti.rr r . Thia reduction to the value obtained from the parabolic expression, the i "'f"ivalcot radius of gyration ry .o. can be obtained by equating the appro-new formulas, by im reasing F, at i O f rom 0.60F, to 2F, 3, i priase expression gNing the ritic.d .-l.istic hemling sire.w for the comprewion provides a continuous stress relationship with the usahraced length , flange of a he.nn t wida deal of an adah laaM colunua.}
" w hen F, is reduced from t he maxiomm pi rmissible value of 0 68)F,. ! '
Whereas the carher smgle Formula i I e apphed even in the r.ange * of' clastic huckling streas ton the assumption that Formula <5) , w oul.1 gos erni, the replacement of F..rnmla (4 ) is hheralued in g
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- - - - - - The s'.cade.seu :stic, K1/r. cl c=mpeu:::: =ne.=>krs aMM not emea ;*
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Ter : .ala n:,c=.bers . . . .. . . . . . . . . . . . . 24 0 .n d as the 6. mage *.htt may For br=&g ed ot.ber se:cmda.h in d=.-s . . . . . . . . 300 E vetutions of stress. Stiesa - crustle:.a. In the eue of a d o the nr.meri:.a] s.:s of SICTION 1.9 WIDTH.THICKNISS RATIOS ]
. .
.ues er the sum of raami= sum 1.L1 Unst!!ene d Elements Under Compre-m5eza given peint tesddg fro : 15.1.1 Ucstlfened tp ojectir.g) ec=presd=n ele dents are those hav.
' :; s need to b4 de. i.ng one free edge para]lel to the d.!rection of i..n., uzien stress. 'The width .r.:.]'
::. tures cer.: c:2y a of unst!! ned pistes shdl h ta'aen f.-em the .'5 edge to the f.rs: row of e- strus :Euctuatier.s. T. h e f:.stenen er wel!.1; the sidth of le gi of angles, c::a:r.n=J and see 'anges anal ,
e leads is too it.fre:tuent to stcr .s cf teu thdl be taken u the full n -J=m1 til:sendon: the wicith of eve:. e-ene rew ys r.ad sup. gangen of 1.shan members and tees shall be itaken a.a cme half the full u n'e c.fian subject to fatirue nomir.nl sidth. The thickntu el a slepi g is :ge shst be zne. surect half.
way wiseen a free edge and the corresponding 5 set of the web.
g- wersien 1.9.1. Ur.ni'ened elements subjeet to x=5a1 -....pressien or q due te WJ:-? =%11 he e-..*A~.J .. r ::e e '..*% when the raisu m
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to "w::ie snickness a not gre.ater t;.an the foUoa=1.ngt t r., it.:!; ut lea di .; a.s de d. .ed
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in aniiy iondey es,preu!ca - When the t~. cal width.te thickneu r:do ez eec:'.4 these va!ues. the ti e ant J s:e.n ve ta e: u tu , , .,, y .e., ei h. J. 14;cr.2 ' ndix C.
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LE. :no : 4, - ete.pt s here, ir. the' czae of cerebine f.ne.7 si de ec:. press.cn dante.ne s@new prevent ne la rement terd dar..of
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3.1.3 3e ein g As med 0.-ev gh: .:t On S;eri.sti:: the n "=.L! r..r- d e.; .d ====N: of these behs arv rabs i.. 1.5.1. E .1 44 which havt f wedd be re-:' t3 to trar c=3 a give= '
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. aer.:.:auly u-ed e: d lth e d to a t s e p'.As e by a r. r;j u bh =.ur.s. - Th.,e e 5ciemey cf th.taded fu.e:Ht: re::~ =e n d e d h a. .: t : .r ua c e pi:.: is :.:t the s.stre u fer d veu. The c=- -6:.a la re duerd when the thre erva' ti>.-u:0.a cf tbt yield nreu cf the pet eenui:izt the F': be tve c= the c===.e: .e d p.a .: In the c: . yj i u!e r.:ed a gaitn i-: ah!My ef tbe phie Wyc:d Oa hele,* thie aca.: s=e u M m are gives: cae ~
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thh faattre i= the cue of A.337 bohs. i .2 Ettets, Ec!ts, and Threaded Puts
.d b tb t'w p'ame a d the allc, r-::: mers, is rrinerd aecerdimrW.
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- me:... rial ; og. .ie s c,f the v=dru ded =::cM ha s W e: fe::d t.s, c' tk pa-i. l= this i vestira Jcn die c re c;:uly p rei;:t the *.c Ce n t:ph cl hrg e r d.u .rtet O/ sa de d p a a :::- 3ir; to the tra d ec:v e ntlen. a2 /
d u ::Jght h u.>ed !: a:ther beha et vpe4t 7 dl e .e: ==d thitiseu cJ the c:inre cte d p:. 1: n: g:id:: ef t'.e you:J:n a g al-r. seich e'er: is the thruii: I.
t, er= zi: g'e .6u h2.rin g a:> d e r:d o s+ ret d by e e rtw.i.t d i:. a' t)g ht c : :g cf hit h* r.re nph icb.1he F.e- O e ,==.:ne:.4e4 ven.i g s ru h ti c :h C::::0 c: PJveie.d c d E.:N d 5t. ; :.- ) ci:u hai nm ==c:5 e d ,% e4: hxi:g. a=d a pp:c:i=.ateI f c:D: ytiativ ely hj gh : s e;c:g nyt:a != te d:: Icr hj;h n e r ph leb.
ctr=a trc :=.:n e :3 e d fc d e te r:i i= E T* - A:y a d d.ldc:.al f a r.4 : e i te _ri:: re r.)ti:g f ::. ; ir.g a :" 3:: dV e to DIS' .
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Allegation Topic No. la <
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I. Sunrnary of Issue
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Use of uncontr'olled design criteria documents in the design of pipe
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supports by OPEG:
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a. Only minimal training.
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II. Sununary of PGandE Position ,
, p.D ( I .- Responsible Engineers .
lb b. e
' a. oman a. teppte . g yu-eo s (o , (/,,.
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, L., Quotation from Affidavit:
(Page 2) "There was only minimal training; the other contract employees and 1-received our initia1 hanger pipe support assignments-on the first day'.. , .
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v. DCP Response ' . ~
' . . All engineering personnel are required to attend Engineering Manual-Survey training within a short time after their assignment to the Diablo "
Canyon Project. This indoctrination is provided to familiarize the
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engineers with the basic requirements of the project quality assurance program as well as the procedures for initiation and control of project Project records indicate engineering documents including calculations.
that Mr. Stokes attended this training on November 8,1982.
Additional engineering training specific to design caiculations was also provided to raview and emphasize compliance with the reqbireme Engineering Manual Procedure 3.3, ' Design Calculations" ,as it
- specifically applies to small bore' pipe support design calculations.
, Project records indicate that Mr. Stokes attended this training on November 19, 1982.
. In addition to these organized training sessions, working f amiliarity with DCP calculational. procedures and pipe support design criteria was acquired by new engineers through the practical experience gained in Supervisory personnel were originating preliminary calculations.
available to answer individual questions and provide clarifications for points of uncertainty.
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Allegation Topic No. Ib ,
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I ., Sumary o'f Issue
.. .. pipe dse of uncontrolled design criteria documents in the design of: , ,
supports.by OPEG: b. Use of uncontrolled xerox copies of DCP criteria, and design No management response to documents from other nuclear projects.
Conclusions that OPES pipe requestsforcontroileddocuments.
support calculations are not in compliance with current criteria.
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i . II. Sumary of PGandE Position . III. Responsible Engineers R. Oman/M. Leppte
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( Quotation f rom Af fidavit ;
(Pages 2 to 5)
'There was only minimal training; the other contract We also received an example of Bechtel calculations to be used as a
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model.
We did not imediately receive controlled documents to accompany our assignments, however, as routinely occurred at most other nuclear sites where I had worked.
" A, controlled document is one whose revisions and updates The are controlle by the Ou.ality Assurance (0A)/Ouality Control (OC) Department.
document symbol.
is either an original or a Xerox that is verified The document is assigned to f ollowed, and relevant codes and standards.an 0A/QC insures that the engineer engineer, who f or his area of responsibility. A major function of receives the latest revisions as they are issued. controlled d requirements.
'I waited around two weeks to ing'uire why I was not working to controlled documents. Not only were we working to unverified Xerox proper paperwork bases. ^ Because the documents did not have (-
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copies, but they were incomplete.all required information, engineer calculations they had brought with them f rom other plants, such as Susquehanna, Davis Bessie, Midland, Mcguire,' Catawba and others.
"It was as if each engineer was working f rom his own ref'erence The library of resource materials drawn f rom his or her unique job history.
Final Saf ety Analysis Reports for.these other plants involved dif ferent comitments to dif f erent issues of various codes than we w to at Diablo Canyon.
design cads f or typical conditions - f or other manuf acturers' equipme that may not have been applicable to Diablo Canyon's design.
"In other words, the engineering calculations at Diablo Canyon were , based on conditions and comitments at other The whole point plants of doing that may or may n separate have been identical with ours. individual engineering calculations is to ta conditions at each plant.'
"I alerted my pipe support design project lead, Leo He Mangoba, supported that th flaw could cause the NRC to close down the design of fice. Mr. ;
me and in December sent a rnemo to Marvin Leppke and Robert Oman.
Oman was the lead of the on-site project engineer group ( Bechtel employee.
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'To my knowledge, there was no written response to the memorandum.Kevin Wau ,
[ early January, I inquired about a response.
assis' verbally: tant, inforned me of the answer which had been passe receive any. This response upset me, because the answer was clearly wrong and assumed we did not care enough to do our job properly.
' Along with Kevin, I went to Messrs. Oman and Leppke's' joint of fice, where I not.ified Mr. Leppke that if the NRC were made aware that we were
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working to xerox copies of documents f rom other plants, the designI office would be clo. sed down.
I , received the necessary controlledWe left.'documents. Mr. Leppke inf ormed me
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that I was not going to get them.
'Within two days, however, bef ore I went to the NRC, I received the necessary controlled documents, in this case the M-9 design guide for engineering. While this allowed me to work to current inf ormation In fact, for an pertaining to the right plant, not everyone else did. extende
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had controlled documents.
bth
"Sporadically, the other employees received controlled documen the latest revision. The significance is that, with exceptions however, until August,1983.
such as my own case, until August 1983, none of the pipe supportTo-( calculations are reliable.
on relevant data, they must be compared to the data on the latest
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revisions of relevant design guides." ,
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V. - DCP Response -
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. Use of Controlled Documents
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Mr. Stokes' assignment to OPEG was part of a substantial staffing
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- increase in the smal'1 bore pipe support group in late 1982 which, by
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- Januarf 1983, brought th'e support group strength to over 35 engineers.
At the time of his assignment, three controlled copies of the project piping design criteria were assigned to personnel'in the support group, who were all located in trailer 19. It was recognized that the number of ' controlled design documents at that time contributed to the - unauthorized use of uncontrolle'd copies of design documents.
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Management Response
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Steps were taken to obtain additional cop'ies of controlled documents for use by the expanded piping group. In December,1982, additional copies of design documents were received from San Francisco for use by the
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expanded pipe support group. However, it was soon realized that these documents were not on controlled distribution. Project records include correspondence to Engineering dated January 28, 1983, requesting controlled copies of instructions and procedures for each of the pipe support group members. This was subsequently modified to request a reduced number of copies in recognition of the f act that not every engineer regularly uses every reference document. A sufficient number of the various controlled documents were requested to provide conveni'lt " reference access 3- reeded by the group, while at the same time not
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( unnecessarily complicating the control of distribution of changes to the
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' Compliance with Criteria
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Pr,ior to May 1983, design calculations originat'ed by the OPEG pipe
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- support group were preliminary in nature since they were based on preliminary assumptions due to the absence of final thernal and seismic design data at that time. All such preliminary calculations have been subsequently reviewed and revised as the final design data has become , available. These revisions of the calculations to final stitus were completed using the latest revision of project criteria and were subjected to IDVP review (ref ') . The calculations 'are based on relevant data contained in the latest revision of the. appropriate reference documents.
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Allegation Topic No. 4 ; ,
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I .- Summary of Issue
. . . . ' ~ , ' Improper pra.ctices'in reviewing / checking calculations. , .
a. Failing calculations were destroyed. . b. Calculation logs were falsified.
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II. SumTary of PGandE Position
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Responsible Engineers
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R. Oman/M. Leppke
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IV. Quotation from Affidavit ,
(Pages 9 and 10) (4a) - "After this point I kept close tabs on what was happening in the Unit 1 trailers, which was easy due to my daily association and friendships with the personnel. I learned that they went beyond routine citation check reviews of the calculations, their official mission. Instead, the engineers redid the calculations entirely for nearly all the systems which h'ad failed. (In the process, the original calculations demonstrating syste'm f ailures vanished.) It is important to realize the scope of this effort to rewrite the results. Prior to the split of the ' trailers, the f ailure rate for the entire small bore ' pipe support group *
had been around 50% or more.
"Those results would have meant that Unit 1 f ailed the reevaluation program and required complete reanalysis. Bechtel and PG&E management could not accept this, so they had the compliant engineers rewrite our prior failing calculations. One problem is that the ensuing packages did not disclose that the calculations had been done over. There is no mention of the original f ailing calculations, which were, destroyed."
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"Further, the calculation logs' do not refer to the packages which had been originally prepared. Someone rewrote the calculation log to delete the references to the original packages which had failed. The logs have *
been falsified. In my own case, the log only reflects five out of more than 100 calculations which I prepared.
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. "The only way to reconstruct an accurate history of the calculations would be to ask each engineer for his or her personal copy of the assignment. Even that may not work. In many cases, engineers did not keep copies, due to Bechtel's policy that the calculations were company property. When Mr. Mangoba told us in early March that Unit I calculations would be redone by others in Unit 1, most of the affected engineers objected to modifications of prior engineering calculations by a .new staffer when the originator was available. This violates normal industry practice. He answered that once we signed off and turned in the calculations, they were Bechtel property and Bechtel could do what they pleased, including destre ing tiem and having someone else rewrite them. I will share examples of the resulting f alsifications of records with the NRC or law enforcement officials."
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Y. DCP Response , j * . In verifying the adequacy of existing designs at Diablo Canyon,
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engineering design practices consistent with both nuclear and
, non-nuclear applications were followed. These engineering practices .. .
utilized it'erative engineering calculations to verify a design that is
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consistent with the acceptable criteria. It is comon practice to do initial calculations using conservative data and simplified methods.
This can save the time and expense associated with more detailed, time consuming, sophisticated calculations (such as computer analyses). When an initial calculation using conservative data demonstrates one or more acceptance criteria are not met, an engineer performs additional trial e calculations that use more precise. input data. Input data can be
-- ( modified by removing unnecessary conservatism or by selecting more .
appropriate boundary conditions.
Typically, engineers are trained to employ the use of rnore sophisticated analy;ical techniques if initial conservative analyses are not acceptable. For example, a hand calculation might be superceded 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.
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
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design parameters. The documentation of such an analysis constitutes
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which are not part of the final calculations need not be retained.-
Quality procedures do not require retention of superceded calculations.
Nowever, the practice at Diablo Canyon has been to retain all calculations which have been signed by an originator, a checker, and an
- approver even when superceded. This allows an accurate reconstruction The cover sheet of each calculation package '
of each calculation.
contains a change sheet which shows the history of all revised' calculations. A review of these records indicates that more 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.
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It is important to point out that it is normal industry practice to censider design calculations the property of the client or his architect-engineer agent. They represent the product for which the engineer is paid and most certainly do not remain the property of the ohiginatingengineer.
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I. Summary o'f Issue
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Improper c,alcu'lation iteration techniques used to achive acceptable
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iesults:-
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"Next, management decided to use new assumptions that would change the results.f rom 'f ail' to ' pass' on the systems. Typically, engineers
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would be instructed to assume gaps that did not exist and vice versa."
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< DCP Response
k- . The use of small gaps in the design and construction of pipe supports is a routine practice that recognizes that piping must have the ability to expand thermally. The gap provides that f reedom, while at the same time
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maintaining the ability of a piping system to saf ely withstand the effects of an earthquake. The first approach is to assume there is a zero gap. Should unacceptable thermal restraint of the pipe result, a plant walkdown is perfonted to ascertain the actual gap configuration.
This more accurate information regarding the actual gap configuration
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often satisfies the design criteria. If not, f urther steps must be taken. The modeling of gaps is simply a matter of more accurately
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Improper calcu'lation iteration techniques used to achive acceptable
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e. Addition of new supports within 6 inches of existing failing
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(- . Quotation f rom Af fidavit: . (Pages 12, 13) "Where this dion't work, management tried still another technique New supports were added within six inches of the existing supports which had failed. The stress group then modeled new support gap assumptions so that the new supports would handle most'of the load for which the failing supports had been responsible. This ' solved' the problem, "
because the f ailing . support could handle the small load that was-left.
Even if they I , doubt that all of the assumptions on gaps are' accurate.
were, however, this approach represents a program of undocumented modifications. The new supports did not have control numbers or document numbers, so they were not reported to the NRC. The new supports officially did not exist. My net reaction to the developments since we ' finished' the calculations in March puts the series of abuses in perspective. I have never before seen management at any plant that was so persistent about accepting work that had been done wrong ' originally." '
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It is not DCP practice-to add a support six inches from an existing support'.' In some cases, preexisting as-built conditions were modeled
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where e structural restraint was present. An example of this was
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several instances in which a penetration was modeled as a seismic
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restraint. When a support'was modeled in the final-analysis, either a
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support / restraint physically existed in the plant or a new support was designed and installed. Any new support had control numbers that assured it would be constructed and documented. The as-built drawing would be sent to Engineering for approval.
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^ A11ecation Topic Wo. 11 ,
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I., Sunrnary of Issue ,
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Deficiency Rep' ort concerning excessive length of angle members used to ,
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build pipe supports. i
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II. Sunn.ary of PGandE Position ;
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l.. E. Shipley/P. Shurer .
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(. Quotation f rom Af fidavit 1 (Pages 17,18) 'The second DR that I authored concerned the angle members which a '
part of th'e piping support structures.
plant were too long for the allowable bending stress standard used a the plant under the AISC code. To research the scope of from 25.33' to 50.67', depending on the angle.
this discrepancy I walked thro' ugh Unit 1 for an hour and Inwrote that down all
- ' the angles that exceeded the maximum code length'which I saw.
short time span, I found around 200 to 300 vilations on .the angle members within approximately 100 f rames out of approximately 300 f checked. The violations were obvious. In one case, a member was longer than the naximum permissible lengths.around nine f eet l To
'There has been some additional bracing, but it is erratic.
illustrate, one oversized member had been repaired, while anothe to it remained untouched. . and replace the angle frames with tube steel. .
"The ef f ect of this violation is that the supports cannot reliably carry Theoretically, the angles could buckle under the pressure.
the loads.
The angles support loads up to 1000 po'unds 2.' or more, such as' a bundle (. small bore piping. The OR is enclosed as Exhibit
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J p1 - ' go S There was only minimal training 1 the other contract employees _ 3r and I received our initial hanger pipe support assignments on ,
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or the first day, He also received an example of Bechtel calcula- i' gg tions to be used as a model. We did not immediately receivs l.
_ controlled documents to accompanv eur assignments, however, as
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pg g) , routinely occurred at most other nuclear sites where I had worked. l
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A controlled document is one whose revisions and updates are f ,c, gg) j controlled by the Quality Assurance (GA) /Ouality Controlf (QC) . Department. The document is either'an original or a Xer,ox that 'f p is verified by a control symbol. The documents contain the areas e to be checked, procedures to be followed, and relevant codes and , standards. The document is assigned to an engineer, who has a ' ec : complete package of all pertinent design material for his area , of responsibility. OA/QC insures that the engineer receives the l latest revisions as they are issued. A maior f" tion of con- l pq trolled documents is to insure that all work i ane to current requirements.
I waited around two weeks to inquire why I was not working to controlled documents. By that point I could not continue without the proper paperwork base. Not only were we worPina to_
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unverified Xerox copics, but they were incomplete. Beca'use the ;
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documents did not have all required information, engineers were working to records of example calculations they had brought
. with them from other plants, such as Susquehanna, Dav'is Bessie, ,
Midland, Maguire, Catawba and others. It was as if each engineer was working from his own reference library of resource materials ___
.[ drawn from his or her unique job history. The Final Safety i Analysis Reports for these other plants involved different l commitments to different issues of various codes than we were l committed to at Diablo Canyon. Similarly, they assumed load ,
oAus east I . ratings -- allowable design ....a for typical conditions _-- l _for other manufacturers' equipment that may not have been appli-_ l / cable to Diablo Canyon's design. ,
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The controlled documents reduce engineering judgments by
1 I spelling out assumptions, based on the nature and location of ' i j , the hardware. To illustrate, the size of gaps between hardware , ,
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controls the definitions, such as whether an item qualifies ( l In the absence of controlled documents, engineers @ _ as a restraint.
, k make assumptions based on their analogous experiences at other l plants. I know from discussions and reviews of ' engineering cal . l
; culations that the Diablo Canyon engineers used assumptions that 1 '
were inconsistent with those on the controlled documents._ B a s ed'- ] on my reviews, the assumptions used in fact differed from those i on controlled documents'in 30% of the cases. This is not a comprehensive assessment of the discrepancies, but an indicator
. that the problem is significant. If the calculations were reviewed now to the latest revision, there probably would not be
more than a handfR where all the assumptions are correct. Even that estimate may be optimistic.
i In other words, the engineering calculations at Diablo Canyon were based on conditions and commitments at other plants that may or may not have been identical with ours. The whole l point of doing separate individual engineering calculations is to ! take into account the unique conditions at each plant.
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* I alerted my' pipe support design project lead, Leo Mangoba_,
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that this flaw'could cause the NRC to close lyt.
down the design l i j
- office. He suppo'rted me and in December sent a memo to Marvin_ '
Leppke and Robert Oman._ Mr. Oman was the lead of the on-site Mr. Leppke
.projec't engineer group (OPEG) and a Bechtel employee.
was his assistant and a pG&E employee. { , Sw cDI M j ff9p.To my knowledge, there was no written response e rh tra nMi to the memo-randum. In early, January', I ingdired about a response. D O , !. ' Waugh,T Mr. M' a ngoba 's assistant, informed me of the answer which__ j had been passed along verbally: we didn ' t recuire cor. trolled l: documents and were not going to receive any. This response upset me, because the answer was clearly wrong and assumed we did not l! ,i
, care enough to'do our job properly, , X : \
Along with Kevin, I went to Messrs. Oman and Leopke's joint
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I4 office, where I notified Mr. Leppke that if the NRC were made k aware that we were working to Xerox copies of documents from other plants, the design office would be closed down. I added that'I would tell the NRC, unless I received the necessary con-trolled documents. Mr. Leppke informed me that I was not going to get them. We left.
' Within two days, however, before I went to the MC, I
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received the necessary controlled documents, in this case the p
, Ji-9 design guide for engineering. While this allowed me to work to current information pertaining to the right plant, not everyone f '
else did. In fact, for an extended period I was the only engineer besides Bechtel su'pervisors who had controlled documents.
Sporadically, the other employees received controlled documents with the latest revision. The process was not complete to my
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knowledge, however, until August, 19832 The significance is that, until Auaust, 1983, none with exceptions such as my own case, i
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5-of the pipe support ca'iculations are reliable. To insure that I the calculations were based on relevant data, they must be compared to the data on the latest revisions of relevant design guides. -
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P g up h I' Further, ,the calculation logs do not refer to the packages which had been originally prepared. Someone rewrot'e the calcula-
,! (/ tion log to delete the ref erences to the original packages which o-had failed _. The logs have bean falsified./d eiY M C43(f 7#( (o6 Y) ~ * Qc.f j#L&TS FvG out of ts*AG TH,%) itc cuenarid artw C pee #4aeb, an accurate history of the p
The only way to reconstruct c calculations would bu to ask each engineer for his or her nersonal copy of the assignment. Even that may not work.
In many cases, the engineers did not keep copies, due to Bechtel's policy.that calculations were ccmpany property. When Mr. Mangoba told us in_ i early March that Unit 1 calculations would be redone by others... in Unit 1, most of the affected engineers objected to modifica _ g,1 f tions of prior engineering calculations by a new staffer when f
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the originator was available. This. violates normal industry
practice. He answered that once we signed off and turned in V
.the calculations, they were Bechtel property and Bechtel could do what they pleased, including destroying them and having scmeoneI ,
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else rewrite them. I will uhare examples of the resulting -
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\'- falsific_ations of records with t'he NRC or law enforcement ~
officials.
To the best of my knowledge, these abuses were limited '
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to the systems which had failed. Those that passed received
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the normal checks on assumptions. ,
.d 3-' / This system was used to implement engineering judgments ,
that could not withstand scrutiny. Management's first approach (w to make Unit 1 look good was to reduce code break spans by _ % This would have reduced the total
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rel ating the code breaks.
d-number of supports in the safety related systems and omitted
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I r 0 / many of those which had failed in'the review program. However,
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management backed off after failing to come up with any plausible explanation for the NRC to justify relocating the code breaks.
,, Next, management decided to use new assumptions that ( would change the results from "fail" to "pass" on the systems.
Typically, engineers woul e instructed to assume gaps that did not exist and vice vers .O nother technioue was to assume joint releases for rigid conn,ections, which means that welds . esc w we which were in place wh we assumed to be nonexistent._ To my knowledge, there was no attempt to actually remove the welds _., Reality was only changed on paper,'not in the plant. Those are just a few of the modeling assumptions that were changed. 3 to seven different.models were tried, with seven different analyses._ l ( o Some of the hangers stil] failed, even after the assumptions had been switched repeatedly. At that po' management request-ed the designers of .the hanger supports o a reverse calcula- -- ro (uu tion trhet. ucui d d e t e rmin e the maximum loads that each hanger could support. I know of this practice because I received such assignments. T-n1 .(f'practice H) almost never occurs in the industry, (.U l d
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e.g ! because it is the eauivalent to providing the solutionc to an
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.a ' fr/ That is exactly what happened, in reve'rse order. _ A f t e.r vO .
the maximum loads were established, results.were returned'to _the stress group._ This group normally supplies the raw data on the loads, so that the support designers'can develop ade- ; I
:00 quate structures to reduce the stress to acceptable levels. tiow ,
the however, the stress group received the results of what a given q I ot support could handle and was asked to change their minds about rf the models and resulting loads that supp. ort needed .to handle.
c' I do not know if the original calculations of the stress group I were destroyed and removed from the official history, as.in our a group. I do know, however, that the stress group wast also suffer gs-o (ges)
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!' ing internal turmoil with three or four changes in the4 pipe ( strest. engineer over a year. The recently disclosed anonymous 52-page document refers to the manipulations within the stress group. 5 " #" # "'
" A Y' hN Wher iis didn't work, management tried still another
. Jew sunoorts were added within six inches of the
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techniqu . existing supports which.had failed. The stress group then modeled n'ew support gap assumptions'so that the new supports would handle most of the load for which the f ailfr3gsupports had been responsible. This "solved" the problem 3 because the , failing support could handle the small load that was left. I doubt that all of the assumptions on gaps are accurate. Even ' if they were, however, this approach represents a program of undocumented modifications. The new supports did not have control numbers or document numbers,_ so they were not recorted to the 11RC. The new supports officially did not exist. ' .M y
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The second DR that I authored concerned the angle members which are part of the piping support structures. Angle members
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throughout the plant were too long for the allowable bending
- The code stress standard used at the plant un g t AIgco requires such angles to range from Zy' to y] , depending on the angle. To research the scope of this discrepancy I g walked through Unit 1 for an hour _and r g p wg ,ll the angles that exceeded the maximum code lengthg In that short time span, I found around 200 to 300 violations on the angle members within _
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V t/ approximaraly 100 frames out of approximately 300 frames checked.
hM'T. The violations were obvious. Angle members were 2 to 3 times h longer than the ma::imum permissible lengths. In one case, a member was around nine ruet long, when the AISC maximum length was
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There has been some additional bracing, but it is erratic.
. To illustrate, one oversized member had been repaired, while I (- j/ another n' ext to it remained untouched. A more practical repair j/ would be to cut out and replace the angle frames with tube steel.
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The effect of this violation is that the supports cannot I re'liably carry th'e loads. Theoretically, the angles could buckle 3 ~ *
- .under the pressure. The angles support loads up to 1000 pounds or more, such as a bundle of small bore piping. The DR is enclosed as Exhibit 2. j .
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+. DR 83-042-S was written by Mr. Stokes and questioned the acceptability of unbraced angle mer.bers that were in apparent conflict with the AISC code. Certain members exceeded the 25" and 50" lengths specified in the
- AISC code. However, the loads carried by these members will be corresporidingly less than the 0.6 F maximum y allowable.
Angle members subject to bending are designed per AISC code , 7th Ed., section 1.5.1.4.6B. Per this section, the maximum allowable unbraced compression flange length, "Ly = is limited by a maximum allowable bending stress of 0.6 yF . Thus a 2" x 2" x 1/4" angle is limited to approximately 25" and a 4" x 4" x 1/2" angle is limited to approximately ( 50".
However, in accordance with "Commentary on AISC Specification" 1.5.1.4.5 and 1.5.1.4.6, a much greater than "Lu " length of angle member in
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bending may be used provided the allowable bending stress is correspondingly reduced. The Bechtel Corporation's Plant Design Manual SFPSM - 3.10.1, Rev. O, lists a number of such commonly used angle members, their greater than "Lu " lengths, and their correspondingly ; reduced allowable bending stress.
Thus the unbraced compression flange length, "L u ", and the allowable bending stress are inter-related. As "Lu " increases, the load that member can accommodate will decrease due to the decrease in allowabl
's bending stress.
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o ists 18 supports identified by Mr. Stokes as discrepant.
[,( All 18 supports were reviewed for compliance with AISC Code, 7th Ed.
l.5.1.4.6b and the related Bechtel Corporation's Plant Design Manual
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SFPSM - 3.10.1, Rev. O. The review concluded that the supports were in compliance Oith the above references and therefore acceptable.
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PROCEDURZ NO: P-11, R-6-s DETE: 10-5-83 . <
, . FILE NO: 146.40 '
BE3E 33 QF 33
#7,, \
g 4.6. Thermal Support Gaps Initial thermal analyses are to be performed without support gaps modeled. If resulting pipe stress or support loads are excessive, support gaps of 1/16" or larger may be modeled. As-built verification must be included in the documentation package
-'"*'* for all modeled support gaps.
4.6.3 Thermal Anchor Movements ;
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Thermal anchor movements must be included whenever anchor movements exceed 1/16" translation or 10*3 radians rotation.
Perform this analysis as described in section 4.5.5. When equipment movements are not available, they may be approximated by referring to the equipment drawing.
5.0 Documentation
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5.1 The LPSA will maintain a list of approv6d analyses. The list will contain analysis numbers, revision numbers, date of issue and whether status is preliminary or final.
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APPENDIX A NOTICE OF VIOLATION Cleveland Electric Illuminating Company Docket Nos. 50-440 and 50-441 Perry 1 and 2 License Nos. CPPR-148 and CPPR-149 As a result of the inspection conducted on September 27 - 30, 1983, and in accordance with the NRC Enforcement Policy, 47 FR 9987 (March 9,1982), the following violation was identified.
10 CFR 50, Appendix B, Criterion V, as implemented by CEI Corporate Nuclear Quality Assurance Program Manual, Section 0500, requires that activities affecting quality be performed in accordance with documented instructions and procedures of a type appropriate to the circumstances.
Contrary to the above, the following examples of' inadequate procedures were identified: 1. There was no evidence of procedures for the development, review,
,. approval and revision of the relay setting sheets within the Nuclear Construction Engineering Section and System Protection Section.
[ 2. The control and distribution of the relay setting sheets was not addressed in approved procedure in both the Nuclear Construction Engineering Section and the System Protection Section.
3. Nuclear Test Section procedure 6-1202 is inadequate because it does not adequately reflect the distribution of the relay setting sheets.
This is a Severity Level IV Violation (Supplement II.D).
Pursuant to the provisions of 10 CFR 2.201, you are hereby required to submit to this office within thirty days of the date of this Notice, a written statement or explanation in reply, including: (1) admission or denial of the alleged viola-tion; (2) the reasons for the violation if admitted; (3) the corrective steps which have been taken and the results achieved; (4) corrective steps which will
- be taken to avoid further violations; and (5) the date when full compliance will be achieved. Consideration may be given to extending your response time for good
' cause shown.
Date: //) 3 1/ S fessard, Director
- / /- Division of Engineer - . * . "% . ' .
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HISTORY OF CONTROLLED DOCUMENT
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DISTRIBUTION FOR P-149. P-126 B-75
*
12/15/83 DOCUMENTS ISSUED RETURN RECEIPT REC'D P-149, P-126 Rev 8 8/83 9/1/83 Rev 9 11/18 Not Returned As of Yet B-75 M 46 - 1 4/15 4/28 Complete manual 4/21 5/2 w/o criteria M 42 - 4 4/25 4/28 M 46 - 2 5/18 6/1 .
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M 46 - 3 6/2 6/7 M 46 - 4 8/15 8/18 M 46 - 5 10/3 10/7 M 46 - 6 11/1 11/8 M 58 - 0 5/2 5/10 M 58 1 6/20 7/29 M 58 - 2 8/3 8/5 , M 58 - 3 8/24 8/29 M 58 - 4 9/20 9/27 M 58 - 5 10/20 10/28 P 11 -4 5/26 6/1 P 11 - 5 8/31 9/6 P 11 - 6 10/20 11/8 P 33 - 0 5/23 6/1 P 33 - 1 7/7 7/15 P 34 - 0 7/26 8/5 P 27 - 3 10/20 11/8
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B - 75 CONTINUED '_. ISSUED RETURN RECEIPT REC'D 1 29 - 1 9/15 9/22 I 41 - 0 7/1 7/12-I 42 - 0 7/1 7/12 1 42 - 1 9/30 10/5 I 50 - 0 8/2 8/5 I 50 - 1 8/12 8/18 I 50 - 2 8/23 8/30 1 50 - 3 9/28 10/6 I 50 - 4 10/27' 10/30-
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, .. . ., ' - ,
B - 75 CONTINUED ISSUED RETURN RECEIPT REC'D I 29 - 1 9/15 9/22 I 41 - 0 7/1 7/12 1 42 - 0 7/1 - 7/12 1 42 - 1 9/30 10/5
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I 50 - 0 8/2 8/5 I 50 - 1 8/12 8/18 I 50 - 2 8/23 8/30 I 50 - 3 9/28 10/6 I 50 - 4 10/27'. 10/30
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December 15,1983
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I k. CFRONOLOGY OF OPEG QA AUDITS Date Audit Item # 9-17-82 PG&E GA Audit 20703, "Approval of implementing Proce-1.
dures" 10-6-82 PG&E QA Activity Audit 20917,"OPEG Drawing Control" 2.
10-14-82 PG&E GA Activity Audit 20813,"incorrect Loods Used in 3. * Honger Calc" 4. Il-3-82 Project GA Audit 28.1-1,"OPEG Design Coleulations" J l' 5. 45:82---Project-GA Audit-28.2-!, "OPEG Cerr+lier.cc :ith eta ir,
-Area of-Drow!ng C=tre!: g. J-2-H-83 Prejst GA Audii 28.4, "OPEC '*t !mt!= :M Trcining" F I;V 6.
-To confirnrOPECrcompliar,cc with4MP-2d ed PE! ! - g bi PG&E QA Audit 83087A,"OPEG Control of Deviation from gv 7. 3-15-83 1 Design"
'
8. 4-18-83 PG&E QA Activity Audit 83159A, "Mointenance of GA !
( Records" ,
9. 4-18-83 PG&A Activity Audit 83161 A, "Document Control Indoc-trination and Training Records" 10. 5-16-83 PG&A QA Audit 8317l A, "OPEG System Interaction Program - Inside Containment" ll. 5-18-83 PG&E GA Audit 83173A, "OPEG System Interaction Program - Outside Containment" 12. 6-15-83 Project GA Audit 28.2-2,"OPEG Design Change Procedure Compliance" 6-20-83 PG&E QA Activity Audit 83259A,"OPEG Problem Report 13.
System" 14. 7-27-83 PG&E QA Audit 83187A, "Review and Approval Cycle of Drawings and Calculations" 8-5-83 PG&E QA Audit 83341 A,"Control of Design Changes" 15.
16. 8-8-83 PG&E QA Audit 83339A, "OPEG Control of Design - Documents Audit 284-2, "OPEG_Colculation-Procedure Y 17, 9-!O 03 P_ro}ect G A hp!!mee" ' 18. B-24-83 Project GA Audit 30.1-3,"OPEG SIP Activities" 19. 9-12-83 PG&E QA Audit 83463A,"OPEG Audits"
_ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ ____ __.
_
-. #- , ) .. * '
HISTORY OF CONTROLLED DOCUMENT
. DISTRIBUTION FOR P-149. P-126 B-75 i 12/15/83 DOCUMENTS ISSUED RETURN RECEIPT REC'D
*
P-149. P-126 Rev 8 8/83 9/1/83 Rev 9 11/18 Not Returned As of Yet B-75 M 46 - 1 4/15 4/28 Complete manual 4/21 5/2 w/o criteria M 42 - 4 4/25 4/28
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.u' INTEROFFICE MEMORANDUM
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DiabloCanyonProject PACIFIC GAS AND ELECTRIC COMPANY BECHTEL POWER CORPORATION To Distribution cree November 23, 1983 From C. A. Huth w w 924 of Onsite Engineering 5,% ,0.P.E.G. Roster 4, Jobsite g, 3506 \
Attached please find the Onsite Project Engineering Group rester dated November 23, 1983 Notify Claudette Huth of any corrections or additions.
,
) (Yb ' SG C. A. Huth ( CAHuth/kas Q Reply Requested: No f (16 $
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I Distribution:
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BNMcAdams RManspat - K3 rickey RHudgins KHodges RGrey - KXrause APryor LMangoba ODuthie ygfg {u;f p y
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Wickard CAMonn g-l RShama JBuson f4 to JMartyn JShryock PSnooks (8 copies) VXraf t (6 copies)
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JBarrett KStokes FHademan G.C. Security (6 copies) SLeslie GSpease JVanKlompenburg TMonti , DCurtis %'
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4L/ - B= BECHTEL B/C = BECHTEL CASUAL P = PG&E V = VALTEK (PG&E AGEECY) III = CODE III (PGAE AGEECT) PACE = PACE (PG&E AGENCY) INE0VA = INIOVA (PGeE AGENCY) PECL = BECHTEL SUBSIDIARY E = ENTECH e
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DUPOOO2
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Noveabor 29, 1983 PGindE GENERAL CONSTRUCTION ON LOAN . a
'., . * * FOSTER, M. -
DPECTOR B (B) d GREEN, E. CONST. ENG. (1) (1) - Unit 1 HUSTON, K.' FLD. ENG. (1) (2) - Unit 2 LANDICE, M. DSPECTOR B~(B) (B) - Unit 1 and 2
) MONN , C.
, SUPV. FLD. CLK. (B) MONTI, T. CONST. ENG. (1) SPINNLER, R. FLD. ENG. (B) BECHTEL CASUAL (Cont'd) YAN KLOMPENBURG, T. FLD. ENG. (2).
PATEL, G. ENGINEER (B) PATEL, K.M. ENGINEER (2) PATEL, M.C. ENGINEER (1) PGandE M&NE
,
PATEL, M.R. ENGINEER (2) LEPPKE, M. PATEL, R. ENGINEER (1) ONS. PROJ. ENG. (B) RAMBELLE, R. ENGINEER (1) RAVINDRANATH, R. ENGINEER (1) REAS. JOHN ENGINEER (B) PACE ASSOCIATES MECINELLA, E. ENGINEER (1) BERKEY, S. SAHA, S. ENGINEER (1) ENGINEER (1) TRIPSIANES, L. SARKAR, A. ENGINEER (1) ENGINEER (2) SCHURER, P. ENGINEER (B) SHOOK, R. ENGINEER (2) SHAH, H. ENGINEER (2) BECHTEL CASUAL SIDDIQUE, M. ENGINEER (1) ALAM, W.
SINGH, H. ENGINEER (1) ENGINEER (2) AMIN, P. SINGH, M. ENGINEER (1) ENGINEER (2) SINHAROY, S. ENGINEER (2) AMIN, R. ENGINEER (1)
* BERGSTROM, H SUTHAR, B.G. ENGINEER (B)
ENGDEER (2) CARD, T.
. THAKAR, T. ENGINEER (2) ENGINEER (2) *
,._
DE ALDAY, J. YAZDANI, R. ENGINEER (2) ENG DEER (B)
'. f .
DHILLON, S. ENGLiEER (B) ' _ J W ANT: M. ENGINEER (2) PECL FARINELLI, T. ENGINEER (B) l GAUTiM, K.V. ENGINEER (1) GHIYA, V. SHIEH, A. ENGINEER (2) ENGINEER (1) GHOSE, A. ENGINEER (1) TVU, W. ENGINEER (B) GUTIERREZ, R. ENGINEER (2) YEN, J. ENGINEER (B) IGOLNIK0Y, T. ENGINEER (1) IZAKSON, L. ENGINEER (2) IZAKSON, V. ENGINEER (2) KANRAH, M. ENGB EER (1)
,PKATCHER, G. EEGINEER (2)
KIM, KI ENGINEER (1) KIM, S ENGINEER (1) LENOV, MAI ENGINEER (B) LIBRE, D. ENGINEER (2) g MAKAM, N. ENGINEER (B) MARDER, A. ENGINEER (1) 7 McGOWAN, M. ENGINEER (B) NAIM, M. ENGINEER (1) l l ONGTENGCO, O. ENGINEER (1) PARIKH, H. ENGINEER (2) PARIKH, R. ENGINEER (2) PATEL, A. ENGINEER (1) PATEL, B.C. ENGINEER (1) PATEL, C. ENGINEEB PATEL, D. ENGINEER _ _ _ _ _ _ -
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/ ** BECHTEL November 29, 1983
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(B) LATOS, M. (B)
i
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A3gLLA, B.
ABOU-FADEL, C.
AGOOT, M.
(B)
(2)
LESLIE, S.
LEVITAS, B.
(B) 2) ALANDY, E. (2) LEVIS, I.
LIN, JOHNSON 1) BARANGAE, P. (B) LONGVORTH, JON B) l BAREFIELD, 7. (2) LOZADA, P. (1) BARKER, R. (2) MANAPAT, R. (2) l
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BHAGAT, R. (1) MANGOBA, L. '
(2)
BOSSELO, F. McCOLLUM, A. (2)
(B} B)
BRABOT, R. (B) MORROW, A.
, BUZON, J. (2) MORST, F. B) : BTERS, J. (2) NAGATO, I. 2) i CHEN, J. (B) NAVARETTE, W. 2) CHESTHUT, S. (B) O'CONNOR, S. B) CUDIAMAT, S. (B) PHAM, LUC 2) CURTIS, D. (B) PIRLAC, E. B) DOKLADAL, J. (B) PINGOL, R. 1) DUTHIE, O. POTOCKI, B.
ESPINELLI, R. (B) QUIOGUE 1. 1) FOSTER, C. (2) RANSON, R. 2) GEHLERT, L. (1) REARDON, E. B) l GRET, R. (B) REMOTIGUE, J. 2) HAAKE, J. (D) RUUD, C. (B) H130VER, M. (2) SHEN, J. (1) HARRIS, S. (B) SHUSTERMAI, A. '
(B) )i HENDERSON, G.
HOOSHANGI, M.
(B)
(2)
SE00HK0, S.
__SOORMA. S.
(B) H17IH, C. (B) SOULE, Y. 2) HUTH, R. (B) SPEASE, G. (B) JAHANGIR, I. (2) STEPANIX, E. (2) JILVAN, E. (2) TAM, P. (B) JOSE, V. (B) VAKILI, M. (B) JUSON, I. (2) VAUGHAI, H. (B) KAHL, C. (B) VIAL, N. (1) IIX, S. (B) VAIJtER, V. (B) KLEIMAN, S. VARREN, D. (B) K0WALSKI, B. (B) VAUGH, K. (2) . KRAUSE, I. (B) WONG, C.
' V00DWARD, V. (1)
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* ' ANDERSON, P. DRAFTERB-1(B) 'BALNARD,J./ ADMD. ASST. SR. (1)
BELLOT, R. ADMIN. CLERK (B) BENN, GARY ADMII. ASST. .(1) ADNIN. CLERK SR. (B) CODE III BERTA, B.
CALDERONE, F. DRAFTER B-1 (B) COLO, R. ENGINEER (2) CARB, C. DRAFTER B-1 (B) CONDRON, K. ENGINEER (1) CECIL, L. ADMIN. CLERE (B) CUMMINGS, T. ENGINEER (1) CHAMBERS, D. DRAFTERB-1(B) ENGINEER (1) COOPER, K. DRAFTER B-3 (B) FAURIA, J.
ADMIN. ASST. SR. (B) GARAY, R. ENGINEER (2) CORNVELL, M.
HANKINS, R. ENGINEER (2) DARLING, M. DRAFTER B-1 (B) * HARE, R. ENGINEER (1) DARLING, R. DRAFTER B-1 (B) HEBERGER, M. ENGINEER (2)
** DEAN, C. AIM. COORD (B)
ADMIN. CLERK (B) JACQUES. S. ENGINEERJ1) HAMES, S.
ADMIN. ASST. SR. (B) KEWALRAMANI, M. ENGINEER (1) HIISON, L. ENGINEER (1) H0DGES, D. SR. CLERE TYPIST (B) KLIMCZAK, R.
DRAFTER B-4 (2) MAHNCKE, A. ENGINEER (B) HOLMES, J.
DRAFTER B-1 +++(2) RYLANCE, K. ENGINEER (2) NORTON, N.
DATA TRANS. A-1 (B) SANT0 TOME, F. ENGINEER (2) JENSEN, S.
ADMIN. ASST. SR.(B) SISON; P. ENGINEER (1) JIMES05, S.
JOHNSTON, B. DATA TRANS. A-2+++(B) JONES, K. DATA TRANS. A-2 (B)
*KEARNS, B. DRAFTER B-1 (B) ~ ,
LA;iGER, C. SR. CLERK TYPIST (B) INNOVA ' i NASON, B. DATA TRANS. A-2 (B) ADMIN. CLERK SR. (B) SHOIKHET, M. ENGINEER (B) MILLER, L.
MOORE, L. SR. CLERE TYPIST (B) - NATE, A. DRAFTER B-1 (B) NISSEN, CHRIS DRAFTER A-2 (B) NISSEN, CINDY DRAFTER B-1 (B) PEOPLES, E. ADMIN ASST. SR. (2) RICHTER, L. ADMIN. ASST. SR. (B) ROBERTS, J. SR. CLERK TYPIST (B) ROBBINS, T. ADMIN. ASST. SR. (B) ROGOVAY, P. SR. CLERK TYPIST (B) SALDIVAR, R. DRAFTER B-1 (B) SCHNEIDER, J. ADMIN CLERK SR.+++(B) SCHOTT, A. DRAFTER B-1 (B) SCHOTT, M. DRAFTER B-1 (B) SPICKERT, K. ADMIN. ASST. SR. A-1 (B) TERRELL, B. ADMIN. ASST. SR. (B) WALLING, S. DRAFTER B-1 (B) VILSON, D. DRAFTER TRAINEE (B)
+++VORKING SECOND SHIFT * On loan to G.C. Services ** On loan f rom G.C. Services .- - -.- -
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0. P. E. G. PERSONNEL ROSTER n/23/83
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OUTSIDE ROIJE
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IAME GROUP LOCATION LINE LINE CO. g Abe n a, Benny Piping 8 541-7488 3488 B B Abou-Padel, Camine Civil Bldg B3 595-7351 3873 B B Agoot, Mario Walkdown P Plant 541-7517 3517 B 2 Alas, Vali S/B Pipe Stress #2 34 541-7064 3064 B/C 2 Alandy, Edgar S/B Pipe Support 32 541-7066 3066 B 2 Amin, Paresh PSDTC Plant 541-7504 3844 B/C 2 h! fAmin,Rajesh S/B Pipe Support 32 541-7066 3066 B/C 1 Anderson, Paula S/B Pipe Support 19 541-7540 3540 V B 9f ff3 --p Amo, Shahram S/B Pipe Support 32 541-7066 X66 B B Ballard, Jayne Ad mi nistration 32 541-7066 3066 V B Barangan, P. Nico Electrical 8 541-7487 3487 B B Barefield, Frank Valkdown E Plant 541-7122 3122 B 2 Barker, Rex Hanger Feasibility 19 541-7540 3540 B 2 Bonn, Gary S/B Pipe Support 32 541-7066 3066 V B Bergstrom, Howard S/B Pipe Support 19 541-7540 3540 B/C 2 Berkey, S. PSDTC 64 541-7071 3071 Pace 1 Berta, Bonnie g Document Control Office 541-7065 3065 V B p f %-.h Bhagat, j Easesh g S/B Pipe Support 32 541-7066 % 66 B 1 I Bossoloo, F. Instrumentation 8 541-7573 3573 B B Braboy, Ralph Civil Bldg B3 595-7351 3873 B 1 Buson, C. John Valkdown ? Plant 541-7517 3517 B 2 Byers, Jeff Valkdown I Plant 341-7122 3122 B 2 Calderono, Fmnk Drafting 2 .541-7065 304" W B S/B Pipe Stress #2 9 541-7492 3492 t/c
[9-kj 9 Card, Carr, Tom Carol Drafting 2 541-7065 % 65 V
B ( .' Cecil, Lis Chambers, Doa Document Control Drefting Office A 541-7065 541-7065 3065 -- V 3065 V B E Chen, Joseph S/B Pips Stress #2 9 541-7492 3492 B B Chesut, Steve Mechanical 8 541 7574 3574 B B Colo, Richard PSDTC Plant S41-7504 3844 III 2 g (p/M Condren, Kelly S/B Pipe Stress 9 541-7492 3492 III 1 Cooper, Kyle Drafting 2 541-7065 3065 V B Cornwell, Mary Document Control Office 541-7065 3065 V B Cudiamat, Serafin Walkdown Plant 541-7577 3577 B B llf G 2.,9 ummings, C Tom S/B '/ipe Support 32 541-7066 3066 III 1 Curtis, Dan PD Group Supv. 34 541-7502 3502 B B Darling, Mike Drafting 2 541-7065 3065 V B Darling, Robert Drafting 2 541-7065 3065 V B S/B Pipe Stress #2 9 541-7492 3492 B/C B kp)9DeAlday,I.
'I/g $PDhillon, Sarjit Jose S/B Pipe Stress #2 9 541-7492 3492 B/C B Dokladal, Jiri gj Arani, Manmohan /
/. Mechanical S/B Pipe Support 8 541-7488 3488 B B 32 541-7066 3066 B/C 2 Duthie, Opel Walkdown E Plant 541-7122 3122 B 2 Espinelli, Rudy Electrical ,8 541-7573 3573 B B Fa en i, Tim Civil Bldg B3 595-7351 3140 PAL 2 auria. Jia 7 PSDTC 68 541-7070 3070 III 1 Foster, Curtis Valkdown E Plant 541-7122 3122 B 2 Poster, Mike Drafting 2 541-7065 3065 P B Garay, Romulo S/B Pipe Stress #2 34 541-7064 3064 III 2 g h' g cautas, K. V. S/B Pipe Support 32 541-7066 3066 B/C 1 Gehlert, Lyle Valkdown P Plant 541-7517 3517 B 2 ! Ghiya, Virendra S/B Pipe Support 32 541-7066 3066 B/C 1 '
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0. P. E. G. PERSONNEL ROSTER 11/23/83 OUTSIDE ROLN
; .IAME GROUP LOCATION LUE L3E CO. UNIT ifs S Ghose, Anal S/B Pipe Support 32 541-7066 3066 B/C 1 Green, irban Piping 34 541-7064 3064 P 1
' Grey, Dick S/B Pipe Stmas 9 541-7492 3492 B B Gutierres, R. S/B Pipe Stress 34 541-7064 3064 B/C 2 Haake, John Mechanical 8 541-7488 3488 B B Haaes, Sandy Administration Office 595-7506 3506 V B Hankins, Rick Piping 34 541-7502 3502 III. 1 (o Ig.s>J)* Hanover, Mark S/B Pipe Stress #2 9 541-7492 3492 B 2 Here, Rich Piping 34 541-7502 3502 III 1 Marris, Steve Mechanical 8 541-7574 3574 5 B ( How eger. M. 3 PSDTC 53 541-7504 3844 III 2 Henderson, Gary S/B Pipe Support 19 541-7540 3540 B B Hixson, Lisa Document Control Office 541-7065 3065 V B Holass, Jim Hanger Peasibility 19 541-7540 3540 V B Holst, Clos Quality Control 2 541-7065 3065 P B Hooshangi, Mehdi S/B Pipe Stress 9 541-7492 3492 B 2 Horton, Mike Walkdown E Plant 541-7122 3122 V 2 Huston, Kim PSDTC 68 541-7070 3070 P 1 Huth, Claudette Administation Office 541-7506 3506 B B Huth, Jeff Mechanical 8 541-7516 3516 B B S/B Pipe Support 32 541-7066 3066 B/C 1 5/jQIgoinikov, Yury Isakson, Ludmila S/B Pipe Support 19 , 541-7540 3540 B/C 2 Isakson, Vi:: tor / S/B Pipe Support 19 541-7340 - 3540 B/C 2 J m --) Jacques, Staa V S/B Pipe Stress #1 9 541-7492 3492 III 1 1 Jahangir, Nosar PSFfC Plant 541-7504 3844 B 2
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Jensen, Steve Computer Services Office 541-7134 3134' W B Jilvan, Emile PSDTC Plant 541-7504 3844 3 2 ! Jimeson, Susan Administration Office 541-7506 3506 V B l Johnston, B. Computer Service Office 541-7134 3134 V B l Jones, Karen Computer Service Office 541-7134 3134 V B ! T M , Jose, Vince S/B Pipe Stress #2 9 541-7492 3492 B B g. p Juson, Ismael S/S Pipe Stress 9 541-7492 3493 3 2 Kahl, Chris Electrical 8 541-7487 3487 B B Kanrar Murari S/B Pipe Support 19 541-7540 3540 B/C 1 Katcher, Gary S/B Pipe Support 19 541-7540 3540 B/C 2 Kearns, Bard Drafting 1 541-7065 3065 V B , y Kowalramani, Mohan S/B Pipe Support 32 541-7066 3066 III 1 ' Kim, Ki C. S/B Pipe Support 19 541-7540 3540 B/C 1 Kim, Sung Electrical 8 541-7487 3487 B B S/B Pipe Stress 9 541-7492 3492 III 1 pgp(p>KHacssk, Kowalski, BrianRich J PSDTC PP.at 541-7504 3844 B B Krause, Konriid Piping 8 541-7488 3488 B B Iendice, Mark Drafting B 541-7065 3065 P B Langer, Constance Administration 9 541-7492 3492 V B
$9 os, Modesto enov, Max S/B Pipe Stress #2 S/B Pipe Stress #2
9 541-7064 541-7492 3064 3492 B B/C B B Leppke, Myron Project Engineer Office 541-7507 3507 P B Leslie, Steve Scheduling Office 541-7063 3063 B B Levitas, Boris Valkdown P Plant 541-7517 3517 B 2 Lewis, Roger PSDTC 64 541-7071 3071 B 1 A Libre, Deogracias S/B Pipe Stress #2 34 541-7064 3064 B/C 2
-> Lin, Johnson 32 541-7066 3066 B -
Longworth, Jon S/B Asst.Pipe Proj.Jtrasar Engr.y p Office 541-7507 3507 B B
% 9 ozada, L Paul Piping 32 541-7066 3066 B 1 Mahncke, Allen S/B Pipe Stress #2 34 541-7064 3064 III 2 N %p b -
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O. P. E. G. PERSONNEL ROSTER 11-23-83 OUTSIDE R0!Jt GROUP LOCATION _ LIBE LINE 0 C0,. g V}
[ ' . NAME .
Makaa, Nagarajagupta Civil Bldg B3 595-7351 3872 B/C B Nanapat, Roland S/B Pipe Support 19 541-7578 3578 B 2 Mangoba, Leo S/B Pipe Support 32 .541-7061 7 B B Marder, Arkady S/B Pipe Support 19 541-7540 3 B/C 1 Nason, B. Computer Services Office 541-7134 3134 V B l l NoCollus, Athens Walkdown E Plant 541-7122 3122 B 2 Riller, Linda Document Control Office 541-7065 3065 V B Nonn, Craig Administration Office 541-7506 3506 P B ,
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Monti, Tom PSDFC 53 541-7508 3508 P 1 Moore, Linda Document Control Office 541-7065 3065 V B Norrow, A. T. PSDTC 53 541-7557 3557 B B Morsy, Pathy Civil 31ds B3 541-7137 3137 3 B Nagayo, Napoleon Valkdown P Plant 541-7517 3517 3 2 S/B Pipe Support 32 541-7066 3066 B/C 1 2hg.d Naim, Moossa Navarotte, Wilfndo Valkdown P Plant 541-7517 3517 B 2 Mayo, Alan Drafting 2 541-7065 3065 V B Eissen, Christine Drafting B 541-7065 3065 V B Iissen, Cindy Drafting 19 541-7540 3540 V B O'Connor Steve Civil Bldg B3 595-7140 3140 B B Ongtengco, Oscar S/B Pipe Support 32 541-7066 3066 B/C 1 Parekh, Harrish S/B Pipe Stress 34 541-7502 3502 B/C 2 S/B Pipe Support 32 541-7066 3066 B/C 1
d Ig - >y Patel, Arvind H. S/B Pipe Support 32 ji,41-7066 3066 B/C 1 1-/ F3 9f. .,Patel, Patel, B.C.
Dahyabbia E. S/B Pipe Support 32 541-7066 3066 B/C 1 S/B Pipe Str:ess 541-7492 3492 B/C 3
~f *Patel,Ghanshyamb 9 g/ , ' apPatel, Kanti M. / S/B Pipe Stross #2 34 541-7064 3064 P/C 2 Patel, Nahesh C. g PSDTC 68 541-7070 3070 B/C 1 P-/dJ,--PPatel, Mahesh R. 3 S/B Pipe Support L 32 541-7066 3066 B/C 2 PSDTC 68 - 541-7070 3070 B/C 1 a W Ramish M. ]
eoples, ava Administration 19 541-7540 3540 V 2 Pham, Luc S/B Pipe Support 19 541-7540 3540 3 2
, .Pingol, Ricardo S/B Pipe Stnes 32 541-7066 3066 B 1 Pinlac y ito S/B Pipe Support 19 541-7540 3540 B B j S/B Pips Stnse 34 541-7502 3502 3 1 S/B Pipe Support 32 541-7066 3066 B 7k)c-Uotocri. >Quiogue, Adello Ranson, Robert BarbaraS/B Pipe Support 19 541-7540 3540 B 2 S/B Pipe Support 32 541-7066 3066 B/C 1 q gy-p Rambelle, Ravindranath, R. R. S/B Pipe Support 19 541-7540 3540 B/C 1 Reardon, Al Civil Bldg B3 595-7351 3873 B B k b Reas, John S/B Pipe Stress #2 9 541-7492 3492 B/C B g/g2_-p,Recinella,Ed S/B Pipe Support 32 541-7066 3066 B/C 2
, ' Remotigue, Julito Walkdown P Plant 541-7517 3517 B 2 Richt er, 14urel Document Control Office 541-7065 3065 V B.
Robbins, Tonya Ad mi ni st ra tion Office 541-7506 3506 V B Roberts, Jill Administration Office 541-7506 3506 V B Rogoway, Patify Administration Bldg B3 595-7351 3874 V B Rylance, Kevin Hanger Peasibility 19 541-7540 3540 III 2 Saha, Swapan S/B Pipe Support 19 541-7540 3540 B/C 2 Saldivar, Roberto Drafting 2 541-7065 3065 V B u Santotone, Permin S/B Pipe Support 19 541-7540 3540 III 2 S/B Pipe Support 32 541-7066 3066 B/C 2
>--y Sarkar, Ami V ,
G nuvidor, not Document Control B 541-7065 3065 B I 1 9. N b$h )h
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. - O. P. E. G. PERSONEEL ROSTER 11-23 33 OUTSIDE ROIJL GROUP LOCATION LINE _ LINE Cf. UNIT NAME 2 541-7065 3065 V 2 Schott, Andrew Drafting Drafting 2 3065 W B Schott, Matt ---- Schurer, Pete S/B Pipe Support 32 (541-7061 541-706T) 3066 B/c B S/B Pipe Stress #2 34 541-7064 3064 B/C 2 Shah, H.N.
S/B Pipe Stress 34 541-7502 3502 B B Shen, Jack S/B Pipe Stress #2 34 541-7064 3064 PECL 2 Shieh, A.M.S. / S/B Pipe Support 541-7066 3066 ~ INNOVA 1
' 0[F t-*> Shoikhet, Moisey V 32 Shook, Richard PSDTC Plant 541-7504 3844 B/c 2 /C[( 0., 9Shustezzan, Asriel S/B Pipe Support 32 541-7066 3066 B B S/B Pipe Support 32 541-7066 3066 B/C 1 g g pr==Siddique, Mohammad S/B Pipe Support 541-7540 3540 B/c 1 Singh, Hardev 19 S/B Pipe Support 32 541-7066 3066 B/C 1 Singh, Monmohan Sinharoy, Sushil PSDTC Plant 541-7504 3844 B/c 2 S/B Pipe Support 541-7066 3066 III 1 6' M' Sison, Pedrito 32 B
Skochko, Steve j Instmaentation 8 541-7573 3573 B Soorma, Suresh v S/B Pipe Stress 9 541-7492 3492 B B (o Plant 541-7122 3122 B 2 Soule, William Walkdown E Spease, Gene Quality Engineer 8 541-7574 3574 B B Spickert, Kara Administration Office 541-7506 3506 v B Spinnler, Robert PSDTC Plant 541-7504 3844 P B Plant 541-7122 3122 B 2 Stepanik, Ed Walkdown E Suthar, B. C. Civil Bldg B3 541-71A0 3140 B/C B S/B Pipe Stress 9 341-7492 3492 B B h('2-# Tam, Peter 8/19 3488 V B Terrell, Bertha Administration 541-7488 Thakar, Yogen S/B Pipe Stress 34 541-7502 3502' B/c 2 S/B Pipe Stress #2 9 541-7492 3492 Pace 2 h , , f 9Tripsianes, Tvu, Wender L. S/B Pipe Stress #2 9 541-7492 3492 PECL B 595-7351 3872 B B Vakili, Mansoor Civil Blds B3 Plant 541-7504 3844 P 2 Van Klompenburg, Jeff PSDTC Va.ughan, Harold Civil Bldg B3 595-7351 3872 B B B Vial, Manuel S/B Pipe Stress 34 541-7502 3502 B 8 541-7488 3488 B B Walker, Willis Civil Varren, Don Instmaentation 8 541-7573 3573' B* B S/B Pipe Support 19 541-7646 3646 B 2 Waugh, Kevin Draf ting B 541-7065 3065 W B Vilson, Dave Woodward, Villiam Walkdown P Plant 541-7517 3517 B g S/B Pipe Stress #2 9 541-7492 3492 B/C 2 NM 'h Yazdani, 5 Yen, JeromeRobert S/B Pipe Stress #2 9 541-7492 3492 PECL B Y l0 g g
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e, S $$ Z MEMORANDUM FOR: D. G. Eisenhut, Director, Division of Licensing, NRR yjg QS FROM: T. W. Bishop, Director Division of Resident, Reactor P@ects, and Engineering Programs, RV --
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SUBJECT: ASSISTANCE IN FOLLOW-UP TO DIABLO CANYON ALLEGATIONS On November 28, 1983 we were provided a copy of a 21 page affidavit relating to concerns at Diablo Canyon. We have reviewed the affidavit and conclude that we require NRR assistance in addressing two of the issues discussed in the affidavit. These are:
' (1) Improper design activities related to the use of U-bolts (pages 5-8 of the affidavit).
(2) Improper design activities related to "code-break" (code boundary) design pages 8-13 of the affidavit).
Excerpted pages of the affidavit are enclosed. This affidavit is being controlled under NRC Allegation Tracking System item number RV83A063.
It is requested that a Task Interface Agreement be initiated on the above.
(Region V contact: D. F. Kirsch FTS 463-3723).
l. h T. W. Bishop, Director Division of Resident, Reactor Projects, and Engineering Programs, RV Enclosure: as stated cc: G. Knighton, NRR J. Knight, NRR S. Ebneter, RI D. Kirsch, RV OHth)$ki ka ! ,( -
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ElKLOSURE [[j/.M'/
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of the pipe support calculations are reliable. To insure that the calculations were based on relevant data, they must he compared to the data on the latest revisions of relevant design f
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guides.
. f D On February 4, 1983, I challenged a second major problem, 'fhedJinvolvingunreliableloadratingsforU-boltsinUnit1.
.e In y pf lay terms, a load rating specifies the maximum amount of force CD - in a given direction that the item can safely withstand, based on a 95% confidence level in the rating. I had learned, however, that the supplier's (ITT Grinne 1 ratings for U-bolts that we had purchased in some cases were ee4y one-third to one-fourth more stringent as those claimed on . Drawing 049243. Drawing 049243 I
, is a PG&E document that specifies acceptable spacing, strengths 1 f and load ratings for small bore pipe supports, including U-bolts.
I It is used on-site as the final authority for relevant analysis. j
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To illustrate, ITT Grinnell only claimed that the load rating was 485 pounds tension for 1/4 inch U-bolts on 1/2 inch pipes - under "design normal and upset conditions". By contrast, Drawing 049243 credited the same U- bolts with a load rating of i 2000 pounds, over.4 times more than the vendor would vouch for.
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I have worked with vendor catalogs for years, and the I ITT Grinnell catalog is particul'arly common.
' I had never before i seen the customer assert that a part was 4 times stronger than the seller claimed. In my opinion, 049243 would represent a false statement if it were furnished to the NRC.
. Obviously, PG&E relied on a series of suspect assumptions in order to exaggerate the load ratings. Three examples out - of many should illustrate my point. The first example involves the interaction equation for the combination of tension and side-loading. The sum of the two factors, which always involves
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[6 4, ;-3:,fj ecimals,.cannot
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d exceed a value of "one". On Drawing 049243,
[' however, there was a handicap to unrealistically increase the odds ~
of passing this test: each factor was squared. Since the factors had values less than one to start with, squaring them produced smaller decimals, distorting the values and reducing'
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the chances that the combined factors would exceed a value of
"one". To illustrate, if two factors each had a value of .'7,
; their sum would be 1.4 and the bolt would fail. But .7 scuared I is .49, which means their sum would pass with a. 98 total due to rigging the equation. ITT Grinnell does not square these,fac-tors, nor does any other plant where I have worked in the nuclear
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industry. At Diablo Canyon, management has declared its own version of engineering reality. I do not believe it is justi-fied in this instance. . Another flaw in the U-bolt load ratings on Drawing 049243 came from the September 28, 1978 PG&E "U-Bolt Test Pro. gram, Revision 1" report. The original test was biased by not reflect.-
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ing actual conditions in the plant. The repcrt concedes dis-crepancies but claims they make the results more conservative.
That is inaccurate. To illustrate, the test report says that stress on the U-bolts was reduced at Diablo Canyon compared to the ITT Grinnell load ratings,due to the absence of shims in certain cases. The report failed to mention what I observed in the field, however: in order to avoid shims, the U-bolts had to be forcibly bent, failing the bolts at installation. In Unit 2, this type of problem was fixed. The U-bolts had been bent so far that new holes had to be drilled to reinstall them.
Management did not want to make these repairs for the same flaws in Unit 1, however. To my knowledge,'the repairs did not occur during my time at the plant.
The report states that the test apparatus to determine load ratings had used "extra heavy schedule 160" pipe. Unfortunately,
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Boa (Itav (y-S) thinner and therefore could create a f ailure due to mcvc=c.w of the pipe itself. As a result, the load ratings for the bolts should be reduced to compensate. They weren't, and are therefore unrealistic. The problem is even worse, however, because the U-bolts also are attached to schedule 40 pipe, - which is thinner yet, and copper tubing so thin it can be bent by hand. The effect is that, even if the U-holt load ratings were accurate,
[pt) Bacc'm& of they would be unrealistic due to(Wthe ifAo possibility of movcncat by the piping thinner than PG&E assumed.
h dy, $A96 A third major flaw is that the PG&E tests were done at ambient, or room, temperature, but the ITT Grinnell ratings were made at 650 Farenheit. Conditions in the plant are hardly room temperature. Well over 50% of the piping lines exceed ambient temperature. Many lines exceed the ITT Grinnell 650 F. assump-tion, with temperatures up to nearly 1000 F. on some lines.
On February 4, I informe Kevin Waugh of what I had . learned.. The flaws described above are st a few f the problems with the U-bolt. load ratings. Afterwards, I.was. permitted.to rely on ITT Grinnell's load ratings which allowed me.to fail some of the U-bolts. To the best of my knowledge, however, the other engineers in the group continued to use the PG&E load ratings.
I I know, because they told me in discussions that I was right, but the group leader had instructed them to use PG&E's ratings.
I I also checked the methodology ~of my critique with a Statistics I j professor at California Polytechnical Institute, who confirmed I that I had used the appropriate formulas to develop my analysis proving that the PG&E load ratings for U-bolts were unrealistic and should have been roughly the same as ITT Grinnell's.
Even if the load ratings for U-bolts were accurate, the hangers to which they are attached would not meet design require-ments. This is because other relevant factors also were defined
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~ - . v s 'N..$ r/.D - / .of the process. The point of the load rating is to determine how .
w-i'/..och force will displace the bolt, and possibly cause the hanger to
* move beyond legal commitments. But, as seen below, other parts of the . '
hanger besides U-bolts can lead to excessive movement. For example, the instructions for review of Unit 1 small bore piping direct the { cngineers to assume that the base plates are rigid for bolts and sup-ports. The problem is that they are not; many of them move. The plates could lead to many times greater displacement than the U-bolts, but l this factor was defined out of existence. Also, the maximum displace- l ment allowed per M-9 for the support was .025 inches. From the U-bolt l report the failure point for the U-bolts was .025 inches. Later, the support steel (not including the base plate) was checked to a .025 inch displacement, ignoring the .025 inch a'lready included in the U-bolt load rating and the displacement at the support point caused by flexi-bility of' base plate.
- ffpf i At the end of February, I learned of a third major problem.
, I was assigned to perform pipe stress calculations on code break sys-
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tems; many were high-temperature lines. This assignment involved determining support requirements for code breaks, which are the boun-daries of Class I critical systems, and other critical seismic systems.
We learned that there was not enough off-set, or space between the valves and the large bore piping, to avoid unacceptable stress on the
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small bore pipe line that branched off the large bore lines. The code
,' break valves we're located on these small bore lines. The group leader, A r. Kim at first did not believe me. But after reviewing my fig- ; ures, he agreed and supported me.
evin Waugh nd I went to se Mr. Leppk to recommend the necessary corrective action -- addition 1 spans of pipe to increase I
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the distance between code break valves and large bore piping. keppkh lI refused to modify the system because tne lines had already been hydro-f tested. He challenged our "M-40" calculations for accuracy, and said ho instead would use the ME 101 computer stress analysis to analyze these lines. He then proceeded, however, to let it slip that the sys-
- tems probably would not work, and explained why. The vendor had not received correct instructions when installing the lines. The vendors p were told to install the piping at room temperature under normal conditions
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n- o wra jffi,/, That would be accurate for a fossil plant but was wrong for the l'. 4 lines in question, which required seismic. supports and had to endure temperature ranges in excess of 650 F. These two fac-tors drastically increased the stress, necessitating greater - distance's between valves and large bore piping.. Leppk ven
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admitted that the lines would probably have been ' installed. right if the vendors had received accurate instructions. Eventually , the computer run confirmed my critique, but I am not aware that
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all critical lines were rerouted In theory, we reached a milestone in Mar'ch, 1983. Supposedly we had finished the cal ~culations for review of Unit 1. In fact, they were generally in a preliminary state. Reverifica-tion for accuracy was necessary for final issuance'. At this time we learned that the personnel who had been doing.the calcu-
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lations would be split. Half would he transferred to a new trailer. , i The transfers displ'ayed.a consistent disturbing trait: those who had not questioned suspect assumptions verbally imp.osed by Bechtel supervisors all'went to the new trailer. Those who ' had been objecting to unreliab'le assumptions stayed in the old trailer. The engineers in the new trailer would continue working on Unit 1. To review Unit 1 calculations, the revamped. crews were filled out with new recruits,.many without background on the issues at Diablo Canyon and without security clearances to enter Unit 1, which is necessary on occasion to check the hardware.
Those left behind received new assignments for Unit 2.
l ' Af ter this point I kept close tabs on what was", happening in the Unit 1 trailer, which was easy due to my daily association and friendships with t.he personnel. I learned that they.went beyond routine citation check reviews of the calculations, their Y official mission. Instead,.the engineers redid the calculations i t
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bW 4j ( ,/ entirely for nearly all the systems which had failed. In the process, the original calculations demonstrating system failures vanished.. It is important to realize the scop,e of this effort
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to rewrite,the results. Prior to the split of the trailers, the failure' rate for the entire small bore pipe support group had been around 50% ot /#or'E, N)
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D Those results would have'mean't,that Unit 1 failed the re-
levaluation program and required complete reanalysis. Bechtel and PG&E management could not accepp.this, so they had the com-cf pliant engineers rewrite our prior failing calculations. One , problem is that the ensuing packages did not disclose that'the There is no mention of the p[H calculations original failinghad been done calculations, over.
which were destroyed.
u I k Further, ,the calculation logs do not refer to the pac ages which had been originally prepared. Someone rewrote the calcula-tion log to delete the references to the original packages which had failed. The logs have been falsified. M W O C 41F j 7H( (o6
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Mt1,AGCLECTS Fidd 007 of M*A6 TH40100 CMcMYth$ atftCH [ PQe#4AcD,
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The only way to reconstruct an accurate history of the calculations would be to ask each engineer for his or her personal copy of the assignment. Even that may not work. In many cases, engineers did not keep copies,.due to Bechtel's policy that the calculations were company property. When Mr. !!angoba told us in early March that Unit 1 calculations would be redone by others in Unit 1, most of the affected , engineers objected to modifica-tions of prior engineering calculations by a new staffer when the originator was available. This violates normal industry practice. He answered that once we signed off and turned in
.the calculations, they were Bech.tel property and Bechtel'could do what they pleased, including destroying them and having someone N, I will share examples of the resulting else. rewrite them.
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fhicifications of records with the NRC or law enforcement officials.
. To'the best of my knowledge, these abuses were limited * to the systems which had failed. Those that passed received i I
{ the normal checks on assumptions.
f This system was used to implement engineering judgments M that could not withstand scrutiny.
' Management's first approach WM to make Unit 1 look good was to reduce code break spans by I{I , gA relocating the code breaks. This would have reduced the total J y Y ) number of supports in the safety related systems and omitted
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many of those which had dailed in the. review program. However, management backed off after failing to come up with any plausible explanation for the NRC to justify relocating the code breaks.
Next, management decided to use new assumptions that would change the results from "fail" to "pass" on the systems.
Typically, engineerr. would be instructed to assume gaps that did
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not exist and vice versa. Another technicue was to assume joint releases for rigid connections., which means that' welds yert (uS . which were in place wher-e assumed to be nonexistent. To my
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knowledge, there was no attempt to actually remove the welds.. Reality was only changed on paper,'not in the plant. Those are just a few of the modeling assumptions that were changed. Up .
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to seven different models were tried, with seven different analyses.
. . Some of the hangers stil.). failed', even after the assumptions had been switched repeatedly. At that point, management recuest-ed the 'desig rs of the hanger supports to do a reverse calcula-re tion t-hed. ,wuld etermine the maximum loads that each hanger could support. I know of this practice because I received such assignments. 7'Nt'S(//S) Tha prac tice almost never occurs in the industry,
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"- / because it is the equivalent to providing the solution tocat,+md an e6]m *
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' undefined challenge. l question in order to satisfy an answer that already has been -
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supplied. .
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That is exactly what happened, in reverse order. After , the maximum loads were established, results. were. returned to the stress group. This group normally supplies the raw data
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on the loads, so that the support designers can develop ade-quate structures to reduce the stress to acceptable levels. Now, however, the stress group received the results of what a given support could handle and was asked to change their minds.about the models and resulting loads.that supp. ort needed .to handle.
I do not know if the original calculations of the stress group , were destroyed and removed from the official history, as in our
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group. I do know, however, that the stress group was also suffer-
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ing internal turmoil with three or four changes in t e x pipe stress engineer over a y' ear. The recently disclosed anonymous 52-page document refers to the manipulations within the stress
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Where this didn't work, ' management tried stil'1 another
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technique. New supports were added within six inches of the existing supports which.had failed. The stress group then modeled new support gap assumptions so that the new supports would handle most of the load for which .the failnysupports had been responsible. This "solved" the problem 3 becau.se the failing support could handle the.small load that was left. I dcubt that all of the assumptions on gaps are accurate. Even
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if they were, however, this approach represents a program of undocumented modifications. The new supports did not have control numbers or document numbers, so they were not recorted p to the NRC. The new supports officially did not exist. .M Y
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not reaction to the developments since wa "finishcd" the calcula-tions in March puts the series of abuses in perspective. I have never before seen mana.gement at any plaat that was so'per- , y sistent about accepting work that had been done wrong originally.( ymT 5 W
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In June, I was transferred to the 3"Qu'ick Fix" group,.offici .
. all k \ nown as the PipegXSubport Dfesign Tolerance Clarification .d i
g 5 (PSD i group. Quick eex was an extension of the pipe support dLil group, ith responsibility to revise drawings so that hangers N could be 'nstalled without later rework.
During is assignment, I continued to learn of inaccurat.e engineering re ords. For instance, in. July, 1983, I learned
. that a pipe was esting on a unistrut -- a light gauge , steel section to which ectrical apparatus is attached. This' meant that it was an unin entional restraint which needed to be removed.
Instead of making th necessary repair, however, the pipe stress designer modeled the u
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str.utasasupportforthepipe[ The plan backfired. en the data on the new support was fed into the computer, the r sults indicated that another' support had to be added for the other ide of the oipe.
' You can't have f & tc one-directionalA supports under. MB-101, the computer pipe stress ,' program. Further, upon inspection, it was obvious that the
& D6MD w7*[dM) 1 on" unistrut was too small to even suppor c.nc cide. j was then remove the original unistrub and add wo full-sized '
support ,w.ich "'r thrce cinc; thc .w u nave..cuL '.sd been
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l trying 'r '"cid. hW l After consulting the stres.s group originato , I dele'ted l the new supports. By this point, the unistrut air ady had been
removed in preparation for thew rvre {6ed I installation. Ther weren't l supposed to be supports tt that location anyway. The xtra work would have been due to the official fiction create :M MS l I l s
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DIABLO CANYON ALLEGATION MANAGEMENT PROGRAM
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NOVEMBER 23, 1983 Approval:
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R. DeYoun A rec , IE
//lL3 Date~
ff3 es- .un H. R. Denton, Director, NRR rDatd fk $ If 7; B. Martin, Administrator, Date
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Region V [-hd13 3 f([Q<fl [ h ,
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,_ Diablo Canyon Allegation Management Program Contents Page
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I Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 II Scope . . . . . . . . . . . . . . ... . . . . . . . . . . . . . 1 III Methodology / Approach . . . . . . . . . . . . . . . . . . . . . . 1 IV Schedule ........................... 2 Y Protection of Individuals . . . . . . . . . . . . . . . . . . . 3 VI Development and Retention of Supporting Data. . . . . . . . . .
. 4 VII Special NRC Staff Actions . . . . . . . . . . . . . . . . . . . 4 VIII Manpower Accounting . . . . . . . . . . . . . . . . . . . . . . 4
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ATTACHMENTS
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1. Diablo Canyon Allegation Management Staff 2. Format For Diablo Canyon Nuclear Power Plant Allegation and/or Investigations Summary 3. Format for Allegation Action Plan !
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DIABLO CANYON ALLEGATION M%NAGEMENT PROGRAM
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I. Purpose l l The purpose of the Diablo Canyon Allegation Management Program is to: l Provide for a systematic examination and analysis of allegations j and expressions of concern pertaining to design, construction, l operation and management of safety-related structures, systems, and components at the Diablo Canyon Nuclear Power Plant; and to Provide for an assessment of those allegations and concerns that question Diablo Canyon criticality readiness, prior to a Commission consideration of restoration of the license for reactor criticality and low power (less than 5%) testing; and to Provide for an assessment of those allegations and concerns that question plant readiness for power ascension testing and full power operations, prior to a Comission consideration of this issue.
II. Scope The Diablo Canyon Allegation Management Program is intended to encompass all allegation, or expressions of concern which may be construed as allegation, which pertain to design, construction, operation, and manage-ment of safety-related structures, systems, and components, at.Diablo Canyon. In this regard, the Allegation Management Program will also address certain concerns raised by Intervenors and members of Congress (e.g. implications of a 1977 audit by Nuclear Services Corporation of a Diablo Canyon Contractor). The program reqrires that all NRC Offices receiving new Diablo Canyon allegations forward those allegations to the Allegation Management Staff in a timely manner.
III. Methodology / Approach The Diablo Canyon Allegation Management Program requires the combined efforts of NRC offices to identify all allegations upon receipt and coordinate resolution of the allegations with the Diablo Canyon Allegation Management Staff. The Diablo Canyon Allegation Management Staff is a coordination group made up of representatives of Region V, NRR, IE and ELD, as identified in Attachment 1.
The Allegation Management Program requires development, updating and maintenance of a comprehensive listing of allegations related to Diablo Canyon and provides for a coordinated assessment and resolution. The management program is not intended to circumvent any other established NRC management program (e.g. the Allegation Tracking System) or relieve
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any NRC office of its normal organizational and technical responsibili-ties. To the contrary, the Allegation Management Program is intended to complement and coordinate the activities of the NRC Offices as they relate to Diablo Canyon allegations.
The Allegation Management Program includes the following elements:
(1) Development of a comprehensive listing of Diablo Canyon allegations.
The fomat for the listing is provided in Attachment 2. The distri-bution of the actual list is c~ontrolled and limited.
(2) Use of procedures to maintain anonymity snd confidentiality of allegers when requested.
(3) Confimation with the alleger (where possible and appropriate) that the NRC has an accurate understanding of the concern.
(4) Definition of an action plan for each allegation which provides an approach to resolution. The action plan shall include such items as: technical approach to resolution; identification of lead / support responsibilities, resource requirements, schedule for completion; and includes an initial assessment of the potential significance of the allegation. The format for an allegation action plan is provided in Attachment 3.
(5) Responsible Lead Office management review and agreement on the action plan for each allegation.
(6) Implementation of the action plan.
(7) Status reporting and preliminary assessments of allegations significance and programatic implications in time to support Comission consideration of Diablo Canyon licensing actions.
(8) Feedback to the alleger, where possible, for further assurance that evaluations accurately address concerns.
IV. Schedule Actions related to Diablo Canyon allegations shall be accomplished in accordance with the following general schedule: Approval of Allegation Menagement Program and Individual 11/23/83 Action Plans
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12/9/83 Individual item status / preliminary assessment reports 12/9/83 ) Pre-criticality project status assessment report . 12/19/83 Updated individual item status / assessment report (later) Pre-power ascension Project Assessment Report (later) It is anticipated that during the course of this pro;; ram additional allegations will be received and that findings resulting from the review or inspection of current allegations may require expanded exa .inations.
Accordingly, new individual allegation action plans will be d!veloped and existing plans may be expanded. The thoroughness of NRC eval n.tions shall not be influenced by the timing of licensing decisions by the Comission. However, full effort shall be made to provide the Comission with as comprehensive and accurate assessments of allegation significance as is possible.
V. Protection of Individuals - The identii.y of persons providing allegations to the NRC shall.not be disclosed as a matter of practice. In addition, for those individuals expressly reouesting confidentiality all efforts shall be made to protect the individuals identity. This will include: limited and controlled distribution of allegation documentation and correspondence; minimal use of names, identifying titles, or position descriptions in written material; enlarged sampling of activities so that it is not evident to persont ;l outside the NRC that the particular sample selection is related to the alleger; and other indirect approaches toward investigation /inspec-tion of allegations. During the course of NRC reviews or inspections, similar efforts shall be made to protect the identity of licensee or contractor employees who provide information which may be construed as being critical of licensee or contractor activities.
Receiving, clarifying, evaluating, and documenting allegations may involve direct NRC staff contact with allegers and will involve representation of the allegers concerns in NRC documentation and other comunications.
These oral and written comunications shall be professional and objective in nature and shall not reflect a pre-judgemental or defensive posture on the part of the NRC staff. Further, these comunications shall not infer criticism of the allegers' motives, or technical expertise.
- Technical review, inspection, and investigation activities for many of the allegations have been in progress for some time. The 11/28-12/9/83 on-site review refers to the consolidated tean effort to address tne bulk of the
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O VI. Development and Retention of Supporting Data An individual record file shall be developed and maintained for each allegation by the lead NRC office. This file will contain or reference-pertinent documentation associated with the allegation. Resolution of each allegation shall be supported with a clearly auditable record trail.
Accordingly, the record file shall contain or make reference to all significant records relied upon to reach resolution.
Records of personnel interviews shall be developed and retained in those instances where interviews are relied upon by the NRC staff for decisions of technical adequacy. Interview records shall, as a minimum contain: the date, time, location of the interview; the name of the interviewer and interviewee; and a sumary of questions asked and relevant information obtained.
Upon issuance of reports which close out allegations, those records which refe are someappropriately other means (e.g.renced in the NRCpermanent meeting transcripts, reports and are traceable licensee records, through etc.) may be removed from the allegation record file.
VII. Special NRC Actions . As addressed in certain allegation action plans, it may be appropriate to initiate actions which are not routinely required for allegation follow-up. These may include: 1ssuing requests for information to the licensee under 10 CFR 50.54 (f), Confirmation of Action letters or orders; or requests to the Department of Justice to grant imunity for certain individuals. Actions such as these, if required, shall be handled in accordance with established NRC procedures.
VIII Manpower Accounting Offices shall implement established manpower accounting systems to monitor resource expenditures related to this ability.
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, Attachment I l
Diablo Canyon Allegation Management Staff I l
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Organization Name Position Region V T. W. Bishop Staff Leader NRR G. W. Knighton Member IE R. F. Heishman Member ELD J. Lieberman Member
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?RitJCIPAL STAFF UNITED 5TATES
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ENf File _ , 01 / MEMORANDUM FOR: John 8. Martin, Regional Administrator, Region V . Harold R. Denton Director. NRR Richard C. DeYoung. Director. IE FRON: William J. Dircks Executive Director for Operations SUBJECT: REVIEW OF ALLEGATIONS OUTSTANDINE ON DIA8LO CANYON NUCLEAR POWERiPt. ANTI
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At the Coninission, meeting of October 28, 1983, theIn staff was directed to addition, the i pursue all outstanding allegations to resolution.
Coasnission diregted the staff to provide a status report "... addressing l these matters..." prior to the authorization:of criticality and low power testing.
IhavedesignatedJackMartin,RegionY.ashavinglineressnsibilityfor this activity. T. Bishop has been further designated by Jack Martin to ' direct this effort. He is charged with the responsibility tot .
* identify in coordination with NRR,.IE and 01 all outstanding - - allegations; ' *- develop a work plan and schedule in concert with NRR and IE which will direct the effort of offices involved to pursue all allegations to - - conclusion; * provide a coordinated status report to EDO approximately 10 days prior-to Comission consideration of criticality and low power testing by Diablo Canyon; and * produca a final report which will provide' final ' conclusions on all ,
outstanding allegations.
In the way of general guidance. I would point out that it is necessary to This plan should begin development of a work plan on aihigh priority basis.
reflect a scoping and prioritization of allegations so that the required status report will be able to provider to the: maximum extent feasible, reasonable infonnation to the Comission- on which they can make a detennination on Diablo Canyon in regard to criticality and low power licensing. Completion of the evaluation of all allegations must be accomplished, but it is more importanc that they be reviewed correctly than that a particular completion time be met.
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I expect all Offices involved to give full support to Mr. Bishop in this effort. The OEDO should be contacted if problems develop or assitance is needed.
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Willia.d.Di - Executdve Director for Operations cc: 8. Hayesi
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G. Cunninghani R. Minogue T. Bishop, RV
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MFt10RA!400M TOR: Mark Hartman . , Mechanical L:nginuaring Branch . TROM: Robert J. Bosnak, Chief Mechanical Fngineering Oranch
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StmJCCT; ASSIG.NMCNT AS TEAM LEADER: DIADLO CANYON SMALL DORL PIPING REVIFW TEAM -
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You are designated as team leader of the Small Rorc Piping Review Tentn.
Ihn taan will consi.e.t of yourself as team leader. I. Yin, Region III as traut member, and tachnical at.21s tance contract personnel from BNL asi * , team members. .
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You and the other team inembers eshJTd make travel plans to visit the
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site (San 1.u f q Obispo), Region V (Walnut Crcok), PGE/0echtel (San i Francisco) and 10VP of'ic% of R. Cinud (flerkelry) nn an itinerary . t:sti.blished by you to ' nycsligate As necessary, the facts required to
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substantiate the allegat f ans which appear below. You should plan to-depart about January 3,1984 and atteapt to concludo your activities b'y - January 14, 1984.
The. principal snill borc piping allegation is number 55. Other allega-tions which influence small bore piping design and aro, therefo're, assigacd to your team for evaluation in addition to number 55 are the following nunbers: /8, 79, 82, 8b, Bri, 37, 88, 89, 9/ and 95..
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In cariyinq out your assignment, the enti product should be an evaluation of the colle etivo effect uf those. of thu allegations which are found to be substantiated on the performance of the small boro piping of Diablo ranyon Unit 1, presenting conclusions regarding such performance, and * reconnendations for operation hl S% and higher levels of power generation.
The ThYP program results should be reconciled with your conclusions sh:iuld they di f fer.
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WASHINGTON. 0/C,20066
% ..... November 22, 1983 MEMORANQUM FOR: Attached List FROM:. T. W. Bishop, Division Director, Region V SUBJECT:
DIABLO CANYON ALLEGATION' FOLLOW.UP - SITE INSPECTION The steff has been directed by Comission'to provide a status report on all outstanding allegations related to the Diablo Canyon Nuclear Power Plant, prior to authorization of criticality and low power testing. A management plan has been developed to accomplish this directive (draft attached as Enclosure 1).
You have been selected to participate in a special team inspection fomed to review certain of the allegations. Mr. Stewart Ebneter, Region I, has been designated Director of the inspection team.
The initial inspection effort will be conducted from November 2'9, 1983 to December 9,1983 at the Diablo Canyon site, and will be u_nannounced. On December 9,1983 a status report and preliminary assessment of each allegation is required. Further inspection activity may be scheduled at a later date.
The initialteam meeting will be held on Monday November 28, 1983 at 7:00 pm at the Howard Johnson's Motor Lodge, San Luis Obispo, California. Reservations have been made for each of you at Howard Johnson!s (Rate $39/ night).
PleasecontactS.B.Ebneter(FTS488e1283) ort. Bishop (FTS 463-3751)ifyou need further infomation on this effort.
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T. W. Bishop, Director Divisicn of Resident, Reactor Projects, and Engineering Programs, Region V cc: J. Sniezek, EDO Region Administrators RI, RII, RIII, RIV. RV R. DeYoung, IE Enclosures: (1) Draf t Diablo Canyon Allegation Management Program Plan' i
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To: I. Yin Region III s' Front T. W. Bishop Region V-(1 page to follow)
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