ML20054K133
| ML20054K133 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 06/28/1982 |
| From: | Bordenick B NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Brenner L, Carpenter J, Morris P Atomic Safety and Licensing Board Panel |
| References | |
| ISSUANCES-OL, NUDOCS 8207010168 | |
| Download: ML20054K133 (184) | |
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Lawrence Brenner, Esq. Dr. James L. Carpenter Administrative Judge Administrative Judge-Atomic Safety and Licensing Board Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 Dr. Peter A. Morris Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear RegulatoryCommission Washington, D.C. 20555 In the Matter of Long Island Lighting Company (ShorehamNuclearPowerStation, Unit 1)
Docket No. 50-322 (0L)
Dear Administrative Judges:
Enclosed are copies of the following documents which the Staff has prepared pursuant to Board requests in this proceeding:
NRC Staff's Response to Certain General Concerns Raised by Some Limited Appearance Statements Status Report on Shoreham SER Open Items NRC Staff Response to Board Unresolved Safety Issue Questions Sincerely, Bernard M. Bordenick Counsel for NRC Staff
Enclosures:
As stated cc: See page two 8207010168 820628 PDR ADOCK 05000322 Q PDR
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cc: (w/ enclosures)
Matthew J. Kelly, Esq.
Ralph Shapiro, Esq.
Howard L. Blau, Esq.
W. Taylor Reveley, III, Esq.
Stephen B. Latham, Esq.
John F. Shea, III, Esq.
Atomic Safety and Licensing Board Panel
- Atomic Safety and Licensing Appeal Board Panel Herbert H. Brown, Esq.
i Lawrence Coe Lanpher, Esq.
Karla J. Letsche, Esq.
Docketing and Service Section Edward M. Barrett, Esq.
Mr. Brian McCaffrey Marc W. Goldsmith
- -David H. Gilmartin, Esq.
Mr. Jeff Smith MHB Technical Associates Hon. Peter Cohalan Mr. Jay Dunkleberger i
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' NAME :B.Bo nick I/s :E. Reis : : : : :
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DATE:06/))/82 : :Ob/j982 : : : :
I NRC STAFF'S RESPONSE TO CERTAIN GENERAL CONCERNS RAISED BY SOME LIMITED APPEARANCE STATEMENTS The Board at Transcript pages 1156 et seq. requested Applicant and the NRC Staff to respond to certain generalized concerns raised during the course of certain limited appearance statements made at this proceeding. Set out below are the Staff's responses.
A. Size of the Containment It was asserted that the power level of the Shoreham unit was raised from 440 megawatts to 820 megawatts, but the suppression pool size was not changed and its volume is considerably les's than some other similar plants. (Tr.1157, citing Tr. 557 and 918)
Staff Response (Prepared by F. Eltawila)
The suppression pool is relied on for condensing the steam that is channeled into it during a loss-of-coolant accident and/or the discharge of one or more safety / relief valves (SRV). To assure these functions and to provide the necessary water source for ECCS, the Staff conducted its review with the following objectives in mind:
- 1. Review of the analysis and conservatism employed to determine the pressure response of the containment during a LOCA;
- 2. Review of the analysis and assumptions used to calculate the suppression pool temperature response during a SRV discharge to establish limits for the Technical Specifications beyond which the plant should not be allowed to operate;
- 3. Evaluate the adequacy of the instrumentaion installed to determine the suppression pool water level and temperature;
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- 4. Evaluate the RHR system design, including strainer sizing, suction point evlevation, compliance with the single failure criterion, leak test requirements and NPSH calculations.
The attached table shows comparisons between the Shoreham containment design parameters and other Mark II containments.
It includes parameters which affect the pressure suppression response, the energy sources, net free volume, pool water volume and active heat removal systems.
As shown in the Table, the Shoreham rated thermal power and the energy contained within the primary system are lower than those of LaSalle County Station (LSCS). It is also clear that net free volume of Shoreham (SNPS) is less than LSCS; this results in vent cleaning and containment peak pressure to occur much sooner for SNPS. In addition, for the long term enugy removal, the SNPS pool is about 60%
that of LSCS; again as shown in the Table, the 2nd containment peak pressure and peak temperature occur much faster than for LSCS.
Staff has performed independent studies for both LSCS and Shoreham.
Staff's analyses which utilize conservative assumptions for both plants, indicate that the containments are capable of accommodating the l
pressure and temperature resulting from a LOCA event.
It is Staff's judgement that the Shoreham containment, which has a higher design pressure to account for the reduced drywell and wetwell volumes relative to the other Mark II plants, is designed to meet the 1
j same acceptance criteria as for the other Mark II containments.
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TABLE COMPARISON OF MARK II CONTAINMENTS Shoreham 'Zimmer LaSalle WNP-2 Rated Thermal Power MW 2550 2430 3293 3324 4
design pres. psig 48 45 45 45 1 Peak Pres. at (sec) 9.8 20.1 Supp. Pool Volume Ft 3 76,839 97,680 128,800 144,184 Energy Contained within Primary System 437+6 BTU 684+6 BTU
- Total Energy Released at time of Peak Pressure 650+6 1139.5+6 4 (14,000 sec) 0 (27,000 sec)
Energy Removed by Hx 315+6 BTU 752+6 BTU Pool Peak Temperature 198 F 200 F
, (15,750 sec) (36,000 sec) 4 l
l B. Artesian Well or Foundation Boil Certain concerns were expressed by Judge Morris about the effects of the foundation boil that was encountered during excavation for the reactor building in 1970. Referring to the NRC Staff conclusion stated in the SER that it was unlikely that foundation materials would liquefy under the SSE, Judge Morris wanted to get a better quantitative handle on what "unlikely" meant. He also wondered if there were "other factors that might be introduced into the performance of the structure from the occasion of this alleged artesian well or foundation boil." (Tr.
1157-59; citing Tr. 614).
Staff Response (Prepared by Raman Pichumani)
Liquefaction of Foundation Materials The Applicant has documented in Appendix 2 J, FSAR, Vol. 4, Rev. 7, that the artesian well had been effectively sealed off by pumping under pressure sufficient quantity of cement grout into the hole. The l
Applicant had also determined the relative density of the foundation soil in the immediate vicinity of the sealed well by drilling new borings before placing the concrete foundation mat.
In response to an NRC Staff request, the Applicant performed a liquefaction analyses by two different techniques. The results of these analyses showed that, even with the somewhat reduced relative density of the sands caused by the foundation boil, the foundation soils would not be expected to liquefy under the SSE. Based on a statistical analysis of
liquefaction reported in the published literature *, the NRC Staff estimates that the probability of error in predicting no liquefaction of the subsoil at Shoreham (with a safety factor close to unity) is about 10% to 15% if an SSE happened to occur. Studies performed for reactor sites in various parts of the U.S. have shown that return periods for the SSE typically are in the order of 1000 year or 10,000 year. Thus, taking a conservative return period of the SSE as 1000 years, the probability of liquefaction is less than 10-4 per year; an "unlikely" occurrence.
Other Factors that Might Affect Performance of the Structure The following are some other parameters that obviously come to mind, and that have to be considered while evaluating the effects of the foundatin boil on the performance of the structure:
- a. Settlement of the Reactor Building,
- b. Tilt of the Reactor Building.
- c. Differential Settlement across the foundation mat.
- d. Changed Soil-Structure Interaction Effects.
- a. Settlement The total measured settlement of the Reactor Building as of September 1981 (about 11 years after construction of the foundation mat) is about 3 inches. This is slightly larger than the Applicant's predicted 40 year settlement of 2.7 inches for the same
-*/ Christian J.T., and Swiger, W.F., " Statistics of Liquefaction and SPT Results", ASCE Jour. Geotech, Eng. Div. , Vol .101, No. GT 11, Nov. 1975.
o location of the building, but is not a cause for concern at this time. The Applicant has verbally reported that all of the anticipated loading has been imposed on the foundation since a year aga. The Applicant will be asked to docket the latest settlement data (since December 1975) along with the loading history. A review of this new information will indicate if the settlement of the structure is indicative of any significant concerns.
- b. Tilt of the Reactor Building The recorded settlement data as of September 1981 indicate no significant differential settlement (Tilt) of the reactor building.
This too will have to be updated and verified after reviewing the latest settlement data recorded after the full loading has been imposed on the foundation.
- c. Differential Settlement Across the Base Mat Since the foundation mat is a very rigid structural member resting on a deformable soil material, no differential settlement across the foundation mat is expected due to the possible nonuniformity of the subgrade support caused by the foundation boil.
- d. Soil-Structure Interaction Effect The reduced relative density of the soil surrounding the
" foundation boil" area will have some effect on the shear modulus of the affected soil which, in turn, will affect the results of the dynamic SSI analysis. This has probably been considered by varying
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., the shear modulus values by 25% or so, as is usually done in such analysis. The relevant section in the FSAR does not, however, indicate if such variation was considered while determining the soil spring constants. The Applicant will be asked to furnish this information when it dockets the settlement data.
Summary Based on the results of statistical analysis of liquefaction at a number of sites reported in the published literature, the NRC Staff estimates that the probability of staff error in predicting that no liquefaction will occur in the subsoil at Shoreham is about 10 to 15 percent if an SSE occurs. A preliminary review of the recorded settlement data indicates that other factors such as settlement and tilt of the reactor building, and SSI prediction are not likely to have a significantly adverse effect on the performance of the structure. This latter conclusion wil have to be confirmed after reviewing the latest settlement data and the details of SSI analysis to be furnished by the Applicant.
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C. Allegations Relatina Tc Supposed Cracking in the Drywell Head Flange (Tr.1159; citing Tr. 614)
Staff Response (Provided by Region 1)
Inspection Reports 74-08 and 75-01, copies attached deal, in part, with the discovery of the Head Flange Cracking Problem as a reportable 10 C.F.R. 6 50.55(e) deficiency. The resolulation of the problem is addressed in Inspection Reports 75-04 and 75-08 copies of which are also attached.
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,, KING oF PRUSSI A, PENNSYLVANIA 194o6 OCT 241974 Long Island Lighting Company License No. CPPR-95
- Attention: Mr. Andrew W. Wofford Inspection No. 50-322/74-08 Vice President 175 East Old Country Road Hicksville, New York 11801 .
Gentlemen:
This refers to the inspection conducted by Mr. Narrow of this office on October 7-11, 1974 at Shoreham, New York of activities authorized by AEC License No. CPPR-95 and to the discussions of our findings held by Mr. Narrow with you and members of your staff at the conclusion of the inspection.
I Areas examined during this inspection are described in the Regulatory Operations Inspection Report which is enclosed with this letter.
Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspector.
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During this inspection, it was found that certain of your activities appeared to be in violation of AEC requirements. The items and references to the pertinent requirements are listed in the enclosure to this letter. This letter constitutes a notice sent to you pursuant to the provisions of Section 2.201 of the AEC's " Rules of Practice", Part 2, Title 10, Code of Federal Regulations.
Section 2.201 requires you to submit to this office within 30 days of your receipt of this notice, a written statement of explanation in reply, including: (1) corrective steps which have been or will be taken by you, and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved.
As indicated in Enclosure No. I to this letter, we note that you have corrected Item No. 1, and therefore, you need not address yourself to this matter in your response.
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l In accordance with Section 2.790 of the AEC's " Rules of Practice",
Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the AEC's Public Document Room. If this report contains any information that you (or your contractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this office to withhold such information from public disclosure.
Any such application must include a full statement of the reasons on the basis of which it is claimed that the information is proprietary, and should be prepared so that proprietary information identified in the application is contained in a separate part of the document. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely,
/f<.th bbert T. Carlson, Chief th W Facility Construction and Engineering Support Branch
Enclosures:
- 1. Description of Violations
- 2. RO Inspection Report No. 74-08 cc: Thomas J. Burke, Project Manager Edward M. Barrett, Esquire Edward J. Walsh, Esquire bec: (w/encis)
' RO Chief, FS&EB '
RO:HQ (5)
DL (4 w/encls plus 9 cpys of report only)
DR Central Files RS (3)
PDR Local PDR RO Files NSIC TIC State of New York OGC Reg Reg Reading Room
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9 ENCLOSURE DESCRIPTION OF VIOLATIONS Long Island Lighting Company Docket No. 50-322 License No. CPPR-95 .
Certain activities under your license appear to be in violation of AEC requirements.
- 1. The following apparent violation is considered to be of Category II severity:
Criterion VI of 10 CFR 50, Appendix B, states in part, " Measures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality. These measures shall assure that documents, including changes, are reviewed for adequacy and approved for release by authorized personnel and are distributed to and used at the location where the prescribed activity is performed."
Contrary to the above, it was observed that although the A-E's specifica-tion SH1-75 had been revised to include Addendum 3, and that this re-vised specification had been issued to the containment construction contractor and that this contractor was performing his work in accord-ance with this Addendum; the contractors QA manual which ostensibly controlled his quality related activities still identified specification SE1-75, Addendum 2 as the appropriate specification.
Prompt corrective action was initiated and the contractors QA manual ,
was revised to designate Addendum 3 as the governing specification.
No response to this item is necessary.
- 2. The following apparent violation is considered to be of Category III severity:
10 CFR 50.55(e) states in part, "...the permit is for construction of a nuclear power plant, the holder...shall notify the Commission of each deficiency...which were it to remain uncorrected, could have affected adversely the safety of...the nuclear power plant at any time through-out the expected lifetime of the plant and which represents: (iii) A significant deficien:y in construction or significant damage to a structure...which will require... extensive repair..."
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a Contrary to the above the licensee failed to report extensive damage to the removable head of the containment structure due to improper post weld heat treatment of certain weld seams.
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U.S. AT0!!IC ENERGY COFD1ISSION
, DIRECTORATE OF REGUIATORY OPERATIONS REGION I RO Inspection Report No: 50-322/74-08 Docket No: 50-322
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Long Island Lighting Company Licensae: License No: CPPR-95 175 East Old Country Road Priority:
Hicksville, New York 11801 A Category:
Shoreham, Unit 1 .
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Location: Shoreham, New York .
Ty e of Licensee: BWR, 819 MWe (GE) -
Type of Inspection: Routine, Unannounced Dates of Inspection: October 7-11, 1974 l
- Dates of Previous Inspection: August 8, 1974 Reporting Inspector.:
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Lewis Narrowi; Reactor Inspector / Da' te Accompanying Inspectors: / Nom 8/(/[ d O M"27fo7s Glenn Walton, Reactor Inspector .
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- N' de$rge'Napuda, Reactor Inspector ' Date Date
. Date Other Accompanying Personnel:
Date Reviewed By: '
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f- s / M7F R. F. Heishman, Senior Reactor Inspector 'Date
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SUMMARY
OF FINDINGS Enforcement Action A. As a result of improper postweld heat treatment of the top head flange of the containment structure it was necessary to remove material which had been damaged adjacent to four field welds. Re- '
fabrication of a portion of the containment head was required in order to replace the damaged material.
Failure of the licensee to report this problem is considered to be a violation of the requirements of 10 CFR 50.55(e) . 0)atails, Paragraph 2) -
B. Failure to control release of changes to the Pittsburgh-Des Moines (P-DM) QA Manual is considered to be a violation of Criterion VI of Appendix B, 10 CFR 50. (Details, Paragraph 3)
Licensee Action on Previously Identified Enforcement Items A. The failure to conform to the requirements for storage of the reactor pressure vessel has been resolved. (Details, Paragraph 4)
B. Noncompliance identified during a vendor inspection of a pump manu-facturer have been resolved. (Details, Paragraph 5)
C. Thin wall conditions in main steam line (MSL) piping and the failure to report these conditions has been resolved. (D& tails, Paragraph 6)
Design Changes None Unusual Occurrences None _
Other Significant Findings
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A. Current Findings
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- 1. The following items are unresolved:
- a. Lack of a procedure to control filing of site audits. (De tails ,
Paragrah 7)
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- b. An inadequate procedure to control distribution of site audit reports. (Details, Paragraph 8) '
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- c. Incomplete weld qualification records. (Details, Paragraph 9)
- d. Incorrect references in P-DM Construction Procedure Sequences.
(CPS) (Details Paragraph 10)
- e. Failure to change the requirement for x-ray film in appendix B of Specification SH 1-75 in accordance with Addendum 3. (Details, Paragraph 11.a) i f. Identification of incorrect addendum to SR 1-75 as latest issued on S&W specification index. (Details, Paragraph kl.b) 3 Incomplete identification of material on QA Examination Check-lists. (Details, Paragraph 12)
- 2. The following areas were examined with no deficiencies identified except as described above:
- a. P-DM procedures for NDE examinations. (Details, Paragraph 13) l
- b. Selected fabrication shop records and mill certifications for the containment floor, cylinder, cone and head. (Details, Paragraph 14) i c. Selected P-DM field records showing NDE' performance, weld procedure qualifications, welder qualifications and NDE per-sonnel qualifications. (Details, Paragraph 15)
- d. Weld electrode storage and controls. (Details, Paragraph 16)
- e. The A-E's QA and QC procedures for control of containment, penetrations and piping installation. (Details, Paragraph 17)
- f. The containment contractors QC procedures for containment and penetrations installation. (Details, Paragraph 18)
- g. Storage of materials and components by the containment con-tractor. (Details, Paragraph 19)
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B. Status of Previously Reported Unresolved Items The following previously identified deficiencies have been resolved:
- 1. The apparent discrepancy in reports of the examination of main
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steam line piping wall thickness. (Details, Paragraph 20)
- 2. Items identified during licensee audit of NSSS. (Details, Para-graph 22)
- 3. Addition of QA personnel to licensee staff. (Details, Paragraph 21) 4 Pepair of voids in reactor pedestal concrete. (Details, Para- ~
graph 23)
Exit Interview An exit interview was held at the licensee's corporate office on October 11, l 1974 with the following personnel attending: -
Long Island Lighting Company W. O. Uhl, Senior Vice-President A. W. Wofford, Vice-President -
T. J. Burke, Project Manager R. E. Black, Resident QA Engineer l E. W. Tesco, Construction Superintendent F. X. Schoner, QA Engineer -
J. M. Kelly, QA Engineer Stone and Webster
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S. J. Stratis, QA Coordinator F. C. Turner, Superintendent of Field QC T. A. Hill, Resident Manager R. L. Cosick, Superintendent of Construction i
The subjects discussed are summarized below:
A. Scope of Inspection The inspector stated that this was an unannounced inspection for the purpose of examining containment and piping QC procedures; QC records and work in progress on containment and licensee actions on previously identified inspection findings. ,
B. Failure to Report a Significant Deficiency During post weld heat treatment of the removable containment head, portions of four field welds were held at temperature longer than the heat treatment time for which the material was qualified and in some cases above the maximum allowable heat treatment temperatures of the procedure.
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Following a series of tests of the physical properties of the material adjacent to the welds, four sections of plate each approximately five feet wide and eight feet long were removed and will be replaced by a single fabricated plate.
The inspector stated that the licensees failure to report this occur-rence as a significant deficiency is considered to be a violation of 10 CFR 50.55(e). The inspector further stated that this is a recur-rence of a similar violation which had been identified during a previous inspection.
The licensee stated that a significant deficiency report will be sub-mitted. He also stated that its omission was inadvertent and due to the fact that during initial considerations of this incident it had appeared to be of minor significance. As additional tests were per-formed the amount of damage was observed to be increasing but its significance was overlooked. (Details, Paragraph 2)
C. Control of Documents -
The inspector stated that Addendum 3 to S&W Specification No. SH 1-75 had been issued to the site on August 10, 1973 and has been distri-buted shortly thereaf ter including P-DM.
The inspector observed that P-DM was apparently complying with this amendment. However, the P-DM QA manual had not been revised and still identified Addendum 2 of SH 1-75 as applicable.and the P-DM site representative had not been instructed to conform to addendum 3. This is considered to be a violation of Criterion VI of 10 CFR 50.
During the inspection a revision to the P-DM QA Manual was received at the site.
The inspector s'ated t that this item is considered to be resolved and no further action will be required of the licensee. The licensee acknowledged this information. (Details, Paragraph 3)
D. Previously Identified Nonconformances The inspector stated that the previously. identified nonconformances listed below had been resolved. The licensee acknowledged this in-formation. --
- 1. Failure to follow established procedures for storage of the reactor l
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v pressure vessel. (Details, Paragraph 4)
- 2. Failure to report thin wall conditions in main steam line.
(Details, Paragraph 6)
E. Deficiencies Identified The inspector stated that the deficiencies listed below had been identified as unresolved items. The licensee acknowledged this information.
- 1. Lack of a procedure to control filing of site QC working docu-ments. (Details, Paragraph 7)
- 2. An inadequate procedure to control distribution of site audit reports. (Details ~, Paragraph 8)
- 3. Incomplete weld qualification records. (Details, Paragraph 9)
- 4. Incorrect references in construction procedure sequences.
(Details, Paragraph 10)
- 5. Incomplete revisions to Specification Index and Specification SH 1-75. (Details, Paragraph 11)
- 6. Incomplete QC checklists. (Details, Paragraph 12)
E. Items Examined with no Deficiencies Identified -
The inspector stated that the items listed below were examined with no deficiencies identified except as identified above. The licensee acknowledged this information.
- 1. P-DM NDE Procedures. (Details, Paragraph 13)
- 2. Selected fabrication shop records and mill certifications for. the containment cylinder and cone sections. (Details, Paragraph 14)
- 3. Selected CPS and QC daily activity reports for field fabrication of the containment structure. (Details, Paragraph 15)
- 4. Weld electrode storage and control. (Details, Paragraph 16)
- 5. The A-E's QA Manual, QA&QC Manual ASNE Section III and PQC Manual.
(Details, Paragraph 17)
- 6. Storage of containment materials / components. (Details, Paragraph 18)
- 7. P-DM's QC procedures. (Details, Paragraph 19) a v.. - - . . _ - . , . , , - . - - , - - .
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F. Previously Reported Deficiencies The inspector stated that the deficiencies listed beloor which had been identified previously were resolved. The licensee acknowledged this information. ,
- 1. Apparent discrepancy in reports of main steam line examination.
(Details, Paragraph 20)
- 2. Items identified during licensee audit of NSSS. 0)etails. Para-graph 22)
- 3. Addition of QA personnel. (Details, Paragraph 21)
- 4. Reactor pedestal concrete. (Details, Paragrapit 23) e p
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DETAILS
- 1. Persons Contacted Long Island Lighting Company T. F. Gerecke, QA Manager ~
T. J. Burke, Project Manager R. E. Black, Resident QA Engineer J. M. Kelly, QA Engineer (site)
E. Bajada, QA Engineer General Electric Company -
R. A. Park, Site QC Representative Stone and Webster F. C. Turner, Superintendent Field QC S. J. Stratis, QA Coordinator S. A. Kalat, QC Engineer R. Bernard, Assistant Superintendent Field QC W. C. Taylor, QC Engineer J. Davis, QC Engineer C. Embler, QC Engineer, Welding J. Poline, QC Engineer E. A. Mac Dougall, Senior QC Engineer
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A. Gerstenlauer, QC Inspector E. Donegan, Project Manager Pittsburgh - Des Moines Steel Company M. Steiger, QA Manager R. Singer, Weld Control Specialist
- 2. Failure to Report a Significant Construction Deficiency i
During post veld heat treatment of the containment liner removable head by P-DM in January 1974, it was held.at temperature longer than the heat treatment time for which the material was qualified, and in some cases above the maximum. heat treatment temperature specified in the P-DM detailed procedure. Subsequently, test material was removed from the weld seam area which experienced the most extreme conditions
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and tests were performed by Pittsburgh De Moines to determine if the physical properties of the material were acceptable. Certain areas of the weld and adjacent base material failed to meet the tension test and Charpy impact values. Further evaluation and test indicated the need to remove all four welds and base material 2' - 6" on each side of the weld and approximately 8' long to ensure removal of the material that was adversely affected during the post weld heat treat-ment.
The corrective action taken by Pittsburgh Des Moines included replace-ment of the removed four sections, by a fabricated replacement piece, 8 feet high x 20 feet wide and making a fif th weld seam.
This information was reported by P-DM to S&W and the licensee by memorandum dated May 21, 1974.
Failure of the licensee to report this occurrence as a significant deficiency in accordance with 10 CFR 50.55(e) is considered to be l a violation of that regulation. -
- 3. Control of Documents j The inspector observed that the P-DM QA manual stated that it was in-tended to meet the requirements of S&W specification SH 1-75, Addendum 2 and that any reference to these specifications shall include Addendum
- 2. However, ~ the inspector pbserved that in the areas inspected, work
, was performed in accordance with Addendum 3 to SH 1-75. As an example, t
Addendum 2 requires use of type 1 film for RT examination when iridium l
I 192 is used as the source. This requirement is deleted by Addendum 3 and type 1 film was not used. Records showed Addendum 3 to have been received at the site on August 10, 1973, to have been received by P-DM at the site on August 13, 1973 and to have been in use by P-DM on August 15, 1973. Records were not available at the site authorizing use of Amendment 3 by P-DM. Failure to revise the P-DM QA manual is considered to be a violation of Section E-2 of Appendix E of the PSAR since it does not conform to the requirements of Criterion VI of 10 CFR 50, Appendix B which states, in part " Measures shall be established to control the issuance of documents, such as instructions, procedures and drawings, including changes thereto, ..." During this inspection a revision to the P-DM QA manual which identified Amendment 3 of SH 1-75 as the applicable specification was furnished to the site. This item is resolved and no further action by the licensee is required.
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- 4. RPV Storage Failure to follow established procedures for storage of the RPV had previously been identified as a violation of Criterion V of 10 CFR 50, Appendix B. The inspector examined S&W procedure " Site Storage of Reactor Pressure Vessel" Revision 2 dated August 13, 1974 which had been reviewed and approved with comments by the NSSS and forwarded for interim use on September 2, 1974. Revision No. 3 of this pro-cedure which incorporates the NSSS comments is available at the site although not approved.
Ermmination of storage history cards showed that Revision 3 of this procedure was used for in-storage inspection effective September 5, 1974.
No deficiencies were identified during examination of this procedure and its implementation as shown on the storage history cards. This item is resolved.
- 5. Pump Manufacturers Procedures Cleaning procedures and lack of signatures on shop operation sequences had previously been identified as noncompliances to the requirements of 10 CFR 50, Appendix B during a vendor Laspection of Byron Jackson.
Documentation was made available by the licensee which identified the cleaning methods and solutions used and their approval by the NSSS as well as an inspection report stating that a number of route sheets in use had been reviewed and found to have been properly signed.
This report also identified written instructions as having been issued by the plant QC Hanager to each inspector for verification of signa-tures. This documentation was examined following the inspection with no deficiencies identified.
l This item is resolved.
- 6. Thin Walls in Main Steam Line Piping Failure to report thin wall conditions in three sections of main steam line piping had previously been identified as a violation of the re-quirements of 10 CFR 50.55(e). These conditions and the corrective actions taken were reported to the Direct.or of Regulatory Operations by letter dated July 11, 1974.
The inspector examined documentation concerning corrective action as follows:
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- 1. N&D reports Nos. 21 and 62 which identified the deficiency and which was closed out by return of the three sections of pipe to the vendor.
- 2. Returned Material Report No. 147 showing return of the pipe to the vendor.
- 3. Records of UT exmaination of the reworked sections of pipe.
This item is resolved.
- 7. Filing of Site Audits The inspector found that site filing of audits is being accomplished, but no written procedure presently controls this filing. Quality Assurance Directive QAD-17.1 (QA site file) was~ presented to the inspector for review and the licensee's representative stated QC-5.3 (QA site files) was in the process of development.
The lack of this control procedure is considered to be an unresolved item.
- 8. Distribution of Site Audit Reports The distribution of site audit reports is not clearly described in the A-E's Field QC Procedures; specifically FQC-20.1. The inspector was informed by the A-E's representative that a revised procedure is in preparation.
This item is unresolved.
I
- 9. Weld Qualification Records - Containment Dome The inspector audited the material certification records contained in records identified as Plate Assembly Book X, Group NAl for the steel containment dome.
This review included an examination of the records of shop fabrication, l of the individual liner components. The following records were re-viewed, chemistry, Quench and temper, stress relief, ultrasonic, radio-graphy, weld material certification, furnace calibration, welder quali-fication and weld procedure-qualification.
The inspector observed that weld procedure qualification 67-1 fails to indicate the thickness for which the welder was qualified and the time
_11 the test sample was held at stress relief temperature. This item is unresolved. '
j 10. Construction Procedure Sequence (CPS)
The inspector reviewed records for the field fabrication sequence of the containment liner, including references to nondestructive examin-ation procedures. The inspector noted that some nondestructive ex ~
amination references did not reference the latest revision of a pro-cedure.
For example, the CPS states that welds will be radiographed per 17.1.2
! of WPS 42 A and RT-2. The inspector reviewed this reference (WPS 42A) and noted it also references RT-2, with no revision number specified.
I The procedure review indicates there is a revision A to RT-2 dated 2/27/73.
The error in identification of the correct procedure revision on the CPS's is an unresolved item.
- 11. Revisions to Documents -
The inspector examined S&W Specification Index and Specification SH 1-75.
The following deficiencies were identified:
- a. Section cc of Addendum 3 to Specification SH 1-75 has deleted the j requirement for use of Type 1 film for RT examination when Iridium 192 i is used as a source for, material 1 inch or less in thickness.
l However, the Summary of Tests in Appendix B of this amendment had not been revised accordingly. ~
l This item is unresolved.
- b. The Specification Index which is issued periodically to identify the latest issue of specifications did not reflect the release of Addendum 3 to Specification SH 1-75.
This item is unresolved.
- 12. Fabrication Shop QC Checklists The inspector examined material certifications, shop welding records '
and certifications and receiving inspection reports for a shipment of steel floor plates. __
QA Examination Checklist (QAEC) showed plate 10A2R as sent to the Pittsburgh plant of P-DK for repair on April 10, 1972 and then rejected
at that plant for excessive surface defects. QAEC dated July 27, 1973 showed plate 10A2R as accepted. Neither QAEC identified the plate except by number and there was no means of establishing that the accepted plate was other than the one previously rejected.
This item is unresolved.
- 13. NDE Procedures ~
l The inspector audited the following P-DM procedures partaining to the field fabrication and installation of the containment liner.
- a. MT-1 Rev. A -
Magnetic Particle
- b. RT-2 Rev. A -
Radiography -
- c. EL-2 -
Balogen Leak
- d. VB-1 -
Vacuum Box
- e. PT-1 -
No deficiencies were identified. .
- 14. Mill & Shop Cer!;fications .
Material certifications, shop welding records and certifications and-receiving inspection reports for two shipments of containment wall liner plates were examined.
No deficiencies were identified. -
L.
- 15. P-DM Field Records '
l ~
The inspector audited P-DM records on file of several completed welds.'
This review included radiography, magnetic particle, vacuum box and -
visual test results. The review also included welder qualifications,_ ,
NDE qualifications and weld procedure qualifications. s 3
No deficiencies were identified.
- 16. Weld Electrode Control j .
The inspector observed the electrode storage controls for welding-materials used in the f abrication of the containment liner bp P-DM. 3 The inspector noted the 7018, and 8018 weld electrodes were kept in *?,
storage in separate buildings. The buildings wereslocked wher? not
(
under direct surveillance. The inspector reviewed corresphndencs ;,
between P-DM and S&W which stated that no weld electeddes 'are to be,\ 's Ns '
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returned by welders for rebaking or holding oven storage.
Rather the
! system in effect require that welders destroy any 8018 electrode after l they are out of the holding ovens for more :han two hours. The 7018 i electrodes may be out of the holding ovens for periods up to four hours.
, One holding oven was inspected and its temperature was observed by the i inspector to be 3150F. The heat sensor device was tagged as being calibrated 9/5/74 and due for recalibration 12/5/74.
~
l The weld electrode record control form was audited by the inspector and no deficiencies were observed. This form contained the following information; Oven number, oven temperature, type of electrodes material, l welders symbol, size and identification; including heat and lot number
- of electrodes withdrawn, the number of electrodes issued and time of I
issue. No deficiencies were identified.
- 17. A-E's Site QC Program 1
The AE's QA and QC procedures for control of containment, penetrations and piping were reviewed for conformance to the licensees PSAR commit-ments.
- a. Quality Assurance Manual .
The QA manual was examined with respect to:
(1) Provision for independence and freedom of action of QA personnel and a description of their duties, and "
(2) Requirements for audits, documentation of audit activities and submittal of audit reports to appropriate management.
No deficiencies were identified. *
- b. QA and QC Manual - ASME Section III
- Sections of this manual applicable to welding and brazing, fabrica-tion and erection inspection, NDE, calibration of measuring and test equipment and personnel qualifications were examined.
~
No deficiencies were identified.
- c. Field Quality Control (FQC) Manual j' Sections of this manual applicable to document control, receiving 4
and storage inspection, welding and weld material control, pipe '
l N: - . - - - - - -.. - - . . - - - - - _ . . _- -.. .-- .-- . - - - - - -
fabrication, structural steel erection and QC personnel quali-fication were examined by the inspector.
- No deficiencies were identified.
(1) General Welding Procedure W-100 (2) NDT 11.1 (PT)
No deficiencies were identified.
- 18. Storage of Material and Components .
The inspector audited P-DM storage of material / components for the con-tainment liner for conformance with procedures and record keeping requirements.
- 19. P-DM QC Program .
This contractors QC procedures as they were applied to erection of the containment structure and in,stallation of containment pentrations were examined for conformance to the licensees PSAR commitmer.ts. ,
e No deficiencies were identified. I
- 20. Discrepancy between NSSS letter and N&D Report The apparent discrepancy between N&D report No. 62 and NSSS letter FW-T-775 concerning NDE examination of main steam line piping had previously been identified as an unresolved item. It was observed that N&D report no. 62 had been prepared on October 15, 1973 af ter a random examination of pipe wall thickness by the A-E which had disclosed three spool pieces with less than the required thickness. The NSSS letter was written after a visit to the site by representatives of the piping supplier and the NSSS on November 13, 1974. During this visit all pieces of main steam line pipe were examined and this was the basis for the statement to that effect in the NSSS letter. ._
This item is resolved. .
e
- ~ -
- 21. Addition of QA Personnel The licensee has added four engineers to his QA staff; two in Hicksville and two at the Shoreham site. An indoctrination program for these em-ployees has been conducted and a training program is planned to be com-plated by December 31, 1974.
This item is resolved. -
- 22. Licensee Audit of NSSS Incomplete action by the NSSS in DDR (Deviation Disposition Request) item had been identified during a licensee audit on November 13-14, 1973.
Following this audit the licensee had prepared a consolidated list of all NSSS DDR's requiring further action. This list was forwarded to the NSSS with a notification of a further audit during August 1974 for review of engineering justification for disposition of DDR's and for verification of completed repairs required by disposition. Licensee's Audit Report No. 3 on August 26-27, 1974 showed satisfactory engineering justification for " Accept As Is" disposition and provision for verifi-cation of completion of repairs required by DDR disposition.
This item is resolved.
- 23. Voids in Reactor Pedestal Concrete Defective concrete in the reactor pedestal had previously been identified by the licensee and reported as a significant deficiency in accordance with the requirements of 10 CFR 50.55(e) . The inspector examined letters FLD 100, 102 and 106 from the licensee approving repair of defective areas which had been cut back to sound concrete. The inspector also examined QC inspection reports during the period of August 6,1974 to August 14, 1974. These reports covered repair of the defective areas in accordance with S&W procedure " Repair of Defective Concrete in Reactor Pedestal" as revised July 31, 1974. No deficiencies were identi-fled.
l This item is resolved.
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UNITto ITATES JE '
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TJUCLEAR REGULATORY C- ~ ~.i;.ti,10N
'i REGloN I
,j 611 P ARK AVI' nut x No or Pnuss A. PENNSYl.VAtH A , 19406
. FEB10 35 lhd Long Island Lighting Companf License No. CPPR-95
' 7 .. ) Attention: Mr. Andrew W. Wofford Inspection No. 50--322/75-01 .
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9 Vice President 175 East Old Country Road
<h '. Hicksville, New York 11801
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.id Centlemen:
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. .- Q This refers to the inspection conducted by Mr. Narrow of this office
.d on January 13-17, 1975 at Shoreham, New York of activities authorized by AEC License No. CPPR-95 and to the discussions of our findings yj
/)d
- held by Mr. Narrow with Mr. Uh1 and other members of your staff at the conclusion of the inspection. ~.
Areas examined during this inspection are described in the office
." of Inspection & Enforcement Inspection Report which is enclosed I.h. with this letter. Within these areas, the inspection consisted of i selective examinations of procedurc; and representative records, l - interviews with personnel, and observations by the inspector.
During this inspection, it was found that one of your activities l appeared to be in violation of an AEC requirement. The item and
! refere :e to the pertinent requirement are listed in the ent.losure to this letter. This letter constitutes a notice sent to you pursuant to the provisions of Section 2.201 of the AEC's " Rules of Practice", Part 2, Title 10, Code of Federal Regulations.
Section 2.201 requires you to submit to this. office within 30 days
..} of your receipt of this notice, a written statement of explanation W in reply, including: (1) corrective steps which have been or will "m. be taken by you, and the results achieved; (2) corrective steps
.q i which will be taken to avoid further violations; and (3) the date when ftfl1 compliance will be achieved.
9aa
'sj In accordance with Section 2.790 of the AEC's " Rules of Practice",
Part 2, Title 10, Code of Federal Regulations, a copy of this
.1
' .31 letter' and the enclosed inspection report will be placed in the
.';, NRC's Public Docur2nt Room. If this report contains any information
.. g that you (or your contractor) believe to be proprietary, it is W necessary that you make a written application within 20 days to this office to withhold such information from public disclosure.
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..$ Any such application cust include a full statement of the C reasons on the basis of which it is claimed that the information .
, , , .y is proprietary, and should be prepared so that propriatary
,, information identified in the application is containco in a 1 separate part of the document. If we do not heer from you in
..o this regard within the specified period, the report will be
'[. ?# placed in the Public Document Room.
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?}j'i Should you have any questions concerning this inspection, we will
, y, i,1 be pleased to discuss them with you.
- vN#@$ .: :
Sincerely, .
'l'f' *,1 ht/f, lc w
)f'.. ; . Robert T. Carlson, Chief
'E Facility Construction and Engineering 1 Support Branch i
j' Enclosurcs:
- l. Description of Violation -
- 2. IE Inspection Report 50-322/75-01 cc: Edward M. Barrett, Esquire
'l'<;.;
Edward J. Walsh, Esquire Thomas J. Eurho, Proje,ct Manager J.,[j ,
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ki bec (w/encis):
If IE Chief, FSt.E3 f%,$ i IE:HQ (5)
N *DL (4 w/cncis plus 9 cpys of report only) t hN et' DR Central Files DE#2 RS (3)
$[f PDR M ';d.s Local PDR IE Files
'- NSIC *
'$j g TIC yl'i State of Neu York 4',.& OGC
' .; REG:1 Reading Room .~
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,-J ] ENCLOSURE
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DESCRIPTION OF VIOL.*. TION
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f,4 Long Island Lighting Company
.y.'!: . Docket No. 50-322 ,
License No. CPPR-95 ,
- ft
- [g Bacc1 on the results of an NRC insper. ion conducted on January 13-17,
'% 1975, it appear. that one of your activities was not conducted in full
.;I compliance with conditions of your Facility License, ar. indicated below:
.n 1.:6 A CriterionV,AppendixB,10CFR50statesinpart,"Activitiesahfecting
, quality shall be prescribed by documented instructions, procedures, or
. .!.]}
..y drawings, of a type appropriate to the circumstances and shall be aces:-
plished in accordance with these instructions, procedures, or drawin;;s."
"1 Se: tion 4.6 of Fieli QC Procedure FQC 16.1 requires that " Component Centrol" i shall periodically 1: sue " Recall Lists" for Measuring and Test Equip ent j *(MOTE) and states in part, "FQC shall issue a Nonconformance and Dis-poc. tion Report, ... for M&TE not returned for cal'.bration by the duc date."
Contrary to the above:
, i It was found that Air Meter'No. 3/12/9 had not been returned for calibratien
i by the due date shown'on the " Recall List" and that a Nonconformance and Disposition Report had!not been issued as required by FQC 16.1.
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9- This deficiency was identified by the inspector.
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1 kE:I. Form'12
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(Jan 75) (Rec)
U.S. NUCLEAR REGULATORY C01D11SSION i
0FFICE OF INSPECTICN AND CITORCC!E"T
.g- . REGION I .
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- ~i, ,
?!I IE Inspection Report No: 50-322/75-01 Docket No: 50-30:
qq .
k Licensce: Long Island Lighting Company License No: CPPR-95 _
gty jef- ; 175 East Old Country Read Priority:
.. d Hicksville, New York 11801 Category: A dht
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- 1 Shoreham, "ev York (Shorcham Unit 1) Safeguards
3 Location:
Group:
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Typ: of Licensee:
- ~M;G Routin'e, Unannounced Type of Inspection:
Dates of Inspection:
January 13-17, 1975 Dates of P evious Inspection: Nove aber 18-2, 1974 m /
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- Reporting Inspector
- -
L. Narrow, Reactor Inspector *Date
$.' Accompanying Inspectors: .. I,b. .'b bs F 7~I ~
df,I E. P. Jefnigan, Rea'ctor Inspector Date
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Dats
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- ' $1 j e Date
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l'JiE Date vp
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3' !,; Other Accompanying Personnel:
3 - Date g [
.M. D 4 "##NAu 3b . Reviewed By: '
- d II. f. Heishman, Senior Reactor Inspector D5te h-o ..
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- n I SU}D!ARY OF FINDINGS
. 'y M Enforcement Actfen V.M
- .Q ; A. Contrary to the requirements of the applicable Field QC Procedurc,
%1 Air Metc.r No. 3/12/9 had not been recalibrated with.in the estab-lished time interval and an N&D report has not been issued.
}/.,,3,1
.':5w.l This is considered to be a deficiency with respect to Criterion V, 5 10 CFR 50 Appendi: B. (Details, Paragraph 5)
'. A
- .j' Lice ~see Action on Previously Identified Enforcement Items
'[.[ A. Failure to report dama;;c to the removable containment head had S previously been reported as a noncompliance with the requirements j.8 of 10 CFR 50.55(e). This item has been resolved. (Details,
-d$ Paragraph 15.a) ll Design Changes
)
None Unusual occurrences None
'! -[ Other SignifLeant Findings . .
V'
,' o A. Current Findings
.eg .
- 1. The items listed below were identified as unresolved.,
4 4:yg ,
6 dd a. Lack of 100% radiographic examination of the casting for f(t the Reactor Recirculation Suction Valve No. B31-F023(A).
(Details, Paragraph 4) .
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?/il b. Follow-up for corrective action of deficient' items identi-fied during the licensce's site audits was inadequate in
@@fj some cast . (Details, Paragraph 6.b) 7 S c.
Discrepancies in do.:umentation for fabrication of pene-
' $,ISl tration X-5. (Details, P ::agraph 7.a)
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$27j 3" ' ;' d. Inadequate data requirements for QC inspection reports
. of concrete materials batching equipment. (Details, j$ '1 Paragraph 8.b) -
s1 2. Performance of work described below was observed and is con-g sidered to be ac,ceptable. : ,
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- a. Pre-placement preparation and concrete placement.
tails, Paragraph 9)
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- b. Implementation of inspection requirements for pre-placement preparation, mixing and placing concrete for primary g
- .-:q containment pour RP-ll. (Details, Paragraph 9) 3fO c. Storage of concrete materials and batch plant operation.
.IM (Details, Paragraph 8.a)
M-% d. Fit-up and welding for installation of t',e reinforcing 1 plate around the personnel hatch. (Details, Paragraph 16) l
- e. Storage of selected items of equipment. (Details ,
l Paragraph 10)
- 3. Documentation described below was examined and is considered j to be acceptabic:
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. . ,d
- a. QC documentatio: for installation of an electrical l 'd penetration and a main steam line penetration. (Details ,
l . (j) Paragraph 17)
.; y rJ b. Licensee site audit schedule, records and selected audit reports. (Details, Paragraph 6.a)
.h.,f Documentation of receiving inspection and QC inspection kN c.
prior to shipment of penetration X-5.
.jyjj% Paragraph 7.b)
(Details,
. ?jb,.
- d. Documentation concerning fabrication and delivery to the N.i! site of Beam Seats 131A and 1313. (Details, Paragraph 7.c)
.wt W1 g- B. Status of Pr'eviousiv Reported Unresolved Items
- c. e .
The following items, previously identified as unresolved, have
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been examined and resolved: ^
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$t$i a. Deficiencies identified during audit of Associated Piping and Engineering Corporation. (Details, Paragraph 11) hf b. Lack of a filing procedure for site audits.
Paragraph 12)
(Details, SE
' 'NU . -.
Xt? c. Improper wall thickness of veld prep area of core spray piping. (Details, Paragraph 13) igl; oy
(;g 2. Documentation of valve wall thickness measurements as examined.
? This item remains unresolved. (Details, Paragraph 14) c.E yh Exft Interview k An exit interview was held at the licensee's office in Hicksville, New York 3
'A on January 17, 1974 with the following attendees:
Long Island Lighting Company
}
8 W. O. Uhl, Senior Vice President T. F. Gerecke, QA Manager I J. P. Novarro, Froject Engineer R. E. Black, Resident QA Engineer N. Falkin, Pro cct a Manager, Jamesport Project
! I Stone and Webster
,,$:$ R. L. Cusick, Superintendent of Construction'
@ R. Bernard, Superintendent, Field QC 2 J. M. DeGeorge, QA Coordinator e,L?
. p'
- 4. General Electric Company 1 ijig)
-R. A. Park, Site QA Representative
' $p:.i v-
- En The subjects discussed are summarized below
\
A. Scope of Inspection
. 47 ;j
- - The inspector stated that this was an unannounced inspection to
',@' examine work in progress on the reactor containment; QC records of f6 containment construction; records of shop fabrication ard field 5 installation .O containment penetrations; and the status of previously identified outstanding items.
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4 73 Cfu B. Cencral
./, The inspector stated that as of this inspection items of noncompliance m would be categorized as violations, infractions or deficiencies in
, g% accordance with the letter dated December 31, 1974 from the Director
.))
of Regulatory Operations to all licensees.
C. Items of Noncompliance
/;.yl4 E-
' i;' m . The inspector stated that the following item of noncompliance had
,j q,u been identified:
.Q my
- l. Instrument Recalibration ,
% Contrary to the requirements of Field QC.Procedu .* FQC 16.1 it 1 hjj was found that Air Meter No. 3/12/9 was in use on safety related work although it had not been recalibrated within the established
,:.y$ time interval and that an N&D report had not been issued for
'{
~{ failure to return the instrument for calibration.
,This is considered to be a deficiency with respect to Criterion V, t 10 CTR 50, Appendix B. The licensee acknowledged this infor-nation ar.d stated th::t the instrument had been withdrawn fro =
service, examined and found to be within the required calibratien l tolerances. (Details, Paragraph 5)
( . . ,
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D. Unresolved Iteme.
- 1. NDE of Reactor Recirculation Suction Valve
.;N;n The inspector stated that examination of fabrication documentatien
$ had revealed thet 100% radiographic examination of the casting
~bM for this valve had not been performed. This is contrary to the
- $,f requirements of the NSSS's purchc.se specification and is con-N sidered to be an infraction of Criterion VII, 10 CFR 50, jf Appendix B.
~ ..i R' The licensce's representative stated that an investigation of j
this matter was under way and that he had been informed that
.Q3 .
the N*.SS did not consider 100% radiographic examination to be a yys
'(W requirement for this valve since it had been purchased in October, 1969 and the applicable code was USAS B31.1.0 and G1 further that the requirement for 100% radiography had been waived by the NSSS. He also stated that he had been informed
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,9 ' ) by '.he NSSS that radiographic examination of three similar M9 valves had shown acceptabic conditions in the areas not radio-
[5[,h graphed on the casting in question.
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'. ",Q The inspector stated that 10 CFR 50.55a did require 100" gh radiographic examination and that the statement concerning the W acceptable results of examination of the other three valves was not supported by the inspectors review of available
%p, documentation. However, this finding will be reviewed at the
. ':y,y Region I Office.
j;.g ) In subsequent telephone conversations the licensee's represent-
.,Q3 ative stated that he was awaiting additional information from the NSSS and further that the licensce's investigation was not 2,.]y
.. f complete. He was informed that this item is unresolved pending
,- d review by Region I of the additional information. (Details, Paragraph 4) i.
O 2. The inspector stated that the items listed below had.been I identified as unresolved. In each case the licensee acknowledged I .this information.
- a. Follow-up for corrective action of deficient items identified dur. tag the licensee's site audits was questionable in some cases. (Details, Paragraph 6.b)
.g b. Discrepancies in documentation for fabrication of pene-tration X-5. - (Details, Paragraph 7.a)
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F c. Inadequate data requirements for QC inspection reports of
,, j concrete materials batching equipment. (Details, Paragraph 8.b) w g .1. E. Ot5er Current Findings
- .a
'(W The inspector stated that documentation and performance of the work
~~Q,j described below had been observed or examined and was considered f.g to be acceptable. The licensee acknowledged this information.
A::
- 1. Pre-placement preparation and placement of concrete for primary I@4 containment pour No. RP-11 and implementation of inspection requirements for this work. (Details, Paragraph 9) 4,1
, a.?,p 2. Batch plant operation and storage of concrete materials. (Details, Paragraph 8.a) h m
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d 3. Fit-up and welding for installation of the reinforcing plate
,9 j around the personnel hatch. (Details, Pcragraph 16)
. c. . . ;
[.'y! 4. Storage of selected items of equipment. (Details, Paragraph 10)
- 5. QC documentation for installation of an elecer.ical penetration J(Q & -
and a main secam line penetration. (Details, Paragraph 17)
- s.3 7' 6.
- 9
'( 2 Licensee site audit schedule, records and selected audit reports. (Details, Paragraph 6.a) y.. 7. QC Examination Checklist and Receiving Report for Penetration
,' X-5. (Details, Paragraph 7.b) 3 3: 8. Documentatic. concerning fabrication and delivery of Beam Seats
- M 131A and 131B. (Details, Paragraph 7.c) gi
- 2. ~ 2 j F. Previously Reported Unresolved Items
- 1. The inspector stated that the items listed below wl.ich h.-d -
.previously been reported as unresolved had been examined and were resolved. The licensee acknowledged this information.
- a. Deficiencies identified during audit of Associnted Piping and Engineering Corporation. (Details, Paragraph 11)
- b. Lack of a procedure for filing site audits. (Details ,
Paragraph 12)
- c. Improper wall thickness of core spray piping in weld
/,E ' prep area. (Details, Paragraph 13)
~
- 2. The inspector stated that documentation of valve wall thickness g?i ' measurements had been examined. No deficiencies were identified I'3 but information was lacking concerning Valve No. C11-F011. This
& item remains unresolved. The licensee acknowledged this infor-
. N;l mation. (Details, Paragraph 14) i
%(;& C. Previously Reported Noncompliance Vf.N: 1 The inspector stated that failure to report damage to the removable
--fQ containment hon which had previc sly been reported in noncompliance 27 7; with the require =ents of 10 CFR 50.55(e) has been resolved.
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g'4 The corrective and preventive actions documented in the licensce's fi.d.- Ict: cr of November 11, 1974 is an unresolved item.
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'h:l 21' i@d DETAILS h 1. Persons Contacted
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c.4' Long Island Lighting Company -
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c24 T. F. Gerecke, QA Manager ~
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.r R. E. Black, Resident QA Engineer 7pj J. M. Kelly, QA Engineer ni
'i Stone and I?ebater
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J. M. DeGeorge, QA Coordinator
",f R. Bernard, Superintendent, Field QC
..g F. C. Turner, QC Consultant
't S. A. Kalat, QC Engineer
.d- W. C. Taylor, QC Engineer
'g K. Kinkela, QC Engineer i
d C. Embler, QC Engineer, Welding 4
j Pittsburgh-Des Mofnt Steel Company l M. T'.iger, QA Manager l .
i i j General Electric Coneany
- ?
R. A. Park, Site QA Reprcr acative
- 2. Status of Project
, :?j-
.i i fgp The licensee reported that the work was 29% complete as of December 31, uQ
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197,4.
, cc.; 3. Organ.zational Changes
- Yfy 3.] The following changes have been made in the Architect-Engineers (A-E's) py! QA organization effective January 1, 1975.
N i ys1 a. Mr. R. Bernard, forectly Assistant Superintendent, Field QC, has been promoted to Superintendent, Field QC, replacing Mr. F. C.
41)
,M Turner who will act as consultant on QC until his retirement in the near future.
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, 45p b. Mr. B. Catlin has been assigned as Assistant Superintendent,
.J[f} Field QC, replacing Mr. Bernard.
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jj c. Mr. J. M. DeGeorge has bcon assigned as QA Coordinator re-g.;,y placing Mr. J. Stratis.
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- 4. NDE of Reactor Recirculatio- Suction Valve
- Paragraph 9.2.1 of Specification No. 21A9200 which 'is referenced M in the NSSS's Purchase Order No. AA-532 for procurement of these valves requires that valves over 4" pipe size shall be 100% radio-id@ graphed in accor ':nce with ASME Section III Paragraphs N323 and Ng N323.4. The casting supplicr's radiographic procedure LSF-RT-C c-dated April 28, 1970 with Amendment 2 dated September 28, 1971, itj . which had been approved by the NSSS requires 100% radiographic -
f coverage. *
- y .
M During review of valve data, the inspector observed that one 28-
. _j inch valve, the Reactor Recirculation Suction (RRS) valve g No. B31-F023(A) had not received 100% radiographic examination.
- The specific areas not examined were the crotch section and seat 1 ring arca on the body and the yoke flange to dome section on the i
bonnet. The lack of radio;raphic coverage was reported and accepted by the NSSS on D. eiation Disposition Request (DDR) No. 5315 dated February 15, 1972.
The inspector examined defect repair charts of castings' for RRS valves Nos. B31 ~023(B), B31-F031(A) and B31-F031(B) . A February 4,
- 1972, revision to the radiographic shooting sketch was used to a 1 obtain extended coverage,of these castings. In the areas generally ji described in DDR 5315 as not radiographed on Valve No. B31-F023(A),
repair of defects disclosed by radiography were observed to have Q}.;D j been required on the other three valves.
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,1lf In accordance with the discussion during the exit interview and
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- a later revicw in the Region I Office this item is unresolved 1,:
' pending review of additional information to be supplied by the M<r.; licensee.
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@j 5. Instrument Recalibration
- /.~
Paragraph 4.6 of Field QC Procedure FQC 16.1 requires that Component 7
3 Control shall periodically issue " Recall Lists" for Measuring and Test i
Equipment (M&TE) to Construction and to the FQC calibration facility ph and at least t..o wecks prior to the calibration duc date, preparc
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, " j. ' l Measurement Instrument Calibration (!!IC) form- and forward them to
,H the calibration facility. It also requires that the calibration
.Aj facility shall notify Field QC M&TE users that calibration is due
'o a and that FQC shall issue a Nonconformance and Disposition (N&D) Report
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for M&TE nc t returned for calibration by the duc date.
A;
...s 1 During observation of concretc placement for Pour RP-ll on January 16, 1975 the inspector noted that Air Meter No. 3/12/9 vcs used by Field K' .} QC in measuring the air content of the concrete. Examination of
,', M, , the calibration record, for this instrument at the calibration JN[ facility disclosed that it had been due for recalibration on
.M January 15, 1975. Ti.e Recall List and MIC form had been issued
%.9 and Field QC had been notified that it was due for recalibratien.
1 However, the item had not been returned for calibration and an NLD
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report had not been issued because of its non-return.
V i.v This is considered to be a deficiency with respect to Criterion V, 10 CFi,50, Appendix B which states, in part, " Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances.and shall be accomplished in accordance with these instructions, procedures or drawings." -
.3 P :ior to completion of the inspection the licensee reported the 9 following: An N&D report had been issued, the instrument had been 9.4 recalled, examined and found to be within the required calibratien
'b tolerances. tic licensee had also determined that failure to cceply cf with the procedure requirements resulted from the absence of both the
. j calibration facility superviso and the responsible QC Engineer.
p The inspector stated that the N6D report and enlibracion documentati:n l
' y(.; viuld be examined during a later inspection together with measures j talsen to prevent similar, failures of the system due to absence of l ;13 certain individuals.
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- 6. Licensec Site Audits l .*p-l vgs The inspector examined the following documentation concerning site
, audits by the licensee:
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. QAP 18.1 Rev. 1 dated May 1, 1974 " Audit Procedures - General" "l. .
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f . QAP 18.2 Rev. 3 dated September 23, 1974 " Quality Surveillance
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of Site Activities e*
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2 . Audit Schedules for July - December, 1974 and January - June
./ 1975 2[.
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l'.? . Record of Status of deficiencies discovered on field audits
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. Record of Status of Corrective Action Requests h' ' A 9 Record of Status of Field Audit Transmittals and the required idg responses .
. Records showing transmittal of monthly reports to QA Manager kf)f eg:
.q showing status of audit findings
..!k . Audit reports Nos. FA 181, FA 192, FA 196, FA 199, FA 200,
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% a. No unacceptable items were identified with respect to schedule, 5
, perfon ace of audits, audit reports and documentation require-
, ments. ,
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- b. Examination of corrective actions revealed that in some cases follow-up by the licensee did not assure that corrective action was accomplished as scheduled. As an examplo: FA 181, an audit
, of welding performar : by tl e piping subcontractor was perforced
,. on October 22, 1974 ar.d identified as an "open item" the failure 4
to determine preheat temperaturcs in accordance with welding j procedure W-100. E&DCR No. F-906 requesting a change in 9 Welding Procedure U-100 was initiated on October 26, 1974 and M,1 requested a response by November 4, 1974.
.; b N') ,
FA 205 another audit of welding performance by the piping sub-
"'! contractor was performed on December 20, 1974 identified the same
[fd nonconformance to procedure W-100 and again identified it as an
' .;j;' "open item" and reported that the piping subcontractor was G awaiting the response to E6DCR No. F-906 before taking action M to correct the deficiency.
h Although welding of safety related piping is not in progress the delay in corrective action shouc inadequate follow-up of g:,. : f deficiencies identified by the licensee during site audits.
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This item is unresolved.
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- 7. Documentation - Penetration X-5
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.-q a. Documentation listed below for fabrication of penetration X-5 for Residual Heat Removal (RHR) system piping was examined.
4] Stone & Webster (S&W) Specification SH1-75 " Shop Fabricatien
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lj]j and Field Erection .of Containment Steel Plate Liner"
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Pittsburgh-Des Maines Company (PDM) Drawing No. 43
- .a PDM Purchase Orders Nos. 601, 1910, 3450, 3480, 38SO 13 and 108271
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PDM Purchase Specifications Nos. MS-7.6.2, MS-10.6, MS-10.7, M.i 31.7 and MS 51.2
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Certifications of material r, ements for the identified J:$b .
materials PDM Fabrication Procedure Sequence (FPS) l (1) No unacceptable items were identified with respect to material certifications and materials meeting the require-I ments imposed by the purchase specifications.
) (2) Discrepancies were observed in identification of materials on FPS for items 43 s and 70 A ap.
9.T Item 43 s was identified on the FPS as type SA 516 Grade 70
- 2. to SA 300 purchased under PO 601. However, PO 601 referencas
,'[y specification MS-51.6 for type SA 516 Grade 70 material.
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Item 70A ap was identified on the FPS as type SA 650 if' Grade LF2 but was identified on Drawing 70 A and PO 108271
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_O' as type SA 181 Grade II.
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- b. The QA Examination Checklist and the receiving inspection repert
' #^L for thi- penetration were examined. These documents verified conformance of the penetration to purchase specification prior kJ (] to its shipment and its receipt with no evidence of damage.
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- This item is considered acceptabic.
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'.1 J,7 c. Documentation listed below for fabrication and delf ary of Beam
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't Seats 131A and 131B was examined.
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'j{1 . Material Certifications for H, cat Nos. C3796 and B7551 y}
(Lukens Steel Compsny)
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. Vendors shop welding records and QA Examination Checklist
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. PDM Receiving Report dated August 12, 1974 .
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.f;g This item is considered acceptable.
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- @ 8. Batch Plant Facilities
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, Specification SHI-42 Rev. 5 dated March 29, 1973 " Specification i for Mixing and Delivering Concrete" was examined. Operation of
.the batch plant and storage of aggregate, cement and air en-trainment admix was obs.rved. The batch plant weighing equip-ment had been calibrated by the Suffolk County Departcent of Weights and Measures on April 2-3, 1974 This item is considered acceptable.
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' [.[ b. QC inspection reports' covering testing of concrete materials 7)te: batching equipment fer accuracy were examined. Tests were eiC;J conducted within the .aquired time intervals of 120 days for
?dY aggregate scales and 60 days for the water and cement scales, 9dE ~ the water meter at.d the admix dispenser; the latest reports fIC being dated October 7, 1974 for the aggregate scales and
.lh{.I December 4, 1974 for the remaining equipment.
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' c.R However, the QC inspection reports merely stated that the
.~ffi results were satisfactory and did not identify the requirements jqp for accuracy or the numerical results of tests.
SE1 This item is unrest.lved.
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,Q 9. Concretc Placenent Observations j?@4 Specifications SH1-64 Rev. 1, Concrete Placement; SH1-66 Rev. 4, g Concrete Testing and Inspection; and Field QC Procedure QC-10.3 3,y Revision B, Concrete Quality Control were examined.
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- $j The incpcetor observed preplacement work in progre<is for contain=ent
,:j wall lift RP-ll including placing forms and reinforcing stecl, cican-it up prior to placing concrete and impicmentation of preplacement
~'?,o inspections.
.y y The inspector also observed placement of concrete; performance of "d tests for slump and air content and preparation of cylinders for dI compress!ve tests; and implementation of inspections during place-
% ment.
d These items are considered acceptable.
i.
1 i 10. Storage of Equipment The inspector observed storage in place of RHR Heat Exchanger No. Ell-E001 and HPCI Pump Ko. E41-P-016. The pu a was stored in plact within a heated enclosure. Storage inspections had been j conducted within the time intervals established on the storage history card.
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-. l This item is considered acceptable.
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l ];>s .,y 11. Associated Piping and Engineering Corporation (AP&E) Procedures T:. -
'jy During vendor inspection of AP&E by Regulatory Operations,
'Ifd deficiencies had been identified in procedures for control of tools
, 'ks and in reporting of deficiencies in accordance with 10 CFR 50.55(e).
l ti. The inspector examined the licensee's letters SNP 332E to the A-E i "?.f . and GEC 38 to the NSSS. These lettcrs require that copies of N&D i 8, reports be forwarded to the licensee with the recom:nended disposition pg and an evaluation as to whether it was reportable under 10 CFR l yll 50.55(e). ,
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' 4.4 The inspector also examined a telegram from the NSSS which stated
'% that AP&E established and posted a shop memo on control of tools.
4 This item is resolved.
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- 12. Procedure for Filing Site Audits
/ The lack of a procedure to control filing of site audits had been d'h..-1, identified as an unresolved item during inspection 74-08.
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$99 The inspector examined Procedure No. QC-5.3 dated'Deccaber 18, 1974
'Nk$h "QA Site File." This procedure establishes an index filing system if t
""fhJ for site records including audits. Paragraph 3.7 established the requirements for filing audits and Attachment 5.1 Section VI provides 79 an audit docucent code index.
I.. .j.f,d, This item is resolved.
dhl 13. Wall Thickness of Core Spr_ay Piping
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,i(lp The licensee had previously reported that core spray piping was below
'qf the minimum specified wall thickness at the weld prep area. One of i the core spray lines had previously been furnished to another 1 site. The NSSS had discussed this problem with the licensee and recomm.:nded that both core e ray lines be replaced since current orders differ in some respects from the lines previously shipped.
The inspector e::amined letter FW-T-1028 from the NSSS to the licensee which summarizes the discussion and recommendation for replacement of both lines and states that following the licensee's concurrence
.., an order 'or two core spray lines had been placed for delivery to g[i the Shor-. m site.
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% e" y B'.-V This item is resolved.'
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- ,y ' 14. Valve Wall Thickness Ifeasurements
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'. Documentation of valve wall thickness measurements were reviewed for
, , , '- conformance to the requirements of Region I letter dated June 22,
't.! '- 1972. Wall thickness measurement data was reviewed for four 28-inch
$h Reactor Coolant Recirculation Suction yalves Serial Nos. B31-F023(A).
B31-F023(B), B31-F031(A) and B31-F031(D) with no unacceptable items
'g@.J 4 ': identified.
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Wall thickness data was not available for a 2-inch Control Red Drive
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drain valve Serial. No. Cll-F0ll.
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> This item is unresolved.
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1 15. Removable Containment Head
.hy a. Failure to report damage to the removable head of the containment structure due to improper heat treatment had been identified as g,"{[ a noncompliance with the regulatory requirements. By letter
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, dated November 11, 1974 the licensee reported this event as a
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significant construction deficiency in accordance with fit 10 CFR 50.55(e). .
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- .d The following actions have been taken by the licensee to prevent
"? . recurrence of such noncompliance: -
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(1) Nonconformance reports issued to date have been re-exanined
]Y.y for reportability with no additional nonconformance found
- to be reportable.
- y (2) At a meating of Lea 1 Engineers on November 27, 1974 the licensco's Project Manager discussed the impor'.ance of I reviewing all deficiencies for reportability under 10 CFR 50.55(e).
< ~(3) The same subject was discussed at a meeting of the.A-E's Project Enginc. ring staff on November 4,1974.
\' (4) On Dece.'er 4, 1974 a training st.asion was Ield for the A-E's endi neering staff on "The Reporting of Significant
,.. ] Problems for Preventive Action," EAP 16.1 and " Reporting
,( Significant Deficiencies to the AEC" EAP 16.2.
7 i j The inspector examined documentation recording these actions and had no further questions.
~.$'[' This noncompliance is resolved.
- b. The corrective and preventive actions documented in the licensce's
[@.I ap 1etter of November 11, 1974 is an unresolved item.
,-/ 16. Personnel Hatch Fit- up and Welding 3.4 7 The inspector observed fit-up and in-process welding of the personnel s, hatch support plates Nos.113B1 and 113B2 to spool piece No.113A2.
Inspection of work in pror,ress disclosed that qualified welding J@. procedures 69-97 and 69-98 were being implemented. The finished Q weld was inspected for appearance and compliance with specifications.
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[2; This item is con [dered acceptabic.
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.;. 17. Documentation - Penetration Inr.tallation,
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.' The irspector audited install -ion documentation for electrical
- .,
- , penetration 24 PIA, and main steam line penetration 42A. Records
{iG.; examined included field receiving inspection forms, construction procedure sequence, material fit-up forms and QA examination check
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- lists.
'Ef]!.-T-This item is considered acceptable.
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. UNITE 3 STATES NUCLEAR REGULATORY COf.1 MISSION REGloN I
- S31 PARK AVENUE ,
- KING oF PRUS$1 A, PENNSYLVANI A 19406 APR 4 ERS Long Island Lighting Company License No. CPPR-95 Attention: Mr. Andrew W. Wofford Insp. No. 50-322/75-04 Vice President 175 East Old Country Road Hicksville, New York 11801 Gentlemen:
This refers to the inspection conducted by Mr. Narrow of this office on March 17-20, 1975 at Shoreham, New York of activities authorized by AEC License No. CPPR-95 and to the discussions of our findings held by Pz.
Narrow with yourself and members of your staff at the conclusion of the
inspection.
Areas examined during this inspection are described in the Office of Inspection and Enforce =ent Inspection Report which is enclosed with this letter. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspector.
During this inspection, it was found that one of your activities appeared to be in violation of an AEC requirement. The item and refer-ence to the pertinent requirement are listed in the enclosure to this letter. This letter constitutes a notice sent to you pursuant to the provisions of Section 2.201 of the AEC's " Rules of Practice", Part 2, Title 10, Code of Federal Regulations. Section 2.201 requires you to submit to this office within 30 days of your receipt of this notice, a written statement of explanation in reply, including: (1) corrective steps which have been or will be taken by you, and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance vill be achieved.
In accordance with Section 2.790 of the AEC's " Rules of Practice", Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public' Document Room. If this report contains any information that you (or your con-tractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this of fice to withhold such information from public disclosure. Any such application must include a full statement of the reasons on the basis of which it is claimed that qoWTio+
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a c4 ;E' i sQ 476 1916
the information is proprietary, and should be prepared so that pro-prietary information identified in the application is contained in a separate part of the document. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely,
/
Robert T. Carlson, Chief Facility Construction and Engineering Support Branch
Enclosures:
- 1. Notice of Violation
- 2. IE Inspection Report No. 50-322/75-04 cc: Thomas J. Burke, Project Manager Edward M. Barrett, Esq.
Edward J. Walsh, Esq.
bec: (w/encis) "
IE Chief, FS&EB IE:HQ (5)
DL (4 w/enci plus 9cy report only)
( DR Central Files i
RS (3)
PDR l Local PDR IE Files NSIC l TIC l OGC REG:I Reading Room State of New York 1
I
License No. CPPR-95 ENCLOSURE 1 NOTICE OF VIOLITION Based on the results of an NRC inspection conducted on March 17-20, 1975, it appears that one of your activities was not conducted in full compliance with your Facility License, as indicated below:
Contrary to 10 CFR Part 50.55a:
Nondestructive examination of the casting for reactor recircula-tion systen valve No. B31-F023(A) did not include 1007. radio-graphic inspection.
This infraction had the potential for causing or contributing to an occurrence related to health and safety.
e 4
3
I
. IE:I Form 12 Qan 75) (Rev) , ,
U. S. NUCLEAR REGUIATORY COMMISSION OFFICE OF INSPECTION AND ENFORCD!EST REGION I IE Inspection Report No: 50-322/75-04 Docket No: 50-327 Licensce: Long Island Lighting Comua_n? License No: CPPE C5 175 East Old Country Road Priority: . . .
- ' Hi c1e w f 11 o v.,. .n v 4 11801 Category: A Safeguards Group: NA Location: c w .s nun s'g..
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Type of Licensee: n"o. 810 m.*e (er)
Type of Inspect. ion: ooutine . L'nannounced Dates of Inspection: March 17-20. 1975 Dates of Previous Inspection: March 3-6, 1975 Reporting Inspector: 3 .3 d'\A in . a YM I.U 'l s L. Narrow, Reactor Insp'ector DATE Accompanying Inspectors: - -et% w-- -
4/4-/7f DATE R. C. Haynes, S e < heactor Inspector chde,o~ k k -7 5 E.JerMgan,RbactorInspector DATE DATE Other Accompanying Personnel: None Reviewed By:
^
2J R. F. Heisluaan, Senior Reactor Inspector # DATE e
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SUMMARY
OF FINDINGS N - -
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j q Enforcement Action s A. Items of Noncompliance s
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- 1. Infraction ) ,, ), 3v
- a. Incomplete NDE of Reactor Recirculation Systen Valve O- '
Contrary to 10 CFR 50.55a, nondestructiva.er.acination of the casting for reactor recirculation system valve,No. B 31-F023(A) did not include 100% radiographic _ inspection.
This infraction was identified by the inspector and had the potential for causing or contributing to an occurrence of safety significance. (Details, Paragraph 3)
B. Deviations None Licensee Action on Previously Identified Enforcement Matters The following previously identified item was resolved.
Calibration of Air Meter No. 3/13/9. (Details, Paragraph 4)
Other Significant Findings A. Current Findings Documentation and performance of work described below was observed or examined and is considered to be acceptable.
- 1. Fit-up of equipment hatch. (Details, Paragraph 5)
- 2. Welding of beam seats Nos. 1382 and 14A-9. (Details, Paragraph 6)
(Details, Paragraph 7)
- 4. Fabrication data package for recirculation pu=p embedments.. (Details, Paragraph 8)
- 5. Fabrication data package for RHR piping spool pieces Nos. E11-UR-210-2-3:
and E11-WR-210-2-4. (Details, Paragraph 9) l l 6. Installation data package for beam seats, main steam line penetration, feed water penetration and CRD penetration.
(Details, Paragraph 10) '
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.. -2 B. Status of Previously Unresolved Items The unresolved items listed below have been resolved.
- 1. Incomplete weld qualification records. (Details, Paragraph 11)
- 2. Correction of Specification Index. (Details, Paragraph 12)
- 3. Site Audit corrective action. (Details, Paragraph 13)
- 4. Documentation for Penetration X-5. (Details, Paragraph 14)
- 5. Data require =ents for QC inspection reports. (Details, Paragraph 15)
- 6. Procedure for control of Engineering and Design Coordination Reports (E&DCR's) in the field. (Details, Paragraph 16)
- 7. Documentation of valve wall thickness measurements. (Details, Paragraph 17)
Corrective actions for resolving the iter- listed below were reviewed.
These items remain unresolved.
- 1. Residual Heat Recoval (RHR) Pumos Rework and test records for two of the RER pu=ps were reviewed and considered to be acceptable. Data packages for the remaining two pumps were not available for review. This ite= re=ains unresolved.
(Details, Paragraph 18)
- 2. Containment Head Documentation concerning revelding and post veld heat treatment of the drywell head flange was reviewed. Information was not available to assure that caterial adjacent to the weld seam had not been adversly affected. (Details, Paragraph 19) i Exit Interview An exit interview was held at the site on March 20, 1975 with the following attendees:
! Long Island Lighting Company i A. W. Wofford, Vice President T. J. Burke, Project .'bnager T. F. Gerecke, QA Manager R. E. Black, Resident QA Engineer l
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Prod. Dept.
M. S. Pollock, Manager Elect.
E. Bajada, QA Engineer J. M. Kelly, QA Engineer E. W. Tesco, Construction Superintendent Stone and k'ebster_
R. Bernard, Superintendent, Field QC F. C. Turner, QC Consultant R. L. Cusick, General Superintendent J. M. DeGeorge, QA Coordinator General Electric Company _
W. A. Shanks, Site Manager R. A. Park, Site QA Represzntative Pittsburg-Des Moines Steel Com:any .
M. Stiger, QA Manager The subjects discussed are sue =arized below.
A. Scope of Inspection this was an unannounced inspection te Observe The inspector stated that examine QC records of shop fab-work in progress on reactor contain=ent, rication and review the status of previously identified outstanding ite=s.
I I
B. Items of Noncompliance Failure to perform 100% radiographic inspections of
! ing review by Region I office and submittal of additional infor:ation by the licensee.
The inspector stated that after review of the additional information, as well as prior interpretations of the regulatory requirements with respect l to codes and standards, the lack of 100% radiographic inspection is con-l
' sidered to be in infraction of 10 CFR 50.55a.
The licensee stated that he did not agree with this interpretation for the following reasons:
4-
- 1. Th,e Code case (N10) which requires 100% radiography is not applicable to USAS B 31.1.0; the code identified as the principal code for this equipment.
- 2. Code Case (N10) which was applicable to ASA B 31.1 had been annulled prior to the date of purchase of the valve. (Details, Paragraph 3) .
C. Current Findings The inspector stated that documentation and performance of work listed below is considered to be acceptable. The licensee acknowledged this information.
- 1. Fabrication data package for recirculation pump embedments. (Details, Paragraph 8)
- 2. Fabrication data package for RHR piping spool pieces Nos. Ell-UR-210-2-3 and E11-WR-210-2-4. (Details, Paragraph 9)
- 4. Fit-up of Equipment Hatch. (Details, Paragraph 5)
- 5. Welding of beam seats Nos. 138E and 14A-9. (Details , Paragraph 6)
- 6. Installation data packages for beam seats, main steam line penetrations, feed water line penetration and CRD penetration. (Details, Paragraph 10)
D. Previousiv Reported Findings The inspector described the status of previously reported findings which had been reviewed during this inspection. The licensee acknowledged this information.
- 1. Enforcement Items The deficiency with respect to instru=ent calibration has been resolved.
(Details, Paragraph 4)
- 2. Unresolved Items
- a. The items listed below have been resolved.
(1) Incomplete weld qualification records. (Details, Paragraph 11)
(2) Correction of Specification Index. (Details, Paragraph 12)
(3) Site audit corrective action. (Details, Paragraph 13) m
~5-(5) Documentation for Penetration X-5. (Details, Paragraph 14)
(5) Data requirements for QC inspection reports. (Details, Paragraph 15)
(6) Procedure for control of E&DCR's in the field. (Details, Paragraph 16)
- b. The items described below remain unresolved.
(1) RHR Pumps Rework and test records for two of the RER pu=ps were re-viewed and are considered to be acceptable. Documentation for the re=aining two pumps were not available for review.
(Details, Paragraph 18)
(2) Containment Head Documentation concerning rework and post weld heat treatment of the drywell head flange was reviewed. Information was not available to assure that material adjacent to the weld seat had not been adversely affected by exposure to original heat treatment plus welding and post weld heat treatment during rework. (Details, Paragraph 19) i I
4
DETAILS
- 1. Persons Contacted Long Island Lighting Company T. F. Gerecke, QA Manager R. E. Black, Resident QA Engineer Stone and Webster J. M. DeGeorge, QA Coordinator R. Bernard, Superintendent Field QC F. C. Turner, QC Consultant
J. Burgess, Sr. Field QC Inspector H. Zassenhaus, Asst. Supt. Office Management J. Davis, Senior Engineer General Electric Company W. A. Shanks, GE Site Manager R. A. Park, Site QA Representative Pittsburg-Des Moines Steel Company M. Stiger, QA Manager
- 2. Status of Proiect The licensee reported that the work was 34% complete as of February 28, 1975.
- 3. NDE of Reactor Recirculation System Valve Failure to perform a complete RT examination of the casting for the 28-inch (nominal pipe size), reactor recirculation system valve No. B31-F023(A) had been identified as an unresolved item on a previous inspection pending review of additional information to be supplied by the licensee.
Prior review of defect repair charts of the other three similar valves had indicated repairs in the area generally described as not radiographed on valve No B31-F023(A); the crotch section and seat ring area on the body and the yoke flange to dome section on the bonnet. The inspector reviewed
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a radiographs of these areas on the three valves in question. No un-acceptable indications were observed on the radiographs examined.
The inspector also reviewed the licensee 's internal memorandum on the radiography requirement for valve No. B31-F023(A), dated March 5, 1975. This memo identified the nuclear steam supplier's purchase order for this valve as No. 205-AA532, dated October 29, 1969 and the purchase specification as No. 21A9200, Revision 4, dated October 14, 1969 and updated to Revision F, on February 5, 1971. In the memo it was stated that the applicable code for the valve was USAS B31.1.0-1967 which does not reqrire 100% radiography of the valve. It was further stated in the me_o that Code Case N-10 (which required 100.';)
radiography) was applicable only to ASA B31.1-1955 and that this code case was annulled as of June, 1969. The inspector informed the licensee that 100% radiography was required and that this position
,, was consistent with past regulatory practices. The inspector further stated that the radiography requirement is clear in view of the in-clusion of the Code Case N-10 radiography requirement in later edi-tions of those codes which apply to such valves.
i The incomplete radiographic examination of this valve is considered to be in noncompliance with the requirements of the regulatory re-quirements delineated in 10 CFR 50.55a.
- 4. Instrument Calibration l During a previous inspection it was observed that contrary to the re-quirements of Field QC Procedure (FQC) 16.1 Air Meter No. 3/12/9 had not been returned for calibration by the established calibration "due date" and that a nonconformance and disposition (N&D) report had net been issued.
The inspector examined N&D report No. 246, which identified lack of re-calibration, and the instrument calibration card. The instru=ent cali-bration card showed no out of tolerance conditions on January 16, 1975.
The N&D report was closed on February 12, 1975 on the basis of these calibration results which confirmed the accuracy of the test results obtained using this instrument.
The FQC Superintendent stated that this incident resulted from a mis-understanding of the requirements for recalibration by certain labora-tory personnel. To prevent recurrence the QA Superintendent provided verbal and written claification of che requirements to all calibration laboratory personnel.
The inspector examined memo dated March 13, 1975 from the Superintendent Field QC to all Calibration Laboratory personnel. This memo included instructions for review of " Recall List" at the time of its receipt, en-taring of exact due dates for all items and initialing of due dates by laboratory personnel.
This item is resolved.
- - . .- . . . -. .- . _ - - ...J
- 5. Fit Up of Equipment Hatch The inspector observed the positioning, fit-up, alignment and tack welding of the outer portions of the equipment hatch, Drawing WL 7.
The weld joint was prepared and aligned in accordance with approved drawings.
This item is considered to be acceptable.
- 6. Welding of Beam Seats
- a. Beam Seat No. 138E The inspector observed the in-process welding of beam seat No.13SE.
The inspector verified that pre-heat, and inter-pass temperatures were checked by the welder and that welder. qualifications and electrode certifications were in confor=ance with appreved drawings and procedures.
This item is considered to be acceptable.
- b. Beam Seat No. 14 A-9 The inspector observed the fitting and welding of the leak chase around beam seat 14 A-9. The inspector verified that the work including pre-heating and use of approved weld electrodes was in compliance with approved procedures.
The inspector audited welder qualifications, and inspection records.
This item is considered to be acceptable.
- 7. Storage of Ecuipment The inspector observed the storage of certain items of equipment and piping.
The CRD Housing Support Beam (Item No. 1-COO 4) had been placed on hold for lack of documentation and was properly tagged. RER piping spool pieces (Pes. Nos. Ell-WR210-2-03 and 04) were . observed to be properly identified and tagged as released for construction.
This item is considered to be acceptable.
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- 8. Recirculation Pump Embedments The inspector examined Specification No. SH1-2'~ 'Hiscellaneous Iron &
Steel" dated May 22, 1974. This specification incorporates the re-quirements of SH1-64, Revised June 10, 1970 and Addenda 1 through S.
The inspector also examined the data packages for the embed =ents items 23 EMB 309 thru 312. The material certifications identified the heat numbers for each item and provided the chemical composition and physical characteristics for the specified materials.
This item is considered to be acceptable.
. 9. RHR Piping
,, The inspector examined the data packages for spool pieces Nos. Ell-WR-210-2-04 and E11-210-2-05. Included were fabrication sketches; material certification identifying heat numbers, actual' chemical analysis and mechanical test results; Form NPP-1 Data Reports; and Material Receiving Reports.
This item is considered to be acceptable. -
- 10. Installation Documentation The inspectors audited documentation covering installation of the CRD penetrations, equipment hatch and beam seat inserts for conformance to the requirements of the drawings, material specifications, velding pro-cedures and repair procedures. Documents examined included inspecticn reports, welder qualifications and welding electrode certifications.
This item is considered to be acceptable.
- 11. Weld Qualification Records Audit of fabrication records for the steel containment dome had previously identified failure of weld procedure qualification WPQ No. 67-1 to in-dicate the thickness for which the welder was qualified and the ti=e the test sample was held at stress relief temperature. The inspector observed that Form Q-1 had been supplied as an attachment to Wpo 67-1 and indicated a qualification thicknuss of 3/16" to 8" and post veld ' heat treat =ent at 11000F for one hour per inch of thickness.
This item is resolved.
- 12. Document Revisions Failure of the specification index to identify the correct addendum to specification SH 1-75 had previously been identified as an unresolved item.
The inspector audited the specification index and found that addenda to specifications were correctly identified for SH 1-75 and other selected specifications.
This item is resolved.
- 13. Licensee Site Audits Review of selected site audit records had previously identified a delay in correction of items identified as "Open Items". During February, 1975 the licensee reviewed site quality assurance activities including audits.
As a result of this review the Resident QA Engineer had been instructed to take corrective actions in certain areas particulary with respect to classification of audit findings, requirements for response to audit findings and submission of reports to the QA Manager.
This item is resolved.
- 14. Penetration X-5 Certain discrepancies in identification of caterial used in fabrication of this penetration had been identified during a previous inspection. These discrepancies had been identified by the fabricator as due to an error in Preparation of the Fabrication Procedure Sequence (FPS) in one case and failure to correct the FPS to conform to a drawing change. The inspector examined the correction to documentation and certifications of material l Provided in accordance with the revised drawing.
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This item is resolved.
- 15. Data Requirceents for Inspection Reports During review of concrete batching equipment inspection reports it had been observed that these reports lacked quantitative data require =ents.
Tha inspector examined QCI 10.3 dated February 13, 1975. These in-structions require that quantitative data be recorded during inspection of the batching equipment and identify the capacity of equipment and allowable tolerances.
This item is resolved.
A e
- 16. Control of E&DCR's in Field Lack of a procedure for identification of E&DCR's applicable to each specification issued in the field had previously been identified as an unresolved item.
i The inspector examined Construction Sita Instruction (CSI) 2.12 "In-formation Requests" dated December 17, 1974. This instruction established a procedure for documenting requests for information and changes to documents. It required that each affected holder of a drawing and specification will receive copies of E&DCR's. Construction personnel audit holders of documents to assure that they are updated and conform to SCI 2.12. However, these are informal audits and are not recorded.
A mini-computer program for control of documents in the field is beint prepared and is scheduled to be in operation by August, 1975. As an
interim measure, Field QC will audit holders of documents for conformance to CSI 2.12 in addition to their normal audit for confor=ance to FQC 5.2.
This item is resolved.
- 17. valve Wall Thickness Verification The inspector reviewed the wall thickness verification data on the 2" control rod drain valve , No. Cll-F011.
The valve wall was measured by Stone and Webster and was determined te be in accordance with drawing requirements.
This completes verification of valve wall thickness for all valves delivered to the site which have been identified by the licensee as requiring such action.
Documentation for all valves to be delivered in the future will include verification of wall thickness by the vendor.
This item is resolved.
- 18. RHR Pumps The inspector reviewed the documentation associated with the i=peller modifications for two RHR pumps. This item was initially reported to the Director of Regulatory Operations as required by 10 CFR 50.55(e)
(Ref. LILCO letter of 4/28/74, Burke to Knuth). All four RHR pump impellers required metal build-up on the vase tips in order to increase the total developed head so that design conditions could be met.
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The documents reviewed by the inspector were contained in two binders, both entitled "LILCO Projects, Addendum 1, Quality Assurance Records Binder RHR Pump Impeller Rework" with one binder marked " Serial Number 691-S1114" and the other binder marked " Serial Number 691-S-1113". The documents included drawings of the required impeller modification, ma-terial certification, heat treatment data, nondestructive test results and pump data including total developed head, flow, brake horsepower and NPSH datum.
No discrepancies were identified by the inspector during his review of the documentation. The inspector observed that the certified pump test data provided by the pump vendor (Syran-Jackson) showed the pu=p met design point requirements.
Since only two pump i=pellers had been returned to the site and the
documentation packages for the modification of the other two pu=p in-pellers and the four drive motors were unavailable, this item is con-sidered to remain unresolved pending a review of said documentation during a subsequent inspection.
- 19. Containment Head The inspector reviewed the corrective action taken by PDM related to the change in physical properties of caterial in certain sections of the containment drywell head. The =aterial physical properties change resulted from i= proper post weld heat treat =ent of the vertical welds in the flange section. (Ref. RO Inspection Report 50-322/74-08, Detail Paragraph 2 and, LILCO letter of 11/11/74 to Knuth, AEC-DRO, from Wofford)
The inspector observed that material adjacent to the previous vertical welds es well as the vertical welds themselves had been removed and a new section of material installed. This was as described in the pre-viously referenced LILCO letter. The inspector also found that post veld heat treat =ent of the new welds had been completed and the flange section was being set up for machining operations.
The inspect.or reviewed PDM drawings "SR 1" and "SR 2", PDM's document entitled " Welding Procedure Qualification Record-Butt Joints" (for joint number 68-12) and S&W Specification SH 1-75, Rev. 3, entitled " Shop Fabrication and Field Erection of Reactor Containment Steel Plate Liner for Shoreham Nuclear Power Station - Unit 1". The inspector observed that the above PDM docu=ent qualified the material for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> of post weld heat treatment at 1100 to 12500F and that the material spect= ens met the physical properties requirements of the S&W Specification SH 1-75.
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The inspector also observed from the details shown on PDM drawings SR-1 and SR-2 (which described the post weld heat treatment system including size and placement of heater elements and insulation) that base material up to 2' - 9" or more on each side of the ver-tical welds could have been adversely affected during the initial post weld heat treatment operation. This condition was more probable for the 2k" and Sk" thick sections because 2' - 6" long "A" heaters were used which were horizontally displaced 3 inches from the weld centerline (2' - 6" + 3" = 2' - 9").
Since the repairs involved the removal of base material 2 ' - 6" on each side of the veld seam, the inspector requested objective evidence which would provide assurance that the material in the as completed condition had the required physical properties. ::o such objective evidence was available at the time of this inspection.
This item is considered to be unresolved.
6
UNITED STATES s' /
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- OLEAR REGULATORY COMMISSION REGloN I
$31 PARK AVENUE KING OF PRUS$1 A, PENNSYt.VANI A 19406 jul 10 B75 Long Island Lighting Company License No. CPPR-95 Attention: Mr. Andrew W. Wofford Inspection No. 75-08 Vice President Docket No. 50-322 175 East Old Country Road
- Hicksville, NY 11801 Gen tlemen:
This refers to the inspection conducted by Mr. Narrow of this office on June 24-26, 1975 at Shcrcham, New York, of activities authorized by NRC License No. CPPR-95 and to the discussions of our findings held by Mr.
Narrow with yourself and members of your staff at the conclusion of the inspection.
Areas examined during this inspection are described in the Office of Inspection and Enforcement Inspection Report which is enclosed with this letter. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspector.
Within the scope of this inspection, no items of noncompliance were observed.
In accordance with Section 2.790 of the NRC's " Rules of Practice", Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room. If this report contains any information that you (or your con-tractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this office to withhold such information from public disclosure. Any such application must include a full statement of the reasons on the basis of which it is claimed that the information is proprietary, and should be prepared so that pro-prietary information identified in the application is contained in a separate part of the document. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Ro'om.
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No reply to this letter is required; however, should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely,
/ A W . 0A 4 =
Robert T. Carlson, Chief Facility Construction and Engineering Support Branch
Enclosure:
IE:I Inspection Report No. 50-322/75-08 cc: Thomas J. Burke, Project Manager Edward M. Barrett, Esq.
Edward J. Walsh, Esq.
T. F. Gerecke, QA Manager bec: (w/encis)
IE Chief, FSLEB IE:HQ (5)
DL (4 w/enci plus 9cy report only)
DR Central Files RS (3) '
PDR Local PDR l IE Files NSIC TIC ELD REG:I Reading Room State of New York I
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!:I Form 12 Ian 75) (P,cv)
U. S. NUCLEAR REGUIATORY Com!ISSION OFFICE OF INSPECTION AND ENFORCDIENT REGION I 50-322 E Inspection Report No: 50-322/75-08 Docket No:
L ng Island Lighting Company CPPR-95 icensee: License No.
175 East Old Country Road p flicksville, New York 11801 A g,
Safeguards Shoreham, New York (Shoreham '1) Group:
BWR, 819 MWe (GE) yp2 of Licensee:
Routine, Unannounced 37a of Inspection:
utes of Inspection:
June 24-26, 1975 .
May 19-22, 1975 htes of Previous Inspect n:
'eporting Inspector:
1 Aff/W L. Narrow, Reactor Inspector 7/7/7d~
'DATE
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- ccompanying Inspectors: /[// - f.' -. N - -? - 1 / 7 s W. F. Sanders, Keactor inspector DATE
/ ha v h7l'7[
'G. A. Walton, Keactor inspector DATE
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- DATE NONE Jthtr Accompanying Personnel:
Reviewed By: _
v (f'/ t r > 7 .)
DATE R. F. Heishman, Senior Reactor Inspector l
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SUMMARY
OF FINDINGS Enforcement Action A. . Items of Noncompliance None B. Deviations None Licensee Action on Previously Identified Enforcement Matters None Design Changes Procedures for modification of the Reactor Pressure Vessel (RPV) feed-water nozzle safe ends were reviewed and are considered to be acceptable.
(Darai]n. Paragraph 8)
Unusual Occurrences l
None Other Significant Findings A. Current Findings
- 1. The following items are unresolved:
- a. Lack of documentation to support the amount of UT experience shown on certification records of a Level II UT Technician.
(Details, Paragraph 2)
- b. Certain items of Shop Test, Inspection and Documentation (TID) records were not signed off by the inspector although the record indicated that they had been verified. (Details, Paragraph 3)
- 2. The following items are considered to be acceptable
- a. QC Program for structural steel installation. (Details, Paragraph 4)
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- b. Material certifications, Receiving Inspection Reports and QC records of installation of the drywell deck at Eleva-tion 76. (Details, Paragraph 5)
- c. Installation of structural steel for the drywell floor at 76 foot elevation. (Details, Paragraph 6)
- d. UT inspection procedure, techniques and work in progress.
(Details, Paragraph 7)
B. Status of Previous Unresolved Items
- 1. The following item has been resolved:
Repair of the drywell head flange. (Details, Paragraph 9)
- 2. Th2 item concerning NDE of the reactor pressure vessel (RPV) repairs which had been reported as resolved (Inspection Report 50-322/75-07) is unresolved pending licensee response to the Division of Reactor Licensing letter dated Jane 11, 1975.
(Details, Paragraph 10)
Exit Interview An exit interview was held at the site on June 27, 1975.
Persons Present Long Island Lighting Company A. W. Wofford, Vice President T. J. Burke, Project Manager T. F. Gerecke, QA Manager R. E. Black, Resident QA Engineer F. X. Schoner, QA Engineer R. E. Hall, QA Engineer W. Hunt, Construction Engineer Stone and Webster R. L. Cusick, General Superintendent, Construction S. W. Baronow, Senior Superintendent, FQC J. D. Davis, Assistant Superintendent, FQC G. R. Mikula, QA Engineer (Boston)
General Electric Company W. A. Shanks, Site Manager Items Discussed The items discussed are summarized below. In each case the licensee, acknowledged the information.
A. Scope of Inspection The inspector stated that this was an unannounced inspection to review the QC program and its application to installation of structural steel, ultrasonic inspection in progress on the RPV, the procedures for modification of the RPV feedwater nozzles, and to review the status of previous fidings.
B. Current Findings
- 1. The inspector stated that no deviations or items of noncompliance had been identified.
- 2. The inspector stated that the items listed below hak been identified as unresolved:
- a. Lack of documentation to support the amount of UT experience shown on certification records of a Level II UT Technician.
(Details, Paragraph 2)
- b. Certain items on the shop TID report on structural steel fabrication had not been signed off by the inspector although they were indicated to have been verified.
(Details, Paragraph 3)
- 3. The inspector identified the items listed under Design Changes and Paragraph 2 of Current Findings in the Summary of Findings and stated that these items are considered to be acceptable.
(Details, Paragraphs 4, 5, 6, 7 and 8) ,
- 4. Previously Identified Unresolved Items
- a. The inspector stated that the item concerning repair of the drywell head flange had been resolved. (Details, Paragraph 9)
i I e i i
- b. The inspector stated that the item concerning NDE of the RPV repairs which had previously been reported as
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resolved will be reopened in view of the additional questions raised by the Division of Reactor Licensing (DRL). (Details, Paragraph 10) 1
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DETAILS
- 1. Persons Contacted Long Island Lighting Company T. F. Gerecke, QA Manager R. E. Black, Resident QA Engineer Stone and Webster R. Bernard, Superintendent Field QC-W. C. Taylor, Field QC Engineer C. Embler, QC Engineer, Welding J. Davis, Senior QC Engineer J. Burgess, QC Engineer T. Arrington, QC Engineer D. Heikkinen, Assistant QC Engineer E. Mac Dougal, Senior QC Engineer C. Schwartz, QC Inspector General Electric Company W. A. Shanks, Site Manager Hardford Steam Boiler Inspection and Insurance Company V. Smith, Authorized Inspector Conam Inspection, Inc.
R. Lucas, Ultrasonic Inspector F. E. Freeman, Ultrasonic Inspector
- 2. NDE Personnel Qualifications The inspector reviewed the personnel qualification records for the licensce's inspection contractor (Conam) and noted that one l individuals record states that he was qualified as a Level II l
in ultrasonics from October 1964 until the present. However, in l
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- an interview with the individual he stated that he qualified in ultrasonics in 1969. Records were not available for the inspectors review to indicate that this individual does in fact have 9h years experience as a Level II UT Technician as the certification indicates.
This is an unresolved item pending review by the inspector of the individuals qualification documentation.
- 3. Incomplete Inspection Report The inspector examined TID report No. 16 dated May 13, 1974. This report lists tests, inspections and documentation requirements of Specification SH1-152 for the structural steel furnished in accord-ing with Vendor (Cives Corporation) Drawings Nos. 861 through 884.
This material has been released for shipment by the inspector and Shipping Release Tag No. 4602 had been issued. However, the indi-vidual requirements had not in all cases been signed by the inspector.
As an example, the requirements for traceability and for transfer of heat numbers on cut stock were identified as having been verified but the record of verification by signature of the inspector was not available.
This item is unresolved.
, and QC instructions listed below:
I
- a. QC-10.4 dated April 11, 1975, " Structural Steel Erection".
- b. QC 9.1, Revision A dated October 28, 1974, " Receiving l
Inspection".
- c. QCI 9.1-004B dated April 4, 1975 " Receipt Inspection of Structural Steel.
- d. QCI 9.1-005, dated May 19,1975, " Documentation Review Portion of Receiving Inspection Process".
The structural steel contractor is committed to perform his work in accordance with the A-E's QC Procedures as well as Specification SH1-152 " Structural Steel and Steel Decking".
This item is considered to be acceptable.
- 5. Structural Steel Documentation The inspector reviewed the QC records listed below concerning structural steel for the drywell floor at the 76 foot elevation.
- a. Cives Drawing Nos. 861 and 870 which identifies the Heat Nos. for Beam Nos. A, D, F and H-861 and A, F and J-870.
- b. E & DCR No. P-157 and letter from S&W to Cives Corporation, File No. 210-47.1 providing requirements for traceability of minor structural members: Material to be furnished from limited number of heats; material to be segregated and controlled and mill test reports to be furnished.
- c. Category 1 Stock List dated June 11, 1975 which, among others,
, identified Heat Numbers of selected 6 x 3 inch clip angles.
- d. Material certifications for heats used in beams and angles identified above.
- e. QC inspection report dated June 16, 1975, witnessing job-inspection torquing of high strength bolts for field connection of beams to headers,
- f. Calibration records for Torque Wrench No. 2-14-06 which was identified as the wrench used for inspection torquing.
This item is considered to be acceptable.
- 6. Structural Steel Installation The inspector observed installation of Structural Steel for the drywell floor at the 76 foot elevation. Specified material was used, the prescribed inspections were conducted and records were in compliance with the established requirements.
This item is considered to be acceptable.
- 7. UT Inspection' Procedure and Techniques l
The inspector reviewed the Inspection Contractors (Conam) pre-service examination activities in progress on the RPV. Examina-tions being performed include only those areas on the RPV which will be inaccessible for examination after the vessel is installed.
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The inspector reviewed the following:
- a. NDE Procedure NIP 631 Rev. O.
- b. Ultrasonic equipment calibration.
- c. Ultrasonic calibration blocks.
- d. Ultrasonic test results.
In addition, the inspector observed test personnel performing ultra-sonic examinations on the RPV support skirt circumferential weld.
This review included instrument calibration, scanning speeds of the ultrasonic transducer and surface finish of the test surfaces.
This item is considered to be acceptable.
- 8. RPV Safe-End Modifications The inspector reviewed the drawings, procedures and specifications relating to the design modification of the reactor pressure vessel feedwater nozzle safe-ends. The present safe-end design shows the feedwater sparger thermal sleeve as a slip fit at the nozzle safe-end interface. The design modification provides for a thermal sleeve attached by welding to the inside diameter of the nozzle safe-end.
The following procedures and specifications were reviewed:
GE Spec 22A-3897 Rev. 0 - Removal of Safe-End, Replacement and Inspection GE Spec 22A-2693 Rev 0 - Installation Specification GE NDE 351N-75A-0001 - Dye Penetrant Examination GE NDE 351N-75A-0002 - Visual Examination t
GE NDE 351N-75A-0003 - Radiographic Examination GE-PS-351N-75A-0004 - Ni Alloy Bare Filler Metal f
' GE-PS-351N-75A-0005 - Ni Alloy Covered Electrodes f
GE-PS-351N-75A-0006 - Stainless Steel Bare Filler Wire l GE-PS-351N-75A-0007 - Mild Steel Bare Filler Wire
(
GE-PS-351N-75A-0008 - Stainless Steel Covered Electrodes l
GE-PS-351N-75A-0009 - Mild Steel Covered Electrodes l GE-PS-351N-75A-0010 - Cleanliness Control l GE-WPSG-351N-75A-0011 - General Weld Process - Mild Steel to Mild Steel ,
GE-WPSD-351N-75A-0012 - Detail Weld Process - Mild Steel to Mild Steel l
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GE-WPSS-351N-75A-0013 - Replacement Safe-End to RPV Nozzle GE-WPSG-351N-75A-0014 - General Welding Ni-CR-FE I GE-WPSS-351N-75A-0017 - Sparger Thermal Sleeve to Safe-End GE-WPSG-351N-75A-0018 - General Welding Process - Stainless Steel to Stainless Steel GE-WPSD-351N-75A-0019 - Detail Welding Process - Stainless Steel to Stainless Steel GE-WPSH-351N-75A-0020 - Repair Weld Procedure GE-SP-351N-75A-0021 - Arc Strike Removal CE-SP-351N-75A-0022 - Etching GE-SP-351N-75A-0023 - Nozzle Safe-End Cut-Off GE-SP-351N-75A-0024 - Counter Bore Weld Prep GE-SP-351N-75A-0024A - Data Sheet and Record Format GE-SP-351N-75A-0026 - Weld Material Storage and Distribution GE-SP-351N-75A-0027 - Feedwater Sparger Installation This modification is intended to prevent vibration of the feedwater sparger which had been experienced at some BWR's; and which resulted from excessive bypass flow through the slip-fit annulus between the thermal sleeve and the feedwater nozzle safe end.
The present safe-end extension will be cut back prior to installation of the RPV and the modification will be completed after the vessel -
has been installed inplace.
This item is considered to be acceptable.
- 9. Drywell Head Flange As a result of improper post weld heat treatment, a change in physical properties of the material had been identified in cartain sections of the containment drywell head flange. This item had been reported in IE Inspection Report No. 50-322/74--08 and by the licensee in accordance with 10 CFR 50.55(e) on November 11, 1974.
Subsequent replacement of a portion of this flange and heat treat-ment after rewelding were reported in Inspection Report No. 50-322/
75-01. However, objective evidence was not available to assure that the material had retained the required physical properties after rewelding and heat treatment.
The inspector reviewed the results of heat treatment and tests per-formed by the fabricator (P-DM) on samples of material from the original flange. The additional post weld heat treatment time
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(PWHT) was determined by establishing the maximum total heat treat-ment time which the material could have experienced. The samples were exposed to at least 80% of this time.
It was determined by P-DM that the maximum PWHT was 28.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the required total.PWHT exposure was 22.56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br />.
The material from which samples were taken had previously experienced 18.03 PWHT time and was heat treated an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for a total of 24.03 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Samples were then taken and subjected to tension, side bend and charpy impact tests.
The inspector noted that test results of these samples were accept-able. This item is resolved.
50-322/75-07) has been reopened in view of additional questions raised by DRL concerning UT examination data.
This item had been identified in 1972 and had been the subject of NDE and metallurgical evaluations since that time. A " Final Report on Reactor Vessel Repair" has been submitted by the licensee (erroneously identified as having been submitted by the RPV manu-facturer in the previous report).
DRL evaluations of this report was forwarded to the licensee with recommendations that the licensee reevaluate certain UT examination data.
This item is unresolved.
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D. Follow-up On Defects Reported As a Result of Region I Inspection and Enforcement Investigations (Tr.1159)
Staff Response (Prepared by J. Higgins)
NRC Region I follows up on every item of the type discussed in the limited appearance statem? "t and documents the resolution in subsequent inspection reports. All items are tracked by Region I so that the inspection reports, where any given item is closed, is easily retrievable and so that a list of the currently open items is available.
The current open itmes list will be transmitted shortly from Region I to NRR as part of the standard letter sent 90 days prior to the licensee's scheduled fuel load date.
The questions raised by Mr. J. McCrystal are similar to those raised in S.C. Contention 12 and the process of NRC follow-up is discussed in more detail in the NRC testimony relating to Contention 12 which is also being filed on this date.
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E. .Steamline Cracks (Tr. 1161)
Staff Response (Provided by Region I)
Inspection Report 79-24, copy attached, discusses this matter in the context of the Staff's overall investigation of that matter and certain other matters relating to alleged construction irregularities at Shoreham (See Allegation 1 in Attached Report.)
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Docket No. 50-322 (
Long Island Lighting Company "w'
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ATTN: Mr. Andrew W. Wofford j
s .: e 4 Vice President
- ^ ~7 175 East Old Country Road -
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Hicksville, NY 11801 .,
I Gentlemen: ,
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Subject:
Investigation 50-322/79-24 c J l This refers to the investigation conducted by Dr. C. Gallina and other. x'. - - .1 personnel from this office from December 11,1979 to March 19,19f.0 at :
' Shoreham Nuclear Power Station, Shoreham New York and other various locations s in Suffolk County, New York of activities authorized by NRC License No..
CPPR-95 and to the discussions of our findings held by Mr. L. Narr6w of. . 1 i this office with Mr. Novarro and other members of your staff at the conclusion ~ -
of the investigation. The investigatio'n was conducted as a result of allegations made regarding construction irregularitiec at the Shoreham i Nuclear Power Station. Thirty allegations were investigated, none were found to be substantiated.
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l Areas examined during this investigation are described in the Office of Inspection and Enforcement Investigation Report which is enclosed with'this letter. Within these areas, the investigation consisted of selective examinations of procedures and representative records, interviews with personnel, independent measurements, and observations by the investigators.
No items of noncompliance with regulatory requirements were found with respect to the allegations investigated. However, during the course of this investigation it was found that certain other of your activities appear not to have been conducted in full compliance with NRC requirements, as set forth in the Notice of Violation, enclosed herewith as Appendix A.
These items of noncompliance have been categorized into the levels as described in our correspondence to you dated December 31, 1974. This notice is sent to you pursuant to the provisions of Section 2.201 of the i
NRC's " Rules of Practice," Part 2. Title 10, Code of Federal Regulations.
Section 2.201 requires you to submit to this office, within thirty (30) days of your recei t of this notice, a written statement or explanation in reply including: corrective steps which have been taken by you and the~
results achieved; corrective steps which will be taken to avoid further' i items of noncompliance; and (3) the date when full compliance will be i achieved.
With respect to Appendix A, we note that you have corrected Item A, and therefore you need not address yourself to this matter in your response.
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Long Island Lighting Company 2
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In accordan~ce with Section 2.790 of the NRC's " Rules of Practice," Part 2 Title 10. Code of Federal Regulations, a copy of this letter and the enclosures will be placed in the NRC's Public Document Room. If this report contains any infomation that you (or your contractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this N. office to withhold such infomation from public disclosure. Any such application must be accompanied by an affidavit executed by the owner of the infomation, which identifies the docurrent or part sought to be withheld, and which contains a statement of reasons which addresses with specificity the items
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which will be considered by the Comission as listed in subparagraph (b)
(4).c1 Section 2.790. The infomation sought to be withheld shall be incorporated as far as possible into a separate part of the affidavit. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.
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Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, r j ,
= _ _ -
Robert T. Carlson, Chief
! Reactor Construction and Engineering l Support Branch l
Enclosures:
- 1. Appendix A, Notice of Violation
- 2. Office of Investigation and Enforcement Investigation Report Number 50-322/79-24 cc w/encls:
J. P. Noverro, Project Manager Edward M. Barrett, Esq.
Edward J. Walsh, Esq.
T. F. Gerecke, Manager, Engineering QA Department bec w/ encl's:
IE Mail & Files (For Appropriate Distribution)
Central Files PublicDocumentRoom(PDR)(LPDR)
Local Public Document Room Nuclear Safety Infomation Center Technical Infomation Center (TIC)(NSIC)
REG:I Reading Room Stata of New York James C. Higgins, Resident Inspector L. Narrow, Region I NRC Case File: NRC-I D-88 l
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' APPENDIX A NOTICE OF' VIOLATION Long Island Lighting Company Docket No. 50-322 :
I This refers to the investigation conducted by representatives of the Region I (Philadelphia) office at the Shoreham Nuclear Power Station, Shoreham, New York, of activities authorized by NRC Construction Permit CPPR-95.
I During this investigation conducted on Decen6er 11,1979 - March 19,1980, the following apparent items of noncompliance were identified:
I A. Criterion XVI of Appendix B of 10 CFR 50 requires that " measures shall l
be established to assure that conditions adverse to quality, such as... nonconfomancas are promptly identified and corrected". The Shoreham Nuclear Power Station FSAR Section 17.1.168 states that the S&W quality assurance program requires that major and recurring conditions adverse to quality, such' as... nonconfomances be identified, the main .,
i causes detemined, and corrective action taken to prevent repetition. ,
I Specifications for concrete work No. SHI-354 and No. SHI-64, Section
! IV state that cold weather curing of concrete will confom to the i
requirements of the Anerican Concrete Institute standard ACI-306
! "Reconinended Practice for Cold Weather Concreting". Tables 1.4.1 and i 1.4.2 of ACI 306 require that " moderately massive sections" of concrete
, be maintained at a minimum temperature of 450 F for either two or l
three days following concrete placement, dependent upon future exposure of the concrete. The S&W Field Quality Control Procedure QC-10.3 i " Concrete Quality Control" requires that the Field Quality Control j Engineer will ensure that concrete is cured for the required time and i that curing temperatures ant in accordance with applicable specifications.
Contrary to the above on December 3,1974, concrete placement No. RS-4-12, which is classified as a moderately massive section, had been
, exposed to a temperature of 380 F on the second day after placement.
This nonconfomance had not been identified by Field Quality Control and corrJctive action'had not been taken to detemine whether the exposure had adversely affected the concrete nor to prevent repetition of such nonconfomance.
This item is an infraction.
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( q Appendix A 2 I
B. Criterion V of Appendix B of 10 CFR 50 requires that " Activities affecting cuality shall be prescribed by documented instructions... '
and shall be accomplished in accordance with these instmetions. The Shoreham Nuclear Power Station FSAR Section 17.1.5A reqJires that suppliers of safety related materials and services are responsible for imposing the above requirements on their internal operations. Courter and Corrpany Welding Procedure Specification NW-100-08011AA, Revision 0, specifies that the filler metal for Weld Joint Nunber 1G-33-WO9 1-FW "D" be AWS No. ER-309.
Contrary to the above, on May 19, 1979 Weld Joint No.1G-33-WD9-3-1-FW "D" was welded using ER-308 filler metal.
This item is an infraction.
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% gpg 2a qq U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
',,,,N REGION I Report No.. 50-322/79-24 Docket No. 50-322 License No. CPPR-95 Priority --
Category B Licensee: Long Island Lighting Company (LILCO) 175 East Old Country Road Hicksville, New York 11801 scility Name: Shoreham Ruclear Power Station, Unit 1 .
Investigation At: Shoreham, New York Investigation Conducted: December 11, 1979 through March 19, 1980 Investigators: % Af O. JAI' _
M 90 C. O. llina,/ h.D., Investigation Specialist date s add +-<,/ W!Y/hdate L. M. Narrow, Reactor Inspector e 'u kkJ W W WO di e p.P.eDurr,ReactorInspector Ob <&
S. D. Reynoldsh, Jr. , ReacTD - Inspector Y*Y 8o date b '
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. C. HiggiiT5, S . R sident Reactor Inspector arte
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- 1) $l . I date ctor Inspector gH. Nicholas, C 11 % ~ l[~ $ D R. K. Christoph Investigation Specialist date OJA)Abur 20v d N-[/-Y Y /P. Remaklus, Ihvsstigation Specialist
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date l Reviewed By: '2/ 4 4 s 943~O R. W. McGaughy/'C6fef, Project Section date RC&ES Branch Region I Form 143 -
(Rev. October 1977) l
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! Investication Summary:
Investigation from December 11, 1979 throuch March 19, 1980 (Investication Report Number 50-322/79-24)
Areas Investicated: 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 allsged 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) e 2
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TABLE OF CONTENTS I. BACKGROUND A. Reason for Investigation B. Identification of Involved Organizations II.
SUMMARY
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 3
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I. BACKGROUND A. Reason for Investigation 1
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 j 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 l presented during the above referenced trial on December 6, 1979 and local I 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. ,
l B. Identification of Involved Organizations b
- 1. LONG ISLAND LIGHTING COMPANY (LILCO) 175 East Old Country Road t 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 A company contracted by the licensee to perform various construction management activities at the Shoreham site.
- 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 Shoreham site.
- 4. GENERAL ELECTRIC COMPANY, I&SE DIVISION (GE) 777 West Putnam Avenue P. O. Box 6850 Greenwich, Connecticut 06830 i
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t A company contracted by the licensee to provide the turbine generator and related components at the Shoreham site.
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- 5. ORAVO CORPORATION (DRAV0)
Neville Island l Pittsburgh, Pennsylvania 15225 l A company contracted by the licensee to perform various construction activities at the Shoreham site.
- 6. COURTER & COMPANY (C0URTER) 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 l San Jose, California 95128 A company contracted by the licensee to perform various construction activities at the Shoreham site.
- 9. PROTECTIVE SPRAY PLASTICS, INC. (P.SP) 1130 Crose Avenue New York, New York 10472 A company contracted by the Itcensee to perform various construction activities at the Shoreham site.
- 10. REGOR CONSTRUCTION CO., INC. (REGOR)
P. O. Box F East Northport, New York 11731 A company contracted by the licenses to perform various construction I activities at the Shoreham site.
- 11. JOHN GRACE & COMPANY (GRACE) 34 Washington Parkway Hicksville, New York 11801 l
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A company contracted by the licensee to perform various construction activities at the Shorehas 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.
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l II.
SUMMARY
OF FINDINGS A. Allegations and Investiaation Findinas This investigation involves allegations that were introduced by 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 Shorenas 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 i
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 boilermaker 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 ifcensee (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 l 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.
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4 Allegations NOTE: The allegations listed beiw have been summarized for clarity. The actual allegations 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 Raactor Pedestal, Primary Containment Wall, and Radwaste Buf1 ding, forms were improparly 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.) Following 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 patched 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 l cadwelds.
l The NRC investigation found no evidence and/or information to substantiate l 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 NRC investigation found no evidence and/or information to substantiate this allegation. (Detafis, Paragraph D.6) l l
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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 form 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. I The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.8) l 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 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.10)
, 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 information 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. (Datails, Paragraph D.12) 13.) Due to the improper 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.
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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 LILCO about the incompetence and corrupt practices of its '
contractors, such as Regor and Courter, which LILCO insisted on using.
The NRC investigation found no evidence and/or infomation 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 l September of 1979 could not have been valid since it occurred too soon l after the gross failure in June 1979 to have permitted proper shutdown and repair.
The NRC investigation found no evidence and/or information to substantiiate 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 0.16) 17.) NDE 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-j 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 l
site is inadequate.
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The NRC investigation found no evidence and/or information to substantiate 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 misaligned that titanium tubing which should 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 in' formation to substantiate this allegation. (Details, Paragraph D.22) 23.) Radiographic tests revealed that the longitudinal seam welds for the con-denser box were improperly done and when opened for rework were found to contain dirt, rubbish and weld rod stubs.
k 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) l 26.) Turbine foundation bolts were installed so far out of alignment that it I 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.
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I The NRC Investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.26) 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 pemitted to remain within the form and the concrete placed on top of it.
The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.27) 28.) Soil percolation testing results were falsified and test results withheld in LILCO'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 pho qualified after performing walds for which they h'ad not been qualified.
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.
The NRC investigation found no evidence and/or information to substantiate this allegation. (Details, Paragraph D.30)
- 8. Items of Noncomo11ance j During the course of the investigation, two (2) items of noncompliance l 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 nonconformance 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) l 2.) (79-24-02) Contrary to Criterion V of Appendix B of 10 CFR 50, the licensee issued two (2) Wald Material Requisitions which indicated ER-308 weld material instead of the required ER-309. (Details, Paragraph D.10) 12
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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. Novarro, 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) l 13
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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 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 with trespassing during an anti-nuclear rally at the Shoreham site in June of 1979.
B. Scope of Investication
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- 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 J .(The Franklin Research Center).
C. Persons Directly Interviewed and/or Contacted During the NRC Investiq4t.io..
During the course of this investigation, representatives of various 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 Shorenas site were interviewed by the NRC in investigating the 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 l required in order to differentiate between the statements of one or more j 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 perso.is being interviewed unless (a) such person was identified by another independent docurrent or person other than the source or (b)
- the person being interviewed indepencently 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 i their choice present during the interview, and the confidentiality provi-sions of this investigation.
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D. NRC Investication 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-u 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- H System may be unfit and may have to be repaired or replaced.
- b. NRC Investigation Findings NRC investigators identified the extent of the N- H System referred to in the allegation by a review of all of the N- u 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 I the isolation valve outside of Primary Containment to the Main Steam Stop Valves for the Turbine. This pipe is 24" 0.0., 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 ASME Section III, Class 2 requirements. The site welding was performed by COURTER with inspections by COURTER and S&W surveillance in-spection groups.
The NRC no.ted 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 l 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 0.0. of the pipe was visually examined during site fabrication, prior to the installation of required insulation material. 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 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 l _ _ __ _
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requirements and included grinding to remove the vast majority of the surface indications and repair welding of some defects shown by ultrasonic testing (UT) to encroach on the minimum wall dimensions. t 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 SMd 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-fied in this area.
The NRC determined that the " cracks" in the main steam line as reported in the allegation were in all probability a misinterpre-tation of the normally 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.
- c. NRC Conclusion The NRC found no evidence and/or information to substantiate 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 form 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.
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- b. NRC Investication Findinos The NRC investigators reviewed applicable sections of S&W specifications and 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 l 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the removal of heat.
The NRC investigators interviewed several LILC0 site QA personnel, S&W QC personnel, S&W construction engineering and supervisory l personnel as well as ORAVO 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 l interviews, the NRC was able to establish that the winter control of curing temperatures was accomplished in the following l manner:
Concrete placements were contained within temporary enclo-sures fabricated prior to placement of the concrete.
-- Heaters within the enclosure were provided. These heaters were maintained within the enclosures by laborers assigned i to each of the areas. In the case of the failure of a heater, it was either repaired immediately or replaced by spare heaters prcvided for this purpose. In the case of damage to the enclosure itself, supervisory personnel were notified immediately and crews were assembled to effect the repairs.
Construction engineers took three (3) sets of temperature l
measurements, (a) outside ambient, (b) ambient within the l enclosure and (c) concrete surface temperatures." These l temperatures were taken daily including weekends at not l
1ess than 6-hour intervals. In addition, the outside t
"These surface temperatures were considered to be conservative since the actual bulk temperatures of the concrete pour would have been somewhat higher than the surface temperature due to the heat of hydration released during the curing process.
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ambient temperature was obtained daily from a thermometer which indicated the maximum and minimum temperature ex- j perienced during that time period. The daily minimum concrete and outside ambient temperatures were recorded on specially designated curing Reports. Construction engineers
, . also stated that during extremely cold weather, additional temperature readings were frequently taken. Temperature readings below the minimum specified were reported to Field QC personnel.
Field QC personnel performed 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 1 of the pours included concrete placements in the Reactor Pedestal, i 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 maintain the required temperature during the curing period.
l Six (6) N&D reports were found in this regard but it was noted 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) N&Ds.
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) 18
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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) l under winter conditions (818 pours) and summer 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.
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) l tests of three (3) shots each were made on each placement. j 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 invest,i-gators noted that the minor deviations identified would only retard the early strength developed by the concrete and not ,
cause any permanent damage, a conclusion further verified by the Windsor Probe tests.
l NRC investigators could find no instances where concrete had been exposed to freezing temperatures, a condition that would l 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 circulating constantly to check on the heaters." He also stated that in critical areas additional laborers were assigned over and abovet' hose making the rounds in order to keep the heaters in operation.
- c. NRC Conclusico Although the NRC identified isolated instances where heaters 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
r 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
NRC investigators examined several curing reports and conducted interviews with several S&W QC, construction engineering and construction supervisory personnel, as well as ORAVO craft and supervisory personnel involved in the placement and stripping of formwork from 1973 through 1976. These records and interviews indicated that wall forms were removed after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and that this time period was closely monitored by QC and engineering personnel.
The NRC noted that there was no ACI requirement or specification requiring these forms to remain in place for seven (7) days, although on occasion, forms were kept on for periods in excess of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> requirement. This latter case was usually dictated by work or location requirements such as when forms could not be removed from the lower areas of the Reactor Pedestal due to space limitations in that area. Although various indivi-duals noted that there might have existed some engineering request for this particular extended support, the NRC could not confirm that fact by any written engineering documentation. No irregularities were noted in this area.
The NRC noted that although the alleger claimed that the seven l (7) day stripping requirement was part of his training as a 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.
- c. NRC Conclusion l
The NRC investigation found no evidence and/or information to substantiate this allegation.
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- 4. Allegation No. 4 . ,
- a. Allegation Following the stripping of forms from concrete placements for the Reactor Pedestal and Primary Contai m t Wall, large cracks, honeycombing deep enough to expose the 7.1,ar and possible through-cracks (cracks appearing at the same azimuth location on both sides of the wall) were all patched over with mortar prior to inspection by QC. It was alleged that the patching was performed 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 alleger worked on site from 1973 through 1976. .
- b. NRC Investication 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 N&D reports of concrete I 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 formwork 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 perform 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 I
(cement finishers) instead of the laborers, the alleger insisted I
that it was the laborers and not the masons who had performed the unauthorized repairs.
l S&W QC personnel informed the NRC that they were required to inspect all concrete surfaces after stripping and that they had all been given verbal instructions to perform the inspection 3
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the forms had been removed. QC inspectors i :
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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.
All personnel interviewed confirmed the repair of any honeycomb or cracks prior to QC inspection would have been obvious to 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 N&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 !
although not requiring a specific N&D report, were repaired under QC supervision in accordance with ACI-301 and utilizing a t special concrete bonding agent.
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 watches everything, how the mortar is' mixed, placed and set -
everything."
The NRC also noted through its interviews of several involved personnel and related N&D reports that through-cracks were l unlikely due to the amount of rebar and due to concrete pouring techniques. No through-cracks were identified at the Shoreham site in the areas examined.
- c. NRC Conclusion The NRC investigation found no evidence and/or information to substantiate this allegation.
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- 5. A11ecation No. 5
- 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 l 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 perfoming 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.
l General Procedure W-300 provides for protective measures to be i 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 i
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.
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! c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.
- 6. Allegation No. 6
- a. Allegation Rubber water stops between concrete layers in the Radwasta Building were not installed properly (not sealed or nailed to keys) and sometimes omitted entirely.
- b. NRC Investication Findings NRC investigators examined applicable S&W specifications, and procedures for preplacement inspections.i.e., inspections performed prior to the placing of concrete (see Exhibit A). ,
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 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.
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NRC investigators examined several Field Data Sheets and Concrete 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.
In order to confirm these findings, NRC investigators interviewed 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 ir.volved in the installation, supervision of installation and inspection of waterstops in the Radwaste Building. The NRC noted that none of the individuals interviewed could recall any problem with the installation of waterstops. Craft personnel described the method of sealing the joints as well as methods of holding the waterstops in position using wooden blocks and wedges. The,e individuals also denied the use of any nails for attaching waterstops. The NRC also noted that training had been given in this area prohibiting the use of nails. QC personnel stated 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 embedments. The NRC noted that several craft individuals expressed mixed frustrations at times because of the several QC 24 l
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 information to substantiate this allegation.
- 7. Allegation No. 7
- a. Allegation The alleger stated that as a carpenter he had been permitted to weld studs to embedment 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 carpentar foreman who provided him (the alleger) with the needed welding rod.
- b. NRC Investigation Findings NRC investigators examined the construction drawings (FC-25F-5 l 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 l be located in the North wall.
NRC invest'igators 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 n ted that there appeared to be no pipe support embedments on the North wall of the Radwaste Building. Further investigation of the Radwaste 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 l type.
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The NRC investigators reviewed the quality assurance program in this area and noted that embedments were normally prefabricated in shops adjacent to the construction area. The embedments 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 uniform 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 determine 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 controlled 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 the material.
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The NRC noted, therefore, that in addition to the fact that there were no embedment plates used to hold pipes on the North 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 he received welding material from a foreman who was not authorized 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 item was documented on Field Data Sheets (Form T-5-31) 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 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 their knowledge no manually welded embedments were ever noted.
- c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.
- 8. Allegation No. 8
- a. A11egation'
! The alleger stated that threaded tie rods were used as form ties for the Reactor Pedestal along with she-bolts attached to the rods to hold the forms in place. It was further alleged that when the she-bolts were removed, in many cases the threaded rods would slide out of the concrete a .1 a small patch would be put on the hole leaving a void in the center of
- a concrete.
, In some cases, it was alleged that when the she solts could not j be removed at one end, the entire tie rod was pulled through the concrete pulling a large amount of concrete off the pedestal l wall. It was alleged that rods were pulled through the concrete approximately twenty (20) to fifty (50) times in this fashion.
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- b. NRC Investiaation Findings NRC investigators examined several slides showing the actual 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-l mation 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 s.tatement 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 Reactor Pedestal and (b) if the rods were installed without sleeves as stated ecphatically 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, nor was the Reactor Pedestal designed to maintain airtight integrity.
- c. HRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.
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- 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 i further alleged that the cracks were so wide one could see through to the other side of the wall.
- b. NRC Investication 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 l it to exhaust air from lesser to progressively greater potentially contaminated areas to a monitored final exhaust. Thus even.if 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.
I In addition, any crack large enough to see through would be I
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.
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- c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.
- 10. Allegation No. 10
- a. Allegation 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 i
I weld rod specified on their weld rod requisitions for scheduled work on dissimilar metal welds. When refused by the weld rod clerk due to the improper designation on the weld red requisition, the welders stated that it was too cold to return to have the requisition corrected to the required ER-309 and that the l
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ER-308 and/or ER-316 would be utilized in its place. It was 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 Investigation Findinas 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 l
be traced to a specific weld. This review identified (12) dissimilar metal welds. The document p'ackages for each of ,
these 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 l 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).
i 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 welds had been identified. Four he..cred and twenty two (422) dissimilar metal welds were identified and cross checked in this manner. The document packages for each of the 422 dis-i similar metal welds were examined in order to verify that l
ER-309 material had been used as required. The results of this examination identified one additional weld (Weld Joint Number 30
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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 irintified 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 stem from paperwork errors related to the similarity of procedure numbers (08011AA for ER-309 and 0811AA for ER-308) and not a l
deliberate attempt by the welder to utilize the improper weld I 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 confiem 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 ana,1yses.
The NRC determined that all welding performed on the control rod drive mechanisms was performed by Reactor Controls, Inc.
(RCI) which maintained its own weld rod issue facilities, 31
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t totally separate from weld rod issues of COURTER for whom the alleger was employed. It was impossible therefore for the alleger 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.
NRC investigators noted that the identification of the two welds containing the improper weld material while constituting an item of noncompliance, did not substantiate the related al-legation due to substantial differences in time frame, scope, 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 The NRC fcund no evidence and/or information to substantiate this allegation.
I 11. Allegation No. 11
- 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 performed in order to covertly use substandard materials.
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- b. NRC Investigation Findings NOTE: The NRC noted that the ASME Code,Section II C, 1 SFA-5.1 does not list an E-6018 electrode. It was assumed that the alleger meant E-6010 electro' des which are listed in the ASME Code and are in common use within the industry.
j The NRC conducted several interviews with S&W QC inspectors, construction supervisors, weld rod control supervisors and weld rod issue clerks. These interviews disclosed that the only E-6010 electrodes used at the Shoreham site were utilized in connection with the carbon dioxide fire system installed by the C. P. Bennett Company. The NRC noted that extremely rigid 32
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 issue was counted and verified by the rod issue clerk, the welder, a LILC0 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 LILCO 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 CO2 System pipe welding."
NRC investigators conducted an additional interview with the C.
P. Bennett General Foreman in order to confirm the above re-ferenced procedures. No discrepancies were noted. The General Foreman stated that ~possibly no more than two (2) 50 pound cans were used for the CO2 System and that o.nly three (3) employees were authorized to sign for the E-6010 electrode issue. ,
NRC investigators noted that the allegation was presented in such a manner as to indicate that the licensee was covertly attempting to introduce substandard material into the con-struction project. Coworkers of the a11eger indicated that this
' was not credible since all involved individuals (including the alleger) were carefully informed that the strict procedure was required in order to comply with the increased QC requirement.
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 these conflicts further as it did not affect the evaluation of the validity of the allegation.
- c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.
- 12. Allegation No. 12
- a. A11ecation 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 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.
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- b. NRC Investigation 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 l to as the "feedwater condenser" was in fact the Main Turbine 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 I 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 weldin'g electrodes for " rep, air 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 immediate 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.
The rewelding procedure would involve a minimum weld size from 1/2" to 5/8" instead of the original 3/8" fillet weld.
1 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 axcessive relative to the volume of weld deposited in the prepared grove.
1 1
^ Includes the cutting and repair of various stiffners which had to be removed in order to gain access to the leaking area i
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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.
, c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation. .
- 13. Allegation No. 13
- a. Allegation 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 estimation of the depth of the water table under the plant.
- b. NRC Investication Findings I 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 question 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 l
indicated that the area had not seen water for some time.
Interviews by the NRC of various LILCO personnel indicated that on infrequent occasions in the past 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.
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Interviews by the NRC of various S&W construction personnel confimed the above observations but added that a minor problem had been experienced in the past with rainwater leaking onto 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 storm 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 remain in effect until the pumps were transferred from the construction to the start-up group at which time automatica1,1y operating systems would he amployed.
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 personnel assumed the QA field inspection duties.
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- b. NRC Investigation Findings t
NRC investigators examined the various changes in the management of the Shoreham project. Among the items 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 tne contract between LILCO and COURTER and CD.
l to a contract between S&W and COURTER and CO. This condition was prompted by the fact that the ASME required QC operations j to be performed by the actual contractor and not the subcontractor.
NRC investigators were informed that at this time LILCO was considering changes in construction management, and in a memoran-dum dated August 11, 1977 recommended that LILCO assume the leading role in construction management with S&W retaining responsibility for engineering and quality assurance. As of September 12, 1977 LILCO did assume the leading role in construc-tion management. This change was effected by having LILCO
( assume leadership of the joint S&W/LILCO unified construction team (UNICO) through the appointment of a LILCO Construction Manager reporting to a LILCD Project Manager rather than the S&W corporate construction organization. LILCO also increased l
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 CD. contract directly rather than transfer it to S&W. Since COURTER was cow the contractor, the ASME would not allow S&W to perform the code work as either LILCO cr 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 inspection of non-safety related mechanical equipment and insulation was corrected by transferring the responsibility l
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 associated with the i transfer.
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l The NRC noted that UNICO was formed prior to March of 1977. The 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 allegation.
- 15. 311ecation No. 15
- a. Allegation At least seven (7) major failures of hydroflushes of the Primary i 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 I ( 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.
- b. NRC Investigation Findinas The NRC reviewed all possible primary system hydrotesting l
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 air. 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 performance of this test, and independently l
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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 ,
The NRC found no evidence and/or information to substantiate this allegation.
- 16. Allegation No. 16 I
- 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, ha.ve shifted, broken and separated from the line itself. It was further alleged that a LILCO QA map documented the separated pipe pieces located in the Sound.
- b. NRC Investigation Findings 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 was able to determine that the fiberglas pipe sections were 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, storms caused various sections to work loose and suffer some 39
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( 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 storm 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 undemater inspection of the outfall verified that it was secure but the inspection indicated minor damage to a diffuser, a condition repained and documented in N&D Report No.
i 1466. The NRC identified no instance whereby secured sections of the outfall had broken loose and separated from the system.
l 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.
l 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.
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- c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.
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- 17. Allegation No. 17
- a. Allegation NDE technicians were not adequately qualified for the jobs they were performing.
- b. NRC Investigation Findings
- NRC investigators examined prior NRC inspection reports for the Shoreham site for the years of 1977, 1978 and 1979. It was noted that from January 1977 through June 1979, NRC inspectors reviewed the qualification records and observed NDE technicians perfoming tests during eight (8) inspections. In addition, it was noted that during welding inspections, NRC inspectors frequently include the inspection of the performance 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-formance of NDE technicians were identified during any of the NRC inspections reviewed.
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- 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.
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l During this particular investigation, NRC investigators examined a random selection of experience, training and qualification ,
records of NDE technicians currently at the Shoreham 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 information to substantiate this allegation.
- 18. Allegation No. 18
- 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 Investigation Findings
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The NRC assumed that since there is no "outake" canal at the Shoreham site, that the a11eger was infact referring to the intake canal. The NRC noted that LILCO 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 i at the storm 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-tives identified the material as a fluorescent dye and identified the individual within the New York State Department of Environ-l 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 investigt.cors examined the intake canal and adjacent shore-fronts on '.ong 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 this allegation. ;
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- 19. Allegation No. 19
- a. Allegation Supervision of trade workers is inadequate and being performed by unqualified individuals. -
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- b. NRC Investigation Findinas*
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 acknc.<1 edged 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 requested one of the contractors (COURTER) 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, Instrumentation and Auxiliary / Yard Area), a supervisory individual l 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 NRC determined that this level of supervision appeared to be l
commensurate with the extent of construction activities being I
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 from the cliegers, the NRC determined the validity of the allegation as it pertained in general to the overall construction site.
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- 20. Allegation No. 20
- a. Allegation Qualifications and training of subcontractor personnel at the Shoreham site is inadequate.
- b. NRC Investigation Findings , ,
During the course of the investigation, several subcontractor personnel were interviewed by the NRC. During the course of these interviews, individuals were informally questioned with respect to their job responsibilities and past experience as 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 1
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 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 pertainea to their activities and responsibilities at the Shoreham site. The NRC reviewed a memo by S&W to all subcontractors dated August 20, 1976 describing the implementation of a site-wide training program, separate ,
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from other training requirements as described in the Field QC Manual. Each contractor was directed to provide a minimum three (3) hour orientation program and to ensure that each new employee attended the courses. The NRC selected the Orientation Manual for DRAVO and reviewed it in detail. No apparent defici-encies were noted. A check of an alphabetical listing of DRAVO employees reflected the fact that all onsite employees were 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.
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 concrete to secure various pipes and equipment. Approximate 1,y 660 workers were given this course over the time period stated
'above. Similar classes were noted in welding and other craft activities on site.
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. Allegation 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.
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- b. NRC Investigation Findings 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-fied by the NRC inspector and the performance of the work as well as the QC inspection associated with it were considered to be acceptable. NRC investigators also reviewed all NRC inspection 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 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.
During this painting operation two (2) types of coating systems were utilized in the suppression pool, an epoxy (K&L) system above elevation 30 and an epoxy phenolic (Plasite) system below that elevation. The NRC noted that painters were qualified separately for each system in accordance with a proposed ASTM qualification method. Qualification for the Plasite system was considerably more difficulty because of the coating system itself as well as the use of a more complex qualification i panel. A large number of painters who had qualified on the K&L system could not qualify on the Plasite system panel. 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.
- the Suppression Pool is the lower portion of the Reactor Primary Containment 1
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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 mathedrine 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 operation itself. -
- c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allegation.
- 22. Allegation No. 22
- a. Allegation Tube support sheets in the condenser box were so misaligned 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. NRC 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 l 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 27, 1970. The NRC noted that the original specification called for 5B111, Alloy CA706 (Cu-Ni) tubes but that the design was modified to improve water-side corrosion resistance by employing ASTM:B338, Grade 2 welded titanium tubes.
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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.
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 tiAN in-service (brush .
type) cleanir.g system by extending the tubes a minimum of 3/8" beyond the tube sheets.
The NRC noted that the unit was reassembled and fabricated in f accordance with S&W Specification SH1-223 with surveillance 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 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, 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 58338 HDE 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 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 reforming 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 i 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 seam welds for the condenser box were improperly done and when opened for reworking, it was found that the welds often contained dirt, rubbish and weld red stubs. It was alleged further that these conditions were sometimes discovered when the condenser had to be cut open to correct misalignment preblems,
- b. NRC 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.
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t The NRC was informed through discussions with the S&W engineer responsible for condenser fabrication that as a result of the Field QC questioning the achievement of full penetration welds, "information only" radiographic and ultrasonic examinations of the welds were performed. The results of these examinations confimed 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 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 arc back gouged and back welded. Inaccessible welds were strenghtened by the welding of a stiffner to the face of the weld in orde,r to provide the equivalent weld strength. Conditions of this type were not found when sisalignment problems were corrected although in certain instances arc back gouging and back welding were performed as part of the realignment of installed partial 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.
- c. NRC Conclusion
- The NRC found no evidence and/or information to substantiate this allegation.
- 24. Allegation No. 24
- a. A11egation' Misalignment of the condenser tube support required rewelding i
so often that in some cases, the " mother material" around the t weld had to be cut out and replaced with a fresh substitute section.
- b. NRC 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
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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 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 reinstallation, 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 I
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 50
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 pe mitted 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 perfomed, 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 acre skilled welders to the higher classification of welds vould mean that the best craftsmen are working on the more sensitive jobs, a process consistent with good practice and ginerally enhancing the safety of the general public.
The NRC observed that if the inspection requirements were pe formed on a random sampling basis rather than the system described above, then the activities as alluded to by the allegation might have the potential of affecting the overall quality of the job by alerting individuals to those safety-re-
!ated welds to be " randomly selected". Since all safety-related selds are inspec+ed, this concern is not justified.
- c. NRC Conclusion _
The NRC found no evidence and/or information to substantiate this allegation.
- 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.
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- 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 2 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 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.
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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 permit 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 l have created the impress' ion of improper alignment which was l responsible fer this allegation. The chipping of the concrete l 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 information to substantiate this allegation.
l 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 fom and the concrete placed on top of it.
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- b. NRC Investigation Findirgs NRC investigators revicted concreta inspection records and N&D repoirts of concrete platements within the reactor building.
There were no records cf any cold joints in the Reactor Contain-ment Wall but N&D No. 514 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 placecent of additional
. concrete and no irregularities were noted. .
The NRC questioned several S&W QC inspectors and construction engineers as well as Of AVO craft and supervisory personnel concerning a cold joint.in the Reactor Primary Containment Wall. Some of the individuals questioned sentioned 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 Primary Containment Wa"1 at or near the 63' level.
The NRC noted that the requirements for preplacement inspections prior to the placing of concrete abose 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 crsft 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 tnere was nothing to support his allegation with respect to the lack of cleanliness.
- c. NRC Conclusion ,
The NRC found no evidence and/or information ta substantiate this allegation.
i 28. Allegation No. 28 1
- a. Allegation ,_ ,
Soil percolation test results were falsified and test results l were withheld frca LILCO's submission of this information to the NRC.
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- b. NRC Investication Findings l
Due to the nonspecificity of this allegaticn, the NRC examined two (2) areas which appeared to have been responsible for the allegation being made. With respect to the soil percolation I
53
4 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 :
LILCO representatives with knowledge of this area and noted 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 !
occasion during the soil liquifaction boring program, and
, subsequent to a heavy rainfall, the building service engineer i asked the geotech engineer to take borings in the vicinity of the septic tanks to find out why they weald 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 so'il liquifaction work itself. The NRC noted that this matter was covered in the licensee's FSAR (Section 2.5) and Appendices 2A, 2I, 2J, 2L and 2M thereof and basically involved tests to obtain additional information on the soil conditions in the j 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 l
already been performed by the, NR:: as part of the overall licensing process.
1 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 I 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 i someone not familiar with the matter again may have misconstrued it as such.
i c. NRC Conclusion The NRC found no evidence and/or information to substantiate this allent on.
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- 29. Allegation No. 29
- a. Allegation -
Welder performance qualification records were postdated for welders who qualified after performing welds for which they had not been qualified.
- b. NRC Investigation Findings _ _
NRC investigators reviewed 1300 COURTER and CO. Nonconformance Reports (NRs) for indications of welding by improperly c,ualified 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 9-8 to P-1 material socket fillet weld in accordance to Weld Procedure Specification (WPS) 08011AA.
The NR was dispositioned to " accept as is" on June 11, 1979' 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 most emphatically that there had not been any postdating of 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.
- 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 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.
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- 30. A11ecation No. 30 .
- a. A11ecation Pressure was applied to construction workers by LILCO, its >
1 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 Findinas In order to determine the validity of the allegation, NRC i investigators provided several pathways by which current or former construction workers could contact the NRC with infor-
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mation in this regard specifically (i.e., intimidation) or or with respect to construction defects in general. The results of these efforts are reviewed below.
- 1) 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 re.nort. The notice provided a minimum of three (3) points of contact with the NRC, one of 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 former 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, i
the NRC received two (2) phone calls. One of the phone calls was from a construction worker at the site who l stated that he had worked at other non-nuclear power plants and that by comparison, Shoreham was "over-designed" j and " super-safe". The other caller was a member of the l general public who had concerns about nuclear power in l 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.
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- 11) Interviews of Construction Workers On-Site 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 determine the validity of this particular allegation.
In the latter case, site representatives for various major 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 randca a sampling of carpenters, laborers, masons, welders, electri-cians, boilermakers, teamsters, mi11 wrights, insulators, steamfitters, sheetmetal workers, weld rod control clerks and QC inspectors. Through the sampling process, shop stewards as well as workers were also selected. Each individual was (a) interviewed in private, (b) informed of confidentiality, (c) provided with information for a 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 as stated in the allegation.
Every individual interviewed informed the NRC that none of 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 construction defects and/or irregularities to the NRC.
Each individual further informed the NRC that they had neither observed nor heard of such actions being executed on any of their friends or other, employees not directly quest.foned. The NRC did not receive any callbacks at other times from the individuals interviewed in this regard.
Several of the workers interviewed volunteered statements to the NRC to the contrary of this allegation stating that if anything, the opposite of the allegation was true.
Some of tb. statements made are included as follows:
Steamfitter A stated: "We never took a shortcut. This is a '
Class A job. Our relationship with LILCO is good and they cooperate fully with our concerns."
Mi11 wright A stated: "Our local is very conscientious with nuclear energy. This (RCI) is the most conscientious outfit I've worked for."
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Boilermaker A stated: "The union would be very supportive of our complaints."
Boilermaker 8 stated: "We've done our job right to the specs."
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 8 stated: "I haven't seen any problems in 5 years."
Electrician C stated: "No threats whatsoever. They (union and contractors) promote safety, especially on a nuclear job." .
Electrician 0 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."
M111wricht 0 stated: "Just the opposite. We are all told to look for problems. I haven't seen any so I don't know how they'll handle it."
Boilermaker 0 stated: "I would go to the NRC without hesitation. I live in the area and I wouldn't want anything to go wrong." ,
Sheetzscal Worker A stated: "They (foraman) tell us to report everything suspicious to them. I wouldn't be afraid to speak up."
Sheetmetal arker B stated: "The company here (LILCO) is far superior to any other company I've ever worked for."
Sheetmetal Worker C stated: "Noone 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."
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- t 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 26 years."
Mason A stated: "This is one of the best built jobs I've worked on."
Mason 8 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) l 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 and holidays, and that to date he had yet to be approached in this regard.
iv) Protective Agreement Attorneys for various al'legers claimed that several other l
workers wished to present information on construction defects but would not do so without a formal protective agreement provided by the NRC in order to ensure their confidentiality. A draft of this agreement was presented by the attorneys to NRC investgators on December 17, 1979.
This agreement was forwarded to NRC Headquarters for legal review and subsequently sent to the allegers' attorneys l for implementation. When no new workers came forth even 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 additional allegers due to potential threats to their jobs and/or persons.
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The NRC could not reconcile this statessent insofar as it had noted in 1), 11) and 111) above that no evidence of any threats whatsoever could be substantiated during this investigation.
- c. NRC Conclusion The NRC found no evidence and/or information that would sub-stantiate this allegation.
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i IV. EXHIBITS Exhibit A: Referenced Codes. Specifications and Procedures ACI 301 Specifications for Structural Concrete for Buildings ACI 306 Recommended Practice for Cold Weather Concrete ACI 347 Recommended Practice fo Concrete Forework ASTM (Draft 14) Proposed Manual of Coating Work for Light Water Nuclear Power Plant Primary Containment Facilities (Subcommittee 0-01.43)
SH1-64 Specifications for Substructure Work SH1-354 Specification for Concrete Work SH1-228 Protective Coatings Within the Primary Reactor Con-tainment Structure SH1-228.700 Procedure for Applying Protective Coatings Within the Suppression Chamber W 300-Section A General Procedure for Cadwelding of Reinforcing Steel QC-10.3 Field Quality Control Procedure - Concrete Quality Control QC-14.2 Field Quality Control Procedure - Cadwelding QC1-10.3.012 Quality Control Instruction - FQC Preplacement Pour Card .
QCl-10.3.013 Quality Control Instruction - Preplacement Inspection QC1-14.2.001 Cadweld Data Analysis 1
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p' ( EXHIBIT B
- UNITED STATES
,( g NUCLEAA. REGULATOftY COMMISSION g a REGION I g S31 PARK AVENUE
(' "g e'g KING CP PRUSSIA. PENNSYLVANIA 19404 NOTICE The (f. 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 l
1
h F. The New Brunswick, Canada Earthquake (Tr.1161)
Staff Response (Prepared by Dr. R. Rothman)
On January 9, 1982 at 12:53:51.8 GMT a magnitude m.. 5.7 earthquake occured in central New Brunswick, Canada epicenter 46.98 N 66.66*W, depth 10 kilometers. This main shock has been followed by over 1000 aftershocks during the past six months. The largest aftershock to date, a magnitude m b
5.4, occurred within three days of the main shock. There have also been several aftershocks in the magnitude range m b 4.5 to 5.1.
However, most of the aftershocks have been much smaller.. The maximum Modified Mercalli (MM) intensity reported for this earthquake, which occurred in a remote and sparsely populated region, is MM VI.
Earthquakes of intensity as high as MM VI have historically occurred in this region. There are no magnitudes for these earlier events because they occurred before the widespread deployment of seismographs. Earth-quakes with maximum MM intensity VIII have usually been associated with magnitudes of approximately 5.7 in the eastern and central U. S. The reporting of a MM intensity VI for this magnitude 5.7 earthquake can mean either that higher intensities were not noted because of the sparsity of population and lack of man made structures or because of some geologic condition in this region which results in lower intensities.
A network of stong motion seismographs has been deployed in the epicentral region by the Canadian Earth Physics Branch and several V. S. ;
Government Agencies and Universities. Some of the larger aftershocks have been recorded by this network and are being analyzed. This should contribute to our knowledge of eastern North American earthquakes by l
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providing near field strong motion records from which peak acceleration and response spectra can be obtained and correlated with instrumentally determined magnitudes.
The January 9,1982 earthquake and its aftershocks appear to have occurred in the New England Piedmont Tectonic Province. The occurrence of these earthquakes has no direct bearing on the seismic adequacy of the Shoreham Nuclear Power Station, since Shoreham is located in a different tectonic province, the Atlantic Coastal Plain Tectonic Province, and its design is based on the conservative assumption of the occurrence of a MM intensity VII earthquake in the site vicinity.
[See also Tr. 4275-78 where Dr. Rothman responded to related questions from Judge Morris.J i
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STATUS REPORT ON SH0REHAM SER OPEN ITEMS On Tr.1167 through 1174, Judge Morris requested a status report on all outstanding SER open items. In compliance with this request, the Staff submits the following report. The input for the report was provided by the Division of Licensing, Office of Nuclear Reactor Regulation. The report consists of a synopsis of the status of the review for each open item that was not resolved in either SSER #1 or SSER
- 2. These are the same open items as those listed in the Staff's March 29, 1982 status report to the Board, with the addition of open item
- 64 - Loose Parts Monitoring.
The first section of this report consists of the items now consid-ered resolved. The resolutions will eventually be formalized in the '
third SER Supplement. The second section of this status report includes those items which remain unresolved or confirmatory.
For the convenience of the Board and Parties, those items which relate to the unscheduled contentions in this proceeding are as follows:
Remote Shutdown Panel (SC 1) Item #62 Environmental Qualification (SC 8/ SOC 19(h)) Item # 9 Qualification of Electrical P.enetrations Item # 9 .
(SC 32/ SOC 19(f))
SeismicQualification(SOC 19(i)) Item # 8 Containment Isolation (SC23) Items #36
- 57-II.E.4.2
- 61 The Staff will submit to the Board and Parties a report, or a copy of the internal memorandum, resolving these contested items as they are closed.
The memorandum on Item #36 is included with this status report.
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I. ITEMS RESOLVED Item #1--Pool Dynamic Loads In Sections 3.8.1 and 3.8.2 of SSER Supplement 2 we indicated that the Applicant had made certain committments on pool dynamic loads as they relate to the containment and concrete and structural steel internal structures. These committments have been met and these items are essentially resolved to the satisfaction of the NRC Staff as identified in Memorandum dated April 9, 1982 (J. P. Knight to R. L. Tedesco). For clarity, the Applicant is submitting a letter explicitly stating that the response spectra used for analyses envelope the design loads established in DAR Rev. 5 and that the plant is constructed to those analyses. -
Item #6--Downcomer Fatigue In Section 3.9.2.1 of SSER #1 we noted that we were awaiting documentation of the final results of analyses relating to downcomer fatigue analysis. This committment has been met and this item has been resolved to the satisfaction of the NRC Staff as identified in a Memorandum dated April 5,1982 (J. P. Knight to R. L. Tedesco). .
Item #15--Inservice Testing of Pumps and Valves In Section 5.2.2 of SSER #2 we noted that LILC0 had committed to submit, prior to licensing, its inservice testing program for pumps and valves for a preliminary review and action by the Staff on any requests for relief from ASME code requirements. LILC0 submitted this information by letter dated January 6, 1982. Based on the Staff's review this matter
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_3-has been granted interim approval in a memorandum and SER input dated Februa ry 19, 1982 (James P. Knight to Robert L. Tedesco). This interim approval does not limit Shoreham to low power operations. Final approval of LILCO's program is scheduled for late 1982.
Item #26--Suppression Pool Bypass In SSER #2 the Staff noted that LILC0 must show that the acceptance criterion (equal to or less than 10% of the bypass capability) for the limiting small break be demonstrated before fuel load. This could be demonstrated by either showing that the leakage from the preoperational high pressure test meets this criterion or by performing a separate test at low pressure (equal to the hydrostatic head in the downcomers). -
By letter dated April 23, 1982 LILC0 agreed to perform such a low pressure test. Assuming successful test results, this item is considered resolved. An SER input reflecting this resolution will be prepared.
Item #36--Containment Purge System Section 6.2.3.1 of SSER #1 identified Staff concerns with a debris screen and the operability of the ,6-inch valves in the vent line. LILC0 submitted details on the debris screen by letter dated January 11, 1982.
The Staff considers this issue resolved and will address this in its next SSER.
With respect to the 6-inch valves, LILC0 in a November 23, 1981 letter committed to demonstrate valve operability by an in-situ test.
This item is now considered resolved. The NRC Staff will perform a confirmatory review of the test results prior to fuel load. The Staff
evaluation of the operability issue is covered in an SER input attached to a March 8, 1982 Memorandum (William V. Johnston to Robert L. Tedesco).
(Attachment 1 to this report).
Item #57-I.A.1.1--Shift Technical Advisor In Section 22.2.I.A.1.1 of SSER #1 we found that the Applicant meets this TMI-item subject to submittal of acceptable qualifications for each Sin candidate before assignment to STA duty. Since this is to be an ongoing requirement, the Staff plans to incorporate this provision as either a license condition as noted in the April 17, 1982 memorandum (J. S. Zwolinski to A. Schwencer) or into the Technical Specifications.
This item is considered resolved.
Item #57-I.D.1--Control Room Design Review SER Supplement 2 reported on five (5) control room review findings considered by HFEB and the Region I office to have unacceptable solutions. LILCO conducted a further review and submitted the results to the NRC in a ic ter dated April 16, 1982. These results are acceptable and we consicer these five issues resolved. This will be documented in an SSER input by June 30, 1982. . .
LILC0 was reminded that SER Supplement 1 states in part that, "All corrective actions are subject to NRC review and verification after they are implemented by the applicant." LILC0 has agreed to submit a letter to the Staff by mid July,1982 indicating the status of this implementa-tion for all corrective actions so that a schedule for a review by the NRC Staff can be established.
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5-Item #57-II.F.1, Attachment 1--Accident Instrumentation - Range Noble Gas Effluent Monitors In Section 22.2.II.F.1 we stated that the Applicant's procedures for monitoring noble gas effluent releases during an accident are acceptable, subject to confirmatory documentation relating to certain items. Through submittals dated January 7, 1982 and February 17, 1982 (Revision 25 to the Final Safety Analysis Report for Shoreham), the Applicant has provided the required confirmatory documentation. In addition, the Applicant has confirmed that installation and calibration of the necessary instrumentaion will be completed in August 1982. Based on our review of the above mentioned submittals and Applicant's commitment to complete installation and calibration of the applicable instrumentation -
in August 1982, we have concluded that the Applicant's response to Attachment 1 of TMI Action Plan II.F.1 of NUREG-0737, " Noble Gas Effluent Monitor", is now complete and is acceptable. This closes the open item relating to Attachment 1 of Item II.F.1. An SER input on this item is attached to an April 5, 1982 Memorandum (R. W. Houston to R. L. Tedesco).
Item #57, II.F.1, Attachment.5--Containment Water Level Monitor .
The Staff has reviewed the design of the proposed bubble system for suppression pool level measurement and finds that it meets the require-ments of II.F.1, Attachment 5. LILC0 has informed the Staff that they may not meet the implementation schedule of NUREG-0737 (fully operational prior to fuel load). The Staff has concluded that delayed implementation is acceptable based upon the following rationale provided by LILC0:
(1) A "best effort" was made to install the system prior to fuel load. The delay was caused by manufacturing delivery delays.
(2) The Applicant currently has installed a water level monitoring system that is comparable in all respects with those of the above required system with the exception of low range monitoring' capability.
Based upon the above, the Staff finds the monitoring system and delay acceptable and the matter resolved. Equipment qualification will be addressed under Item # 9.
Item #57-II.K.3, Item 28--Verify Qualification of Accumulators On Automatic Depressurization System Valves LILC0 has committed to perform periodic testing of the ADS valves to meet concerns raised by the Staff in response to LILC0's letters of July 21, 1981 and revision of August 7, 1981. As a result of this commitment the Staff finds this item closed satisfactorily as described in a memorandum dated May 11, 1982 (William V. Johnston to Robert L.
Tedesco).
II. ITEMS REMAINING UNRESOLVED Item #8--Dynamic Qualification As noted in Section 3.10 of SSER #1, the Staff concluded that the Shoreham program, when fully implemented, will provide adequate assurance that seismic category I mechanical and electrical equipment will function under the specified loads. However, we noted that based on the first SQRT audit, completion of the program was not far enough along to reach a O
conclusion in implementations without returning for a second, confinnatory audit. By letter dated June 11, 1982 LILCO has provided additional information on implementation which is currently under Staff review to determine whether degree of completion is now sufficient to warrant committing Staff resources to that second confirmatory audit.
The Staff expects to be able to make such a determination in early July.
Item #9--Environmental Qualification Section 3.11 of the SER noted that the Staff would provide its evaluation on environmental qualification of electrical equipment in an SER supplement. In June the Staff conducted a site audit of electrical equipment qualification records and a trip report and SER input is in ~
preparation. There are some minor open items requiring correction by LILCO. The trip report and SER input should be available by early July, 1982.
The Staff has also requested LILC0 to provide information on mechanical equipmeat. LILC0 has indicated that the requested information will be available in mid July. Staff review and evaluation is likely to l
l take 1 to 2 months to evaluate and prepare an SER input based on that .
evaluation. Therefore, SER input is currently being targeted for mid August, 1982.
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Item #10--Reactor Seismic and LOCA Loadings l
The major part of the Staff's evaluation of seismic and LOCA loadings on fuel assemblies has been completed as described in Section 4.2.3.4 of SSER #1. For the reasons stated in SSER #1, although fuel l
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assembly lift-off is not expected to be significant for this plant we have required that the Applicant submit a detailed evaluation, which is being performed by General Electric, before fuel loading. The Applicant expects to submit this confirmatory report in August, which will meet the Staff's requirement.
Item #11--Supplemental ECCS Calculations With NUREG-0630 Model In Section 4.2.3.2 and 4.2.3.3 of SSER #2, we noted that, prior to restart after the first refueling outage we would require LILC0 to submit an ECCS reanalysis for NRC approval to show adequate peak cladding temperature margins will exist in the fuel.
For the low fuel burnup associated with first cycle operation we -
concluded that compensating PCT margins associated with thermal hydraulic model improvements can be used to offset ballooning and rupture
, uncertainties.
. We consider this matter resolved even though LILCO does not agree with the Staff that such a reanalyses is warranted. The Staff position remains as stated in SSER #2. We plan to condition the license to require NRC approval of an ECCS reanalysis prior to plant operations l
after the first cycle operation.
Item #19--Preservice Inspection Preservice Inspection (PSI) for the reactor coolant pressure l bounda.ry is discussed in Section 5.2.7 of the SER. PSI for Class 2 and 3 I
l components is discussed in Sections 6.6 of the SER and supplemented in l Section 6.6 of SSER #1. We noted in these sections of the SER that when l
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LILC0 submitted requests for relief'feb.a ASME Code requirements the Staff would evaluate such requests. LILC0 has indicpted it will submit such requests in early August 1982. After seeing the responses the Staff will then be able to estimate the time required for its review.
In Section 6.6 of SSER #1 the Staff also noted that LILC0 had committed to perform a PSI examination of a representative sample of ASME Code Class 2 welds. LILC0 has indicated that results if the Class 2 weld examinations will be available around the end of June 1982. Unless significant defects are discovered from these examinations, no further
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NRC Staff evaluation of this PSI item will be requirad.
Items #28 and #32--Steam Condensation Loads In Section 6.2.1.8 of SSER #1 we noted that LILC0 had committed to use final load specifications to assess / confirm the adequacy of the design of the Shoreham plant with respect to steam condensation oscillation loads and steam condensation chugging loads. LILC0 had indicated orally that this has been done and will confirm this is a letter to the NRC Staff in June 1982. This letter should provide the basis for closing these confirmatory items.
Item #39--Emergency Procedures In Section 6.3.1 of SSER #1 we noted that pending satisfactory review of the emergency procedures for the leakage return system, that system is acceptable. By letter dated May 2, 1980 LILC0 provided identification and isolation procedures. We require that LILC0 would incorporate the detailed procedures into the plant operating procedures
s ..
for use by plant operators. Neither operating procedure SP.23.702.04 Rev. O nor the alarm response procedures submitted by LILC0 letter dated April 30, 1982 included these identification and isolation procedures.
The Staff will require these identification and isolation procedures be included in the plant procedures in a position to be forwarded to LILC0 the week of July 5, 1982.
In Section 6.3.2 the Staff noted that with respect to scram system piping the Applicant, following guidance of NUREG-0803, must implement revised emergency procedures for pipe breaks in the scram system. This information was forwarded to the Staff by LILC0 on May 13, 1982 and is currently under review.
Item #46--IE Bulletin 79-27 In Section 7.5 SSER #1 LILC0 committed to conduct a failure effects analysis of the Class 1E and non-1E uses to determine whether emergency procedures are adequate for dealing with resulting plant conditions.
LILC0 indicates that the study and any procedure revisions will be complete by early August 1982. The NRC Staff will confirm that this has been done. .
Item #47--Control System Failures In Section 7.7 of SSER #1 we noted that LILC0 committed to conduct a review to determine that power sources or sensors that supply power or signals to two or more control systems will not fail or malfunction in such a way as to result in consequences outside the bounds of the FSAR Chapter 15 analyses or beyond the capability of operators or safety
_ 11 -
systems. LILCO has indicated that it will provide the results of this study by the end of July 1982. The Staff will require, as noted in SSER
- 2 that all identified problems be corrected prior to full power operations.
Item #48--Environmental Effects of High Energy Line Break On Control Systems In Section 7.7 of SSER #1 we noted that the Applicant committed to conduct a review to determine whether the harsh environment caused by high energy line breaks could cause control system malfunctions that could result in consequences more severe than those of FSAR Chapter 15 analyses or beyond the capability of operators or safety systems. LILCO has indicated that it will provide the results of this study by mid August 1982. The Staff will require that all identified problems be corrected prior to full power operation.
Item #52--Management Organization Section 13.1 of SSER #1 identified several committments which require confirmatory documentation by LILCO. LILC0 had indicated that it .
plans to submit a Chapter 13 FSAR revision by the end of July 1982. We will review this revision to determine that the confirmatory documentation is provided.
Section 13.2 of SSER #2 indicated that LILC0 was providing additional information needed to complete our reviews. Additional information has been provided in a January 11, 1982 LILCO submittal.
LILC0 has indicated that additional information to resolve outstanding
e, s
- review areas will be submitted in June 1982. The Staff will complete its review following receipt of this information.
Section 13.4 of SSER #1 indicated the need to update the FSAR in several areas including documenting consistency with the guidelines of Regulatory Guide 1.33, Revision 2. As noted earlier LILC0 plans submittal of a revised FSAR Chapter 13 by the end of July. LILC0 has also indicated it will be in a position to commit to the guidelines of RG 1.33, Rev. 2 in June 1982. The Staff will complete its review following receipt of this additional information. Resolution is expected to take place during completion of the Technical Specification review.
Section 13.5 - The Staff review of this area is essentially complete
~
as noted in Section 13.5.2 of SSER #2. The submittal of the Revised Chapter 13 and confirmation that Regulatory Guide 1.33, Rev. 2 guidelines will be met, as noted above, should complete this area of the Staff's review. .
Item #53 and Item #57-III.A.1.1 and III.A.2--Emergency Planning and Preparedness -----
In Section 13.3 of SSER #2 the Staff concluded that no operating .
license would be issued unless a favorable finding is made by the NRC with respect to both onsite and offsite emergency preparedness. The Staff identified several items that needed to be completed.
On June 11, 1982, the Commission approved a final amendment to 10 C.F R. Part 50 and Appendix E, concerning energency planning. The rule provides that for the issuance of an operating license authorizing fuel loading and operation to 5% of rated power, no NRC or FEMA review,
- .. .. . _ _ _ . . _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ . _ . _._ ________.________.__________J
findings, a'nd determinations concerning the state or adequacy of the offsite emergency preparedness shall be necessary. The rule also specifies that emergency preparedness exercises are part of the operational inspection. The exercise must be conducted prior to operation above 5% rated power but is not required for any initial licensing decision process. The Staff is now attempting to resolve all matters except those related to offsite emergency preparedness issues.
The Staff has scheduled an on-site appraisal for July 19,1S32. An SER input is targeted for late August or early September 1982.
As noted in Section 22.7.III.A.2, the Staff's position requires satisfactory completion of an integrated emergency response exercise.
This is currently unscheduled, The joint exercise planned by FEMA for -
June 16, 1982 was cancelled.
Item #57, I.A.2.3--Administration of Training Programs for .
Licensed Operators In SSER #1 the Staff has required that confirmation be completed prior to fuel loading that all permanent members of the station staff who teach systems specific to BWRs, integrated responses, transients and simulator courses to licensed operators or licensed candidates have completed an SR0 exam.
LILC0 expects to have this qualification completed in August, 1982, at which time the verification by the NRC Staff can be completed.
Item #57, I.C.7--NMSS Vendor Review of Procedures
Following NRC Staff review of low power test, power ascension, and emergency procedures, the Applicant is required to have GE review these procedures to further verify their technical adequacy. The Staff is awaiting submission of the vendor comments to verify that the review has been completed. This LILC0 submittal is expected in early July 1982.
Item #57, I.C.8--Pilot Monitoring of Selected Emergency Procedures for NT0Ls An audit of emergency operating procedures had been conducted. The Staff is awaiting submission of the final version of the procedures, which is to include plant-specific data, and the final versions of the graphs to be used in the procedures. This LILC0 submittal is expected in early July 1982.
Item #57, II.B.3--Post Accident Sampling Capability The NRC Staff has evaluated LILC0's post accident sampling capabilities. LILC0 has committed to provide chemical procedures for offsite informatory analysis of pressurized grab samples. The Staff has required for review three items: .
(1) That information be submitted to demonstrate that reactor coolant and suppression chamber samples are representative.
(2) That a procedure be provided to relate specific radionuclide concentrations to the estimated core damage.
(3) That electrically operated components associated with post-accident sampling be capable of being supplied with power and operated within 30 minutes assuming loss of offsite power.
The Staff is reviewing the provided documentation to determine if additional information is required. Verification of the above requirements will satisfy the Staff's concerns in this area. Imple-
! mentation of the requirements is not necessary prior to low power operation (greater than 5%) because only small quantities of radionuclide inventory will exist in the reactor coolant system and therefore will not affect the health and safety of the public. Prior to exceeding 5% power operation LILC0 must demonstrate the capability to promptly obtain reactor coolant samples in the event of an accident in which there is core damage.
Item #57, II.D.1--Performance Testing of Boiling Water Reactor '
and Pressurized Water Reactor Relief and Safety Valves LILC0 has comitted to participate in the BWR Owners Group program for testing of safety and relief valves. LILC0 is reviewing the BWR .
program description and scope to ensure that it is applicable to Shoreham plant specific valves and piping. LILC0 has reported satisfactory test results for Shoreham plant specific safety and relief valves based upon l preliminary review of the generic.BWR test program results in a transmittal dated December 9,1981.
The NRC Staff is currently reviewing the above test results to i
verify satisfactory completion of this requirement. The Staff has
- '.dentified additional information which must be addressed by LILC0 before l
the review of this item can be completed; specifically the applicability i
of the generic test results to Shoreham's safety / relief must be justified. This information has arisen from the review of report NE
l 1
DE-24988-P as discussed in an internal memorandum dated June 21, 1982 ;
(Zoltan R. Rosztoczy to Albert Schwencer). This request for information will be transmitted to LILCO the week of July 5, 1982.
l Item #57, II.E.4.2--Containment Isolation Dependability In Section 22.2.II.E.4.2 of SSER #1 we noted that LILC0 had met all l requirements for this item except for valve operability and the provision for isolation due to a high radiation should this occur while a purge l l
line is in use. By letter dated November 23, 1981 LILC0 proposed an in-situ valve test satisfactory to the NRC (see discussion under item 36). By letter dated January 1,1982, LILC0 agreed to the addition of a high radiation signal to isolate the purge line when in use during ~
operating conditions 1, 2 and 3 (power operations, startup, and hot shutdown). This is acceptable to the Staff. However, the Applicant's proposal to install the isolation signal after commerical operation has not been justified. The date for installation and operability for this isolation signal capability remains to be resolved. LILC0 has indicated it will address the date of installation by August, 1982.
Item #57, II.F.2--Instrumentation for Detection of Inadequate Core Cooling The Staff has found LILCO to be in partial compliance with this requirement and has found in the review the need for the following:
(1) Incorporation of thermocouples into the ICC monitoring system prior to June 1983 in accordance with Regulatory Guide 1.97; m
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(2) Documentation addressing the inclusion of thermocouples in the fuel ICC monitoring system; (3) Performance of a study to confirm that the LPRM assemblies are the most suitable location for incore thermocouples or propose an alterrate location and/or number at thermocouples to detect inadequate core cooling.
LILC0 is scheduled to submit documentation to address the above
/
concerns in August, 1982.
Item #57, II.K.1, Item 5--Assurance of Proper Engineered Safety Features Functioning ,
LILC0 has c'mmitted to review procedures to ensure that safety -
related valves are returned to their correct positions following ,
necessary manipulations. Surveillance testing and maintanance procedures are also being reviewed to verify that they also meet the requirements of this item. The NRC will review these procedures to verify that the intent of this item is met. LILC0 proposes to be prepared for such review in July, 1982.
Item #57, II.K.1, Item 10--Safety-Related System Operability Status Assurance LILC0 has committed to review plant surveillance testing and maintenance procedures for removinr T tm end returning to service safety related systems. The NRC wi): ' r t ew .hese procedures to verify that they meet the intent of this item. LILC3 intends to have the procedures ready for review in July, 1982.
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- Item #57, II.K.3, Item 13--Separation of High Pressure Injection and Reactor Core Isolation Cooling System Initiation Levels The Staff has determined that separation of High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiation levels is unnecessary at this time but notes that LILC0 is subject to the results of the ongoing generic evaluation of the topic.
LILCO has committed to make an earnest effort for installation prior to fuel load of the automatic restart of RCIC on low water level. Once installed, the Staff will verify that the system meets the requirements of this item.
Item #57, II.K.3, Item 16--Reduction of Challenges and Failures of-Relief Valves-Feasibilty Study and System Modification LILCO is a participant in the ongoing evaluation by the BWR Owners L
Group of possible ways to reduce challenges to safety / relief valves.
l That study encompasses the 2-stage Target Rock safety / relief valve which is used at Shoreham. LILC0 has provided the results at the evaluation as prescribed in II.K.3, Item 16. We conclude that no modifications are necessary for Shoreham at this time.
This item is subject to generic review by the Staff. This review is scheduled to be completed in December,1982. No action is required by LILC0 until the generic resolution is provided at which time LILC0 will be provided with a resolution implementation schedule.
l Item #57, II.K.3, Item 45--Evaluation of Depressurization With l
Other Than Automatic Depressurization System I
LILC0 reported the results of a BWR Owners Group study in a letter dated April 15, 1981. Based upon the information presented in this letter, the Staff has concluded that no change to the current mode of depressurization is necessary for Shorcham but that LILC0 is subject to the results of a generic review and lab data confirmation of this item.
Item #57, III.A.1.2--Upgrade Emergency Support Facilities LILC0 has proposed a two phased approach to meeting the upgraded emergency response facility requirements as detailed in NUREG-0696. The first pher.e prcvides for compliance with all of the criteria of NUREG-0696 except for the data acquisition system by the fuel load date of September, 1982. LILC0 is modifying their plant computer to provide -
recording and display of critical plant parameters as an interim data acquisition system. LILC0 has committed to have facilities in full compliance by April,1983. The Staff has approved this approach as a result of discussions with LILCO. The Staff is currently reviewing LILCO's latest emergency response facility submittal of January 4,1982.
Acceptability of the facilities will be evaluated at an onsite post implementation review. In addition an onsite emergency preparedness appraisal will be conducted prior to fuel loading at which time the adequacy of the emergency response facilities will be evaluated against the interim criteria in NUREG-0654 and NUREG-0578. All major deficiencies indentified during this appraisal must be corrected before the Staff can make a favorable finding under the provisions at 10 C.F.R. 50.47(s)(2).
Item #57, III.D.3.4 Control Room Habitability In Section 6.4, " Habitability Systems," of the Shoreham Safety Evaluation Report, NUREG-0420, issued in April, 1981, the Staff concluded: (1) that the proposed control room habitability system meets the dose guidelines of General Design Criterion 19 and taat the radiation protection provisions are acceptable; and (2) that there is no risk to control room personnel from toxic gases and that the design of the control room is acceptable from this standpoint.
In a submittal dated May 15, 1982, the Applicant responded to post-TMI requirement III.D.3.4, " Control Room Habitability," as promulgated in NUREG-0737, " Clarification of THI Action Plan Requirements" issued in November,1980. In the submittal, the Applicant confirmed the adequacy -
of the control room habitability system, and committed to include provisions for carbon dioxide detection and alarm. This conclusion is consistent with the Staff findings as described above.
With respect to control room habitability, the Staff concludes that, when the carbon dioxide modifications have been completed, the Applicant will have satisfied the requirements for a full power license. Final verification of proper implementation is scheduled for July,1982.
Item #60--Station Blackout-Generic Letter 81-04 LILC0 will have station procedures to deal with Station Blackout completed in early July,1982. Operation training for operators who will be involved with fuel load will be completed by early July,1982. The Staff will review the procedures and training to verify their adequacy to
meet the requirements of the generic letter upon notification by LILC0 of their completion of this item.
Item #61--Scram System fiUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," as described in SSER #1 provides guidance for plant specific resolution of issues related to a postulated break in the scram discharge volume. The Staff is currently reviewing LILC0's response to this item provided in their letter dated May 13, 1982 (J. L.
Smith to Harold R. Denton).
Item #62--Remote Shutdown System -
The Staff identified concerns regarding the capability at the Shoreham Remote Shutdown Panel (RSP) to comply with the Staff position (including compliance with 10 C.F.R. 50, Appendix A (GDC-19), Appendix.K and Appendix R) which was transmitted to LILC0 on August 31, 1981. LILC0 responded to this transmittal on llovember 23, 1981 providing additional, design information regarding the RSP. The Staff raised additional questions concerning the design transmitted on February 3,1982 to which LILC0 responded on April 20, 1982 and June 2, 1982. A meeting was held with LILC0 and Stone and Webster on June 14, 1982 to discuss additional design details regarding the RSP design. As a result of this meeting the Staff has concluded that the present RSP design does not comply with the Staff position. The Staff has delineated additions to be made to the RSP design and requested the submittal of formal documentation. This information was included in an internal memorandum dated June 22, 1982
(Themis P. Speis to Robert Tedesco) which will be transmitted to LILC0 the week of July 5, 1982.
Item #63--Design Verification In Section 1.7 of SSER #2, the Staff noted that we are seeking an additional level of assurance that Shoreham has been designed and constructed in accordance with the application. On March 15, 1982 LILC0 presented an overview of its QA program to the Staff. On April 19, 1982 LILC0 documented the material presented at that meeting and provided additional information and clarification of serveral items. The April 19, 1982 letter also indicated that LILC0 had contracted with Teledyne Engineering Services (TES) to undertake a limited scope '
independent design verification similar to one undertaken for LaSalle.
On May 26, 1982, LILC0 provided the Staff with a Program Plan for the TES effort. This effort is scheduled for completion in August 1982. The i
Staff intends to monitor this effort and may suggest changes or additions if appropriate.
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1 j Item #64--Loose Parts Monitoring System l
LILC0's Loose Parts Monitoring System (LPMS) has been ietermined to be not in conformance with their commitment to meeting the guidelines of Regulatory Guide 1 133. At a meeting held on June 8, 1982, LILC0 was requested to provide an amendemnt to the Shoreham FSAR containing the following information:
(1) A formal submittal of the LPMS descriptive material previously provided to Dr. Y. Hsii of the
! e Staff in draft.
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(2) Documentation and detailed justification of all deviations from R.G.1.133 for LPMS design and operation.
(3) A commitment to providing a more detailed report of the LPMS installation calibration and -
operating program after startup testing is complete and prior to operation above 5 percent power.
This request for information will be fomarded to LILC0 the week of July 5, 1982. LILC0's response will be reviewed for its adequacy.
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GAR 0 8 M Docket No. 50-322 MEMORANDUM FOR: Robert L. Tedesco, Assistant Director for Licensing Division of Licensing FROM: William V. Johnston, Assistant Director Materials and Qualification Engineering Division of Engineering 4
SUBJECT:
SHOREHAM NUCLEAR POWER STATION UNIT 1 (SNP-1)
INPUT FOR SUPPLEMENTAL SAFETY EVALUATION REPORT Plant Name: Shoreham Unit 1 Licensing Stage: OL Responsible Branch: Licensing Branch No. 1 Responsible Project Manager: J. Wilson i
The enclosed Supplemental Safety Evaluation Report (SSER) was prepared by DE: MQE, Equipment Qualification Branch. This SSER input covers II.E.4.2 as. it relates to the operability of the purge and vent system isolation valves. ,
I William V. Johnston, Assis nt Director Materials and Qualification Engineering Division of Engi.neering ,
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Enclosure:
As stated cc: R. Vollmer Z. Rosztoczy "
W. Butler B. Youngblood J. Wilson G. Thomas T. Restivo, BNL M. Haughey CONTACT: M. Haughey, NRR Ext 49-28386 e
P 11.F.4.? Contaie ..t 1,olation D r pi n d.,t i t i t .
[This evaluation 6ddresses the issue of valve operability on the purge system isolation. valves. The Containment Systems Branch will address the remainder of this issue].
Requirement Containment purge valves that do not satisfy the operability c
criteria set forth in Branch Technical Position CSB 6-4 or the ,
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Staff Interim Position of O c t ob e r 23, 1979 must be sealed closed as defined in SRP 6.2.4r item II.6.f (NUREG-800) during operational conditinns. Furthermore< these valvas must be verified to be closed at least every 31 days. .
Applicants must be in compliance with this position before they receive their operating licens'e.
Discussion'and Conclusion j The Shoreham primary containment purge system purge valves are 18" valves as follows:
1T46*A0V038A 1T46*A0V038B /
1T46*A0V039B ,
1T46*A0V039A 1T46*A0V039C 1T46*A0V039D 1T46*A0V038D 1T46*A0V038C, These valves are to be operated in cold shutdown and re f ueling modes only. The valves listed above are t o b6 sealed closed per SRP Section 6.2.4 item II .6. f (NUREG-800) in modes other than cold' shutdown or refueling and verified closed every 31 days.
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1ht p r i t..,r y c orit a ineru rit p u r ry e t, y s c it. vent vslvea ase 4" ;4 ri af 6" copes Vulcan Valves with D-100-100 operators as follows:
4" 1T24*A0V004 A B
IT24*A0V001 A B
6" IT46*A0V078 A <
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IT46*A0V078 8 IT46*A0V079 A IT46*A0V079 8 The applicant has committed in their letter of November 23, 1981 to perform a test on a representative 6" valve as follows:
During the containment st ruct ural integrity acceptance tpst while the containment is pres'surized to approximately 55.psig, one o f t he 6" vent valves will be openede maintained open fo'r .
a time sufficient t o at t ain steady state flow conditions, and then closed. Cont a inment pressure'will be mpnitored to ensure ,
that pressure is maintained above 48 psig-for the duration of this i s test. The applicant has st at ed' t h e peak calculated pressure of ~ t he drywell following the design basis accident is 41.9 psig allowing ,4 a margin of 8.1 psi during the test. Flow direction during the test will be in the conservative di re c t i on, the di rect i on that .
tends to open the valve. Closure time of the valve will be monitored to occur within 5 seconds. At the completion of the test the valve is to be visually inspected.
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! In the applicant's let t er oi fl o v e rals e r 23r 1981 it was s t'a t e d that the 4" air operated valves are the same model as the 6"
, valve and use the same size 100 air operator. The tests performed on the 6" valve therefore will be a conscrvative demonstration of operability of the 4" valves. ,
Based on the applicant's commitment to perform this test .
the staff finds the applicants program to meet this requirement satisfactory contingent upon a, confirmatory review of the results of this test prior to fuel load.
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NRC STAFF RESPONSE TO BOARD UNRESOLVED SAFETY ISSUE QUESTIONS (Prepared by Generic Issues Branch, Division of Safety Technology, Office of Nuclear Reactor Regulations)
The Board at Tr. 1163 et. seq. made an inquiry into the status of certain unresolved safety issues. The Staff's response follows:
INTRODUCTION The Staff's basis for the acceptability of licensing Shoreham pending resolution of outstanding USI's is more fully discussed in the following' questions and responses. Generally for those USIs that are generically unresolved the staff has discussed in Appendix B why it believes that the current NRC regulations and guidance in the standard review plan are .
adequate to permit continued licensing while the longer term generic USI review is underway. These current requirements are supplemented with additional interim requirements in some issues when this is believed to l
l be necessary. The discussion in Appendix B also provides a reference to .
the section(s) of the Shoreham SER where a plant specific assessment is conducted for each issue in accordance with the regulations, standard review plan and any additional interim requirement. This provides the basis for safe operation and continued licensing of nuclear power plants prior to the ultimate resolution of the unresolved safety issues.
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Question Page 1166 of the Shoreham hearing transcript questions the basis for the Staff's conclusion that the Shoreham plant can be operated safely prior to the ultimate resolution of USI A-40.
Response
The Staff's basis for concluding that Shoreham can be operated safely, though USI A-40, Seismic Design Criteria, is generically unresolved, is provided in NUREG-0420, and specifically for Shoreham in the Safety Evaluation Report in Section 2.5.2. Basically, the basis of the conclusion in Section 2.5.2 is that the design spectral shape used for Shoreham is more conservative than a Housner sepctrum, and that a -
Mercalli Intensity VII earthquake is adequately represented in the Housner shape.
Question Page 1165 of the Shoreham hearin5 transcript indicates a question regarding the status of USI A-44 as it relates to the Shoreham plant.
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Response
The A-44 Station Blackout issue remains generically unresolved at this l
l time. The Staff's basis for licensing operation of Shoreham prior to generic l
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,e resolution is found on pages B-12 and B-13 of the Shoreham Safety Evaluation Report, NUREG-0420. In summary, as noted in the SER, the Staff has concluded that a combination of design, operating, and testing requirements that have been imposed on the Applicant will assure that Shoreham will have substantial resistance to a loss of all alternating current and that, even if a loss of all alternating current should occur, there is reasonable assurrance that the core will be cooled until power can be restored. To add to this assurance, the Staff has imposed the additional requirement on the Applicant of interim emergency procedures and operator training for safe operation of the Shoreham plant and
' restoration of alternating current power.
In NUREG-0420 Supplement No.1 on page 1-9 item 60, we have identified Station Blackout-Generic Letter 81-04, as resolved pending confirmation.
This item relates to our above described requirement for imterim emergency procedures and operator training. The Applicant plans to submit a letter to NRC in early July,1982 connitting to our requirement. (See the status of SER open item #60.) We will follow up on this item ofter review of Appli. cant's submittal.
Question Page 1165 of the Shoreham hearing transcript the Board questioned the status of the Staff's review of Shoreham for issues A-46, Seismic Qualification of Equipment in Operating Plants and A-47, Safety Implications of Control Systems.
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Response
The Staff in NUREG-0420 Section 1.7 has identified both of these items as outstanding issues for Shoreham. In the case of A-46, this issue literally pertains to seismic qualification of equipment in operating plants only. However, the Staff has identified the seismic qualification of equipment at Shoreham as an outstanding item requiring additional information as outlined in Section 3.10 of NUREG-0420, Supplement No.1.
The Staff has conducted one on-site audit regarding this equipment and requires additional information to enable the staff to conduct a second audit before this item can be completed. We anticipate receipt of this information in June 1982. (See status report on SER open item #9).
Similarly, for the A-47 issue the Staff in NUREG-0420 Supplement No.1 Section 7.7 identified the need for confirmatory information. We anticipate receipt of this information in July 1982. Upon receipt of .
this information related to A-46 and A-47 we will report the result of our review in a Supplement to NUREG-0450 the Shoreham SER.
l Question .
l On Page 1167 of the Shoreham hearing transcript the question is raised as l
to how the River Bend Appeal Board criteria relate to Shoreham for the generically resolved USIs (i.e., A-8, A-10, A-24, A-31, A-36, A-39, and A-42). The three River Bend criteria are as follows:
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...the furnished information would likely shed light on such alternatively important considerations as whether: (1) the l
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i problem has already been resolved for the reactor under study; (2) there is a reasonable basis for concluding that a satisfactory solution will be obtained before the reactor is put in operation; or (3) the problem would have no safety implications until after several years of reactor operation and, should it not be resolved by then, alternative means will be available to insure that continued operation (if permitted at all would not pose an undue risk to the public."
Response
Page B-5 of NUREG-0420 contains a table of generically resolved USIs.
This table inadvertantly did not include cross references to the body of a the Shoreham SER where the Shoreham implementation of the generic resolution is provided. Instead these cross references were provided in
- Section 1.7 of the SER and in the SER Supplements 1 and 2. These cross references address the above question. This information from Section 1.7 of the SER and the SER supplements is summarized in the following table "Shoreham Status of Generically Resolved USIs."
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e Task Number A-8 NUREG-0487,"@
Evaluation anc Supplement 1 1 Supplement 2 NUREG-0808, "t.
and Acceptance A-10 NUREG-0619, "I Return Line Ni l
l A-24 HUREG-0588, "
Qualification A-31 SRP 5.4.7 and incorporate ri A-36 NUREG-0612,"I A-39 NUREG-0487 an<
NUREG-0802** '
Report BWR Ma' A-42 HUREG-0313, R1 Selection and Boundary Pipi
- The Shoreham plant has been upcoming supplement.
- This report is scheduled f!
proposed by General Electric l
SHOREHAM STATUS OF GENERICALLY RESOLVED USIs NUREG Report and Title Shoreham Status SER/SER Supplement Sections II Containment Lead Plant Program Load See Status of SER Open Itea #1 Supplement 1 ;
eptance Criteria," October 1978 3.8.1 REG-0487, October 1980 3.8.2 REG-0487, February 1981 f II Containnent Program Lead Evaluation iteria", August 1981 Feedwater Nozzle and Control Rod Drive Closed Supplement i e Cracking" 4.6.2, 5.2.1, 5.2.7
- rim Staft Position on Environmental Second Audit Completed Supplement 1 Safety-Related Electrical Equipment" See Status of SER Open Item #9 3.10 5-1," Residual Heat Removal Systems," Closed 5.4.2 rements of USI A-31 rol of Heavy Loads at Nuclear Power Plants" Closed Supplement 1 9.1.4
'pplement 1 to NUREG-0487 (See A-8 above) See Status of SER Open Item #1 Supplement 1 ety/ Relief Valve-Quencher Loads Evaluation (June 18, 1982) 3.8.1 I and III Containment" 3.8.2 ion 1, " Technical Report on Material Closed Supplement 1 vscssing Guidelines for BWR Coolant Pressure 5.2.6.2 siewed against the applicable SRP and BTP and found acceptable. This will be documented in an publication in the third quarter of 1982. The report documents NRC acceptance of the SRV loads Report # 22A7000 Rev.1, Appendix 3B for GESSAR II 238 Nuclear Island dated 11/25/80.
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