ML20133G197

Responds to Re Sixty Minutes 850324 Broadcast Alleging Poor Workmanship & Organized Crime Involvement in Facility Const.Program Identified No safety-related Concerns Not Previously Investigated or Under Investigation

ML20133G197
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	07/25/1985
From:	Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	Lieberman J CONNECTICUT, STATE OF
Shared Package
ML20133G201	List: ML20133G197 ML20133G211
References
CON-#385-059, CON-#385-59 OL, NUDOCS 8508080598
Download: ML20133G197 (4)

Text

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Docket No. 50-322 Joseph I. Lieberman, Attorney General l State of Connecticut 30 Trinity Street Hartford, Connecticut 06106

Dear Mr. Lieberman:

I have been asked to respond to your July 8,1985 letter to Chairman Palladino regarding the transcript of a segment of "60 Minutes" broadcast by CBS on March 24, 1985. As you stated, the main thrust of this program was alleged poor workmanship and organized crime involvement in the construction of the Shoreham Nuclear Power Plant. You suggested in your letter that NRC initiate an investigation into the charges made during the program.

We are well aware of the problems LILCo has experienced during the past 12 years in constructing the Shoreham plant. In some instances, construction work had to be redone several times before it was right. Nonetheless, we have concluded, on the basis of our extensive inspection program that, in spite of these problems, the Shoreham plant meets NRC regulations.

Our analysis of the information presented in the 60 Minutes program is that no issues were identified that were not previously evaluated or under evaluation by the NRC Staff at the time the program was aired. Furthermore, the inspec-tions and evaluations we have completed to date have not revealed any quality or safety-related deficiencies in the Shoreham facility as a result of alleged negligence or criminal activities or any other cause. The NRC has encouraged the identification of potential problems at Shoreham and has aggressively and objectively pursued such concerns when they have been raised. The following paragraphs briefly summarize the staff's efforts with regard to all known al-

' legations regarding Shoreham and in particular the concerns expressed in the 60 Minutes program. Also, the routine inspection program' conducted at the Shoreham facility, which forms a large part of the basis for the NRC's conclu-sion that the facility is built in substantial accordance with commitments and NRC regulations, is described briefly.

To date, over 125 allegations regarding Shoreham have been received and evalu-ated by NRC technical staff and management. The results.of these evaluations are documented in published inspection reports. Of special interest with respect to the 60 Minutes program are our recent contacts with two former Shoreham workers (one of whom, George Henry, was interviewed in the program),

and an extensive investigation conducted at Shoreham during the period December 1979 through March 1980 which relates to issues discussed in the 60 Minutes program.

Mr. George Henry's concerns regarding Shoreham were first brought to NRC's attention in p January 17, 1985 newspaper article published in The Suffolk Times. An inspection at the Shoreham site into Mr. Henry's concerns was 8508080598 850725 PDR ADOCK 05000322 U PDR

Joseph I. Lieberman 2 i

conducted'in late January. The results of that inspection were documented in Inspection-Report No. 50-322/85-10 issued on February 19, 1985 (Enclosure 1).

While Mr. Henry described problems or events that did occur, it was found they had' been identified by LILCo in the normal conduct of their quality assurance program. NRC found that the issues were properly evaluated by LILCo's Startup and Engineering personnel, and received acceptable technical disposition. None of the technical issues described were found to represent a serious operational or design problem. In February, the NRC met with Mr. Henry and discussed his concerns. During those discussions, Mr. Henry provided additional information and raised a few new concerns. Additional inspections have been conducted and will be documented in an inspection report to be issued shortly. When our in-spection program finds problems in the quality of plant construction, or when they are brought to our attention by others, we have not hesitated to require that those problems be fixed by LILCo. The record of NRC actions at Shoreham over the years shows numerous instances where we have found such problems and required corrective actions. This is not uncommon in the construction of complex facilities such as nuclear power plants.

Investigation Report No. 79-24 (Enclosure 2), issued on April 28, 1980, addressed 30 allegations. That investigation encompassed the three-month (410 on-site-hours) effort of three NRC investigators and five inspectors. Areas covered included containment concrete, weld materials, welder qualifications, and intimidation of LILCo inspectors. During that time, public notices were posted by NRC for a period of 70 days. A 24-hour phone number was also pro-vided for points of contact with the NRC, in addition to the onsite interview of Shoreham workers. No evidence was found which could substantiate: (1) the use of defective concrete; (2) the employment of unqualified workers; (3) the supply of inappropriate weld materials; or (4) the intimidation of construction workers.

The NRC staff's routine inspection program directed at verifying an acceptable level of construction quality at Shoreham has been extensive. Over 300 inspec-tion reports and 24,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of inspection time have been devoted toward those ends since a construction permit was issued in 1973. A senior resident inspec-tor was initially assigned to the site in October 1979, and there have been

' four resident inspectors assigned at various times since the inception of the resident program at Shoreham.

! In summary, the 60 Minutes program identified no safety related concerns that were not previously investigated or under investigation by NRC at the time the program was aired. Furthermore, based on extensive inspections conducted to date, no quality or safety related deficiencies resulting from alleged poor workmanship or organized crime involvement in construction activities have been I identified at Shoreham. The NRC's routine inspection program and investigation of allegations at Shoreham have been fully documented in inspection reports.

1

Joseph I. Lieberman 3 1

Per your request, we have included your name on the mailing list to receive NRC Region I correspondence regarding Shoreham from this time on. We request you notify this office if you decide that you no longer want to receive this i correspondence. We will continue to be receptive to, and promptly follow-up

on, all safety allegations made by members of the public with regard to l Shoreham and other reactor facilities.

Sincerely, .

i

/lb <,-

Thomas E. Murley ' -

Regional Administrator

Enclosures:

1. Inspection Report 50-322/85-10

2. Inspection Report 50-322/79-24 s

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s Joseph I. Lieberman 4 Distribution w/encls:

H. Denton

, G. Kerr J. Taylor T. Murley OCA EDO 000829 SECY 85-559 R. Starostecki, Region I H. Kister J. Strosnider J. Berry, Senior Resident Inspector, Shoreham E. Conner R. Fuhrmeister Docket No. 50-322 Public Document Room (POR)

i. Local Public Document Room (LPDR) i Region I Docket Room l Shoreham Hearing Service List m

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A' A I REGION I lNx  ! ui PA:n avtNur Enclosure 1

'%N 0040 C / misc cr P :ussia. rENNSYLVANIA IM06

. Docket No. 50-322 Long Island Lighting Company ATTN: Mr. John D. Leonard, Jr.

Vice President - Nuclear Post Office Box 618 Shoreham Nuc1* ear Power Station Wadin,g;R1ver, New York 11792 Gentlemen

Subject:

Inspection 50-322/85-10 A special inspection was conducted by Mr. E. M. Kelly of this office on January 28-31, 1985, regarding allegations made in a newspaper article by a former LILCo employee relative to the Shoreham facility. The concerns were expressed in a newspaper article published by The Suffolk Times on January 17, 1985. Although our efforts to-date have not identified any problems with the as-built safety systems and hardware, we are continuing our efforts to better understand and define the nature of the concerns mentioned in the article.

There is one ites regarding the Wildwood Substation decontamination catch basin that appears to require further evaluation. We are pursuing this issue and will inform you as to the need for any information from you in this matter. No violations of NRC requirements were found, and a reply to this letter is not required.

Your cooperation with'us'is appreciated.

St- rely, f

W Richard W. Starostecki, Director Division of Reactor Projects

Enclosure:

NRC Region I Inspection 50-322/85-10 cc w/ enc 1:

W. Steiger, Plant Manager J. Smith, Manager, Nuclear Operations Support R. Kubinak, Director, QA, Safety and Compliance E. Youngling, Manager, Nuclear Engineering Edward M. Barrett, Esquire Jeffrey L. Futter, Esquire T. F. Gerecke, Manager, QA Department Shoreham Hearing Service List Public Document Room (PDR)

Local Public Document Room (LPDR) l Nuclear Safety Information Center (NSIC)

NRC Resident Inspector l State of New York 85ReTYhW '

U.S. NUCLEAR REGULATORY COMMISSION Region I t

I Report No. 50-322/85 Docket No. 50-322 License No. NPF-19 l Licensee: Long Island Lighting Corspany 175 East Old Country Road t

Hicksville, New York 11801 Facility: Shoreham Nuclear Power Station Inspection At: Shoreham, New York Inspection Conducted: January 28-31. 1985 Prepared by: els L,-

E. M 4 fly, Project Eng er

• dafe Reviewed by: .

m J. Strosni -

t Chief, Projedts 'Section 1C 0-IT'iI cate Approvec by 'w Harry B. Kister, Chief, Projects Bra ~nen No. I d~89-ff Division of Reactor Projects cate Summary:

A specia', inspection by a region-based project enginee- (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) of allega-tions relatec to the design, inspection and testing of the Shoreham Nuclear Power Station was conducted. The allegations were made by a former LILCo Operational Quality Assurance (00A) insoector (Mr. George Henry) and presented in a Janua n 17, 1985 newspaper article published in Tne Suffolt Times wnich providec the casis for tne inspection.

The newspaper article cited Mr. Henry as expressing essentially eight tecnnical problems:

(1) tne backup TDI diesels are not reliable; (2) lack of recundancy in off-site power circuitry for the plant; (3) a critical check valve has problems which neec te be corrected; (4) a valve in the emergency core cooling system (HPCI) would not operate properly anc it couic result in a serious accident; (5) tnere are defects in fuel rods; (6) a ca :n-::asin at a vehicular decontamination area af ter an accident is not lined with fiberglass; (7) plant staff in an emergency cannot be relied upon for evacuation advice because of oistakes in plotting prevailing wind during a drill; and (8) people (e.g. ,

inspectors quality from LPL standards arehired on by laid off a contract LILCo. basis) who are strict on enforcing

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Summary (Cont'd. )

Based on information cited in the article and on site inspections, interviews and evaluations, there was no new information identified that would affect prior conclusions regarding engineering problems or construction defects.

Items 3, 4 and 5 were found to be accurate descriptions of situations or conditions which did occur, but were properly documented by LILCo and appro-priately dispositioned. Item I was the subject of prior NRC enforcement, special inspections and an agency task force evaluation; specific diesel testing problems in 1982 were the genesis of other NRC actions that were taken to assess both testing and reliability. Item 2 may be the subject of a mis-understanding; there is no single switch through which all offsite power is routed, although the temporary backup diesel generators, but not the gas turbine or the designated emergency diesels, are associated with a single switch. The catch basin in item 6 refers to mitigating features for automobile decontamination after an accident at the plant that could release airborne contaminants off-site in an uncontrolled manner; the detailed provisions and features of the decontamination facility are under evaluation. Item 7 refers to a practice drill held in 1982. Based on our review of the specific concern cited, the problem was identified during a practice drill, and appropriate corrective measures were taken at the time in response to the comments of the drill observers. The specific example regarding the use of contractors in enforcing quality standards in item 8 was pursued based on the available records; pending the availability of more detailed information, the actions taken in the matter of LpL appear reasonable, due to the completion of work, and not indicative of problems with enforcing quality standards.

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l DETAILS

', 1. Principals Contacted D. Crocker, Onsite Emergency Preparedness Coordinator  :

R. Gauthier, Lead Power Engineer (S&W)  :

G. Gisonda, Regulatory Affairs Supervisor J. Kelly, Quality Assurance Divission Manager J. Leonard, Jr., Vice President - Nuclear Operations A. Muller. Quality control Division Manager J. Reilly, Operations Manager (GE)

W. Schiffmacher, Manager - Electric System Operations

. 2. Backcrounc This inspection addresses allegations made by a former LILCo quality con-trol inspector (George W. Henry) in a newspape article written by Karl Grossman of the Suffolk Times on January 21, 1985.

Mr. Henry was a QC inspector assigned to LILCo's Operational Quality Assurance (00A) Section from July 1981 until August 15, 1983. He was certified as a Level II mechanical / electrical inspector in accorcance with ANSI Standard N45.2.6 on July 27, 1982.

His principal duties involved: review of procurement documents (estimated to be 50% of his work activity) material reuipt inspections in the ware-house (15% of time); review and comment upc. selected Station Procedures and logs (15L') and participation in 00A aues ts and witness of startup activities (20%). All of the alleger's work was st.uject to the re' view and approval of a 00A Engineer. None of his assigned duties involvec con-structior (he was assignec to 00A, not Field Quality Control); m: reeve ,

, at the time of his Level II certification, plant cor.struction was ap:rcai-

! mately 95% complete and less than 25% of all plant systems remainec to be turnec over from the LILCo Startup Group to the Operations staff. He observec maintenance, preoperational testing, and repair /rewerk act',vities; performec 00A surveillance; and originated " violations" (as alleged) in the form of LILCo Deficiency Reports (LDRs) for deficiencies found during conduct of his assigned duties.

Tne aseve information was previced by Mr. Hen y's fo-mer superviso , and was Dasec in part or. a peer review of his werL activities which was occu-mentec in an interoffice mem:-andum (J. Rose te A. Muller) dated Novemser 7, 1952. Tnat revie,, inclucec a sameling of 50; repair / rework re:uests and all LDRs originated by OQA during the period January-August 1953.

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3. A11ecations .

', 3.1 TDI Diesel Reliability Mr. Henry indicated in the article that the " backup diesels could not be counted on to function properly" and that he "would not trust them to shut the reactor down". The article indicated that he was present when the diesels were tested in August 1983 and " wrote several viola-tions" which pointed to problems' with the units.

3.1.1 Findings Emergency Diesel Generator (EDG) 1C2 failed on August 12, 1953 due te a fra:turea crank shaft after approximately 1 3/4 hours of testing at the 3900 kW overload rating. The test was to demonstrate. load carrying capability following the replacement of all eight cylinder heads with a newer design due to previously experienced leaks in the cooling water area. Inspection Report 50-322/83-28 documents NR followup and review of that failure.

Mr. Henry was apparently present on the backshif t (3 pt-midnight) and observed diesel testing during the time of EDG 102 crankshaft failure at approximately 5:15 p.m. on August 12, 1983. No record of repair / rework or LDR docu-mentation originated or processed by the alleger, and asse-ciated with the cranksha ft failure, could be found.

However, two LDRs originated by Mr. Henry were found which covered EDG 103 testing performed in June, 1983 (pT 307.004-C-1, test run #6):

LDR-1417, 6/20/83; Turbocharger Lube Oil Pressure Low; closed 11/23/83.

LDR-1424, 6/22/83; Failure to Shutdown fror Control Room; closec 7/26/E3.

Both of these LDRs were reviewed and found to re:eive pro-pe- LILCo Sta-tu: anc ensi nee-ing evaluations an:: 00A approval and closecut.

Anether LDR relatec te the installation of new cylincer hea::s (unce repa'e/ rework request 43-1030) for all three engines was reviewed anc signed by Mr. Henry.

LDR-1545, 8/3/83; Cylinder Head Stuc Fitue; closed 2/18/54

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That LDR was dispositioned " accept-as-is", with justifica-tion provided by an August 3, 1983 letter (McHugh to Rudikoff) from TDI to LILCo which clarified proper cylinder head stud length, thread fit and engagement. The LDR re-ceived appropriate review and approvals, and was closed out by OQA.

3.1.2 Conclusion No evidence was found of nonconforming or deficient diesel conditions, identified in LDRs by Mr. Henry, which were contributing factors in the August 12, 1953 ED3 102 crank-shaft failure.

NRC Region I staff was closely monitoring the diesel pre- ,

operation test prograr and associated mechanical problems.

Escalated enforcement action associated with acceptance of a preoperational test at less than the full-load carrying capability was taken in April 1983. Consequently, the qualification and test program for the diesels was an issue addressed in numerous published reports prior to the crank-shaft failure. A meeting had been held on June 30, 1983 at the Region I office to discuss that status, including cor-rective action for known problems such as: turbocharger bearings, piston modifications, cylinder head replacements, and vibrational problems, A summary of that meeting was issued by Region I on July 27, 1983. Since the crankshaft failure, a Recovery Test Program was developec and imple-mented by LILCo, and an NRC Task Force has studied the acceptability of TDI design. The qualification of these engines as emergency power sources at Shorehar is currently under litigation with an ASLB.

In summary, no new information was identifiec which would be relevant to the evaluation of TDI engines by either the NRC or the appropriate Shorehar ASLB. All of the

" violations" (LDRs) ioentified by Mr. Henry while he was working as a Quality Assurance Inspector at Snoreham have been satisfactorily diseositioned.

i 3.2 Of# site Elect-ical Powe- Redunda cy .

Mr. Henry indicated in the article that "there is only one switch theough which offsite electricity would come". This was based on

" documents mapping the electric grid to Shorenam". This was cited as being indicative of a lack of recundancy or bactup.

3.2.1 References LILCo Drawing No. F-48570-7, September 25, 1982; One Line Diagram, 69 kV Switchyard

6 FSAR Figure 8.2.1-1, Revision 24, December 1981; 7 Main One Line Diagram LILCo Supplemental Motion for Low Power Operating License, filed on March 20, 1984; Affidavit of William G. Schiffmacher; pp 3-19 and Exhibits A through C Technical Specification 3/4.13.1, Alternate AC Sources January 29, 1985 (SNRC-1140) LILCo letter to NRC Board Notification 85-009 cated Februa y 1, 1935.

3.2.2 Findinos The sources of AC power available to Shorehar were de-scribed in detail in LILCo's March 20, 1984 Supplemental Motion for a low power Operating License, in an affidavit by William Schif fmacher, Manager of Electrical Engineering.

Mr. Schiffmacher was contacted and explained to the in-spector the four separate 138 kV circuits and three sepa-rate 69 kV circuits which feed Shoreham, on two separate and independent rights of-way. The reliability of the LILCo offsite distrioution syster is enhanced by at least 8 offsite gas turtines with "bl a cksta rt capability. In addition, a dedicated 20 MW gas turbine has been installed at the Shoreham site (in the 69 kV switchyard) for dead-line, blackstart backue power to the reserve station ser-vice transformer (RSST) via the 69 kV bus. Finally, .a block of four 2.5 W General Mctors EMC mobile diesel generators are directly connectec to the plant's 4.16 kV bus network. This arrangement of offsite power was litigated before the ASLE and founc to be acceptable in the Board's Decision issued on October 29, 1984.

The inspector could identify no single switch through which off site power is supplied. This is physically impossible due tc: (1) seoarate rights-c' way; (2) two different transmission voltages; and (3) incepencent feecs, by the E9 and 138 KV buses, to the normal station service (NSS) anc RSS transformers, from which are fee the plant's 4 kV bus ne work.

Tne inspector walkee down all breaters in the 69 kV switch-yard using Single Line Diagrar F-48570-7. Tne following breake* positions anc functions were verified:

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7 No-mal

. Breaker Tyjte Function Position 640 DE-Ion Grid OCB Isolates Yard Closed Type GO-4-B 623 ABS-D3183 RSST Supply Closed -

Type MO-10 613 ITE Type MD-10 Gas Turbine Supply Closed ABS-W4136 These supply / isolation breakers are associated only with the 69 kV bus, one of 2 separate and independent offsite sources to Shoreham which feeds the RSST. No single switch exists through which both sources of offsite power are routed.

Onsite, proposed by LILCo as a temporary and alternate source of emergency power (should offsite power be lost),

are the four 2.5 MW mobile diesel generators whien directly supply the plant's 4 kV bus network. These are an addi-tional source of emergency power should both the NSS and RSS transformers (ie. offsite power) as well as the three TDI diesels be lost or not available. At power levels up to 5'e, two of the four mobile diesels would be needec, within 30 minutes, to supply power to emergency loads. All four are fed, through a single supply breaker (112), to the 4.16 kV switchgear bus numoer 11. However, this is not an independent offsite source of power -

rather, it is an alternate emergency sou ce for low reactor powe operation (up to 5*. rated) and is bacr.ed up by the 20 MW gas turbine.

On January 25, 1985 the NRC staff aetermined that there dic exist the possibility of a single equipment failure (breaker fault) that could disable both alternate sources of AC power (e.g. , the 20 MW gas turbine and the four 2.5 MW mobile diesel generators). However, this situation does not involve the supply of "off site power"; ir fact, for this scenario it is assumed that all offsite power is lost.

This issue was the subject of Boarc Notification 85-009 dated February 1, 1925 which states that an acceptable resolution (racking out of the subject breaker), meeting the single f ail ur.e proof crite-ion, has beer oeveloped.

3.2.3 Conclusion The inspector could not identify any single switch or

( breaker plied.

through which offsite electrical power was sup-Given the separate 69 LV and 138 kV circuits, fed from various independent transmission facilities, the existence of such a switch is a physical im ssibility, r

8 This is based on consideration of: (1) the 'three 69 kV circuits which feed the Wildwood substation; (2) the four 138 kV circuits which feed the 138 kV Shoreham switchyard; and (3) their respective tie-in to the RSS and NSS' transformers.

Shoreham's offsite electrical power distributi6r network was litigated before an ASLB. A decision in favor of LILCo's testimony was issued by that Board on 0:tober 29, 1984 3.3 Fuel Rod Defects Defects wnien coulc leac to serious conse:uences were aP.eged te exist in the new fuel rods. Mr. Henry statec that he was involved in fuel rod inspections which found problems with "zirconiurr claading, gouges, and improper spaces for water flow". Those problems were alleged to " lead to hot spots ... and a breakdown in the roc itself

... if the rod becomes distorted and is not being cooled".

3.3.1 References SP No. 58.001.01, Revision 7, July 21,1982; Receipt, Inspection and Channeling of Unirradiated Fuel; Procedure Step 8.1.6, New Channel Receipt Appendix.12.1, Fuel Inspection Check List NRC Inspection Report Nos. 50-322:

82-15, Detail 7 (p. 12), issued August 30, 1982 B2-34, Detail 13 (p. 71), issued January 3,1983 83-03, Detail 3 (p. 3), issued February 15, 1983 83-33, Detail 10 (p. 44), issued Nove:ser 20, 19Ba LILCo Field Aucit No. FA-1519; September 3, 1982 LILCe Def t:iency Rep:rt (LDR) Nos. 0771, 0781 tnrough 0817, 0911 and 1588 3.3.2 Findinos initial Fuel Receiot ODA coverage of new fuel receipt and handling was main-

' tained round-the-clock during July-August 1982. Detailed procedural controls, based or GE soecifications, were pro-viced in Sp 58.001.01 anc its A::s cir 12.1 che:Liist. A

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9 total of 26 LDRs were originated by 00A during the initial receipt and channeling of all 560 new fuel bundles. LILCo Field Quality Assurance Division audited (FA-1519) the OQA surveillance of those activities, and found 00A to be in general compliance with applicable procedures and instruc-tions. There has also been NRC Region I inspection cover-age of LILCo's receipt and handling of new fuel with no problems or discrepancies noted.

All LDRs associated with new fuel receipt and initial in-spection and channeling were reviewed during this inspec-tion. None were found to be representative of fuel rod defects, nor were any left uncorrected or in:prope-ly d's-positioneo such tnat fuel rod performance would be affec-ted. Half (13) of the LDRs were written for scratches which were found on the upper and lower tie plates of the fuel channels (not the rods themselves). These were all dispositioned " accept-as-is" by the GE representative (J. Wnitman) present onsite for fuel handling operations, based on the fact that the scratches were in a "non-func-tional" area of the tie plate (ie. not load-bearing or a significant stress area). An example is LDR No. 0785 which detailed a scratch (h" inch long and 1/16" wide) found on the lower tie plate of assembly LJH-874 The assembly was channeled, a hold-tag was attached, and the condition dis-positioned accept-as-is since it was described to exist in a "non-functional" area. Clarification of that disposition was provided by the GE Operations Manager on January 29, 1985, following his verification wi+h GE's Fuel Division in Wilmington, NC.

Five of the LDRs involved channel spacer or fastener camage i which required replacement or return to General Electric:

Dates LDR No. Originate-Closed Condition Discosition

0783 8/13/82 10/21/82 1/64 inch Replace spacer spacer dent and reinspect s

0790 8/18/82 10/8/82 Channel 83464 Returned to GE camage during unused unpacking 079E 8/19/82 10/8/82 Channel Replaced with fastener spare damaged curing installation

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Dates LDR No. Originate-Closed Condition Disoosition 0801 8/20/82 10/12/82 Gouge in top Returned to GE of channel unused greater than

.006 inches 0802 8/20/82 10/8/82 Gouge in Returned to GE bottom of unused channel greater than .004 inches Two instances (LDR Nos. 0789 and 0911) were icantified where channel fastener bolts jammed during installation.

The bolt in one case was broken-off by a GE-representative during removal - that upper tie-plate was replaced, rein-spected and accepted. The other instance was similarly dispositioned. Another LDR (0787) involvec a dent upper tie-rod finger spring which was replaced. LDR-0771 was originated to clarify the serial numbers of metal shipping containers (MSC) which were contained in the outer wooden shipping containers (WSC). There were 2 MSC in each WSC.

This was the only LDR found to be originated by Mr. Henry.

Re-Insoettion for Fretting LDR-1588 was originated on August 19, 1953 and described

" fretting" of 0.0035 inches founc on rod HE at twc spa:er locations on bundle LJH73E. This condition was ioentified af ter a 10*, sample (53 bundles) of assemblies we e inspe:-

te: as a result of a similar probier. experience: with WDP55 Hanford 2 fuel. The condition was disposi t ioned as at eptable based on three letters:

GE to LILCo (DRJ 83-118) dated 9/26/83 LILCo NE Mem: (N Q-83-156) dated 10/27/E3 S. M. Stoller Corp. letter te LILCc datec 11/29/S3  :

Ine condition was evaluated by GE as having n; impact on- l fuei performance (cesign stress margin), anc was believed i: de asso:iated with shipment from tne faorication site to the plant. LILCc commissioned the 5. M. Steller Corp. to pe-form an indepen::ent study of the ex.ert an: effect o' the clac-spacer wear experienced. It was conclu ed tnat a ciac thickness recu: tion of up to 0.006 in:nes in fretted areas wouic result in no significant change in material k

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11 properties and no significant local stress concentration.

Also, the random sam:le was determined to be sufficient, based on a probabilistic analysis performed by GE which used the results of LILCo's sample, such that a full core re-inspection was not done. Therefore, the maximum wear depth found (0.0035 inches) on the Shoreham fuel rods was acceptable for use, and no additional fuel rod de-channel-ing or re-inspections were recommended. LDR-1588 was closed on December 27, 1983.

3.3.3 Conclusion None of the 26 initial _DRs rescived by LILCo 00A described uncorre: ec problems irvolving improper soa:ing or otner conditions which woulc affect cooling water flow through a bundle or impair rod heat transfer. Approximately half of the documented deficien:y reports involved minor tie plate scratches which were a:ceptable as-is and do not affect fuel performance. Five LDRs described gouges, dents or other channel damage (not on the fuel rods) wnich resulted in replacement or return (unused) to General Electric.

The coverage of new fuel receipt and handling by OQA was found to be thorough and well-documented. All 560 bundles were inspected per pro:edures and a statistically based sample inspection for fretting was performed. Observed defects were identified and properly dispositioned in OQA LDRs.

One instance of documerted damage to a fuel rod was found.

"Frettin5" of 0.0035 in:hes was founc at two soots whe e spacees come in conta:: with the zir:oniue ciaoding, and was dispositioned acceptably -base: on GE clad stress analyses and an independent enginee-ing evaluation by S. M. Stoller. -

3.4 Velar Check Valve It was alleged that a critical eneck valve in the hydroger re:ombinc system dic not seat properly, and was sent ba:k to the manuf acturer

( Ve '. a n Corpora-ion) for rework and returnec to LILCo " worse than bef:re". It was ce:ided the "6C.300 dolla " valve woulc n: . De sent ba:s. *c actitional work, and was kept be:ause of "the rush to get l Shorenar into operation".

3 . t. . I References NRC Inspection Repo-t Nos. 50-322:

82-34, Detail 13.3 (p. 71) issued January 3, 19E3 83-33, Detail 2 (p. 10) issued November 20, 19Bc

12 l LDR No. 0781; originated August 24, 1982, closed

,.. April 26, 1983

/ ~

Velan Field Service Repo-t to LILCo (J. Kuhner to l D. Borska) dated September 23, 1982 LILCo Receipt Inspection Reports of 24-inch Swing Check Disc (Serial No. 2205) dated August 6,1982.and April 26, 1983 f

LILCo Purchase Order No. 364883 issued to General Electric Tecnnical Specification 3.4.3.2.c anc Tacle 3.4.3.2-1 LILCo Receipt Inspection Report of 24-inch Swing Check Disc (Serial No. 2E37); inspected August 6, 1982 Repair / Rework Request E11-296, September 23, 1982 Velan Certificate of Compliance for Disc Serial No.

2837 dated August 5, 1982 LILCo Drawing No. NFSK-20B Sheet 2 (Revision 8);

RHR System P&ID 3.4.2 Findings The statement that the -valve in cuestion is part of the hydrogen recombiner system, coule not be corroborated in that: (1) Inspection Report 50-322/83-33 adcressed an allegation related to the only check valves in that systerr (6-inch Velan Serial Nos. 218 and 455), neitner of which were sert to Velan for repair; (2) the original purchase price for each of those valves was 1,525 doila-s; and (3) neither are containment isolatior valves for which leakage or seating is critical.

Du-ing initial containment isclation valve local leak rate testing (LLRT) perfomec during July-Decemser 1982, low pressure coolant injection (LPCI) testable eneck valve ADV-  ;

BIE was disassembled for inspection anc repair / rework. Tne l valve is a 24-inch Velan swing eneck, locatec inside the drywell (inboard isolation valve) where the LPCI injection line ties in to the "B" reactor coolant recirculation 1000.

Valve ADV-81B is normally seatec against reactor coolant pressure during piant operation, anc is' critical in the sense that its. failure / leakage could contribute towards an i " inner-system" loss of coolant atticent in the Reactor Building. The original purchase price of that valve (1976 ,

l

( quote) was 62,280 dollars. l l

l l

g. --- - --

Dise Repair Disc serial number 2208, which had been installed in valve ADV-81B, was sent to General Electric on July 27-31, 1982 for seat repair which consisted of grinding-out and local weld repair of a crack in the stellite seating surface.

General Electric, in turn, sent the valve disc to Velan to complete that repair (re-stellite seat). Non-destructive testing of the completed repair showed the stellite seating surface to be defect free; however, the base metal indicated some minor surface indications which were 4

eliminated by grindouts in the presence of Stone & Webster and GE inspectors on August 2-4, 1982. Disc serial number 2208 was then shipped back tc the Snoreham site.

LILCo 00A Receipt Inspections were performed on August 6, 1982 by Mr. Henry for two 24-inch Velan swing check valve discs (Serial Nos. 2837 and 2208). While disc 2837 was found to be satisfactory (and is presently installed in LPCI check valve ADV-81B), the other disc 220E which had been sent to Velan for repair was rejected. Mr. Henry noted unsatisfactory " physical damage and properties, work-manship, and weld preparations". He also originated LDR-0781 on August 24, 1982, which described the condition as " gouges noted in the base metal and on the side of the stellite seat".

LDR-0781 was reviewed by Stone & Webster Site Engineerins Office (SEO) and dispositioned " accept-as-is" on September 24, 1982. LILCo -Startup concurred in that disocsition on September 28, 1982. Final 00A closecut was datec April 26, 1983, and was based on an April 13, 1983 mem -anoum f rom Startup to OQA (Nicnolas to Muller) which referenced a Velan Field Service Report dated September 23,_1982. The Velan service representative (J. Kuhner) examinec disc 220E I with SEO personnel anc confirmed the twc minor grindouts in the base metal area of the disc "at 6 and 9 o' clock positions". These were stated to be acceptable and, as previously noted by Velan, not within the steilite seating surface. The disposition of LDR-07El statec that:

"The gouges do not affect the structural integrity or the seating surface of the cisc".

Re-inspection was documented by Mr. Henry via Receipt Inspection Report dated April 26, 1953 (with a ncte that it superceded the earlier inspection), and LDR-7E* was closed. .

(

14 a

f Current Valve Status o

LPCI A0V-81B has had at least four LLRTs performed since October 1982. The valve does not have disc 2208 installed; rather, disc serial number 2837 .is installed. Leakage data recorded for penetration X-6B, which is a three valve arrangement that includes A0V-81B (as well as MOV-81B and MOV-378) are as follows:

Date Measured Leakaoe (sefd) i 10/25-26/82 32.84 5/4/83 19.92 1/21/84 6.72 1/23/84 31.92 .

The individual limit administratively imposed on this valve is 115 standard cubic feet per day (sefd), and is based on a set fraction of the total Technical Specification limit for all Type B and C penetrations of 4045 scfd. In addi-tion to the LLRT limit, there is a more restrictive limit placed upon LPCI valve A0V-0818 by plant Technical Specif-ications of 1.0 gpm reactor coolant system. leakage.

The inspector observed disc serial number 2208 on January 30, 1985. The disc is currently stored in the site ware-house, with an " Accept" tag, and is a usable spa *e part.

The two surface grindouts were observed to be net on the stellite seating surface. .

3.4.3 Conclusion l

The alleged critical check valve sent back to Velan for repair was . confirmed to be LDCI testacle check valve E11* '

ADV-81B. The valve's original disc (serial num:>er 2208) was repaired by GE and Velan, anc while initially rejectec during receipt inspection, was later dispositioned as-

. acceptable for use. Disc 220E is currently not installed, and is a cualified spare part sto-ec in the Snorenam ware-house. The two minor surface grincouts are no: on the stellite seating surface, anc are an acceptacie condition.

3.5 HPCI Valve Stroke Time It was alleged that a "below standard" valve in the high pressure  !

coolant injection (HPCI) system did not meet "ooenine anc closinc criteria" wnich could result in a serious acc1 cent. Mr. hen y statec tnat he hac written an inspection repo-t rejecting that valve, anc tha. his supervisor criticizec that report and "dispositioned it with

-s' a date several months hence, meaning the valve was approved in the future".

15 3.5.1 References I --

GE Design Specification Data Sheet for High Pressure Coolant Injection System (Document No. 22A1362AC)

• Requirement 4.5.12, Vacuum Breaker Isolation Valve (MPL No. E41-F079);

Revision 6, June 11, 1974 and Revision 12, January 13, 1984 GE Preoperational Test Specification (Document No.

22A2271AU)

. Acceptance Criterion Bi.E.4.f GE Field Deviation Disposition Request (FDDR) KS 1159, dated January 20, 1984 S&W Specification No. 253, Valve Data Sheets dated August 3, 1977 (pp. 5-21 and 2E)

Engineering and Design Coordination Reports (E&DCR):

F-41799; requested June 30, 1982, approved July 14, 1982 F-41799A; reauested November 1, 1982, approved November 8, 1982 L-0413; recuested Fes-ua ry 14, 1984, approved February 17, 1984 LILCo Inter-Dffice Correspondence (Barnett to Kammeyer) dated September 3C, 19S2 GE Letter tc Stone & Webster (Lebre to Gauthier) datec February 10, 1982 Stone & Webste- Letter to LILCc, LIL-24:09, (Hclder tc Project Engineer) dated Augus: 25, 1953 Shoreham FSAR Sec-ion 6.3.2.2.1, Tatle 6.2.4-1 and Figure 6.2.4-2 Snoreham Techr.ical Specifica:ior Table 3.6.3-1 ASME Section X: Inservice Tes:ing. Valve Su . mary anc Trenc Data Sheets for Syster E41 (HD:I), Valve MOV-049 is

e .

16 Shoreham Station Procedure No. 24.202.03-1, Revision 4, HPCI Valve Operability Test Data Sheet Shoreham preeperationa.1 Test Package PT 202.001-1 (HPCI), C&IO Data Sheets (pp. 114-116.c), Test Except-ion Nos. 11 and 16 LILCo 00A Aucit Finding 82-36-37; issued August 26, 1982, approved September 7, 1982 LILCo Corrective Action Request (CAR) No. 040; orig-inated October 25, 1982, response Novembe- 8, 1982, approvec Ma-cr 27, 1954 OQA Surveillance Plan No. 83-27 (suomitte: 5/28/83) and 83-28 (submitted June 30,1983)

NRC Region I Inspection Report Nos. 50-322:

81-20, Detail 2.a (unresclved item 81-14-02),

issued December 9, 19El 83-05, Detail 2.2.4 (unresolved iter 82-15-01),

issuec March 30, 1983 83-03, Detail 2 (unresclved item 83-03-01),

issued February 15, 1983 80-14, Details 11.b and c (unresc'ved items 80-14-05 and 06), issued October 8, 198:

3.5.2 Findines Functional Recuirements HPCI motor operated valve (MOV)-049 is a 2-inch Velan check valve which serves as an outboard containment isolation valve for the HPCI turbine exhaust vacuum breaker line.

Tne valve nas nc HDCI system functional recuirement -

its only safety relatec function is te isolate containment within 36 seconcs after HPCI is ne longer rec.irec (ie.

drywell pressure g eater than 1.69 osig anc reacter steam come pressure less than 110 psig). Tne valve is normally open during plant operation, and automatically closes upon the above coincider conditions. The 2-inen vacuum creater line, in wnich MOV-049 is situated, connects tne 18-inch HPCI turcine exnaust line di rec .ly with the suppression enameer vapor space. The line prevents water fror the sup-pression poci from being drawn u:: into the turoine exhaust line, anc also prevents concensed steam from remaining in the turbine exhaust line (if HDCI had been secueed) which

. could cause hycraulic damage (if HPCI were re-initiated).

y .

Original Stroke Time Criteria f[

Technical Specification Table 3.6.3-1 lists IE41*MOV-049 as prima ry containment isolation valve with a maximum a

isolation time (closing) of 36 seconds. The valve's actual stroke time has been documented, at various times during the preoperational test program, in a range of from 15.39 seconds (March 16, 1984 inservice testing) to 18.15 seconds (October 27, 1981 C&IO test). The opening time has no cur-rent operational limit imposed since the valve is normally open during operation and has no functional requirement to open within any set time. The valve was originally pro-cured fro- Velan unde- Specification SH1-253, and the asse-ciatec valve cata sneet cetailec the requirec open/close 4

times as: "By ve ndo- ( Mf g . S td. ) 210%" .

The standard applied by LILCo Startup personnel during Checkout and Initial Operation (C&IO) testing, which pre-ceded preoperational HpCI system testing (on at least 3 separate occasions), was the General Ele:tric Design Specification for HPCI which at that time (Revision 6) required the vacuum breaker motor operated isolation valve to:

Open and/or close against a differential pressure of 200 psi at a minimum rate of f 12 inches per minute However, that time recuirement was based on the assumed use of a gate valve, anc was a general inoustry stancard as such. The corresponcing rate for motor-opera ed globe valves is 4 inches per minute, or 15 se:ones per inen of travel.

Since MOV-049 is a globe valve with an actual stroke (or travel) of 3/4-inch, the prescribed closing time used for C&IO testing was less than 11.25 se:onds. The opening time was prescribed with an approximate 20% tolerance on that lirit, or 13+0% secon:s.

Preoperational Test Results C&IC testing of MOV-049 perforce: on 0:tomer 27, 1981, in-dicated that the valve opened in 18.0 se: ends and closed in 15.15 se:cncs. These values we-e reviewe: an: ve-ifiec by LILCo Startup personnel on June 28, 1982, as part of HpCI Presperational Te s- Package PT-202 . 00.' -1, and test

, exception numbers 11 and 16 were written. 1

(

18 i

E&DCR F-41799 was initiated on June 30, 1982 to resolve

/ those preoperational test exceptions, and a solution was approved on July 14, 1982 for LILCo Startup representatives to ... inspect, clean, repack valve, verify proper motor rpm, and retest". If the valve's stroke times were still not satisfactory, then the Stone & Webster Site Engineering Office (SED) was to be contacted.

Current Stroke Time Criteria Following the recommended repairs to MOV-49, the retest on  !

September 22, 1982 found strokes times still in excess of prescribed limits; namely, 16.3 seconds for both open and closing. SEO was contacted by memorendum dated September 30, 1982, requesting disposition of the unsatisfactory stroke times. The Stone & Webster SEO reply dated November .

1, 1982 stated:

In accordance with Boston Engineering, operating time for MOV-49 has been changed to 18 seconds. A revision to ,

FSAR and Tech Specs to reflect this j change is in process.

1 In order to properly resolve the: (1) preoperational test i exceptions; (2) E&DCR-41799; and (3) acceptable valve oper-f ating times, another E&DCR-41799A was initiated on i November 1, 1982, to verify the revised 18.0 second limit for MOV-49. That E&DCR clarifiec the stroke time reouire-ment for MOV-049 as 18 seconds +20% or November 5, 1982 and was approved by the S&W Project Engineer on Novecer 8, 1982, by reference to a previous GE to S&W letter which d

concurred in an acceptable closing time of 18 seconds i

(+20%). A later E& DOR L-0413 revisec tnat time to 18 i seconds +10*.. The August 25, 1983 S&W 1etter, upon which j that E&DCR was based, approved the maximum allowable time  :

currently in Technical Specifications of 36 seconds - l doubling the norinal limit of 15 seconds, a#ter cor.sidera- l tion of HPCI functional recuirements, environmental qualification, radiological and other conditions.

00A Audits Concurrent with LILCo Startup's efforts to resc1ve an acceptable stroke time for MOV-049, 00A performed an Audit of the HPCI o'ecoerational test package. and issued finding 82-36-37 on August 26, 1982. Startup's response to that finding on September 3,1982 referrec to E&DCR F-41799 and the preoperational test exceptions. Tnat response was k

l i I

-..-.-_\

19 approved by OQA on September 7,1982; however, 00A Correc-

/ tive Action Request (CAR) No. 40 was later originated on October 25, 1982 when review of the retest of MOV-049 (approved by Startup on 9/22/82) still showed unsatisfactory closure time.

The Startup response to CAR-040 on November 8, 1982 was two-fold: (1) the CAR should not have been issued since 00A had already approved the proposed corrective action (ie. retest and contact of SEO if still greater than 13 seconds); and, (2) the approval of the 16- second test results was appropriate since the 18-second time had already beer authorized by General Electric's r eo uary 10, 1952 letter to S&W.

Six months later, during an OQA Surveillance Nc. 83-27 (by Mr. Henry) of outstanding CARS, CAR No. 040 was noted on May 28, 1983 as " outstanding. ..not closed" and checked on the Surveillance Plan as unsatisfactory with respect to corrective action. Mr. Henry's supervisor (the Manager of 00A) wrote a caustic note to the alleger on the submitted plan sheet (but unapproved) dispositioned on "11/9/83". which stated that CAR-040 was This was explained to the inspector as an error and was meant to be datec 1982 to i

reflect the Startup response to CAR-040. The ncte further stated that he (Mr. Henry) should "not play games on QA l .

documents" and should rewrite the Surveillance Plan sheet, A copy of that (unapproved.) plan was placed in Mr. Henry's personnel file.

had: The rewritten sheet for Surveillance 83-27 i

the unsatisf actog notations cressed out, anc checks placed in the satisfacto y boxes regarding corrective action.  !

1 asterisks next to the crossed out "unsats", which were explained on the sheet below as "see CAR surveillance 7/83 #E3-28". l final 00A aporoval on July 18, 1953.

i Surveillance Nc. 83-25 was submittec by 00A or June 30, ISE3 anc app-oved by the Manager of 00A on Ju y 15,1983. i 1

It was performed as part of the normal weekly 001 surveill-

{

ance of outstan::ing CARS, anc feliovec up the respor.se to 1 CAR-40 involving a draft Techrical Specification change

( tne r. proposed as 22 seconcs) to properly reflect the approved (15 second) stroke time for E41-MOV-045. . j The final closecut of this entire issue occurred witn the review and approval of CAR-040 on March 27, 1984

20 Independent NRC Inspection Various NRC open items related either directly (or were similar) to the problems experien:ed in defining an accept-able stroke time for MOV-049. All have been satisfactorily resolved, and they include:

Item . Report In Which Closed 81-14-02 81-20 (page 3) 80-14-05 82-13 (page 2)

S2-15-01 E3-05 (page 8) and 84-27 83-03-01 83-23 (page 5)

Also, during .this inspection, LILCo Operations Staff stroked MOV-049 on January 29, 1935, and the inspector in-cependently timed the valve as closing in 14.60 seconds and

opening in 16.85 secones. Further review of IST data, col-i lected on five occasions since March 25, 19E3, shows an average closing time of 15.3 seconds -

that same IST 1

procedure lists a limiting value of 18.7 seconds for closure. The limit was established on October 23,1984 and is based on a mean value of previous stroke times (plus some standard deviation). The valve would be tagged out, and 'HPCI considered as inoperable, if quarterly stroke tests exceed the 18.7 se:ond closure time. i l

Finally, a comparison cf simila valve applicati'ons in the HPCI turbine exnaust va:uum breaker lines at Limerick anc Susquehanna found their Tech Spe: isolation times to be comsarable to Shorehar's, but larger in valve size and different in valve type:

Plant Size Tyoe TS Isciation Limit Limerick 4-in:b Gate 40 secones 1

Susquenanna 3-inch Gate 15 se:oncs 3.5.3 Conclusion There was difficu'ty or cor. fusion on the part of LILCo Startup during preope-a .ional testing of HDCI as to what

- were a::ep able ocen/ciose times for MOV-049. That time was later authorizec oy Stone & Webster and GE to be nominally 18.0 se:cncs (-10'4). The maximus isclation time listec in Technical Spe:'fications is double the nominal s value, or 36 seconcs, anc is ac:eptable since:

v . .

21 MOV-049 serves no HPCI system functional requirement.

Containment leakage past MOV-049 would be contained within the HPCI turbine exhaust piping. {

For a design basis reactor recirculation line rupture, I suppression pool air space pressure is predicted to peak at approximately 35 seconds after the drywell pressure peak, and approximately 20-25 seconds after logoc for MOV-049 initiates its closure.

't The HPCI turbine exhaust pressure is typically ex-pected to be in the range of 10-50 psig; but, the pip-ing is designed to 175 psig, wnich is well above the predicted peak containment pressure of 33.7 psig.

The OQA audit of HPCI preoperational testing correctly identified the discrepant MOV-049 stroke times; however, that same audit, the subsequent CAR-040, and the original Surveillance No. 83-27 all failed to fully recognize the concurrent resolution of the valve's stroke time by LILCo startup. , OQA was originally unaware of the basis

( February 10, 1982 GE letter to S&W) for the approval of the 16-second C&IO test results. However, that basis was explained on November 8, 1982, in the LILCo Startup response to CAR-040. It was that date (not "11/9/83") which the 00A Manager apparently intended to refer to in his caustic note. The CAR was therefore not dispositioned "6 months in advance." In fact, tne approval of E&DCR F-41799A on Novemoer 8,1952, appropriately adcressed the concerns of CAR-040.

In spite of the large documentec effort on the part of both LILCO Startup and OQA to . define an acceptable open/close criterion for valve MOV-049, the nominal and maximum times i of 18 and 36 seconds are reasonable and appropriate. The long hi story of test exceptions, E&D:Rs, memoranda anc l

letters, and tne DQA audit finding and CAR all indicate LILCO's efforts to appropriately disposition this issue.

3.6 Vehicle Decontamination A-sa M. Henry indicated in the article that there is a LILCc map c' an area where there would be " contaminated vehicle parking - where such ver.icles would be hosed oowr., radioactive debris washed off them."

He alieged that the area was "supposec to be lined with a catch basin of ' finerglass", that the catch basin ooes not exist, and that when venicles are washed down, the " radioactivity will go right inte the l s,. ground water".

22

.- 3.6.1 References Shoreham Emergency Plan Implementing Procedures (all Revision 2, July 15, 1984):

EPIP 2-18, Vehicle Monitoring EPIP 2-19, Vehicle Decontamination EPIP 2-20, Offsite Personnel Monitoring /

Decontamination EPIP 2-21, Offsite Decontamination Facilities Activation 3.6.2 Findings Shoreham Emergency Plan Implementing Procedure (EPIP) 2-21 details the activation of the Wildwood Substation Offsite Decontamination Facility Station, if a site evacuation has i been initiated following an accident, but only upon the j direction of the Shoreham Radiological Assessment or Protection Coordinators. The facility (a trailer) is located within the fence surrounding the 69 kV Wildwood Substation and is approximately one mile from the plant site, on LILCo property, and just off of .the South Property

, (access) Road and Route 25A. Cabinets in the trailer con-tain one 1-inch and two 5/8-inch rubber garden hoses.

Attachments 1 anc 2 to EPIP 2-21 depict the location and layout of the oecontamination station / trailer.

EPIP 2-18 ensures that, during a site evacuation, all

, vehicles leaving the site boundary will be monitored in the South Property Road as they leave the site if the decon-tamination facility has been activated. If a vehicle is found to be contaminated te a level 100 cpe above back-ground, ther. it's directed into the area off of the road at the 69 kV substation. Additional surveys are then done, recorded, anc the vehicle and occupant (if contaminated) pre directed to the trailer.

EPIP 2-19 prescribes metnods for vehicle decontarination in steps 5.1.3 through 5.1.5. These steps are progressive measures which include:

First attemots with wet gauce pads, vacuur cleaning ano sweecing.

Careful cleaning ~with a mild detergent or solvent s solution.

l l

23 4

Detergent in hot water with a scrub brush (if contam-ination with 100 cm 8 smears still shows a limit of 500

, dpm or greater).

Parking the vehicle, if still contaminated, at a designated area for "further decontamination efforts at a later date when it is practical".

' The inspector verified that no catch basin or collection system are located near the Wildwood substation.

Subsequent phone conversations with LILCo representatives on February I and 5,1985, indicated:

Their intent tc acc precautionary statements in EPIP 2-19 and 20 which will require personnel to minimize the use or generation of water at the facility and ensure that contaminated. solids are disposed of properly.

Of 12 operating plants surveyed regarding Emergency Plan procecures for vehicle decontamination after an accident: 9 specify no spe:ial collection provisions of liquids; I has provisions for dry materials; I has a decontamination center located 13 miles away with a truck bay that is lined with plastic; and I decontar-inates vehicles onsite in an area which is capable of containing any liquids generated.

The licensee's representatives stated that no commitment or J olan for a fiberglass caten basin was ever made, and that no such basin currently exists.

3.6.3 Conclusion EPIP 2-18 through 21 de not call for " hosing-down" of venicles which are con aminatec, and no cat:n basin for contaminated liquids is requirec by Shorehan's Emergency Preparedness Program. A survey of Emergency Plan post-ac:icent vehicle de:ontaminatior measures at operating nuclear stations indi:ates that Snorehar's Off site Decon- I tamination incustry practice.

Facility is typical with respe:t to accepted i Wnile no NRC regulatory requiremert or documented recommen-datier for

providing a catch bas'n exists, the acequa

y of Shoreham's Emergency Preparedness Program (including the measu es for containing liquid cer . amination) is currently uncer evaluation by NRC. Also, the issue of offsite emer .

gency preparedness at Shoreham is currently being litigated by an Atomic Safety and Licensing Board.

24 3.7 Emeroency Drill Plume Plotting During his participation in an " emergency drill", in which plant radiochemistry and health physics personnel were monitoring prevail-ing winds and predicting the heading of the (ficticious) radioactive plume, Mr. Henry stated that an error was made upon which evacuation recommendations were being made. Because plant personnel were

" plotting the plume totally wrong", he stated that they were " calling for an evacuation of the Nortn Shore of western Suffolk County".

This exercise recommendation, which should have been " premised on prevailing winds from the West" was "off-by 180 degrees".

3.7.1 References f

Scenario dated July 2,1982 for Emergency Preparedness (EP) Exercise No. Il conducted on July 7,1982 Critique of July 7,1982 Exercise No.11 NRC Region Inspection Report (EP Appraisal Followup) 50-322/83-37, issued February 6,1984 3.7.2 Findings s

LILCo EP Drill No. 11 of the Shorehae Onsite Emergency Preparedness Plan was concucted on July 7, 1982. The exercise was intended to activate the control roor., tech-nical support center (TSC) and opera 1! ions support center.

' This exercise was also the first to use a pre planned and prepared scenario. LILCo 's initial training session  ;

i involving emergency prepareaness was conducted on May 23,  ;

1982. As of January 31, 1925, 109 training sessions have  ;

been held. '

LILCo personnel identified four sessions in which Mr. Henry was listed as a' participant. During the July 7, 1982 drill, he was an 00A representative assigned as a "TSC Coordinator". His duties involvec being a phone person whc was provided information adces as tne scena*ic developec which be relayee to the TSC.

At a time approximately 215 hours0.00249 days <br />0.0597 hours <br />3.554894e-4 weeks <br />8.18075e-5 months <br /> into the . exercise, there was confusion associatec with wind direction, plume heading, and evacuation recommendations. This was later noted during the criticue of the drill by its members (on 4 the same day of the crill) as "no announcement in TSC of wind shift during drill ... crucial to TSC dispatcher ...

resulting in celay of survey teams." The scenario cover page, uncer " Meteorological Data", listed the following winc directions:

s

25 t ..

l l

Time Wind

.f 0:00-2:30 From 315 degrees elevated 2:30-6:30 From 215 degrees elevated The inspector was informed that this printed wind shift 2k, hours into the exercise (from 215 or southwest) was a typo-graphical error. The readings at selected detector i

locations, which were listed on subsequent ~ pages of the scenario for the last four hours of the exercise, correctly reflected the intended wind shift as from the 35 degree (northeast) direction.

The intended wind conditions in that exercise were 10 mph from the northwest for the first 2k, hours (stability Class C), shifted to 6 mph from the northeast for the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (stability Class E) - an 80 degree shift. The confu-sion as to plume heading and evacuation direction was fur-ther compounded by the comsass headings (marked in oegrees) on the posted sector map in the TSC. Instead of the normal compass markings (eg. 90-degrees being due East), the map

, was rotated by 180 degrees, such that the West sector was marked 90-degrees. This was intended for the simplicity of correlating wind direction to plume heading. Stated diffe-rently, when a meteorological condition of 90-degree (or

" easterly") winds was given', the drill personnel could then easily locate the plume passage on the map at the location marked 90-degrees (actually a west sector). This map has since been changed to indicate true compass headings (ie.

90-oegrees is the eastern sector).

3.7.3 Conclusion There was confusion at the midpoint (wind shift) of the

, July 7,1982 exercise, as alleged. Wind direction and the associated dispatch of survey teams and evacuation recom-mendations were in error for a shon time. LILCo personnel responsible' for the drill notec that error, and discussed it during the subsecuent crill critique. Similar errors have not been experienced with the apcroximate 100 Eme gency Plan. training exercises that have since been con:: acte::.

The error in that exercise is attributed to:

A typographical error-in the scenario.

Compass / degree markings on the TSC map.

. N.

26 l

,- Those map markings hav'e been corrected, and subseq0ent training exercises have not had similar problems > NRC inspections have observed and reviewed Emergency Prepared-ness Program training at Shoreham and no significant prob-  :

lems have been identified. t 3.8 LPL Insoectors Mr. Henry apparently stated that quality control inspectors from Lpl Technical Services of Great Neck (New York), contracted by LILCo, were " routinely laid off for being strict on standards".

3.E.1 Findines Personnel files were reviewed, in the presence of the LILCo Quality Assurance and Control Division Managers, for all

employees provided by LPL Technical Service, Inc. A total of 19 inspectors, all with former ANSI N45.2.6 Level II certification from at least one other nuclear facility, were considered for assignment as OQA inspectors. Twelve 4

of those 19 individuals were eventually certified seven were never certified, for various reasons. Certification was governed by LILCo procedure QAP-5-2.3, " Certification of Inspection Personnel".

LPL contractors were present at Shoreham as OQA inspectors during the 2 year period of April 1982 - April 1984 Du--

ing that time, construction progressed from approximately 95% to full completion, and the remaining 25%~ of precoeea-i tional testing was finished. Also, the majority of activ-ity during that period focused upon the TDI generators' preoperational anc recovery test diesel programs.

The original contract with LPL Technical Service, Inc. was -

a lump-sum bid, billed on time / material and originally t,

based on 6 individuals for 18 months. LILCo's Manager of Quality Control Division provided the following data on the complement of DQA inspection activity by LPL inspectors:

Their assignment was to handle peak work loads asse-ciated with preoperational and C&IO testing, and the TDI diesels.

The maximum number of LPL personnel, which suople-mentee the 8 00A personnel who were full-time LILCo employees, was six at any one time.

A large attrition rate was experienced with LPL per-sonnel, such that the average residence time of any i

.\ one LPL contractor was estimated to be 2 months.

T 6

- - - - - - . - , - , . -----,--,w-,-- ---r-w,ny,-- - , . - - - - - - - , . , - - - - . , -

27 Of the total 12 LPL employees who were certified as i Level II inspectors at Shoreham, three.were terminated by LILCD - the remaining 9 left on their own accord.

The three who were terminated by LILCo were apparently let go because their contract expired, and LILCo indi-cated that they had no further need of their services.

The last LPL contract inspector assigned to LILCo 00A, left in April 1984 3.8.2 Conclusion Based on a review of the LPL staffing histcry no evidence was found to seppert the aliegation that LPL inspe: tors were routinely laid off for being strict on standards.

There were a total of 12 employees from LPL, over a 2 year period, assigned to LILCo 00A. Only 3 were stated to have not left on their own accord; apparently their services were not needed and their contract had expired. There has not been an LPL contract inspector at Shoreham since April 1984 3.9 Exit Interview The inspector discussed the preliminary findings of this inspection

' with licensee personnel on January 30, 1985. Phone conversations have been held with other LILCo personnel, during the period January 31 - February 7,1985, to clarify details of this inspection. The LILCo Vice President - Nuclear Operations, J. Leonard, Jr. , was also apprised of preliminary inspection findings in a phone conve-sation hele on January 31, 1985.

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6 U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION I Report No. 50-322/79-24 Docket No. 50-322 License No. CPPR-95 Priority --

Category 8 Licensee: Long Island Lighting Company (LILCO) 175 East Old Country Road Hicksville, New York 11801 Facility Name: Shoreham Huclear Power Station, Unit 1 Investigation,At: Shoreham, New York Investigation Conducted: December 11, 1979 through March 19, 1980 Investigators: ^ L 0. JAl> __

V-il- So C. O. 11ina,' h.D., Investigation Specialist date aMw L. M. Narrow, Reactor Inspector Wiv/hdate ,

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1 r,. /sk L WWT p.P.4urr,ReactorInspector '

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s S. D. Reynold%, Jr., ReacD r Inspector Y

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date b

. C. Higgijf5, S . R sident Reactor Inspector '

/N at.e 4] 14 #o

a. + M H. Nicholas, ctor Inspector I date R. K. Christopn s ' ww1M - dn_ 9-//-8d

' Investigation Specialist date 1& % _s W 0~ ll'V

                                      . 'P. Remaklus, Ihvestigation Specialist                       date i

Reviewed By: ') / -td  % 9 '13-- O k R. W. McGaughy/C&ef, Project Section date l RC&ES Branch l Region _L_Eam_143__ 77,, 1 TRev. October 1977) % Q - g g I

                                      -                                ,v .
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g k e (  ! Investigation Sumary: Investigation from December 11, 1979 through March 19, 1980 (Investigation Report Number 50-322/79-24) Areas Investigated: The investigation covered thirty (30) allegations related to construction irregularities at the Shoreham site. The allegations were made via court testimony, personal interviews, magazine articles and alleged phone calls by third parties to one of the known allegers. The investigation involved 410 man-hours on-site by three (3) NRC investigators and (5) NRC in-spectors. , Results: Of the thirty (30) allegations investigated, none were found to be substantiated. During the course of the investigation, two (2) items of noncompliance were identified: (1. Infraction - failure to identify nonconfor-mance, Paragraph D.2; 2. Infraction - improper weld rod requisition forms, Paragraph D.10)

            /

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• TABLE OF CONTENTS I. BACKGROUND A. Reason for Investigation B. Identification of Involved Organizations II. SUPMARY OF FINDINGS A. Allegations and Investigation Findings B. Items of Noncompliance C. Management Meeting III. DETAILS A. Introduction B. Scope of Investigation C.

Persons Directly Interviewed and/or Contacted During the NRC Investi-gation D. NRC Investigation Findings and Conclusions Related to Allegations IV. EXHIBITS A. Referenced Codes, Specifications and Procedures B. Notice of Investigation s 3

m I. BACKGROUND A. Reason for Investication The Nuclear Regulatory Commission (NRC), Office of Inspection and Enforce-ment, Region I, was first informed of potential construction irregularities at the Shoreham Nuclear Power Station (SNPS) on December 1,1979 when the NRC's Resident Inspector at the Shoreham site was contacted by Mr. Leighton Chong, one of the defense attorney's for an individual charged with trespassing at the Shoreham site during an anti-nuclear demonstration in June of 1979. During the week of December 3 through 10, 1979 additional information concerning these irregularities was presented by Mr. Chong and Mr. John Hall, a local independent TV producer. In addition, testimony presented during the above referenced trial on December 6,1979 and local newspaper coverage thereof on December 7,1979 was also made available by the NRC's Public Affairs Office. Based on the information received, the NRC initiated an investigation into alleged construction irregularities at the Shoreham site on December 11, 1979. B. Identification of Involved Organizations

1. LONG ISLAND LIGHTING COMPANY (LILCO) 175 East Old Country Road Hicksville, New York 11801 An electric utility licensed by the NRC to construct a nuclear power plant under NRC Construction Permit No. CPPR-95. (Docket Number 50-322) 2, STONE AND WEBSTER ENGINEERING COMPANY (S&W) 245 Summer Street P. O. Box 2325 Boston, Massachusetts 02107 l A company contracted by the licensee to perform various construction sanageeant activities at the Shoreham site. i

3. GENERAL ELECTRIC COMPANY (GE)  !

175 Curtner Avenue San Jose, California 95125 A company contracted by the licensee to provide the nuclear steam supply system and related components at the Shorehan site.

4. GENERAL ELECTRIC COMPANY, I&SE DIVISION (GE) 777 West Putnam Avenue P. O. Box 6850 Greenwich, Connecticut 06830 4

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l l A company contracted by the licensee to provide the turbine generator and related components at the Shoreham site.

5. DRAVO CORPORATION (DRAVO) l Neville Island Pittsburgh, Pennsylvania 15225 A company contracted by the licensee to perform various construction activities at the Shoreham site.

6. COURTER & COMPANY (C00RTER) 317 West 13th Street New York, New York 10014 A company contracted by the licensee to perform various construction activites at the Shoreham site.

7. L. K. COMSTOCK & CO. INC. (COMSTOCK/ JACKSON) 155 East 44th Street New York, New York 10017 A company contracted by the licensee to perform various construction activities at the Shoreham site.

8. REACTOR CONTROLS, INC. (RCI) 1245 South Winchester Boulevard San Jose, California 95128 i

A company contracted by the licensee to perform various construction activities at the Shoreham site.

9. PROTECTIVE SPRAY PLASTICS, INC. (PSP) 1130 Crose Avenue New York, New York 10472 A company contracted by the licensee to perform various construction activities at the Shorehan site.

10. REGOR CONSTRUCTION CO., INC. (REGOR)

P. O. Box F East Northport, New York 11731 A company contracted by the licensee to perform various construction activities at the Shoreham. site.

11. JOHN GRACE & COMPANY (GRACE) 34 Washington Parkway Hicksville, New York 11801

(. - 5 l

A company contracted by the licensee to perform various construction activities at the Shoreham site.

12. C. P. BENNETT/F&G CO., INC. (C. P. BENNETT) 231 Russel Street Brooklyn, New York 11222 A company contracted by the licensee to perform various construction activities at the Shoreham site.

13. KTA-TATOR ASSOCIATES (KTA-TATOR) 3020 Montour Street Coraopolis, Pennsylvania 15108 A company contracted by the licensee to perform selected QA/QC activities at the Shoreham site.

l l l 1 l l l l  % 6

II. SLM4ARY OF FINDINGS A. Allegations and Investication Findings This investigation involves allegations that were introduced tiy various methods by several individuals. The investigation was initiated on December 11, 1979 as a result of the testimony of two individuals testi-fying on behalf of the defendant in a trial related to trespassing charges incurred at the Shoreham site in June of 1979. On December 12, 1979 an allegation was received in connection with the Shoreham site from an unidentified alleger through a local shopkeeper. This allegation is described and numbered 1 below. On December 17, 1979, NRC investigators met with Witness A in the above referenced trial. Witness A's allegations are described and numbered 2 through 9 below. On December 17, 1979 NRC investigators met with Witness B in the above referenced trial. Witness B's allegations are described and numbered 10 through 18 below. On December 17, 1979, defense attorneys in above referenced trial presented NRC investigators with allegations from other witnesses who were not called to testify and who wished to remain anonymous. These allegations are described and numbered 19 through 21 below. On December 17, 1979, the defense attorneys also presented NRC investigators with allegations from a former boilersaker at the site. These allegations are described and numbered 22 through 25 below". On January 9,1980, NRC investigators met again with Witness A who presented three (3) additional allegations allegedly received by anonymous phone calls. These allegations are described and numbered 26 through 28 below. On February 26, 1980, NRC , investigators met again with Witness B at his request at which time another allegation was introduced. This allegation is described and numbered 29 below. Throughout the initial investigation, defense attorneys reported that pressure was being applied by the licensee (LILCO) and/or related unions in order to prevent workers from coming forth to the NRC with information. This matter was covered as a separate allegation and numbered 30 below.

                  " Knowledge of the existence of these allegations was made known to the NRC Resident Inspector at the Shoreham site on or about December 12, 1979 and appeared, in part, in an article published in Seven Days, Volume III, No.12,
 ,(                dated October 26, 1979.

7 L .

          . ~   .

Allegations NOTE: The allegations listed below have been sumearized for clarity. The actual allegation: are cited in detail in Section III of this investi-gation report. 1.) Inspection of the N-11 steam lines revealed cracks which may require that the entire system be replaced. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.1) 2.) Following concrete placements for the Reactor Pedestal and Reactor Building Primary Containment Wall, heaters were not used as required for curing during the winter months of 1973-1974. Similar conditions were allowed to occur in the Radwaste Building during the winter months of 1974-75 and 1975-76. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.2) 3.) Following concrete placements for the Reactor Pedestal, Primary Containment Wall, and Radwaste Building, forms were improperly stripped on the day following concrete placement instead of the required seven (7) days. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.3) 4.) Foliowing the stripping of concrete forms for the Reactor Pedestal and. Primary Containment Wall, large cracks, honeycombing deep enough to expose the rebar and through-cracks were patchec over with mortar prior to inspection by QC. ' The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.4) 5.) Cadwelds for rebar in the concrete located in the Reactor Building were in some cases found to be loose with concrete poured over the loose cadwelds. The NRC investigation found no c.Vidence and/or information to substantiate this allegation. (Details, Paragraph D.5) 6.) Rubber waterstops between concrete layers in the Radwaste Building were not installed properly and sometimes omitted entirely. The NitC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.6) - (

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8

7.) A carpenter was permitted to weld studs to embedment plates used for pipe supports in the North wall of the Radwaste Building even though he had failed the welding test seven (7) times. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.7) 8.) Threaded tie rods without sleeves used as fore ties for the Reactor Pedestal were pulled from the concrete after it had set, leaving a void in the wall. In some cases, large amounts of concrete were pulled from the pedestal wall in this manner. The NRC investigation found no evidence and/or infomation to substantiate this allegation. (Details, Paragraph D.8) 9.) Several through cracks were observed in the Turbine Building wall that separate it from the Reactor Building. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.9) 10.) During the first three weeks of February 1979, several dissimilar metal welds were made with ER-308 and/or ER-316 weld rod instead of the required I ER-309 weld rod because welders claimed that it was too cold to return to their foreman and have incorrect weld rod requisitions changed. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.20) 11.) In a covert attempt to use substandard materials, E-6018 electrodes were ' used rather than the required E-7018 electrodes. The NRC investigation found no evidence and/or infonsation to substantiate this allegation. (Details, Paragraph D.11) 12.) A large scale repair on the feedwater condenser jacket was performed by a Regor boilermaker instead of the usual Courter and Company steamfitter in order to avoid having the crack reported to Courter QA personnel which would have raised the issue as to the integrity of the entire jacket. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.12) 13.) Due to the imoroper estimation of the depth of the water table by LILCO, salt water is seeping through the Secondary Containment wall at the 8 foot level and around-the-clock efforts are being undertaken to pump the water out. 9

The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.13) 14.) Stone and Webster lost its general contractor duties when it repeatedly complained to LILCD about the incompetence and corrupt practices of its contractors, such as Regor and Courter, which LILCO insisted on asing. l The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.14) 15.) In addition to seven (7) major failures of hydroflushes of the primary closed loop piping system, a gross failure on or about June 15, 1979 caused valves to pop and a section of pipe to be ejected 50 feet into the air. It was also alleged that the hydrostatic test of the system in September of 1979 could not have been valid since it occurred too soon after the gross failure in June 1979 to have permitted proper shutdown and repair. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.15) 16.) The outfall pipes for the circulating water systems have never been properly anchored and due to the tidal action in Long Island Sound, have shifted, broken and separated from the line itself. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.16) i ' i

      ,       17.) MDE technicians were not adequately qualified for the jobs they were performing.

The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.17) 18.) Large quantities of green dye used for dye penetrant testing were being discharged by LILCO without proper approval and are polluting Wading River shellfish. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.18) 19.) Supervision of trade workers is inadequate and being performed by unquali-fied individuals. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.19) 20.) Qualification and training of subcontractor personnel at the Shoreham site is inadequate. 10 -

The NRC investigation found no evidence and/or information to substantiate j

this allegation. (Details, Paragraph D.20) '

21.) Workers painting the inside of the Reactor Primary Containment were not qualified and when discovered by the NRC, most workers were layed off but

          -            the substandard work was not inspected and allowed to remain. It was further alleged that the remaining workers completed the jobs and on one occasion worked a 30-hour shift on methedrine.

The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.21) 22.) Tube support sheets in the condenser box were so aisaligned that titanium tubing which thould fit loosely, was hammered into place resulting in damage severe enough to cause a tube to break with the possibility of a ' radioactive spill. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.22) 23.) Radiographic tests revealed that the longitudinal seas welds for the con-denser box were improperly done and when opened for rework were found to contain dirt, rubbish and weld rod stubs. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.23) 24.) Misalignment of the condenser tube support sheets required re welding so often that in some cases the " mother material" around the weld had to be cut out and replaced with a fresh substitute section.  ! The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.24) 25.) Welds to be inspected are pre-marked by QA in order that the best welders can be assigned to these' jobs while other welds are made by lesser quali-fied welders and never inspected. This resulted in a degradation of the ) overall quality and resultant safety factors at the Shoreham site. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.25) 26.) Turbine foundation bolts were installed so far out of alignment that it was necessary to chop out the concrete around the bolts, heat the bolts and bend them into a "Z" shape in order to fit them to the foundation plates.

      \

11

t The NRC investigation found no evidence and/or information to substantiate

             ,                this allegation. (Details, Paragraph D.25) 27.) When concrete was placed in a cold joint on the 63' level of the Reactor Primary Containment, a large amount of rubbish and trash was permitted to remain within the form and the concrete placed on top of it.

The NRC investigation found no evidence and/or information to substantiate j this allegation. (Details, Paragraph D.27) 28.) Soil percolation testing results were falsified and test results withheld in LILC0's submission of this information to the NRC. The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.28) 29.) Welder performance qualification records were postdated for welders who qualified after performing welds for which they had not been qualified. 4 The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.29) 30.) Pressure was applied to construction workers by LILCO, its subcontractors and/or related construction unions in order to prevent and/or discourage workers from coming forth to identify construction defects and/or irreg-ularities to the NRC. - l The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.30) B. Items of Noncomoliance During the course of the investigation, two (2) items of noncompliance were identified related to concrete records and dissimilar metal welds. Item No. I was corrected prior to the completion of the investigation.

1. ) (79-24-01) Contrary to Criterion XVI of Appendix B of 10 CFR 50, the licensee failed to identify the nonconfomance of Concrete Placement RS-4-12 with respect to curing requirements and consequently failed to take appropriate corrective actions at the time. (Details, Paragraph D.2) 2.) (79-24-02) Contrary to Criterion V of Appendix B of 10 CFR 50, the licensee issued two (2) Weld Naterial Requisitions which indicated ER-308 weld material instead of the required ER-309. (Details, Paragraph D.10)

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i

   . .                                            s C. Management Meeting A management meeting was held on March 19, 1980 with licensee representa-tives at the conclusion of the investigation in order to discuss the NRC's investigation findings. The following individuals were in attendance.

L. Narrow, Reactor Inspector (NRC) J. C. Higgins, Sr. Resident Inspector (NRC)  ! J. P. Movarro, Project Manager (LILCO) W. J. Museler, Assistant Project Manager (LILCO)

                   T. F. Gerecke, QA Manager (LILCO)

W. Hunt, Project Engineer for Construction (LILCO) J. M. Kelly, FQA Manager (LILCO) T. Arrington, FQC Superintendent (S&W) A. F. Earley, Attorney (Hunton and Williams) i l , 13

III. DETAILS A. Introduction This investigation was initiated as a result of the NRC-Region I being informed of various alleged construction irregularities at the site of the Shoreham Nuclear Power Station (SNPS). The exact number of allegers could not be determined due to the manner in which the allegations were l ' received, i.e. , court testimony, personal interviews, magazine articles and alleged phone calls by third parties to one of the known allegers. The primary initiating event of the investigation was the testimony provided on December 6, 1979 by two (2) former construction workers at the Shoreham site who were defense witnesses for an individual charged 3 with trespassing during an anti nuclear rally at the Shoreham site in June of 1979. B. Scope of Investication This investigation included an examination of pertinent documents and records at the Shoreham site and at the NRC Regional Office; interviews and contacts with several licensee and contractor personnel (present and former employees); observations by the investigators as well as independent ^ testing as applicable by the NRC through an outside testing laboratory (The Franklin Research Center). C. Persons Directly Interviewed and/or Contacted Durina the NRC Investigation

During the course of this investigation, representatives of various i subcontractors at the Shoreham site were contacted in order to set up interviews with their employees. The subcontractors contacted have been delineated in Section I-B of this report. The principal licensee repre-sentatives were identified in Section II-C of this report.

Several licensee and subcontractor personnel, present and former workers, at the Shorehaa site were interviewed by the NRC in investigating the i allegations contained herein. In order to protect the identity of those individuals, alphabetic designations have been assigned to individuals noted within the context of the report only when such designations are required in order to differentiate between the statements of one or more of the individuals. Otherwise, descriptive designations (e.g. QC inspector, welder, etc.) are used. Throughout the investigation, sources of information were not identified by name to persons being interviewed unless (a) such person was identified by another independent document or person other than the source or (b) the person being interviewed independently acknowledged the identity of the source to the NRC. All individuals interviewed were notified of the voluntary nature of the interview, the right to have another person of their choice present during the interview, and the confidentiality provi-sions of this investigation. l 14

P D. NRC Investigation Findings and Conclusions Related to the Allegations

1. Allegation No. I

a. Allegation A random visual inspection of the Main Steam Line System (N-11) revealed cracks which had to be repaired. No further inspections of the N-11 System have been undertaken, and it is alleged that there are many other cracks in the system. It was further alleged that the entire N-11 System may be unfit and may have to be repaired or replaced.

b. NRC Investigation Findings

NRC investigators identified the~ extent of the N-11 System referred to in the allegation by a review of all of the N-11

System isometric drawings,_ including a review of the history of

the piping from the steel mill to site erection. QA documents on pipe irregularities were also reviewed as well as an inspection of the piping itself.

The subject piping is that piping which carries main steam from the isolation valve outside of Primary Containment to the Main . Steam Stop Valves for the Turbine. This pipe is 24" 0.D., 900 psig rated SA-106, Grade B seamless carbon steel material , obtained for spool piece fabrication by DRAVO from U. S. Steel. The piping system was designed by S&W to ASNE Section ! III, Class 2 requirements. The site welding was performed by i COURTER with inspections by COURTER and S&W surveillance in-spection groups. The NRC noted that manufacturing linear surface indications are

                                 ,to be expected in pipe of this large a diameter and these indications were acknowledged along with acceptance and repair methods in Paragraph 20 of the material specifications for the pipe. ASME NC-2550, 2551 and 2558 also indicate the acceptance and repair methods for surface indications in this type of pipe. NDE surface examination is not required by the code.

The NRC noted that the 0.D. of the pipe was visually examined during site fabrication, prior to the installation of required

insulation waterial. A total of fifty eight (58) COURTER Deficiency Correction Orders (DCOs) were written describing the surface irregularities as linear indications. These documents were included into COURTER Nonconformance Reports (NRs) NR-466 and NR-466A. The disposition and correction of the problem was i

also documented in S&W Engineering and Design Coordination Report (E&DCR) F-18716 and F-22478. The NRC determined that the disposition was in accordance with SA-106 and NC-2250 code { 15

1

  .-                                                                                                                         j
                       .                                                                                                     l requirements and included grinding to remove the vast majority                          1 of the surface indications and repair welding of some defects shown by ultrasonic testing (UT) to encroach on the minimum wall dimensions.                                                                        j NRC investigators interviewed the cognizant engineer who had provided the disposition of E&DCR F-18716 and who was knowledge-able of E&DCR F-22478. The NRC also interviewed one of the i,                                    COURTER welders who worked on the repair of the linear indications.

Both individuals independently stated that the linear indications identified were not cracks, but rather those types of metal forming irregularities acknowledged in the material specification. The S&W engineer indicated that the steam lines were completely examined by visual methods and all linear indications were dispositioned without any difficulties being encountered. An NRC investigator visually examined all N-11 piping not covered by insulation and found no signs of " linear indications" but a large number of ground out areas as expected from the DCOs. One complete steam line had been previously inspected by the NRC and this inspection was documented in NRC Inspection

Report No. 50-322/78-03. No items of noncompliance were identi-4 fied in this area.

t ! The NRC determined that the " cracks" in the main steam line as reported in the allegation were in all prot ability  ! a misinterpre-tation of the nomally occurring seams 'and laps found in material of this kind and were not cracks, per se. The disposition of these visually observed linear indications more than satisfied the minimum code requirements and no other deficiencies were observed by the NRC in the N-11 System. j c. NRC Conclusion l The NRC found no evidence and/or infomation to substantiate i i this allegation. '

2. Allegation No. 2

                                                                                                                             )

a. Allegation Following concrete placements for the Reactor Pedestal and Reactor Building Primary Containment Wall during the winter months of 1973-1974, heaters used to maintain the required curing temperatures were either not used, or when used were

( permitted to go out for extended periods of time during the night shifts. It was also alleged that ice was allowed to fore on the concrete, a condition noticed when workers arrived in the morning. It was also alleged that the same conditions were allowed to occur in the Radwaste Building during the winter months of 1974-75 and 1975-76. . 16

l l .

                .                    b.        NRC Investigation Findings The NRC investigators reviewed applicable sections of S&W specifications ard QC procedures, including American Concrete Insititute (ACI) standards referenced in these specifications.

These, as well as other selected specifications, standards and procedures reviewed during this investigation are referenced as Exhibit A of this investigation report. The documents reviewed, established as a requirement for the

;                                              winter curing of concrete that the temperature be maintained at

! 40* F or higher for mass concrete (pours in excess of 30" in thickness) and 45' F for other concrete placements (pours in excess of 8" in thickness). The time period for this temperature maintenance ranged from 2 to 3 days depending upon the concrete's ! exposure to the elements, with a maximum allowable drop of 20* F within 24 hours after the removal of heat. The NRC investigtt. ors interviewed several LILCO site QA pers inel, S&W QC personnel, S&W construction engineering and supervisorj personnel as well as DRAVO craft and supervisory personnel, all of whom had been involved with concrete curing from 1973 through 1976. From a review of procedures, documentation and these interviews, the NRC was able to establish that the winter control of curing temperatures was accomplished in the following manner: Concrete placements were contained within temporary enclo-sures fabricated prior to placement of the concrete. 1 Heaters within the enclosure were provided. These heaters were maintained within the enclosures by laborers assigned to each of the areas. In the case of the failure of a heater, it was either repaired immediately or replaced by I spare heaters provided for this punose. In the case of ! damage to the enclosure 'itself, supervisory personnel were I notified immediately and crews were assembled to effect

the repairs.

Construction engineers took three (3) sets of temperature measurements, (a) outside ambient, (b) ambient within the enclosure and (c) concrete surface temperatures.* These { temperatures were taken daily including weekends at not less than 6-hour intervals. In addition, the outside i

                           "These surface temperatures were considered to be conservative since the                                      l actual bulk temperatures of the concrete pour would have been somewhat higher                                ,

i; than the surface temperature due to the heat of hydration released during the I N curing process. l l 17 _-___a

I ! .' ambient temperature was obtained daily from a thennometer j - which indicated the maximum and minimum temperature ex-t perienced during that time period. The daily minimum j concrete and outside ambient temperatures were recorded on i specially designated Curing Reports. Construction engineers i also stated that during extremely cold weather, additional i temperature readings were frequently taken. Temperature 3 readings below the minimum specified were reported to i Field QC personnel. l Field QC personnel perfonned periodic reviews of all curing records to assure that the temperatures of the concrete met curing specification. requirements. Nonconforming conditions were written up on Nonconformance and Disposition (N&D) reports for evaluation and disposition by Engineering. Nonconforming temperatures reported by Construction Engineering personnel were also written up on N&D reports. NRC investigators reviewed in detail over 150 written curing reports for the time period from 1973 through 1976. The areas , of the pours included concrete placements in the Reactor Pedestal, the Reactor Building Primary and Secondary Walls and the Radwaste ! Building. During the course of this review, one (1) concrete ! placement in the Reactor Secondary Wall (RS-4-12) was identified ! with a recorded temperature of 38' F on December 3, 1974, the second day of curing. The NRC investigators reviewed in detail

all N&D reports for 1973 through 1976 written for failure to j maintain the required temperature during the curing period.

l Six (6) N&D reports were found in this regard but it was noted t that Placement RS-4-12 had not been identified as nonconforming. The dispositions of the six (6) N&Ds identified required the removal of defective concrete if necessary and the testing of the concrete with a Windsor Probe in order to demonstrate that , the compressive strength of the concrete conformed to the construction specification. The NRC identified no problems in the disposition of these six (6) M&Ds. 4 j The NRC investigators noted, however, that the failure to

identify the nonconformance of the RS-4-12 Placement with respect to curing requirements and the failure to take necessary corrective action was considered an item of noncompliance with

! respect to 10 CFR 50, Appendix B, Criterion XVI which states, in part, " Measures shall be established to assure that conditions adverse to quality such as...nonconformance are promptly identified and corrected." (79-24-01) l .. l 18

4 Prior to the completion of the investigation, the licensee ! - reviewed the curing reports for all concrete placements made from November of 1973 through February of 1980 (2156 in number) under winter conditions (818 pours) and suemer conditions (1338 pours). Eight (8) additional placements were identified where either (a) no temperature was indicated for a given day (b) temperatures were identified which did not meet specification. i The lowest temperature recorded during the time periods of the

 ,                                                                                 referenced winter placements was 37' F. The referenced place-ments were identified on N&D No. 2877 and 2909 and Windsor Probe tests were performed and witnessed by Field QC. Two (2) tests of three (3) shots each were made on each placement.

i Calibration procedures and test results were reviewed by NRC investigators and no irregularities were noted. In each case, the average of three (3) shots showed the compressive strength of the placements in question to range from 5200 psi to 6900 psi, well in excess of the design strength of 3000 psi. The item of noncompliance was considered resolved and NRC investi-gators noted that the minor deviations identified would only j retard the early strength developed by the concrete and not l j cause any permanent damage, a conclusion further verified by j the Windsor Probe tests. I NRC investigators could find no instances where concrete had been exposed to freezing temperatures, a condition that would have been evident even after the fact as the freezing would ! cause the surface of the concrete to chip and flake away. Of all the individuals interviewed, none could remember any circum-

stances relating to freezing conditions on the concrete and/or

, forms. One laborer foreman stated " Occasionally one (heater) ! would fail but it would be repaired quickly. Laborers would be l circulating constantly to check on the heaters." He also stated l that in critical areas additional laborers were assigned over and above those making the rounds in order to keep the heaters i in operation.

c. NRC Conclusion l

Although the NRC identified isolated instances where heaters l had failed, this condition was to be expected and was corrected ! by licensee contractors in a timely manner. The NRC investigation found no evidence and/or information to substantiate this allegation.

3. Allegation No. 3

a. Allegation Following concrete placements for the Reactor Pedestal and

            '-                                                                      Reactor Primary Containment Wall as well as concrete placements in the Radwaste Building made from the end of 1973 through the 19 l

i i beginning of 1976, forms were improperly stripped on the day following the concrete placement instead of the required seven (7) days after placement,

b. NRC Investigation Findings NRC investigators examined applicable S&W specifications and ACI standards as referenced in Exhibit A of this report. It was noted that under ordinary conditions, when form removal is not controlled by specification, wall forms may be removed within 12 to 24 hours following concrete placement. S&W specifi-cations state that wall forms may be removed when the concrete has achieved a minimum compressive strength of 500 psi and that this strength should be achieved within one (1) day.

NRC investigators examined several records of the 24-hour tests of concrete cylinders of 3000 psi concrete. These tests indicated that a range of from 789 psi to 1497 psi and an average strength of 1173 psi had been achieved after 24 hours. NRC investigators examined several curing reports and conducted interviews with several S&W QC, construction engineering and construction supervisory personnel, as well as DRAVO craft and supervisory personnel involved in the placement and stripping of forimpork from 1973 through 1976. These records and interviews indicated that wall forms were removed after 24 hours and that . this time period was closely monitored by QC and engineering personnel. ~ The NRC noted that there was no ACI requirement or specification i requiring these foms to remain in place for seven (7) days, j although on occasion, forms were kept on for periods in excess 1 of the 24 hour requirement. This latter case was usually l dictated by work or location requirements such as when forms i could not be removed from the lower areas of the Reactor Pedestal t due to space limitations in that area. Although various indivi-i duals noted that there might have existed some engineering request for this particular extended support, the NRC could not l confim that fact by any written engineering documentation. No i irregularities were noted in this area. I

The NRC noted that although the alleger claimed that the seven (7) day stripping requirement was part of his training as a I

carpenter's apprentice, no substantiation of this fact could be obtained in the interviews of various craft personnel. As noted earlier, the seven (7) day requirement did not exist. j

c. NRC Conclusion I

[ The NRC investigation found no evidence and/or information to

       't               substantiate this allegation.                                               l l

20 l L ..- \

(.

4. Allegation No. 4

a. Allegation Following the stripping of forms from concrete placements for the Reactor, Pedestal and Primary Containment Wall, large cracks, honeycombing deep enough to expose the rebar and possible through-cracks (cracks appearing at the same azimuth location l

on both sides of the wall) were all patched over with mortar prior to inspection by QC. It was alleged that the patching was perfomed by laborers in order to complete the work before, QC had an opportunity to inspect the concrete surfaces and the patch work was done improperly, covering the honeycomb or crack with a loose mortar mixture. It was further alleged that patching in this fashion was a common, almost daily practice during the time that the a11eger worked on site from 1973 through 1976.

b. NRC Investigation Findings NRC investigators examined applicable S&W specifications, ACI standards and procedures for the repair of concrete (see Exhibit A). The NRC also examined in detail all M&D reports of concrete related defects in the Reactor Support and Primary Containment Walls covering the period in question (1973 through 1976) and interviewed several S&W QC inspectors, construction engineers, construction supervisors as well as DRAVO craft and supervisory personnel involved in the stripping of concrete fornwork and concrete repair. ,

All of the individuals interviewed were emphatic in stating that the concrete repairs could not have been performed by laborers. The NRC noted that repair work of this type was under the jurisdiction and contract of the cement finishers and that the assignment of this work to laborers, or any attempt on the part of laborers to perfors this work, would in all probability have resulted in a jurisdictional dispute which in turn would have caused the job to be shut down by the involved unions. Although the NRC realized the possibility of these conditions and questioned the alleger if possibly he had meant the masons (coment finishers) instead of the laborers, the alleger insisted that it was the laborers and not the masons who had performed the unauthorized repairs. S&W QC personnel informed the NRC that they were required to i inspect all concrete surfaces after stripping and that they had i all been given verbal instructions to perfom the inspection within 24 hours after the forms had been removed. QC inspectors I 21

stated that due to the nature of the repair work and mortar used, any attempt at an unauthorized repair would " stick out like a sore thumb" and lead the inspector to chip into the patch in order to determine if a significant probles existed. t All personnel interviewed confirmed the repair of any honeycomb or cracks prior to QC inspection would have been obvious to J anyone experienced in concrete work because of the difference in surface color and texture between the poured concrete and the patched area. This was later confirmed by NRC investigators as they examined various concrete pours throughout the plant. Investigation of the aforementioned concrete related M&D reports

indicated that a report had been written for any honeycomb which exposed rebar to one half of its diameter. A review of the disposition of the N&D reports indicated that they included a detailed repair procedure approved by Engineering. The repairs were inspected throughout by QC inspectors in order to insure the removal of all defective material down to sound concrete and subsequent repair in accordance with the specified

. repair procedure. Defects which did not expose the rebar within the concrete were considered to be surface defects and l although not requiring a specific N&D report, were repaired under QC supervision in accordance with ACI-301 and utilizing a l special concrete bonding agent. ~ i j The above repair requirements were confirmed by interviews with ' involved cement finishers who stated "(There was) no way that ' repair work could have been done without QC seeing' it as soon as the forms were raised. QC would be all over it." Another cement finisher stated "We won't touch a thing until QC has looked at it. We never do any repair work on our own. QC < l watches everything, how the mortar is mixed, placed and set - everything. " The NRC also noted through its interviews of several involved  ! j personnel and related N&D reports that through-cracks were unlikely due to the amount of rebar and due to concrete pouring techniques. No through-cracks were identified at the Shoreham l site in the areas examined.

c. NRC Conclusion The NFC investigation found no evidence and/or information to

! substantiate this allegation. l i i i l 22 ,

          ~

1

5. Allegation No. 5 l

a. Allegation Cadwelds of the rebar in the concrete located in the Reactor Building were never x-rayed and were in some cases found by the alleger to be loose. It was further alleged that several loose cadwelds were found in the Reactor Building Outer Wall (Secondary Containment).

b. NRC Investigation Findings NRC investigators examined applicable S&W specifications, General Procedures and QC procedures for cadwelding (see Exhibit A). These documents require that a specific number be assigned to each cadweld and this number is stamped on the sleeve of the individual cadweld. Each cadweld is subsequently inspected by QC and marked to identify it as either satisfactory or unsatis-factory and the sleeve number, welders symbol and result of the final inspection are marked on a Cadweld Control Record. In addition to the inspection of completed cadwelds, QC is required to monitor in process cadwelding activities on a random basis.

The location of all cadweld splices are noted on drawings by Field QC and these drawings are maintained in the QC file. NRC investigators noted no irregularities or deficiencies in these areas. NRC investigators interviewed QC inspectors'who had performed inspections of cadwelding during 1973 through 1976. It was noted that each individual interviewed stated independently that they had inspected each completed cadweld in addition to performing several in-process inspections on a random selection of cadwelds throughout the time period in questions. The NRC noted that problems identified were infrequent and when identified were corrected in accordance with accepted procedures. The NRC also noted that although no x rays of cadwelding was required, the integrity of the cadwelds could be determined satisfactorily by mechanical and visual means. The NRC also determined that the control, inspection and documentation procedures made it unlikely that a loose cadweld would have been missed. General Procedure W-300 provides for protective measures to be taken if cadwelding which is in process cannot be completed by the end of the shift. At first, it appeared that perhaps the alleger may have observed a partially completed cadweld left for completion on the next shift and assumed that it had been completed and accepted. This assumption had to be discounted as a possible explanation for the alleger's observation since it was determined that such occasions were infrequent and the

       '              cadweld would have been wrapped in plastic clearly identifying it as "in process" and incomplete.

23

i .. ,

i. .

i '.' c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation. 3 , 6. Allegation No. 6 l a. Allegation Rubber water stops between concrete layers in the Radwaste Building were not installed properly (not sealed or nailed to ' keys) and sometimes omitted entirely.

b. NRC Investigation Findines 4

NRC investigators examined applicable S&W specifications, and l procedures for preplacement inspections i.e., inspections

 ;                                 perforised prior to the placing of concrete (see Exhibit A).

i These documents require that prior to release for concrete placement, the Field QC inspector shall complete the preplacement ' inspection portion of the Field Data Sheet and sign the " Release l for Pour" block of the Concrete Pour Card. The "Preplacement Inspection" portion of the Field Data Sheet includes waterstop

installation as one of the items to be inspected.

. NRC investigators examined several Field Data Sheets and Concrete i Pour Cards for Radwaste Building concrete placements for 1973 through 1976, the period of employment of the alleger. The NRC identified no deficiencies or irregularities in this area as all waterstop inspections had been performed as required and ] the pour cards had been properly signed by QC. 4 In order to confirm these findings, NRC investigators interviewed I j, several S&W QC inspectors, construction engineers and construction supervisors, as well as DRAVO craft personnel and supervisors.

All of the individuals interviewed had been directly involved i in the installation, supervision of installation and inspection of waterstops in the Radwaste Building. The NRC noted that none j of the individuals interviewed could recall any problem with the installation of waterstops. Craft personnel described the

~ i method of sealing the joints as well as methods of holding the waterstops in position using wooden blocks and wedges. These

individuals also denied the use of any nails for attaching i

waterstops. The NRC also noted that training had been given in ! this area prohibiting the use of nails. QC personnel stated i that the waterstop installation for each concrete placement was ! checked against the appropriate construction drawings as part ! of the preplacement inspection in the same manner as other } embedmonts. The NRC noted that several craft individuals expressed mixed frustrations at times because of the several QC i 24 i .-_-_-.- .- -_-_-_- - - .-

( inspections which had to be performed as the overall concrete placement operation progressed. None of the individuals, however, reported any irregularities in these operations whether they pertained to water stops or other involved inspection criteria.

c. NRC Conclusion The NRC found no evidence and/or infomation to substantiate this allegation.

7. Allegation No. 7

a. Allegation The a11eger stated that as a carpenter he had been permitted to weld studs to embement plates used for pipe supports located in the North wall of the Radwaste Building. It is further alleged that the welding was allowed even though the alleger had failed to pass the welding qualification test seven (7) times. The alleger also stated that the work was ordered by a carpenter foreman who provided him (the alleger) with the needed welding rod.

b. NRC Investigation Findinas NRC investigators examined the construction drawings (FC-25F-6 and FC-25K-6) which show the wall elevations and details of the.

Radwaste Building in the area described by the alleger. Drawing , FC-25F-6 depicts the "V" line wall as the North wall. The NRC noted that the drawings do not specify any embedment plates to be located in the North wall. NRC investigators visually examined the North wall of the Radwaste Building and all adjacent walls and surrounding areas and noted only three (3) locations where embedments were required, two (2) were for waste treatment equipment and one (1) was for a roll up door. The NRC noted that there appeared to be no pipe support embedmonts on the North wall of the Radwaste Building. Further investigation of the Raowaste Building revealed that pipe support embedments were used very sparingly in the entire building and that most of the pipe supports in the Radwaste Building were of the concrete expansion anchor type. I

                                                       .                         25

1 )*. f l l The NRC investigators reviewed the quality assurance program in j this area and noted that embedments were normally prefabricated in shops adjacent to the construction area. The embedments i were made of structural shapes such as plates, channels or angles and manufactured studs were welded in rows to these shapes. The NRC noted that the welding was performed with a

                                            " gun" which secures the stud and welds it under controlled electrical conditions. These controlled electrical conditions provide a fairly unifom and repeatable weld. In contrast, a manually performed weld on studs of this relatively small diameter would be extremely difficult to make in a uniform manner even by an experienced welder much less by an individual who could not pass the welding test.

The NRC noted that the quality assurance program not only limited the number of people authorized to perform this welding but also limited the number of people authorized to receive welding rod and to sign the welding rod withdrawal form. From the standpoint of welder qualification, a review of all appli-cable records and interviews with several DRAVO craft and supervisory personnel indicated that (a) the number of qualified carpenter / welders was very limited," (b) very few carpenters were ever selected to take the test and (c) an individual could only fail the test twice, for if he could not pass it on the second attempt, he was not allowed to take the test again. The NRC was also able to detensine that the alleger was never selected to take the welders test and consequently could not have failed it one time much less seven (7) times as alleged. , From the standpoint of weld rod issue, the NRC noted that the quality assurance system closely controlled the issue of welding materials to qualified personnel only. The contractor involved in these areas had only four (4) individuals authorized by letter to sign welding material withdrawal slips. The issue of welding material was contro11ec by an independent group who, without proper authorization, would not issue welding materials. The NRC noted that the signature authorizing the withdrawal of welding materials must be authenticated. This was considered important due to the fact that the carpenter foreman who al-legedly ordered the unauthorized welding and who allegedly provided the welding material was not authorized to withdraw I the material.

                            "only thirteen (13) carpenters had been qualified in the past seven (7) years 26

The NRC noted, therefore, that in addition to the fact that there were no embedment. plates used to hold pipes on the North j- wall of the Radwaste Building, the alleger was not selected to take the welding test, could not have failed it seven (7) times, would have had difficulty obtaining welding rod because he was unqualified and finally, determined that it was unlikely 1 he received welding material from a foreman who was not authorized i to sign weld rod withdrawal slips. In order to further confirm the above findings, the NRC noted that the Radwaste Building was designated as a Seismic Category I structure, meaning that the quality assurance program applied for safety-related structures would require a "preplacement inspection" for every concrete placement performed. This ites was documented on Field Data Sheets (Fom T-S-31) and reviewed by the NRC. One of the attributes requiring quality control inspectors sign off was "Embedments". The NRC interviewed five (5) of the original QC inspectors for the Radwaste Building who , stated that all embedments were visually inspected for location, conformance to drawing details, and restraint to avoid movement during the actual placement of concrete. Although interviewed j individually, all of the QC inspectors independently agreed that a manually welded embedment would be readily recognizable and would have been immediately questioned. To the best of j i their knowledge no manually welded embeenents were ever noted.

c. NRC Conclusion 3 l The NRC found no evidence and/or information to substantiate this allegation. l

8. A11ecation No. 8 i

a. A11ecation 1

The alleger stated that threaded tie rods were used as fora l ties for the Reactor Pedestal along with she-bolts attached to l the rods to hold the forms in place. It was further alleged 1 that when the she-bolts were removed, in many cases the threaded i rods would slide out of the concrete and a small patch would be { put on the hole leaving a void in the center of the concrete. i In some cases, it was alleged that when the she-bolts could not l be removed at one end, the entire tie rod was pulled through the concrete pulling a large amount of concrete off the pedestal wall. It was alleged that rods were pulled through the concrete i approximately twenty (20) to fifty (50) times in this fashion. t l . i ) 27

l

              .                                                                                                                                       l l

l I

b. NRC Investication Findings '

NRC investigators examined several slides showing the acteal concrete forms in various stages of installation for concrete placement for the Reactor Pedestal. The NRC also interviewed several QC and construction personnel in order to obtain infor-nation concerning the use of threaded bolts for form ties. The information obtained indicated that in general, standard manu-factured (Williams) form ties and she-bolts were used. Although threaded rods and she-bolts may have been used in certain locations where the standard form ties did not fit the required configuration, interviews with the above referenced individuals could not determine these locations exactly. The NRC noted that either type of tie rod was acceptable, hence the matter was not pursued further. The NRC also noted that the alleger's statement that the threaded rod would slide out of the concrete when the forms were stripped appeared to indicate that the rods were installed in sleeves, a fact which the alleger denied, stating that threaded rods not in sleeves were pulled from the concrete after the concrete had set. This was in apparent conflict with all of the individuals interviewed who stated that even when threaded rods were used, they were not in sleeves as the alleger stated but that they definitely remained in the concrete after the forms were stripped, in conflict with the alleger's statement. The NRC did not attempt to pursue the matter further in order to resolve the conflict due to the fact that (a) either method of installation (with or without sleeves) was acceptable in the i Reactor Pedestal and (b) if the rods were installed without sleeves as stated emphatically by the alleger, it would have been physically impossible to remove them from the concrete once it was set. The NRC determined that the alleger's state-ment that tie rods without sleeves were pulled out of the concrete could not be considered as credible. The NRC also noted that had sleeves been used, patching the holes would not have affected the structural integrity of the Reactor Pedestal, i nor was the Reactor Pedestal designed to maintain airtight integrity.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.

s. f , 28 i I- _ _ , _ . . . _ . _ _. , . _ _ _ .,. _

o

9. Allegation No. 9

a. Allegation Several through-cracks were observed in the Turbine Building wall that separates it from the Reactor Building. It was further alleged that the cracks were so wide one could see '

through to the other side of the wall.

b. NRC Investigation Findings The NRC noted that although the Turbine Building has a sheet metal roof and upper wall which are designed to blow off during a design basis tornado, the overall Turbine Building is not seismically designed nor is any credit taken for leak tightness.

, The design of the Turbine Building Ventilation System calls for it to, exhaust air from lesser to progressively greater potentially l contaminated areas to a monitored final exhaust. Thus even if i a crack did exist, its significance would be minimal. The NRC noted further that there was no common wall between the Turbine Building and the Reactor Building so a crack in the Turbine Building wall would not mean a crack in the Reactor Building. In addition, any crack large enough to see through would be difficult to patch on any permanent basis without the crack opening up periodically. , On February 21, 1980, the NRC inspected the entire South wall of the Turbine Building, the wall facing the Reactor Building. Although several temporary openings were noted (openings scheduled to be closed at a later date), no significant cracks were identified.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.

10. Allegation No. 10

a. Allegation l During the first three weeks of February 1979, welders were asking for ER-309 weld rod in lieu of the ER-308 and/or ER-316 l weld rod specified on their weld rod requisitions for scheduled

! work on dissimilar metal welds. When refused by the weld rod j clerk due to the improper designation on the weld rod requisition, the welders stated that it was too cold to return to have the l requisition corrected to the required ER-309 and that the i

       ~

29 '

ER-308 and/or ER-316 would be utilized in its place. It was l alleged that these instances occurred two to three times each day during the three week period in February of 1979. The systems alleged to have been involved were the Control Rod

    .                           Drive System at Elevation 78' in the Reactor Building; the Flow and Pressure Instrumentation Lines in the T-48 Primary Control /At-mospheric Control System, the G-33 Reactor Water Clean-Up System; and the G-41 Fuel Pool Cooling System,

b. NRC Investication Findings NRC investigators reviewed in detail the flow diagrams of the four (4) alleged systems involved in the allegation as well as seven (7) additional systems. These flow diagrams depict material specification changes and pipe line numbers which can be traced to a specific weld. This review identified (12) dissimilar metal welds. The document packages for each of the.se welds were examined to verify that ER-309 weld rod was,.

in fact, used to make the specified welds. The result of this examination revealed that one of the welds (Weld Joint Number 1G-33-WD9-3-1-FW "D") was welded on May 19, 1979 using ER-308 material. No utilization of ER-316 was identified during this review. The licensee was informed that 10 CFR 50, Appendix B, Criterion V requires in part, that " Activities affecting quality...shall be accomplished in accordance.with these instructions, procedures, or drawings". The COURTER and CO. Welding Procedure Specifi-cation NW-100-01-08011AA, Revision 0, specifies that the filler metal for this weld be AWS Class ER-309. Contrary to the foregoing, Weld Material Requisition No. 55780 was issued to weld joint 1G-33-WD9-3-1-FW "D" specifying ER-308 material. This was considered an item of noncompliance (79-24-02). Based on this finding, two (2) separate and distinct reviews were initiated by both the NRC and the licensee. All Category I piping isometric drawings were reviewed and all dissimilar metal welds were identified. In addition, all Task Engineering Component Checklists were checked and it was found that all dissimilar metal welds were listed. A cross check of these two { sources of information indicated that all dissimilar metal 1 welds had been identified. Four hundred and twenty two (422) i dissimilar metal welds were identified and cross checked in this manner. The document packages for each of the 422 dis-similar metal welds were examined in order to verify that ER-309 material had been used as required. The results of this axamination identified one additional weld (Weld. Joint Number 30

M50-CW3-3-99) welded on April 12, 1979 that also had been welded utilizing ER-308 material. None of the dissimilar metal welds were identified as involving ER-316 material. As an independent verification of the above findings, and due to the fact that ER-308 material was found in two (2) of the 422 welds examined, NRC investigators listed all of the identi-fiable ER-308 weld rod issues for the month of February 1979 from the weld rod control log. This log listed the date and use for each weld rod issue. The weld rod issues were cross referenced against'the appropriate piping isometric drawing which would identify any dissimilar metal welds. Even if the pipe line did not represent a dissimilar metal weld joint, on isometric drawings which show more than one pipe line number, any dissimilar metal weld depicted on the drawing was examined in order to verify that ER-309 was in fact used. This independent verification of over six hundred (600) log entries dispositioned indicated no further discrepancies. Independent cross checks were also made by the NRC investigators on the completeness of the dissimilar weld testing and no discrepancies were identified in this area. NRC investigators interviewed all available principals involved in the above referenced item of noncompliance. Interviewees included the Assistant Construction Manager, the welders who made the welds, involved weld rod checks, welding supervisors,

                     ' area supervisors, assistant area supervisors and deputy foremen.

As a result of these interviews, the NRC was able to determine that the misuse of ER-308 weld rod materials were relatively isolated occurrences rather than any significant breakdown in the licensee's construction QC program. The problem appeared to stas from pape mork errors related to the similarity of procedure numbers (08011AA for ER-309 and 0811AA for ER-308) and not a deliberate attempt by the welder to utilize the improper weld rod. The two welds in question were cut from the system in the presence of the NRC on March 4 and 5, 1980. The removed welds were bisected and analyzed independently by both the licensee and the NRC to confirm the composition of the weld rod material contained therein. By chemical analysis, performed at the Franklin Research Center, the NRC determined that the weld material was in fact ER-308. Similar results were obtained by the licensee using spectrographic and chemical analyses. The NRC detemined that all welding perfomed on the control rod drive mechanisms was performed by Reactor Controls, Inc. (RCI) which maintained its own weld rod issue facilities, 31

4 e '. , l totally separate from weld rod issues of COURTER for whom the j' alleger was employed. It was impossible therefore for the

a11eger to have dispensed weld rod utilized for the control rod drive mechanism. This system was included nonetheless in the NRC review and no irregularities were noted.

i NRC investigators noted that the identification of the two welds containing the improper weld material while constituting l an item of noncompliance, did not substantiate the related al- t ] ,, legation due to substantial differences in time frame, scope, 2 causality and magnitude. The NRC also noted that the use of ER-308 in place of ER-309, while not in compliance with specifi-cations, would not have a significant adverse affect on the structural integrity of the two welds in question, a conclusion confirmed by the Franklin Research Center which stated that both welds were sound and that they were deposited without excessive base metal dilution.

c. -NRC Conclusion i

The NRC found no evidence and/or information to substantiate this allegation. l 11. Allegation No.11 l

!                               a. Allegation
 !                                    Although taught that only E-7018 electrodes were to be used at the Shoreham site, E-6018 electrodes were issued to a single

welder who continually drew weld wire separately from other welders. It was further alleged that the above practice was '

;                                     perfomed in order to covertly use substandard materials.

b. NRC Investigation Findings NOTE: The NRC noted that the ASME Code, Section II C, SFA-5.1 does not list an E-6018 electrode. It was assumed that the alleger meant E-6010 electeddes j which are listed in the ASME Code and are in common

 !                                                use within the industry.                                   ;

l The NRC conducted several interviews with S&W QC inspectors, i construction supervisors, weld rod control supervisors and weld rod issue clerks. These interviews disc',osed that the only E-6010 electrodes used at the Shoreliam site were utilized in connection with the carbon dioxide fire system installed by the

C. P. Bennett Company. The NRC noted that extremely rigid i

32 l l

'/ controls had been imposed on the issue, use, handling and return of the E-6010 electrodes in order to assure that their use was limited only to this fire system. Each weld electrode i issue was counted and verified by the rod issue clerk, the welder, a LILCO QC representative and a representative of C. P. Bennett. A similar procedure was followed when the unused weld electrode and weld electrode stubs (used electrode) were returned. A LILC0 Work Directive (WD-PS-4894) directed the Bennett Company to hire "...one Quality Control Inspector. Inspector's sole responsibility will be to ensure that the E-6010 Series Welding Electrode is used exclusively on the CO System pipe welding." i NRC investigators conducted an additional interview with the C. ! P. Bennett General Foreman in order to confirm the above re-forenced procedures. No discrepancies were noted. The General i Foreman stated that possibly no more than two (2) 50 pound cans I were used for the C0: System and that only three (3) employees j were authorized to sign for the E-6010 electrode issue. I NRC investigators noted that the allegation was presented in ! such a manner as to indicate that the licensee was covertly l attempting to introduce substandard material into the con-i struction project. Coworkers of the alleger indicated that this I was not credible since all involved individuals (including the

a11eger) were carefully infomed that the strict procedure was

;                                         required in order to comply with the increased QC requirement.

l In fact, one supervisor stated that the alleger was fully aware { of the purpose of E-6010 electrode and he (alleger) complained ! that the procedures were too strict. The NRC did not pursue i these conflicts further as it did not affect the evaluation of l the validity of the allegation. i

c. NRC Conclusic The NRC found no evidence and/or information to substantiate this allegation.

12. Allegation No. 12

! a. Allegation

A large scale repair of a 10 to 20 foot crack was performed on the feedwater condenser jacket by a REGOR boilermaker instead of the usual COURTER and CO. steamfitter in order to avoid having the crack reported to COURTER QA personnel which would i

have raised an issue as to the integrity of the entire jacket. It was also alleged that an exceedingly large quantity of weld electrodes had been drawn in order to effect this repair. 4 ~. ! 33

L.

b. NRC Investication Findings NOTE: The NRC was informed by the alleger of the welder's symbol and the Weld Rod Requisition Number, which was 170376.

An NRC review of plant equipment disclosed that what was referred to as the "feedwater condenser" was in fact the Main Turbine j Condenser. This condenser is located in the Turbine Building and parts are located at the 63 foot level where the alleged irregularity occurred. In reviewing the time period in question, the NRC noted that the condenser was hydrostatically tested and a leak was found on or about August 25-28, 1978. The leak was located in the general area of the interconnect between the condenser shells. Weld Rod Requisition Number 170376 was confirmed by the NRC to have been issued on August 29, 1978 to the welder identified by the alleger. The requisition was for 100 each, 3/32" diameter, E-7018 weldiog electrodes for " repair weld condenser." The NRC review indicated that the 3/8" fillet weld on the interconnect had been leaking and approximately 10 linear feet of this weld had been removed in order to repair the leak. The NRC noted that there was no " crack" in the strict sense of the word. Due to the fact that the area which was leaking was inaccessible from the outside of the interconnect and that the exact location of the leak could not be pinpointed, the weld in

                                                           .the imediate area of the leak as well as additional weld material on either side was removed by air carbon are gouging.

It was noted that the removal of the weld by air carbon are gouging would also remove some of the base material therefore adding to the volume of weld metal needed to replace the weld. L The rowelding procedure would involve a minimum weld size fros

                                                     /      1/2" to 5/8" instead of the original 3/8" fillet weld.

NRC investigators performed calculations in order to determine

          ',                                                the number of 3/32" diameter welding rods needed to fill a 3/8" X 10 foot

• weld groove. The NRC determined that the issuance of 100 each, 3/32" diameter electrodes was not excessive relative to the volume of weld deposited in the prepared grove.

                                           " Includes the cutting and repair of various stiffners which had to be removed in order to gain access to the leaking area s.

e 34 L

D The NRC noted that since REGOR was contractually responsible for the work on the condenser, they were assigned to perform the repair work and that the job would not have been assigned to COURTER in any event. The NRC noted no deficiencies and/or irregularities in the repair operation. l

c. MRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.

13. Allegation No. 13

               ,                  a. A11ecation Salt water is seeping through the Secondary Containment Wall at the 8' level and around-the-clock efforts are being undertaken to pump the water out. It was further alleged that the seepage occurred due to LILCO's improper estination of the depth of the water table under the plant.

l b. NRC Investication Findings NRC investigators questioned several individuals working at the Shoreham site concerning the allegation and were unable to obtain any confirmation that any seepage,had occurred. The NRC reviewed construction descriptions of the various waterproofing methods and other means of preventing water seepage and deter-mined that the possibility of seepage in this area was highly unlikely. In order to confirm the above finding, the NRC personally examined the areas in questioti during high tide on January 31, 1980, a date where according to local marina owners, tides were expected to be their highest due to the full moon. The NRC noted that there was absolutely no indication of leakage or any signs (e.g. , water marks) that seepage had occurred in the past. In fact, the NRC noted that dust accumulations in some parts of the 8' level floors of the Secondary Containment indicated that the area had not seen water for some time. l Interviews by the NRC of various LILCO personnel indicated that on infrequent occasions -in the part that some water might have reached the 8' level from within containment due to leakage of the temporary drain system or leaks from various hydrotests above that level within the plant but that to the best of their knowledge, no through-wall seepage problems had occurred in this area. L ,

                                                                                                         \

35

Interviews by the NRC of various S&W construction personnel confirmed the above observations but added that a minor problem had been experienced in the past with rainwater leaking onto i the 8' level floor through an unsealed spare penetration during periods of heavy and sustained rainfall. The water was not extensive and merely flowed to one of the several permanent sumps located on that level for such purposes. S&W personnel , stated that when sump pumps were operated, the water from the ' sump was held up for chemical analysis and filtering prior to discharge to the store drain system in accordance with EPA guidelines. The NRC Resident Inspector was present at this penetration during a recent heavy rainstorm and no leakage problems were identified. S&W representatives stated that when the pumps had to be operated, they were operated by a repre-sentative of the operating engineers' union in accordance with union agreements and this condition would in all probability I l remain in effect until the pumps were transferred from the construction to the start-up group at which time automatically operating systems would be employed. From the standpoint of other pump operations, the NRC noted that pumps outside of Secondary Containment were merely pumping water from the temporary drain system within the building into exterior drains and no irregularities were observed with respect to this practice. The pumping operation was due to be gradually phased out as the permanent drains within the building were integrated into the overall drainage system. The NRC also noted that extensive pumping operations were undertaken when the concrete was initially poured several years ago but that this was part of the normal dewatering operations conducted during construction work of this kind and did not indicate any seepage or water leakage problems or problems with any codes or other requirements.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.

14. Allegation No. 14

a. Allegation Stone and Webster-(S&W) lost its general contractor duties when it repeatedly complained to LILCO about the incompetence and corrupt practices of its contractors, such as COURTER and REGOR, which LILCO insisted on using. In August of 1978, S&W was relegated to design, drafting and general QA duties while a dummy corporation, UNICO, assumed QC duties, and COURTER QA s

personnel assumed the QA field inspection duties. 36

 'I                  b. NRC Investigation Findings NRC investigators examined the various changes in the management of the Shoreham project. Among the itees reviewed was a memoran-dus dated March 8,1977 which documented a meeting between representatives of LILCO, S&W and the ASME Subcommittee on Certification. The subject of the meeting was S&W's application for NA and NPT Certificates of Authorization which would have allowed S&W to perform ASME Code work at the Shoreham site.

The ASME Subcommittee agreed to grant the Certificates of Authorization subject to certain conditions, among which was the transfer of the contract between LILC0 and COURTER and CD. to a contract between S&W and COURTER and C0. This condition was prompted by the fact that the ASME required QC operations to be perfomed by the actual contractor and not the si& contractor. NRC investigators were informed that at this time LILCD was considering changes in construction management, and in a memoran-dum dated August 11, 1977 recommended that LILC0 assume the leading role in construction management with S&W retaining responsibility for engineering and quality assurance. As of September 12, 1977 LILCO did assime the leading role in construc- l tion management. This change was effected by having LILC0 assume leadership of the joint S&W/LILCO unified construction team (UNIC0) through the appointment of a LILC0 Construction Manager reporting to a LILCO Project Manager rather than the , 54W corporate construction organization. LILC0 also increased I its participation in the UNICO organization by the transfer of a number of qualified construction supervisors to Shoreham from other LILCO departments. As part of this change, LILCO decided to retain the COURTER and CO. contract directly rather than transfer it to S&W. Since COURTER was now the contractor, the ASME would not allow S&W to perform the code work as either LILCO or COURTER had to obtain the ASME Certificate of Authorization. Therefore, COURTER was directed to obtain the ASME Certificate and in order to do so was given the responsibility for quality assurance for their work on site. The ASME transition date was January 1,1978. The only reference to any changes occurring in August of 1978 was a memo dated August 28, 1978 where some duplication in the l inspection of non-safety related mechanical equipment and insulation was corrected by transferring the responsibility from the Construction Inspection Program (CIP) and FQC to the CIP alone. There were no complaints involved in the memo, only non-safety related equipment was involved and the NRC determined that there was no problem or irregularity assc,ciated with the transier. 9

                                                                       -37

The NRC noted that UNICO was formed prior to March of 1977. The i . change of QC functions from S&W to COURTER was occasioned primarily as a result of ASME requirements and not due to any actions by S&W and/or COURTER. NRC investigators ~also interviewed ' several individuals involved in the S&W QC and COURTER QC organizations and the referenced transition and were not able to identify irregularities or deficiencies, either past or present, resulting from this transfer of responsibility.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate

• this ellegation.

15. Allegation Nc. 15

a. Allegation At least seven (7) major failures of hydroflushes of the Primary Closed Loop Piping System have occurred and during the flushing of the system on June 15, 1979, a gross failure of the primary system occurred involving valves popping and a section of pipe being thrown fifty (50) feet into the air. It was also alleged that the hydrostatic test of the system in September of 1979 could not have been valid, since it occurred too soon after the gross failure in June of 1979 to have permitted proper shutdown and repair of the system.

l b. NRC Investigation Findings The NRC reviewed all possible primary system hydrotesting records and interviewed several involved personnel and could not identify any major failures of the Primary Closed Loop Piping System during the time period up through and including June of 1979. Examining the time frame around the alleged June 15, 1979 date, the NRC noted that on June 13, 1979, one minor incident was identified where during the flush of the Core Spray System, a temporary gasket in a bolted joint failed and sprayed water over a large area. There were no major failures ' in the system, no valves failed nor were pipes thrown into the l ai r. This review was documented in NRC Inspection Report No. 50-322/79-20. The NRC investigators noted that the primary system hydrostatic test was not conducted until September 21-22, 1979. The NRC reviewed all test procedures prior to the test, was present and witnessed the actual perfomance of this test, and independently l l sa

                                                                   --   -   __           _ . . , _ . ,         .--,_..___-_._m.

verified the acceptability of the test results. The NRC noted no gross failures and the details of the test were documented ' in NRC Inspection Report No. 50-322/79-15. ' The NRC reviewed documents referred to them by the alleger and found them to be outdated copies of normal check off sheets for the hydrotests which he (the alleger) had taken off site prior to his termination in May of 1979. The NRC was unable to reconcile the alleger's claim of having seen the gross failure (June 1979) since he had terminated one month prior to its allegedly having occurred. The NRC was also unable to reconcile the alleger's statement that exclusive of repairs it would have taken 4 to 6 weeks to shut down from the alleged failed test due to the fact that the shutdown operation can be performed in a matter of hours or at most days, depending on the procedures utilized and the system involved.

c. NRC Conclusion i The NRC found no evidence and/or information to substantiate this allegation.

16. A11ecation No.16

a. Allegation .

The diffuser (outfall) pipes for the circulating water system have never been properly anchored and due to the tidal action in Long Island Sound, have shifted, broken and separated from the line itself. It was further alleged that a LILCO QA mp documented the separated pipe pieces located in the Sound.

b. NRC Investigation Findings 1

The NRC investigators noted that a similar allegation had been made by a contractor employee in 1978 and had been investigated by the NRC in October of 1978. The results of this former investigation are documented in NRC Inspection Report No. 50-322/78-16. NRC investigators pursued this matter further and interviewed various individuals associated with the' placement and securing of the Offshore Discharge and Diffuser Pipe System. The NRC l was able to determine that the fiberglas pipe sections were l ' placed on the floor of the Sound and then covered with a minimum of 3 feet of crushed stone (approximately 1.5" in size) and then covered with a layer of armour stone (6" to 2' in size). The NRC identified the fact that while the outfall was being placed, prior to securing the outermost sections with stone, (, stores caused various sections to work loose and suffer some 39

4 damage. A weekend storm of October 2-3, 1976 caused damage to four (4) sections of the outfall. This occurrence was covered in N&D Report No.1000. Another stom from May 6 to 9,1977 caused damage to two (2) sections of the outfall, one of which was replaced as documented in N&D Report No.1253. In October of 1977 an underwater inspection of the outfall verified that it was secure but the inspection indicated minor damage to a

diffuser, a condition repaired and documented in N&D Report No.

1466. The NRC identified no instance whereby secured sections of the outfall had broken loose and separated from the system, a In attempting to determine the validity of the allegation with respect to the map depicting the location of separated outfall pieces, the NRC determined that a map of the outfall was posted when the system was being placed and being secured by rock. The map and diagram indicated which sections had been covered with the various sized rock and the flags indicated the location of the barges containing the rock used to secure the pipes and not section of pipe which had broken loose. . c. NRC Conclusion The NRC found no evidence and/or infomation to substantiate this allegation.

17. Allegation No.17

a. Allegation NDE technicians were not adequately qualified for the jobs they were performing,

b. NRC Investigation Findings

• 1 NRC investigators examined prior NRC inspection reports for the  :

Shoreham site for the years of 1977, 1978 and 1979. It was i noted that from January 1977 through June 1979, NRC inspectors l reviewed the qualification records and observed NDE technicians  ! performing tests during eight (8) inspections. In addition, it l was noted that during welding inspections, NRC inspectors frequently include the inspection of the perfomance of nonde-structive tests and also review the results of these tests, observations that would not necessarily be reported unless nonconfoming conditions were identified. The NRC evaluation indicated that no deficiencies in the qualifications or per-fomance of NDE technicians were identified during any of the NRC inspections reviewed. .

      .       *Due to the fact that no specific NDE technicians were identified by the

(- alleger, the NRC evaluated this allegation as-it pertained in general to the

              -overall construction operation.                                                              l

, 40 _

                                                                                               , _ _ _ _ _)

        **                                                                                   \

l

         -                During this particular investigation, NRC investigators examined a random selection of experience, training and qualification records of NDE technicians currently at the Shorehas site.

This examination showed the men in question to have been ade-quately qualified for the level of performance for which they are certified and to a degree commensurate with their responsi-bilities.

c. NRC Conclusion The NRC found no evidence and/or infomation to substantiate I t

this allegation. '

18. A11ecation No. 18 l

l

a. Allegation .

Large quantities of green dye used for dye penetrant testing were being discharged by LILCO without proper approval and that this green dye was visible on the "outake" canal and polluting Wading River shellfish.

b. NRC Investication Findings The NRC assumed that since there is no "outake" canal at the Shoreham site, that the alleger was infact referring to the intake canal. The NRC noted that LILC0 had discharged green dye to the Sound and that this matter had been identified by the NRC and documented in NRC Inspection Report No. 50-322/78-16.

During that inspection, the NRC inspector observed the green color of the water in the intake canal at the screen wall and at the stom drain discharge in the canal. The NRC requested identification of the substance and examination of the discharge permit provision which allowed the discharge. LILCO representa- l tives identified the material as a fluorescent dye &nd identified the individual within the New York State Department of Environ- i mental Conservation (NYS-DEC) with whom approvals had been

      '                 coordinated. NRC environmental inspectors were notified and subsequent contacts with NYS-DEC verified their cognizance of the discharges and their acceptability.

NRC investigators examined the intake canal and adjacent shore-fronts on Long Island Sound, including the most susceptible areas of Wading River, and found no visible evidence of green dye.

c. NRC Conclusion -

The NRC found no evidence and/or information to substantiate l ( this allegation. 41

4

19. Allegation No. 19

a. Allegation Supervision of trade workers is inadequate and being performed i by unqualified individuals.  !

b. NRC Investication Findings" l l

During the course of this investigation, several construction , workers from various trades and crafts were interviewed privately ' by NRC investigators. During these interviews, selected workers  ! were asked their opinion not only of their own supervisors but of other supervisors within their own craft. The NRC noted that the majority of workers believed their supervisors to be qualified for their positions and in general providing an adequate amount of supervision. Although some workers did not ' feel qualified or able to evaluate their supervisor's performance, they did state that they did not have any problems in this area at this time. The NRC noted further that even in areas where workers admitted to personal differences with their supervisors, they still acknowledged their supervisor's ability to meet his supervisory responsibility. A random evaluation by the NRC of various supervisor's qualification did not reveal any inadequacies. In order to indirectly determine the adequacy of supervision on  : site independent of workers interviews, NRC investigators i requested one of the contractors (CDURTER) to provide an indi- - cation of the number supervisory and non-supervisory personnel currently at the Shoreham site in construction related work areas. The NRC noted that the information provided indicated ' that in the four construction areas reviewed (Radwaste, Reactor, Instrtmentation and Auxiliary / Yard Area), a supervisory individual was provided for every 2.4 craft individuals. Supervisory personnel included Welding Department Supervisors, Deputy Foremen, Area Foremen, General Foremen, Assistant Area Supervisors and Area Supervisors. If COURTER QC personnel were included, there would be one (1) supervisor for every 1.9 workers. The l NRC determined that this level of supervision appeared to be commensurate with the extent of construction activities being performed at the Shoreham site.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.

                "Due to the fact that no specific instances and/or individuals could be obtained i

t ('- from the allegers, the NRC determined the validity of the allegation as it pertained in general to the overall construction site. i s 42

l l l

20. Allegation No. 20

-l l a. Allegation , 1 l l Qualifications and training of subcontractor personnel at the ] Shoreham site is inadequate. i

b. NRC Investigation Fin' dings During the course of the investigation, several subcontractor personnel were interviewed by the NRC. During the course of I

these interviews, individuals were informally questioned with respect to their job responsibilities and past experience as l well as how various aspects of their particular jobs were performed. The NRC noted that the workers interviewed appeared to have a degree of knowledge (by experience and/or training) commensurate with the job responsibilities given them, noting also that this level of knowledge covered a wide area ranging from the novice apprentice to the experienced journeyman. Some workers noted frustration in not being able to perform their jobs as they were accustomed due to the several QC checks which were required and which would delay the job until the inspection was performed and proper standards were met. Although some workers admitted not knowing all the standards in detail, they stated that they had no problems in this area due to the fact 1 that their supervisors and QC inspectors would handle these areas. The NRC found no_ instance where a job requiring a high degree of skill was performed by an individual not having the qualifications and skills commensurate with the job.  ; The NRC also interviewed several QC inspectors and supervisory personnel in a similar manner as described above and were able to determine that their knowledge of the appropriate standards and procedural requirements was also commensurate with their responsibilities. No irregularities or deficiencies were identified (see also Paragraph 17 of this section). From the standpoint of the actual training of subcontractor personnel by their own crafts, the NRC did not review the actual apprenticeship program of the various crafts as this was

                            .outside the purview of the NRC's responsibility. 'The NRC did, however, review the training provided by the subcontractors to these individuals as it pertained to their activities and responsibilities at the Shoreham site. The NRC reviewed a memo l by 58M to all subcontractors dated August 20, 1976 describing     '

the implementation of a site wide training program, separate l

                     .                             43

I from other Manual. training requirements as described in the Field QC Each contractor was directed to provide a minimum ! three (3) hour orientation pro employee attended the courses. gram and to ensure that each new The NRC selected the Orientation Manual for DRAVO and reviewed it in detail. No apparent defici-encies were noted. A check of an alph-'a+4r=1 listing of DRAVO employees reflected the fact that all onsite employees were  ; t recorded as having attending this particular training.  ! The NRC reviewed Orientation Manuals for other subcontractors and determined that the DRAVO Manual was generally reflective , of all submitted manuals in terms of content and course design. i

The NRC also conducted a random inspection of several of the training records in various areas as given by various contractors.  !

For example, from April 26, 1977 through May 31, 1977 a one (1) ' hour course was given several times on the installation and inspection of wedge type anchor bolts which were installed in l concrete to secure various pipes and equipment. Approximately 660 above. workers were given this course over the time period stated Similar classes were noted in welding and other craft activities on site. i

                                                                                                   )

The NRC was able to determine, therefore, that based on personnel '

 .                         interviews and training records reviewed, that the training of subcontractor personnel appeared to be adequate for their duties at the Shoreham site.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.

21. Allegation No. 21

a. A11ecation Workers who commenced the task of painting the inside of the Reactor Primary Containment were not qualified. It was further alleged that an NRC inspection of worker qualifications resulted in a majority of the workers being ordered off the job; however, the substandard work that had been done to that point was allowed to remain and was not inspected by the NRC inspector; the remaining workers were ordered to complete the task on overtime and on one occasion they worked a 30-hour shift while on methedrine.

4 44

p. i

b. NRC Investigation Findinas l NRC investigators noted that painting specifications, procedures and records were reviewed, as well as an in progress inspection performed, by an NRC inspector as documented in NRC Inspection Report No. 50-322/79-20. No items of noncompliance were identi-fled by the NRC inspector and the perfomance of the work as well as the QC inspection associated with it were considered to be acceptable. NRC investigators also reviewed all NRC inspection l reports for 1975 through 1979 and found no record of an NRC inspector questioning painters qualifications at any time during this period nor any report of painters being layed off as a result of the NRC having questioned their qualifications.

The NRC interviewed S&W construction and QC personnel responsible l for painting as well as QC personnel working for KTA-TATOR Associates, the consultant responsible for the inspection of painting within the Suppression Pool". The NRC noted that during July through November of 1978, the painting of the Suppression Pool required a large amount of overtime, with men working 12-hour shifts, 7 days a week, using two (2) shifts per day with a day off every two or three weeks. The NRC could not identify any 30-hour shifts worked at any time.  ; I During this painting operation two (2) types of coating systems 1 were utilized in the suppression pool, an epoxy (K&L) system I above elevation 30 and an epoxy phenolic (Plasite) system below l that elevation. The NRC noted that painters were qualified separately for each system in accordance with a proposed ASTN oualification method. Qualification for the Plasite system was considerably more difficulty because of the coating systes itself as well as the use of a more complex qualification panel. A large number of painters who had qualified on the K&L system could not qualify on the Plasite system pa'nel. As the work in other areas neared completion, the men who had qualified for both system were retained while other painters with lesser qualifications were layed off. l

. "the Suppression Pool is the lower portion of the Reactor Primary Containment I

( 45 i _. . ._ _ _ _ . _ _ _ _ ,_.- _ _, _ ._ _ .._, .__ _ _

Individuals interviewed by the NRC stated that the most visible effect of the long hours was a decrease in productivity. They stated further that the effect on quality was not identifiable and the reject rate was little if any greater. The NRC noted that the requirements for inspection and testing before, during and after application were sufficient to identify any decrease in quality. During this investigation, NRC observation of in progress painting showed no irregularities or deficiencies. As mentioned earlier, 30-hour shifts could not be identified by NRC investigators. From the standpoint of the methedrine allegedly used during this shift, the NRC found no indications of drugs being used in this manner whether in the area of painting or other construction areas covered in other aspects of this investigation report. Although the NRC was informed that the labor problems engendered by the aforementioned layoffs may have initiated this allegation, this matter was not pursued inasmuch as problems were not identified with the painting l operation itself.

c. NRC Conclusion l

1 The NRC found no evidence and/or information to substantiate l this allegation. l

22. Allegation No. 22

a. Alle'oation .

                                                                                                                         ~

Tube support sheets in the condenser box were so misaligned l that titanium tubing which is supposed to fit loosely through  ! the support holes was often hammered or twisted to fit through ' the sheets. It was alleged further that the damage was extensive enough to cause a tube to break with the possibility of a - radioactive spill.

b. MRC Investigation Findings The NRC noted that the " condenser box" or Main Surface Condenser is not a safety related unit, nor is it classified as an ASME Code Vessel as the shell side is under vacuum. The unit itself is pre-assembled without the tubes, disassembled for shipment to the site, where it is reassembled, welded and tubed. This particular condenser was designed and shop assembled by Inger-soll-Rand (IR) in accordance with S&W Specification SH 1-6 dated June 16, 1969 and revised on August 15, 1969 and February i 27, 1970. The NRC noted that the original specification called for 58111, Alloy CA706 (Cu-Ni) tubes but that the design was modified to improve water-side corrosion resistance by esploying

( ASTN:B338, Grade 2 welded titanium tubes. l l

1 The NRC noted further that at the time of the material change , to the improved titanium tubes, assembly of the condenser was ' partially completed to the Cu-Ni design. Since the maximum unsupported span for the titanium tubes was approximately 20% less than for the Cu-Ni tubes, the total number of supports had to be increased by approximately 25%. Due to the fact that the original tube support plates had been installed, the decreased unsupported span was achieved by the installation of additional partial tube support plates between those already installed. S&W stated that such a procedure was a standard solution for a retubed or partially built unit. This was confirmed by the NRC. l The resultant tube support plate system was reported by S&W to  ! contain approximately 60-70% more tube support plates than ' would normally be necessary for a completely new design with some tube support plates as close as 18 inches apart. The unit was also designed to accommodate the MAN in-service (brush type) cleaning system by extending the tubes a minimum of 3/8" beyond the tube sheets. t The NRC noted that the unit was reassembled and fabricated in . acco."ance with S&W Specification SN1-223 with surveillance , j inspection recorded for nine (9) attributes on the " Condenser ' Tube Installation Inspection Checklist." The NRC noted further that the increased number and decreased span of the tube support l . plates was recognized early as potentially producing a tube ! insertion problem prior to the tubing of the unit. This fact was documented in meeting minutes on the subject " Erection of the Surface Condenser" dated November 13, 1975 with LILCO, S&W, I IR and REGOR (the condenser erector) in attendance. It was ' recognized at this point that the tubes would be scratched during the tubing operation and a maximum 0.004" scratch was permitted. S&W indicated that the 0.004" was acceptable based

on equivalency to the current SB338 NDE Calibration Standard (Paragraph 10.2.1.2) of 0.004" and the UT Calibration Standard in the tubing Purchase Order Specification SH1-299. As indicated

! in an S&W 1etter dated July 22, 1976, a series of calibration comparison scratches were prepared for use during tubing in order to the able to determine acceptable scratch depths. The NRC'noted that as anticipated, there was a tube insertion problem due to alignment difficulties. This fact was acknowledged in the " Condenser Tube Installation Inspection Checklist" review which indicated that tubes with scratches exceeding 0.004" were removed from the condenser and scrapped. A detailed examination of these records indicated that the A-2 section of I f k 47

the condenser was the most difficult to tube. S&W stated that the unit was tubed from the outlet side and consequently the outlet tube ends extending into the water box for the A-2 section of the condenser were visually examined by NRC investi-gators. Only minor inconsequential nicks were found at the ends of these tubes. S&W stated that no tube ends had required , refoming for the insertion of the closely fitting rolling tool, a further indication that the tube ends were not deformed during tubing. Tubes accessible for visual examination on the shell side of the A-2 section of the condenser were also visually inspected by the NRC investigators. Only minor scratches were observed and these appeared to be in the 0.002" dimensional range.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation insofar as the tube insertion problem had been anticipated and was performed in an acceptable manner without any significant damage to the tubes. The NRC noted that a tube failure would not normally result in a radioactive spill. The condenser is under vacuum, and a leak would result in water from Long Island Sound leaking into the condenser rather than radioactive water leaking out.

23. Allegation No. 23

a. Allegation .

               ,                Radiographic tests revealed that the longitudinal seas welds for the condenser box were improperly done and wnen opened for reworking, it was found that the welds often contained dirt, rubbish and weld rod stubs. It was alleged further that these conditions were sometimes discovered when the condenser had to be cut open to correct misalignment problems.

b. MRC Investigation Findings The NRC noted that the condenser is not part of any safety-re-lated system and is not clasified as an ASME Code Vessel as the shell side is under vacuum. The unit was reassembled on site by a LILCO constructor (REGOR) in accordance with S&W specifi-cations and with S&W performing surveillance quality control.

The specification requirement for nondestructive examination of the welds was for visual examination only. a ,

V l The NRC was informed through discussions with the S&W engineer responsible for condenser fabrication that as a result of the l Field QC questioning the achievement of full penetration welds, "information only" radiographic and ultrasonic examinations of the welds were performed. The results of these exaainations confirmed the lack of full penetration welds. Field QC reported that the radiographic examination identified welding technique problems including indications of slag, porosity and lack of fusion, but showed no indication of foreign objects such as i rubbish and weld rod stubs. The NRC noted that an S&W 1etter (SNPS No. 3850) dated April 19, 1976 describes the welding deficiencies and S&W E&DCR No. F-5104, dated November 5,1976 submits the welding nonconfor-mances to Engineering for review. Resolution of the E&DCR required that all accessible welds be air carbon are back i gouged and back welded. Inaccessible welds were strenghtened j by the welding of a stiffner to the face of the weld in order i to provide the equivalent weld strength. Conditions of this i type were not found when misalignment problems were corrected I although in certain instances arc back gouging and back welding ' were performed as part of the realignment of installed partial l' tube support plates. The NRC noted further that in the majority of cases involved, the orientation of the welds would have inhibited the inclusion of any foreign objects. l

c. NRC Conclusion '

                                                         ~

The NRC found no evidence and/or information to substantiate. 1 this allegation. l

24. Allegation No. 24

a. Allegation i l

Misalignment of the condenser tube support required rewelding so often that in some cases, the " mother material" around the weld had to be cut out and replaced with a fresh substitute section.

b. MRC Investigation Findings The classification, fabrication and field assembly of the condenser have been described in Paragraph 22 of this section, together with the tubing changes and the installation of addi-tional partial tube support plates. Discussions by the NRC with S&W personnel and a letter dated October 14, 1975 by Ingersoll-Rand (IR) the condenser fabricator to their site erection supervisor, established that the partial tube support 49 -

plates do not provide any structural support for the condenser box. These support plates are installed in order to mitigate the flow-induced vibration of the titanium condenser tubes. ' The NRC noted that the partial tube support plates were cut

                ,,             into pieces at the shop, reassembled in the field and installed by welding them to support hangers which spanned the distance     l between the original tube support plates. The difficulty encountered in the tube insertion revealed some misalignment problems between the original tube support plates and the additional partial tube support plates. The partial tube support plates were then cut from the support hangers, recut into sections and reassembled in order to improve the alignment.

In order to facilitate the reassembly of the partial tube , support plates and its attachment to the support hangers, flat plates, similar to stiffner/ backing straps were installed for ease of welding. The NRC noted that this practice is in ac-cordance with standard practices for the industry. The NRC concluded that it was the stiffner/ backing strap-like plates that were mistaken as the " fresh substitute section" referred to in the allegation. The NRC noted no unusually large amount of rewelding associated with the realignment of the partial tube sheet support plates. The flat plates were installed to facilitate reinsta11ation, the work was approved by engineering and inspected.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation. '

25. Allegation No. 25

a. Allegation Welds are marked by quality control before welding in order that the best welders can be assigned to these jobs. This is done to assure that they pass examination. Other welds are made by lesser qualified workers and never inspected, implying a degradation in overall quality and resultant safety factors at the Shoreham site.

b. NRC Investigation Findings The NRC noted that welds in safety related systems are required by the NRC to meet varying levels of quality based on their importance to the safety of the reactor. These welds are

! so

_ _ _ _ _ ~ l l l required to conform to the applicable codes and standards. For example, at the Shoreham site, all safety-related pipe welds must meet ASME III and ASME IX Codes. The reactor coolant pressure boundary must meet ASME III Code, Class 1 by the NRC regulatory requirement. Welds of lesser sensitivity are permitted to meet ASME III Code, Classes 2 and 3, again, based on their relative importance to safety. The ASME III Code specifies what testing methods are to be used for each class of weld; surface examinations for Class 3, radiography for Class 2 and a combination of these for Class 1 ' welds. The execution of these welds is controlled by an approved Nuclear Quality Assurance Manual which directs the project engineer to predetermine the inspection requirements for each weld on the component checklist. These checklists are posted at each weld where work is being performed, providing an addi-tional indication of the classification of the weld involved further insuring that the more sensitive welds receive appro-priate attention as required. The NRC noted that welders working on safety-related pipe are qualified to the ASME IX Code. This code only recognizes

                           " qualified" or " unqualified'? welders with no grading system within the " qualified" classification. The assignment of the more skilled welders to the higher classification of welds would mean that the best craftsmen are working on the more serisitive jobs, a process consistent with good practice and j                           generally enhancing the safety of the general public.

The NRC observed that if the inspection requirements were performed on a random sampling basis rather than the system described above, then the activities as alluded to by the 1 allegation might have the potential of affecting the overall quality of the job by alerting individuals to those safety-re-lated welds to be " randomly selected". Since all safety-related welds are inspected, this concern is not justified. l

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegatic.. ,

26. Allegation No. 26

a. Turbine foundation bolts had been installed so far out of alignment that it was necessary to chop out concrete around the bolts, heat the bolts and bend them into a "Z" shape in order to fit them to the foundation plates.

51

\'.

b. NRC Investigation Findings NRC investigators interviewed S&W construction personnel and GE Installation and Service Engineering Department (GE I&SE) employees who were responsible for the turbine installation.

The NRC also examined relevant GE procedures for the setting of the foundation plates. These procedures require that 24 to 3 inches at the surface of the concrete foundation be chipped out prior to setting of the foundation plates in order to permit packing of grout under the plates. GE representatives explained 1 that the centerline of the turbine must be aligned within hundredths of an inch. In order to meet this requirement the bolts were installed within sleeves so that they are free to move within the sleeve once the concrete has set. This is a conventional method of setting anchor bolts. The GE and S&W personnel who were responsible for establishing the alignment and for setting the foundation plates all stated that there had been no problem with the alignment of the bolts. The only problem encountered had been the cleaning out the space between the bolts and sleeves to pemit adjusting the bolt location within the sleeve. NRC investigators examined slides showing the actual chipping of the concrete in progress and the installation of the foundation plates. No significant alignment problems were identified in these slides. The NRC noted that the planned chipping away of 2% to 3 inches of concrete was an extensive operation and may have created the impression of improper alignment which was responsible for this allegation. The chipping of the concrete I was performed by laborers who would not be fully cognizant of the foundation plate installation procedure further leading to the misunderstanding.

c. NRC Conclusion The NRC found no evidence and/or infomation to substantiate this allegation.

27. Allegation No. 27

a. Allegation When concrete was placed in a cold joint on the 63' level of the Reactor Primary Containment, a large amount of rubbish and trash was permitted to remain within the form and the concrete placed on top of it.

k . 52

  '                             NRC Investigation Findinas b.

NRC investigators reviewed concrete inspection records and N&D reports of concrete placements within the reactor building. There were no records of any cold joints in the Reactor Contain-ment Wall but N&D No. 594 identified a cold joint in the shield wall at the 63' level. A preplacement inspection of the area had been made and documented prior to the placement of additional concrete and no irregularities were noted. The NRC questioned several S&W QC inspectors and construction engineers as well as DRAVO craft and supervisory personnel concerning a cold joint in the Reactor Primary Containment Wall.. Some of the individuals questioned mentioned one or two cold joints, including the one in the shield wall but were unanimous in stating that there were no cold joints in the l Primary Containment Wall at or near the 63' level. The NRC noted that the requirements for preplacement inspections prior to the placing of concrete above a cold joint are the same as for any concrete placement. A review of preplacement records by the NRC and interviews with several S&W and craft personnel identified cleanliness as an item of major importance during preplacement inspections to the extent that craft personnel were frequently delayed in placing the concrete in order to satisfy QC cleanliness requirements. The NRC acknowledged that while it was probable that the alleger mistakenly identified  :

                     .           the shield wall cold joint as being in the primary containment wall.there was nothing to support his allegation with respect           1 to the lack of cleanliness.                                             '

c. MRC Conclusion The NRC found no evidence and/or information to substantiate this allegation. f

28. Allegation No. 28

a. Allegation Soil percolation test results were falsified and test results were withheld from LILCO's submission of this information to the NRC.

b. NRC Investication Findings I

Due to the nonspecificity of this allegation, the NRC examined two (2) areas which appeared to have been responsible for the allegation being made. With respect to the soil percolation ( 53 _. _ , _ O

o. t tests themselves, the NRC noted that these tests are utilized to size and locate septic systems and are not used with respect to any structural design considerations. The NRC interviewed l LILCO representatives with knowledge of this area and noted i that the installation of the Office and Service Building septic system coincided with investigations into the potential for soil liquifaction under the service water system. On one L occasion during the soil liquifaction boring program, and subsequent to a heavy rainfall, the building service engineer

asked the geotech engineer to take borings in the vicinity of the septic tanks to find out why they would not drain. Three

                                 '(3) borings were taken and given to the building service engineer, but the logs /results of these borings were not included in the liquification information for the FSAR. The NRC determined that this request to withhold the nonrelated information might have been misconstrued as a willful attempt to withhold infor-nation from the NRC.

The other area evaluated by the NRC investigation was the soil liquifaction work itself. The NRC noted that this matter was , covered in the licensee's FSAR (Section 2.5) and Appendices 2A, 21, 2J, 2L and 2M thereof and basically involved tests to obtain additional information on the soil conditions in the vicinity of the various components making up the service water system in order to determine the liquifaction potential in the event of a design basis earthquake. The various components making up the service water system included the service water  !

                 .                piping between the Reactor Building and the Screenwell, the Screenwell, the wingwalls adjacent to the Screenwell and the intake Canal itself. During the investigation, the NRC investi-gators did not evaluate the results of these tests as this had already been performed by the NRC as part of the overall licensing process.

The NRC did note, however, that the licensee's response to Request 324.5 of the FSAR provided LILCO's response (March 1976) as to why test boring data had been changed from the PSAR results. The investigators noted further that the changes primarily appeared to be minor accidental omissions of data which occurred when unnecessary data was intentionally removed. The NRC investigators found no intentional attempt on the part of the licensee to withhold information from the NRC, although someone not familiar with the matter again may have misconstrued it as such.

c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.

( . i

l . . . .

1

        .-                                                                                                    I iI                             29. Allegation No. 29                                                          ;

a. Alleg~ation Welder performance qualification records were postdated for l welders who qualified after performing welds for which they had not been qualified.  ;

b. NRC Investigation Findings i t

NRC investigators reviewed 1300 COURTER and CO. Nonconformance l Reports (NRs) for indications of welding by improperly qualified welders. One (1) NR was identified which showed welds performed by an unqualified welder. NR-837 dated May 22, 1979, stated that in accordance with the Procedure Qualification Test Method ) (PQTM), the welder was required to qualify to Performance Test (PT) C-1 in order to weld a P-8 to P-1 material socket fillet weld in accordance to Weld Procedure Specification (WPS) 08011AA. 1 The NR was dispositioned to " accept as is" on June 11, 1979 i since the welder was qualified to (PT) C-11 for welding P-8 to P-8 material. ASME Section IX, QW 310.5 permits a welder qualified to weld P-8 to P-8 to also weld P-8 to P-1. Since  ! ASME Section IX is the authority for welder performance qualifi-cations, this permitted the welder to be authorized to weld on the P-8 to P-1 joint in question. The exception to the PQTM was authorized as a specific case and the welder was not added to the list as having passed the (PT) C-1 test'. The NRC identified no problem in the disposition of this NR. ~ NRC investigators also interviewed the UNICO individual responsi-ble for maintaining the listing of qualified welders who stated i most emphatically that there had not been any postdating of 1 welder performance qualifications. Although stating that an

)                                         authorized mechanism existed for post qualifying a welder in accordance with an E&DCR*, that this mechanism had never been
 ,                                        utilized at the Shoreham site and in any event would not involve postdating any existing records. To do so, she stated, would require (in accordance with QAP-7.2) postdating and signing authorizations on five separate documents prior to inclusion on the qualified welders list and then the postdating of the qualified welders list itself. The NRC noted that such actions would be easily noticed if attempted.

c. NRC Conclusion The NRC found no evidence and/or information to support this allegation.

J

      ,                 "This mechanism is different from the disposition of the Courter NR as described      '
        \                 earlier. In this later case, the welder would have to be actually requalified i

on the weld type in question and if he could not be, the weld would be removed from the system and performed by a qualified welder. 55 l

s l [' 30. Allegation No. 30

a. Allegation i Pressure was applied to construction workers by LILCO, its subcontractor and/or related construction unions in order to prevent and/or discourage workers from coming forth to identify construction defects and/or irregularities to the NRC.

b. NRC Findings q

In order to determine the validity of the allegation, NRC investigators provided several pathways by which current or former construction workers could contact the NRC with infor-nation in this regard specifically (i.e., intimidation) or or with respect to construction defects in general. Tae results of these efforts are reviewed below. i) Public Notices From January 3, 1980 through the end of the investigation on site on March 12,1980 (a period of 70 days) notices were placed at several frequented locations at the Shoreham site. This fact was confirmed by the NRC Resident Inspector and NRC investigators on site throughout the course of the investigation. A sample of. this notice has been attached as Exhibit B of this report. The notice provided a minimum of three (3) points of contact with the NRC, one o~f which was manned on a 24-hour basis. The 24-hour number was also made public via a local newspaper, covering therefore, off-site and/or for1eer employees still remaining in the area. In each case, full confidentiality was offered and workers had the option to call anonymously with their concerns if they so desired. During the 70 day period that this notice was published, the NRC received two (2) phone calls. One of the phone calls was from a construction worker at the site who stated that he had worked at other non-nuclear power plants and that by comparison, Shoreham was "over-designed" and " super-safe". The other caller was a member of the general public who had concerns about nuclear power in general and specifically Shoreham because he lived in the area of the plant. Although his concerns were addressed, he provided no information related to a construction defect at the Shoreham site. l \ .,. 56

Interviews of Construction Workers On-Site 11) , .During the course of the investigation, several workers were interviewed. All of the workers were interviewed in private, some during the investigation of the other al-legations, some specifically selected at random in order to detemine the validity of this particular allegation. In the latter case, site representatives for various major l subcontractors on site were contacted by the NRC and asked ]' to provide a list of all current employees to NRC investi-gators. The investigators then selected at random a sampling of carpenters, laborers, masons, welders, electri- , cians, boilermakers, teamsters, millwrights, insulators,  ! steamfitters, sheetmetal workers, weld rod control clerks ' I and QC inspectors. Through the sampling process, shop stewards as well as workers were also selected. Each 4 individual was (a) interviewed in private, (b) informed of l confidentiality, (c) provided with information for a j callback to the NRC if desired and (d) was allowed to express his concerns in any area (i.e., construction defects) in addition to the specific area of intimidation i as stated in the allegation. l 1 Every individual interviewed informed the NRC that none of i the workers had been directly or indirectly intimidated by any individual, union, contractor or licensee, in an attempt to prevent them from coming forth to identify I construction defects and/or irregularities to the NRC. i Each individual further informed the NRC that they had

neither observed nor heard of such actions being executed 1 on any of their friends or other employees not directly i questioned. The NRC did not receive any callbacks at

!                                                other times from the individuals interviewed in this regard.

I Several of the workers interviewed volunteered statements

;                                                 to the NRC to the contrary of this allegation stating that l                                                 if anything, the opposite of the allegation was true.

Some of the statements made are included as follows: Steamfitter A stated: "We never took a shortcut. This is a l Class A job. Our relationship with LILCO is good and they 4 cooperate fully with our concerns."

Mi11 wright A stated

"Our local is very conscientious with i

nuclear energy. This (RCI) is the most conscientious . outfit I've worked for."

      '\_

57

Boilermaker A stated: "The union would be very supportive , l ( of our complaints."

 ,                     Boilermaker B stated:      "We've done our job right to the     l specs."                                                         l Boilermaker C stated: "If we ever found a problem it would be taken care of. They (LILCO) check out everything."

Electrician A stated: "Just the opposite. Everyone is concerned with QC. I've seen some of the better workmanship in the trade on this site." Electrician B stated: "I haven't seen any problems in 5 years." Electrician C stated: "No threats whatsoever. They (union a.ui contractors) promote safety, especially on a nuclear job." Electrician D stated: "We are told to bring up safety-re-lated issues. Everyone is doing a Class 1 job. We are all going to suffer if something happens (when the plant is operational) so we are all supersafe about our jobs." Millwricht D stated: "Just the opposite. We are all told l I to look for problems. I haven't seen any so I don't know how they'll handle it."

                     ' Boilersaker D stated: "I would go to the NRC without hesitation. I live in the area and I wouldn't want anything to go wrong."

Sheetmetal Worker A stated: "They (foreman) tell us to report everything suspicious to them. I wouldn't be afraid to speak up." Sheetmetal Worker B stated: "The company here (LILCO) is far superior to any other company I've ever worked for." Sheatmetal Worker C stated: "Moone is afraid to come forward. In my opinion the job is being done better than it has to be." Sheetmetal Worker D stated: "Something like that (threats) would get around fast, therefore no way they would do it. This job is going slow because all the safety checks you have to go through." v_ 58

s. -

Sheetmetal Worker F stated: "I'm a shop steward. We watch who we send out there because it is a nuclear job. If they don't want to give quality work, they won't be kept here. I'm a shop steward and I have no fear of coming forth to the NRC."

                         .                 Carpenter B stated: "I never saw a better job in my life and I've been in concrete for 25 years."

l Mason A stated: "This is one of the best built jobs I've worked on." Mason B stated: "I've been encouraged to bring problems to management. There would be more white hats (QC, S&W, NRC) than we knew what to do with. The QC here is tough." iii) Resident Inspector ' NRC investigators were informed by the NRC Resident Inspector that during his entire assignment at the Shoreham site (from October 1, 1979) even prior to the initiation of the investigation, no workers at the site (licensee or contractor) had ever approached him with any form of problem relative to construction defects and/or pressure not to present their concerns to the NRC. The Resident Inspector stated

   .                                       that he had made several tours of the site during this                                         -

time, not only during normal working hours, but also during off-hours, weekends an~d holidays, and that to date  ; he had yet to be approached in this regard. iv) Protective Agreement i Attorneys for various allegers claimed that several other ! workers wished to present information on construction defects but would not do so without a fomal protective agreement provided by the NRC in order to ensure their confidentiality. A draft of this agreement was presented l by the attorneys to NRC investgators on December 17, 1979. 1 This agreement was forwarded to NRC Headquarters for legal review and subsequently sent to the allegers' attorneys for implementation. When no new workers came forth even 4 with the protective agreement, the NRC investigators inquired as to the reason. The attorneys stated that they (attorneys) had decided not to risk the lives of these 1 additional allegers due to potential threats to their jobs ) and/or persons. t 59 l

4. r t

    .             The NRC could not reconcile this statement insofar as it had                   {

noted in i), ii) and 111) above that no evidence of any threats  ! whatsoever could be substantiated during this investigation. '

c. NRC Conclusion The NRC found stantiate no evidence and/or information that would sub-this allegation. )

                                                                                              .i ,

l I 60

NUCLEAR REGULATORY COMMISSION N .q k RESICN I g "j s21 PARK AVENUE

        #                    xiso ce rnusstA PsuusytvAniA is4os N0TICE The U. S. Nuclear Regulatory Comission is conducting an investigation into alleged improper construction practices which may have been undertaken at the Shoreham site. Any worker having information concerning these allegations or other concerns is urged to contact the NRC (collect) at 215-337-5000.       NRC investigators will be on site from February 11-15,1980*and may be contacted through the NRC Resident Inspector (extension 83-221) or at the Holiday Inn at Riverhead, New York, (516-369-2200). The NRC will maintain the confidentiality of the source of any information received during this investigation and in any resultant reports.
    *new notices were posted prior to each on-site visit and the appropriate dates inserted at this location

\

      +

f LE s

      . .p+n y,        uc ,70.,                                  UNITED STATES

{5 , .< g

                   , E NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
                   /

EDO PRINCIPAL CORRESPONDENCE CONTROL

                              --- _               ----                = = -_ = - - - - - - - - =___ _ - __

FROM: DUE: 07/30/85 EDO CONTROL: 000829 DOC DT: 07/08/85 JOSEPH I. LIEBERMAN, ATTORNEY GENERAL FINAL REPLY: STATE OF CONNECTIENT TO: CHAIRMAN PALLADINO FOR SIGNATURE OF: ** GREEN ** SECY NO: 85-559 MURLEY DESC: ROUTING: SEGMENT OF "60 MINUTES" DEALING WITH ALLEGATIONS DENTON OF POOR WORKMANSHIP AND ORGANIZED CRIME KERR,SP INVOLVEMENT IN CONSTRUCTION OF SHOREHAM TAYLOR DATE: 07/16/05 ASSIGNED TO: RI CONTACT: MURLEY ,

                                                                                                                                            )

SPECIAL INSTRUCTIONS OR REMARKS: 4 4. Q h

f , , Joseph Lieberman [ CORRESPONDENCE CONTROL TICKET SECY NUMBER: 85-559 LOGGING DATE 7/16/85 0FFICE OF THE SECRETARY ACTION OFFICE: ED0 AUTHOR: Joseph Liberman AFFILIATION: Attorney General--Conn LETTER DATE: 7/8/85 FILE CODE ADDRESSEE: Palladino

SUBJECT:

Segment on 60 minutes dealing with allegations of organized crime, etc; at Shoreham ACTION: Appropriate DISTRIBUTION: RF, Docket SPECIAL HANDLING: None SIGNATURE DATE: FOR THE COM ISSION BAC Rec'd Off. EDO} la 'W ( - Date. . .) . ~, Time. . ./. :#:"" ::"-- EDO --- 000829}}