IR 05000322/1987010
| ML20237L619 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 08/21/1987 |
| From: | Bettenhausen L, Eapen P, Finkel A, Florek D, Hunter J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237L602 | List: |
| References | |
| 50-322-87-10, NUDOCS 8709090008 | |
| Download: ML20237L619 (29) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
Report No.
50-322/87-10 Docket No.
50-322 License No. NPF-36 Licensee:
Long Island Lighting Company 175 East Old Country Road Hicksville, New York 11801 Facility Name:
Shoreham Nuclear Power Station Inspection At: Shoreham, New York Inspection Dates: July 13 - 17, 1987 Inspectors:
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8 /9![f 7 Dr. P. K. Eapen, Tea 4n Leader
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5/Mh7 g.J'Florek,AptistantTeamLeader date Jb Vf/f 97 A. E. Finkel, Le d Reactor Engineer
/ /date WW9 Wn/n y.G.HunterIII,ReactorEngineer
~date FM #/
Approved by:
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Dr. E H. Bettenhausen, Chief, Operations date Branch, DRS Inspection Summary:
Special, announced team inspection conducted on July 13 - 17, 1987 (Report No. 50-322/87-10)
Areas Inspected:
Effectiveness of the licensee's engineering support to the plant and other such site organizations as maintenance, training and QA.
i Results: Two violations (Failure to establish isotopic composition of Boron in Standby Liquid Control System Tank prior to entering Operational Condition 2 l
and installing Raychem end caps in the drywell contrary to licensee's own
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procedure) were identified.
Except for these violations, the engineering support was observed to be effective.
8709090000 870902 PDR ADOCK 03000322 G
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TABLE OF CONTENTS PAGE NUMBER 1.
Introduction & Summary....
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Engineering Support'0 organization................2 3.
Modifications Program
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Engineering Assessments of Bulletins, Information Notices and Licensee Event Reports........
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Routine Engineering Support to the Site
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Plant Tours
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Exit Interviews
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Attachment A Attachment B
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I 1.D. Introduction and Summary I
The purpose of this inspection was to assess the effectiveness and quality of engineering support provided to the Shoreham Nuclear Power Station.
Through a carefully selected sample of design changes, engineering-
evaluations of plant occurrences and transients, technical information
provided by engineering to the Operations, Maintenance, QA and Training l
departments and routine activities, the inspectors assessed the technical
quality, timeliness and effectiveness of engineering activities in the Nuclear, Electrical, Mechanical and Instrumentation and Control disciplines.
The activities were also selected to assess the Systems Engineering, Fuels, Engineering Assurance, Project Engineering and Nuclear Analyses Divisions of the licensee's Nuclear Engineering Department.
The activities were chosen to assess the effectiveness of key elements in the licensee's design change process, the organization's ability and technical depth to provide engineering support and the licensee management support provided to engineering.
Each selected activity was reviewed against NRC regulat.ans and industry codes and standards endorsed by the licensee.
In general the team noted excellent management support to assure the quality and timeliness of engineering support provided to ohe nuclear station. The licensee is in the process of developing the technical depth for performing engineering support functions independent of the Architect Engineer (A/E).
At the time of this inspection about half of the engineer-ing activities were being conducted by the licensee.
The technology transfer program to assume design responsibility from the A/E was pro-gressing well and in accordance with the established schedule.
The team identified two violations (Failure to verify Boron isotopic composition in the Standby Liquid Control System Tank prior to entering Operational Cordition 2 and ir. stalling Raychem end caps in the dry well when it was not permitted by licensee's own procedures).
The inspectors concluded that the licensees engineering support activities were conducted effectively and in accordance with the NRC regulations, licensee pro-cedures and the industry codes and standards endorsed by the licensee.
2.0 Engineering Support Organization The Nuclear Engineering Department (NED) is headed by a Manager who reports to the Vice President Nuclear Operations. This department consists of six Divisions (Nuclear Systems Engineering, Nuclear Fuels, Engineering Assurance, Nuclear Project Engineering, Radiation Protection and Nuclear Analysis Divisions).
Each Division is headed by a Division Manager and is further divided into sections under Section Heads with staffs of two to ten engineers.
The Nuclear Engineering Department is authorized to have a complement of 114 engineering personnel. At the time i
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of this inspection, the department had a staff of 102 degreed engineering personnel with an average nuclear experience of 9.6 years.
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Additional personnel at the Office of Engineering reporting to a. separate
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.Vice President are dedicated to support the Shoreham site.
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.special. Divisions are dedicated.in Electrical Engineering (13), Power
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Engineering (.13) and Engineering Design and Mapping (20) Departments of the Office of Engineering.. The wor.k done by the.0ffice of Engineering is coordinated through the Nuclear Project Engineeri_ng Division in NED.
The licensee has developed the Shoreham Nuclear Power Station Engineering Assurance Program specifically to cover engineering work done by.the Office of Engineering and Nuclear Engineering Department.
The licensee also maintains a pool of dedicated personne1'from the A/E on a' quarterly basis by issuing quarterly staffing authorization letters.
For the. third calendar quarter of.1987, the letter authorized a total of 85 personnel from the A/E. Additional personnel can be obtained from the A/E for contingencies on an as needed basis.
The' licensee relies en plant staff engineers to implement modifications and respond to daily plant ne'eds it, the Reactor Engineer, Modifications Engineer and Systems Engineer groups.
Examples of unique licensee programs and activities to assure the adequacy of engineering support to the site are discussed below:
A. Status of Technology Transfer Program The licensee in USAR section 13.1.1.3 stated: "To ensure a continuing high level of engineering support, QA, and design control', an interim management control. program for station modification (called the Interim
. Station-Modification Program) has been implemented and will cover the period through approximately the first refueling.
This program has been implemented in accordance with approved administrative procedures. The qualified architect-engineer of record for the plant construction will be retained under this program to supply the necessary assistance to maintain the safety and operability of the plant." This program was in effect at the time of this inspection. -The licensee has developed a Technology /
Design transfer program to manage and control the transfer of engineering and design from the A/E.
This transfer program consists of ten elements which include staffing and
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training of personnel, development of the Station Modification program and I
the elimination of dependence on the A/E. The A/E developed and
. implemented a training program for the licensee. personnel.
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lectures were given. The A/E's procedures, standards and technical l
documents have already been transferred to the licensee.
Sixteen
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procedures have already been approved and nine others are in the approval cycle.
Each design document is assigned to a responsible licensee division.
This division is required to identify and develop the procedures that are prerequisites to transfer.
Licensee engineers
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identify documents to be' transferred and organize them into work packages.
The A/E provides the status and additional work required for each of these packages. Upon completion ofLthe task.the division manager performs a selfr assessment and announces readiness for transfer. The Engineering
. Assurance division performs an independent assessment of each division that announces readiness as a. prerequisite to authorize transfer.
At the time of this inspection the 1icensee transferred 50 of the 172 work packages and 17,000 of-the 70,000 documents identified for transfer. The licensee was performing about fifty percent of all design changes for the plant. At the time of this inspection the technology transfer was scheduled to be completed by December 31, 1987.
.In the interim, the A/E retdns verification authority for safety related n
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design;-however, the licensee performs its own verification of safety
- related designs in addition to the verification by the A/E.
B. Application' of Probability Risk Assessment (PRA)
The licensee recognized the importance of PRA in the design and licensing process and' initiated a study in 1981.
This study was designed to provide insights into the operation of the pla,.t and to allow future planning and docision making by employing assessment techniques diverse from those used in the original design. The engineering insights gained from the use of PRA have aiready been applied to emergency procecure development, training, containment design and detailed system design.
The study was
- conducted by two independent contractors with the inputs from the Nuclear Steam Supply System supplier (GE), the A/E and other utilities. A peer review groLp was established to assure that the study was done using
" state-of-the-art" PRA methods.
.The' Shoreham PRA employs the same basic approach and techniques utilized
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in the NRC's Reactor Safety Stody (WASH-1400).
The PRA for Shoreham includes three principal phases, namely, (1) The identification and quantification of low frequency accident sequences using event tree and
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fault tree logic models, (2) The evaluation of containment response and radionuclides releases during a postulated accident using improved physical models developed since WASH-1400 and, (3) The assessment of the
consequence of radionuclides released to the environment.
The results of this study are documented in the final Report of the Shoreham Probabi-listic Risk Assessment dated June 24, 1983.
At the time of this inspection, the licensee had established a dedicated section with a section head and two engineers to manage the PRA efforts contracted to several outside contractors. The licensee used PRA
techniques and results to independently verify the results of the Independent Plant Evaluation-(IPE) and to analyze the performance of the licensee's supplemental containment design. Other plant activities affected by PRA included (1) The use of 90% enriched Boron in the Standby Liquid Control System, (2) Raising the Reactor Core Isolation Cooling
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(RCIC) System trip set point from 25 psig to 50 psig for increased l
post-LOCA availability, (3) Modification of Automatic Depressurization
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System Logic in light of a NUREG 0737 Post TMI action Item, (4) Traversing Incore Probe reliability study, (5) Reactor water level design study, and (6) Comparison of risks associated with various emergency diesels. A good understanding of PRA application in design was evident in engineering, l
maintenance and operation personnel.
The recommendations resulting from PRA were implemented with good management and staff support. The open items generated by the PRA studies are tracked to completion by the Nuclear Engineering Department.
From the above, this inspection concluded
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that the licensee uses PRA techniques very effectively to identify and minimize risks associated with various plant systems, structures and components.
C. Engineering Assurance Activities The Engineering Assurance program was developed in 1984 in light of a commitment made to the NRC by the licensee and to perform independent design reviews required by ANSI N45.2.11.
The Engineering Assurance Division is headed by a manager who reports to the Manager NED.
The EA function is primarily conducted by the EA section consisting of a section head and three experienced engineers. The EA division monitors on a routine basis the work performed under the Interim Station Modification Program, work performed to support the plant Technical Specification, work required to support license proceedings, safety evaluation,10 CFR 21 evaluations, equipment qualification, fire protection, security and emergency planning. As stated before, the EA division also plays an important role in the technology transfer program. A Readiness Assessment by EA is mandatory prior to authorizing any division to assume design responsibility from the A/E.
EA performs this independent assessment using pre-established criteria made known to the organization being assessed.
The inspector reviewed the EA assessment for the readiness of the Instrument and Control Section.
This assessment was conducted on February 17-20, 1987. The EA assessment identified a lack of clarifica-tion of interfaces with other organizations and a lack of a permanent section head. There was also a lack of backup engineers for the Seismic Monitoring and Loose Part Monitoring System.
EA restricted the I&C Section from assuming design responsibilities until the above deficiencies were corrected. At the time of this inspection, the licensee selected a permanent section head for the I&C section and the vacancies in the sec-tion were beino actively recruited.
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Similarly during the assessment of the Engineering Mechanics Section, EA identified that hydrogen detonation concerns in the off gas system were not discussed by the A/E.
The inspector reviewed the follow up actions in this regard. On June 1,1987, the assessed division established the basis for the design of the off gas system. As of July 10, 1987, the Engineering Mechanics and the Nuclear Process engineering sections jointly J
assumed the responsibility to resolve this concern.
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The above are examples of effective assessments conducted by EA to identify organizational and design deficiencies and obtain timely actions.
D. Reliability Group The reliability group within the QA department is responsible for ensuring that, (1) The nuclear plant is maintained at an acceptable level of reliability by evaluating the operating history of plant systems and equipment and experience from other operating plants, (2) Reliability goals are established for plant systems and components and availability goals are set for operation, (3) Computer software and procedures for reliability assessments are developed, maintained and used, and (4) Action plans to enhance equipment reliability and plant capacity factor are developed and recommended. The reliability group is headed by a supervisor and has a staff of four engineers and a staff specialist.
Various site organizations submit ideas for study to the reliability group supervisor who prepares project definition and with the approval of QA management assigns the project. The completed report after careful review and approval is distributed with specific action items assigned to individuals and groups.
The inspector randomly selected recently completed projects for Low Pressure Coolar:t Injection System, Reactor Water Clean up System, Control Rod Drive System, and Standby Liquid Control System to assess the effectiveness of the studies undertaken by the reliability group. The reports were well written and the conclusions and recommendations were clearly stated.
The results section of the report discussed the availability, reliability and maintainability data in detail. The analyses are performed using UNIRAM computer software package.
The fault trees and the availability data are generated by the computer using plant spectiic and industry data input by the user.
At the time of this inspection 14 reports were issued.
Twenty nine Engineering Evaluation and Assistance Requests (EEARs) were generated and eighteen were approved.
The reports were well received by the organiza-tions requesting the study as the reports contained beneficial information to the group.
All action items resulting from the study are tracked by the reliability group personnel to closure.
From the above, it is ap-parent that the Reliability Group is beginning to function effectively by interfacing with plant groups and providing timely analyses for identified concerns.
E. Independent Safety Engineering Group (ISEG)
Within the QA organization the licensee has established the ISEG in fulfillment of a commitment made during the licensing process and to comply with a technical specification requirement.
The function of ISEG is to assess and report the effectiveness and quality of nuclear operations and related safety and environmental programs.
The ISEG reports to the QA organization and interfaces with the Nuclear Review Board (NRB) and other licensee departments, as appropriate.
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The ISEG is a permanent organization with a dedicated supervisor and an authorized stafT of seven engineers and.one clerical aide.
Supplemental staff may be loaned from other licensee organizations on an as needed basis.
ISEG members meet the qualification requirements of Technical Specification 6.2.3.
The ISEG routinely examines plant opeda;.ing characteristics and industry and NRC issues.
The operating experience from the station and other similar facilities are assessed to identify areas needing improvement.
The findings and recommendations resulting from the above examinations, reviews and assessments are provided to applicable departments and divisions. This group also evaluatet 'the operating experience feed back program.
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The inspector randomly reviewed ISEG evaluations for HPCI check valve failure, response time of scram instrument volume level detector, Radiation Monitoring System, and improper installation of heat shrinkable tubing. All these evaluations were well planned with clear statements of purpose and work scope. The final report contained details of the methodology, research of the concern and avai hble solutions. Conclusions and observations were stated in readily understandable format. Specific recommendations were also provided in the report when appropriate.
Open items resulting from ISEG reviews.are tracked by the group to closure.
In summary the inspector conchdet that the ISEG is functioning effectively in accordance with its charter and the Technical Specification requirements.
F. Training And Qualification The Nuclear Engineering Deoartment personnel training includes self study
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of licensee procedures, supplemental technical and professional develop-ment training classes, sem3rars and lectures.
The personnel training records are maintained by NED.
The Modification Engineer maintains training records for the Cognizant Site Engineers (CSE).
The. training program for CSEs includes inhouse training seminars on Jolicies and procedures and required formal training in the area of safety evaluations, veld inspection, examiner certifica-tion, QA indoctrination and BWR fam.liarization.
Certain CSEs have also received supplemental training in. 'he areas of ASME Section XI Repair, analog controls, vibration analys;s, BWR design, BWR simulator and advanced supervision.
The Medif' cation Engineer also utilizes a guideline to provide a listing of the minimum training required to enhance the skills of the new CSE and to maintain the skills of the experienced CSEs.
A new CSE is assigned to an experienced CSE for assistance during training and indoctrination. This guide will also walk the new CSE through some station modifications and test his/her knowledge.
The new CSE is required to maintain the Training and Indoctrination card current.
This card i
documents the courses taken, procedure familiarization and writing of a practice station modification, for a previously implemented design output package.
Upon completion of the qualification card, the new CSE and the L--_--__--
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guide sign it and submit to the M % :fication Engineer for review. The qualification is typically completed within the first year of duty and ensures that the new CSE is adequately trained to perform his/her duties.
The inspector reviewed the training records and qualifications for selected NED and Modification engineers and determined that the training and qualification were commensurate with the duties to be performed.
From the above, the inspector concluded that the training and qualification programs received adequate support from the licensee management.
G. QA Audit of ISMP The inspector reviewed the licensee's Quality Assurance (QA) Department Audit No. QC86-11 of the Interim Station Modification Program (ISMP)
conducted from October 27 through November 7, 1986. This audit included a performance oriented evaluation of station modification activities through a review of in process and completed station modifications.
The audit findings included concerns in record retention, drawing updates, design output package requirements for minor modifications, system description and lesson plan updates, drawing approval, 10 CFR 50.59 submittals and tractability of materials. The resolutions of the audit findings that required program revisions were adequately incorporated in the revisions of ISMP as applicable.
The licensee's QA organization is actively involved in identifying problem areas or weaknesses in the modification control area as indicated by the above audit.
The inspector specifically reviewed the audit finding that discussed the scheduled date of completion and issuance of a report that documented 10 CFR 50.59 changes completed between January 1, 1985, and December 21, 1985. The licensee's Nuclear Licensing and Regulatory Affairs Manager committed to complete this report and submit it to the NRC by January 12, 1987. The NRC inspector reviewed the licensee's program to report 50.59 changes. This program assures that a report is submitted annually.
However, it does not assure that the changes made during each calendar
' year #will be reported at the end of that calendar year. At the exit meeting, the licensee committed to submit 10 CFR 50.59 changes made in
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< < :alendar year 1986 by December 31, 1987. The changes made in calendar year 1987 will be reported in the first quarter of calendar year 1988.
Subsequent reports on 50.59 changes made in each calendar year will be reported during the first quarter of the next calendar year.
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3.0 Modification Program The essential elements of the licensee modification program are illustrated in Figure 1.
The program involves both the Nuclear Engineering Department and the plant staff. Once a modification is requested using the Engineering Evaluation and Assistance Request (EEAR),
it is evaluated and a plan is developed.
If this request is approved, the Nuclear Engineering Department develops the Design Output Package (DOP).
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s 1,his process involves preparation, review and approbl~. ij a y also l
'lnvolve use of personnel from'o'.her organizations including use of outside I
contractors. Once the packageiis approved by NED, it is forwarded to the
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plant staff to prepare a Statio.t tiodification Packageo(SMP). The SMP
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provides the details for implementing the modification.
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developed under the cognizance of the Modification Engineer (ME) who
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final closeout of the modification.
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The inspectors reviewed the modifications in tn Me anialiI&C, Nuclear
and Electrical areas as discussed in the following p>ragra'piis.' The selected modifications were reviewed for technical adequacy and to assure.
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tj that implementation;was in accordance with the established procedi.res.
The inspectors contacted the.verfous design engineers and CSEs responsible for the modification to understad their awareness of the x difidation program and to assess the extent of interfaces among work [rouos inycived r
in the modifications.
The inspector reviewed such items as' materials of construction, environmental qualification, seismic qualifications, test
procedures, classification analysis techniques and ALARA as appropriate for the modification as fetailed below.
k" A. Evaluation of 4160 Volt Circuin E,reakers Equipped with Chain Drive Spring-u.
si~draing Mechanisms (DORifCSPJ57).
i 4160 Voll circuit breakers supplied to,the licensee by Seneral Electric Company (GE) per specification SH-139 he1' ratchet and' pawl mechanisms for breaker spring charging.
Currently, GE is not making the ratchet and pawl mechani.w.
Eight spare 4160 Volt GE breakers received recently by the 1icensee use 4 chain drive spring charMoj mechanism insteild: Lf the
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ratche't and pawl mechanisn for spring charging, The licensee reviewed the chain drive design and detemined that ibis' acceptable fr use. The insper; tor reviewed the licensee's engineering evaluation as documented in Rept. t No. SH-1-039, dated February 5,1986., and determined that the ana?ysis and maintenance program addressed the plant spec"!c conditions,
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The aspector also verified that'the licensee's engineering data was v
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consistent with the data providec by GE in various letters and reports.
This replacement did not require a modification package as it was installed as a " Form, Fit and Function" item. The NRC inspector found
" Form, Fit and Function" classMication to be acceptabla for this replacement.
B. Modification to Trip Panel 1011-PNL-N2 (DOP No.35-027 Modification-PackageNo.85-065}
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h Panel ID11-PNL-N2 is a nonsafety related distributf en panel which p'ovides power to various componenn of the radiation momtorino,syste'n. -The
original design war to feed this panel from IR24*NCL-103 vie transformer 1D11-T-N2. To meet the load profile for TDI Diesel 102, panel 1011-PNL-N2
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has to be shed frnrr the eejergency bus during a Loss Of Cool. ant Accident
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(LOCA) in conjunction vPS Loss of Offsite Powec.
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The modification was to feed 1011-T-N2 and 1011-PNL-N2 from the circuit
' breaker and contactor.that supplies the TSC panel 1R24-PNL-113C (IR24*MCC-1133).
This.contactor sheds the load for ten minutes upon receipt of a'LOCA. signal coincident with a loss of offsite power signal.
.The inspector reviewed the test procedure and test data and verified that the circuit breaker for this panel tripped upon receipt of coincident LOCA and loss of.off site power' signal during test. The inspector has no further.questionLin this regard.
C. Automatic Start Logic Defeat of Reactor Buildino Service Water Pumps (DOP 84-230 Modification Package No.86-034)
The original design of the service water system was to have the operation of two out of four service water pumps to mitigate a DBA LOCA.
This modification was.to allow three of the four service water pumps to auto-start with the' fourth pump'in manual standby.
The licensee selected service water pumps IP42*P003C or_IPH1*P003D to have the capability, for the control room operator, to. switch the auto / manual initiation. The
. modification consisted.of the addition of-a two position switch, indicating
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lights:and-wiring.added to the Main Control Board (MCB).
The inspector verified that the installed switch, lights and wiring of the MCB were in conformance with modification package No.86-034. The test results were within the acceptance ranges specified in the test procedure and USAR. ~ The inspector also verified that the Station Modification Review Form Appendix 12.7 of procedure 12.010.02 was completed and signed by engineering,.A discussion with the CSE and MOD Engiraer indicated that they accepted the test results on January 23, 1987, and.that the an:1ysis f
of the data verified that test resuits complied with the D0P and the USAR requirements.
D. Tracking of Modification Impacts On Diesel Generator Load The licensee has established a program to assure that the loads connected to the TDI Emergency Diesel Generator (IR43*G 101,102 and 103) do not exceed the limits specified in the Technical Specifications (TS).
Electrical Engineering Department procedure No. EED-2, Revision 2, February 5, 1987, requires all DOP's, temporary modifications, or calculations which propose a change to the TDI loading scheme be reviewed by the NPEED Supervisor and' the Nuclear Projects Division Manager (NPDM)
prior to implementation. The inspector also verified that the NPEED maintains the bus loading status of the system including proposed changes to the. system.
The review of proposed DOP's by NPEED and NPDM assures that potential overload problems associated with modifications will be identified and corrected prior to installation. The inspector reviewed the bus loading calculations and verified that at the time of this inspection'the TDI buses were within their design rating.
Data sheets reviewed by the inspector are listed in Attachment A.
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'E. Replacement of the-Reactor Recirculation Pump Discharge Valve Limitorque Motor Actuator SMB-3-100 with SB-3-100 TheLpurpose of this change was to replace unqualified Limitorque SMB-3-100 actuator with SB-3-100 actuator that was being environmentally and seismically qualified. The licensee issued Engineering and. Design Coordination Report (E&DCR) No. H-00035A to request an A/E evaluation of
'the-impact of this~ change on the Reactor Recirculation System piping.
The A/E's evaluation indicated that the change has negligible impact.on piping system. The A/E reviewed the design as required by the Interim Site Modification Program. The Final Design Output Document has been approved
~for issue.
The design has been developed by qualified engineers with good knowledge and understanding of the design and regulatory requirements.
The inspector independently reviewed the design bases and found them to be consistent with applicable NRC regulations and licensee commitments.
This design was done under the cognizance of the Power Engineering Department-of the Office of Engineering.
F. Hydrogen Sensor Replacement This design was initiated to replace an ob wlete sensor with a more-reliable sensor.
The original' components were difficult to calibrate and needed to be replaced. The new components will provide ease of calibration and operate within the specifications for the original system.
This change was also designed by the Power Engineering Department of the Office of Engineering and was independently verified by the A/E. The Final Design Output Document was issued on November 21, 1986.
The inspector independently reviewed the design assumptions and the final
. design and had no further questions concerning this subject.
G. High Pressure Coolant Injection (HPCI) System Check Valve Replacement The HPCI Turbine Exhaust Carbon Steel horizontal swing check valves experienced premature wear during start-up testing due to operation below recommended turbine rated speed.
This resulted in low flow in the exhaust line and valve disc slamming during quick turbine starts.
These valves were manufactured by Anchor-Darling Valve Company. This replacement was to resolve the above concern.
The replacement of the HPCI Turbine Exhaust Check Valves (DOP 85-185 and SMP 86-021) performed by Stone and Webster Engineering involved removing the two existing valves and welding in the two slightly longer and much heavier Anchor-Darling Y-Globe Lift check valves, modification of pipe supports, test and instrument connections and insulation alteration to accommodate the larger valves. The licensee also incorporated two vendor recommended modifications to the valve to stiffan the valve disc skirt assembly to absorb the impact during quick turbine starts and to use a new molded type ethylene propylene seat.
In order to implement the HPCI Check Valve modification, piping, pipe supports and electrical conduit modifications were required in the
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immediate area of the valves ( Non safety related DOP 85-270 was issued for this purpose.) The inspector reviewed the Design output Packages (DOPs) for the HPCI check Valve Replacement (85-185), Relocation of Components'to support the check valve replacement (85-270) and the implementing Station Modification Package (86-021) for the HPCI check valve replacement as detailed below.
A Safety Evaluation was properly conducted, documanted and reviewed by the Review of Operations Committee (ROC) in accordance with 10 CFR 50.59 and the technical specification.. The evaluation considered the applicable areas of changes to the SAR, technical specifications and the existence of'
an unreviewed safety question.
Since the new valves weigh over 1000 lbs more than the-old ones, analyses were performed to verify the adequacy of the pipe supports.
The A/E revised base plant calculations using the dead weight of the new valves. The inspector reviewed selected portions of the
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calculations, listed in Attachment A.
These calculations were professionally done and verified with clearly documented assumptions and conclusions. The inspector reviewed the Engineering and Design Coordination Reports (E&DCRs) associated with the above DOPs and
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determined that the changes were generally minor or issued for clarifi-
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cation. The dispositions adequately addressed the identified concerns.
The minor scope.of these changes indicate that the original design was adequately developed.
The design verification process utilized in the safety related DOP to replace the check valves included an independent review in accordance with ANSI N45.2.11 by-the Stone and Vebster Engineering personnel.
Supplemental multidisciplinary reviews were performed by both applicable Stone and Webster and LILCO personnel. The multidisciplinary technical review comments were dispositioned and incorporated into the D0P, as applicable.
The inspector performed a walkdown of the completed modification and verified that the installation was as designed.
The inspector verified that the required control room drawings (11600.02-FM-25A and MFSK 25A)
were physically red lined to show the applicable changes and that procedures SP23.202.01 and SP23.654.011 were revised to reflect the modification prior to returning the system to service.
The post modification testing was reviewed by the inspector. These tests included a hydrostatic test and a HPCI Pump Operability and Flow Rate
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Test.
The inspector verified that the hydrostatic test met the Hydrostatic Test Pressure Requirements of ASME Section III, NC-6220 and that the operability and flow test met the acceptance criteria.
The DOPs and the Station modification package reviewed were well organized
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and complete..The engineering design analyses and verifications by the I
A/E were effectively developed as evidenced by the lack of major E&DCRs to
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DOP 85-185. The safety evaluations were adequately performed with clearly stated bases. Overall the implementation of the Interim Station
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k modification Program was effective and the interfaces ~between the A/E and-the responsible licensee modifications groups were effectively maintained.
H. Installation of' Californium 252 Neutron Source (DOP 86-121)
This DOP was to increase the time interval between neutron source replacement. :This modification involved the replacement of existing neutron ' sources with Californium 252 sources.
Californium 252 has a longer life than the existing source.
The licensee has already replaced the neutron sources once. The Nuclear Core Analysis Section was the responsible NED organization for this modification. The licensee utilized the Form, Fit.and Function modification portion of their administrative procedure.
A nuclear engineer from the Reactor Engineering Section of the Plant Staff served as the _ Cognizant Site Engineer (CSE) for the modifi-cation. Based on a review of the modification package with the responsible licensee personnel in both NED and Reactor Engineering, the inspector determined that the personnel were aware of the reason for the modifica-tion and they adequately assessed the. impact of the change. The licensee witnessed similar activities at another facility, prior to developing this modification. Adequate interfaces with operations and health physics were evident.,.The licensee replaced the source and changed the associated procedures. The inspector also determined _that the licensee considered several options and decided to replace the sources as it was the most beneficial option.
I. Standby Liquii__ Control System Modifications (DOP 86-1679)
The purpose of this modification was to upgrade the Standby Liquid Control System to meet the requirements of 10 CFR50.62. The licensee exceeded the requirements of the regulations.
They used Boron-10 enriched to 85 atom percent in the sodium pentaborate solution. The licensee's probabilistic risk analysis indicated that the risk could be reduced significantly by maintaining Boron-10 enrichment at 85 atom percent.
This would allow additional time for operations personnel to assess the
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initiation of the SLCS, It will also decrease the time for the boron to reduce reactor power when initiated. The modification involved replacing the current contents of the SLC storage tank with enriched B-10 sodium pentaborate solution and reestablishing new limits for the tank inventory.
This change also involved a change to the plant technical specifications.
Amendment No. 6 to the Shoreham license was issued on May 18, 1987, to establish new limits for the tank inventory. This amendment also required additional technical specification surveillance to assure that the contents of the tank had the required B-10 enrichment.
The inspector reviewed the modification package and associated documents with the licensee's cognizant personnel in NED and the CSE. The licensee developed procedures to assure that all of the contents of the tank were removed and the tank properly cleaned prior to introducing the enriched boron into the tank.
The inspector also noted that the control room documents were red lined per the modification procedure.
The inspector
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also discussed the change with control room personnel.
The control room personnel were aware of this change.
The inspector reviewed the licensee's activities to assure compliance with technical specifications for the SLCS. Technical Specification Surveillance-4.1.5.e.1 was added to assure that the B-10 enrichment is always greater than 85%.
It requires analyzing a sample taken from the tank prior to entering Operation Condition 2 and at each refueling outage L
thereafter.
B-10 enrichment determination is important as the higher enrichment allows a reduction in the volume of the sodium pentaborate solution in the tank. The sodium pentaborate is a powder and is mixed with water in the Standby Liquid Control (SLC) tank. The licensee records indicated that the concentration and mass of the contents of the tank were analyzed. However, the B-10 enrichment was checked off as N/A. Further investigation by the inspector indicated that the licensee had not sampled the contents of the SLC tank for B-10 enrichment after the sodium pentaborate had been added to the tank. This was based on a licensee decision to use the results of'the analysis of the samples from the batches of sodium pentaborate powder that was added to the tank instead of an analysis of a sample of the sodium pentaborate solution taken from the tank. The samples of the batches were taken by the supplier anc were included with the shipment. The supplier had also provided the results of his analysis which indicated that the B-10 enrichment was in the range of 90.31-90.63 atom percent. The licensee utilized an independent laboratory to analyze these samples and obtained results ranging from 90.52 to 90.63 atom percent..The licensee performed a vendor surveillance (QSD-87-081)
-in May 1987 to witness blending, packaging, sampling, and testing of the enriched sodium pent 4 borate associated with the procurement of a new order. A sample was also taken by the licensee to do an independent analysis. The vendor's analysis yielded 90.64 atom percent and the licensee's independent analysis yielded 90.53 atom percent.
' tie licensee had high confidence that the contents of the tank met the B-10 enrichment requirements. However, he did not analyze a sample of the sodium pentaborate solution taken from the tank to verify the enrichment of B-10.
The SLCS was returned to service on May 20,1987, and the reactor
.was placed in Operational Condition 2 on May 22, 1987 and was operated in that conditinc since then. Operational Condition 2 requires operability o f.S LCS.
Technical specification 4.0.4 requires that entry into a Operational Condition shall not be made unless the surveillance requirements have been performed.
Not taking a sample from the tank and analyzing it for B-10 enrichment as required by TS surveillance 4.1.5.e.1,
.is a violation (50-322/87-10-01).
At the exit meeting the licensee stated that a sample from the tank would be taken and analyzed off site. Subsequent to this inspection, the licensee informed the Resident Inspector that the required analysis has been performed.
The results of this analysis indicated that the B-10 I
. enrichment for the solution in the SLCS tank was 90.47 atom percent. This result is. consistent with the results obtained from the batches of sodium pentaborate powder and the technical specification requirement.
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The documents. required to close out this modification were not complete.
The licensee tracking system for modifications properly reflected the situation.
Personnel involved in the Modification program are aware of their duties and responsibilities. The modifications were technically adequate and implemented acceptably with the exceptions discussed above.
3.1 Modification Feedback Process The licensee utilizes a feedback process to evaluate the effectiveness of the modification program implementation.
The Cognizant Site Engineer (CSE) completes a Feedback File questionnaire and assigns points depending on the adequacy of the design, procurement, installation, testing and closecut. The CSE also maintains the Feedback File. Any negative trends or problems (with the probable cause if apparent) encountered in meeting the objectives of the program are documented in the Feedback File Memoranda.
In addition to the aforementioned, applicable LILCO Deficiency Reports (LDR), NRC inspection findings, QA audit findings or any other
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documentation relating to modification program deficiencies are maintained in the Feedback Files.
The licensee periodically reviews the Feedback File of the station modifications to identify problem areas to insure necessary corrective activ.1s.
The inspector reviewed the results of the licensee's review of the Feedback Files from October 1, 1985, through October 1, 1986, and noted that the review committee identified three common concerns: (1) the need for a field walkdown during the design phase; (2) the need to clearly define procurement responsibilities; and (3) the existence of numerous warehouse stock and issue delays.
The first two concerns were already addressed in the revision of the ISMP and the third concern was addressed by hiring a full-time specialist to work with the warehouse to eliminate the problems.
The licensee's management is utilizing the Feedback File effectively to monitor and enhance the station modification process in a uniform way.
4.0 Engineering Assessment of NRC Bulletins, Information Notices and Licensee Event Reports The following NRC Information Notices (ins) and Licensee Event Reports (LER's) were reviewed by the inspector to assess the adequacy of the technical resolution for the identified concerns:
A. Information Notice (IN) 86-53: Raychem Heat Shrinkable Tubing Installation On June 26, 1986, NRC issued this IN to identify improper installation of heat shrinkable tubing at four nuclear facilities. The identified concerns at these facilities included improper tube dimensions, improper overlap onto wire insulation, use of tubing over fabric, improper bending of wires and tubing inside junction boxes, and insulation damage.
Additionally, this information Notice informed the licensees that if the installed configurations are not consistent with the configurations used during qualification, the equipment qualification status was to be declared indeterminable.
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- The inspector reviewed licensee actions to address the concerns identified
.in the above information Notice.
On January 13, 1987, the licensee's A/E completed a detailed review of the licensee s use of heat shrinkable d
tubing.against the information notice. The site controls installation, inspection and application of Raychem heat shrink products through the use of twelve Engineering and Design Coordination Reports (E&DCRs).
The Shoreham Criteria were compared to the latest Raychem criteria for application of heat shrinkable tubing.
The Shoreham requirements did not include the following Raychem requirements:
1. Remove non qualified or braided jacket material 2. Remove sharp edges on connectors 3. Minimum bend radius of 5 times 0.D.
Shoreham minimum cable bend radius were 2.5 time 0. D. for instrument cable, 4 times OD for control cable and 2.5 - 5.5 times OD for 600V power cable.
E&DCR L-1453 was issued to endorse Raychem requirements for installation at Shoreham on April 27, 1987.
The review by the A/E identified that the following concerns of Information Notice 86-53 were not fully covered by Shoreham procedures:
1. Tubing over fabric cover on wire 2. Small' bend radius, except visual inspection by QC 3. Absence of inner tubing On March 3, 1987, the licensee identified the equipment installed with heat shrink kits.
From this, a sample inspection program was developed to physically verify the condition of Raychem heat shrinkable tubing used in all safety related cable applications.
MWR 87 -2357 was issued to cover this QC inspection activity. The inspector witnessed the QC inspector completing this inspection program on July 14, 1987.
If no concerns were
identified in the sample, the inspection requirement was to accept the entire population.
However, if one defect is noted, one hundred percent inspection of the population was required. The inspector verified that the sample selection criteria was in accordance with the military standard MIL-STD-105D.
In response to Information Notice 86-53, the licensee arranged special training sessions for personnel involved in the installation and inspection of Raychem heat shrinkable tubing.
Ten persons were trained at Lyndhurst, New Jersey by the Raychem personnel.
Additionally twenty two L
people were trained by Raychem personnel at the site. The training seminar included suggestions for field engineering, general procedures for inspection, accept-reject criteria and corrective actions.
The inspector noted that the engineering and QC personnel involved in the Raychem heat shrinkable tubing application were knowledgeable in the recent Raychem concerns and steps necessary to address the concerns satisfactorily.
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The NRC inspector independently selected sixteen terminations from the total population and visually examined the conditions of the heat shrinkable tubing installations. With the exception of the item discussed l
below the NRC inspector found these installation to have addressed the I
concerns of Information Notice 86-53 adequately.
On July 14, 1987, the NRC inspector identified Raychem end caps on the splices to the Solenoid Operator Valve (SOV) leads for IT46*S0V-39A located in the Drywell.
The Licensee's subsequent inspections identified Raychem end caps on splices to the leads for IT46*S0V-38A also located in the drywell. The use of Raychem end caps in any harsh environment is prohibited by E&DCR F-7320D.
Failure to follow the requirements of this ER&DCR for the Raychem heat shrinkable tube application in the drywell is a violation (50-322/87-10-02).
This is related to an existing NRC open item 50-322/86-12-03 regarding improper installation of heat shrink tubing.
Subsequent, the above identification of end caps in the harsh environment, the licensee inspected all other similar devices in the drywell and identified one additional installation of end caps as oiscussed above. A LILC0 Deficiency Report (LDR-87-189) was written to address this concern and the disposition was to to remove the identified end caps from the drywell and replace them with Okonite tape in accordance with E&DCR F-73200. A separate LDR will also be issued to address all
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other concerns identified during the QC sampling inspection program.
The licensee is also pursuing the differences between the as installed configuration of end caps and the configuration used during qualification.
The resolution of the concerns identified during QC sampling inspections and the resolution of the end cap qualification issue will also be tracked under NRC Open item 86-12-03 and reviewed during future routine NRC inspections.
Except for the violatior, discussed above, the engineering activities at the site related to technical assessments were observed to be conducted effectively and in accordance with the NRC regulations and licensee commitments.
B. Information Notice No. 87-30: Cracking of Surge Ring Brackets in large General Electric Company Electric Motors.
General Electric, as part of their 10 CFR Part 21 program, notified the licensee that during a routine motor inspection in a BWR 4, a fatigue failure was found in the support brackets of the Residual Heat Removal (RHR) and Core Spray (CS) pump motors.
Inspection results indicated bracket failures in 5 of 6 motors inspected. Modified support brackets were installed in 4 motors and tested. The test results showed that significant cyclic loading was caused by starting and not by normal operation.
General Electric (GE) recommended an inspection to establish the baseline condition of the installed motor configurations. A modification kit has
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been developed by GE to replace the existing configuration, if necessary.
GE also suggested an increased inspection frequency for these motors.
The inspector verified that the Geeral Electric letter MC391, March 23, 1987, is being tracked by the licensee.
The licensee will track the action items for this IN in the NOSD tracking system.
The inspector verified that the maintenance organization was familiar with the General Electric letter, fne licensee has been exploring a method to inspect these motors without disassembly.
C. Information Notice No. 83-44 - Potential damage to redundant safety equipment as a result of back flow through the equipment and floor drain system.
The licensee review of this information notice generated an engineering study by the Power Engineering Department. The licensee determined from this study that drain lines for the TDI Emergency Diesei Generators (EDG's) in rooms 101 and 103 did not have adequate backflow protection.
An emergency modification (ED0P-86-176) was issued to install a check valve in the drain line header for these diesel rooms.
On January 16, 1987 the licensee issued LER 87-046 to. notify the NRC of the conu rn and the actions taken to resolve this concern.
The licensee has been tracking the IN in their system, but delayed any system modification work until the engineering study of the plant lay out was completed.
This engineering study was completed December 9, 1986. The inspector reviewed the following sections of this study to determine the completeness of the work and that the study was performed using the existing as built drawings of j
the plant.
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The licensee issued Engineering Evaluation and Assistance Request (EEAR)
No.86-176 which described the work to be performed to comply with the intent of IN No. 83-44. The licensee committed to the NRC on February 19, 1987, to develop a preventive maintenance scheduled activity work sheet (PM-SAWS) to assure operability of Check Valves installed to prevent backflow from drain system.
The required PM-SAWS has been issued with a i
24 month inspection frequency.
The first inspection after the modifica-tion is scheduled for March 15, 1989. An engineering assessment will be
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performed after the first inspection and the inspection cycle will be s
reevaluated.
The inspector has no further questions in this regard.
D. Licensee E,ent Report (LER) 87-003: Loss of offsite power and isolation of
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the Normal and Reserve Station Transformers due to a procedure inadequacy in a modification package.
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The root cause' of this event was an inadequate description of a step in a
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station modification procedure. The Cognizant Site Engineer (CSE) was
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unaware of recent-circuit modifications _when he wrote a procedure for a te s t.' As discussed in LER-87-003, when this test-procedure was implemented the plant experienced a loss of off-site power.
'To correct'this' condition, the licensee' revised'their Station Procedure'SP.
12.010.02, Station Modification Activities, to require a technical review
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and check off sheet to be completed prior to a sign-off and acceptance of the DOP package.- This check off sheet contains' specific sign-offs for various attributes of the Station Modification Implementation (SMI)
package,-including a step which requires verification that an analysis.was performed to ensure that the modification will not adversely affect the-safety or operation of the plant.
In addition, the Station Modification Procedure was revised to_ require that the modification package technical reviewer is independent lof the' originator (s). The licensee committed to the'NRC to' prepare and train their personnel using a new lesson plan. At the time of this inspection the.above lesson plan was being reviewed by the licensee for approval.
The licensee is' sued procedure No. 12.035.02, Revision 5, May 22, 1987, titled " Control of Temporary Modifications". The inspector reviewed this procedure and identified several areas requiring clarification. The-license modified the procedure to clarify specific areas identified by the inspector prior to the completion of this inspection. The revised procedure was'also scheduled for review by the Review of Operations Committee (ROC). This item will be tracked by the NRC under the existing
' violation-(50-322/87-05-01).
'E. LER 86-030: Unplanned Automatic Actuations of ESF Systems caused by Power Spikes on the Grid Voltages Due to Thunderste.w This LER was the' result of two auto actions of ESF systems from power spikes. caused by thunderstorms. The LER indicated that the Engineering organization would be requested to evaluate the events to determine if a common mode failure exists. The inspector reviewed the licensee actions as
.a result of this LER. The licensee performed a root cause analysis of the
. events.
Other licensee actions are also being taken.
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Routine Engineering Support to the Site The NRC inspe:. tors randomly selected routine engineering support provided to the plant and other organizations on-site. The selected activities included engineering dispositions of licensee deficiency reports, engineering support to operations, QA, maintenance and other plant organizations.
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A. Engineering Review of Deficiency Reports The licensee utilizes the LILCD Deficiency Report (LDR) to document a nonconformance and to assure proper disposition, closecut verification and
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timely notification, of the affected organizations.
The Nuclear Engineering Department (NED) either provides dispositions with documented
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technical justification or approves technical justification documented by other organizations.
The technical justifications provide the assurance that deviations will not adversely affect the safety functions, repaired items will meet prescribed requirements and that repair procedures are delineated or referenced.
The inspectors reviewed the LDRs, listed in Attachment A, for technical
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adequacy of the disposition, timeliness and overall quality. The dispositions in all cases reviewed, adequately addressed the identified concerns and were timely.
The inspector discussed the engineering
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t dispositions of LDRs with the QA and Maintenance organizations and received positive feedback that the engineering resolutions were adequate and timely.
The licensee's QA organization publishes a monthly report in which the status of open LDRs is tracked.
The " Quality Assurance, Safety and Compliance Monthly Report" for June 1987 shows that the Nuclear Engineering Department (NED) is responsible for 36 open LDRs with the Power Engineering Department responsible for three and the Electrical Engineering Department responsible for one. Of the 36 open LDRs assigned to NED, eight remain undispositioned.
However, extensions have been granted as there was no safety significance involved.
The amount of engineering support afforded to the LDR process war adequate as evidenced by the absence of any backlog of undispositioned LDRs.
B. Engineering Support to QA
The Nuclear Engineering Department (NED) routinely interfaces with the QA
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organization through LDR dispositioning, receipt inspection hold tag
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dispositioning, and the Management Awareness Program (MAP).
The NED currently is responsible for dispositioning six hold tags and has recently become involved in a MAP report to evaluate the condensate /feedwater system for the Surry Nuclear Power Plant erosion /corrcsion problem. The QA organization has also requested NED expertise to help evaluate liquid penetrant examinations of the 1R43*EDG-101 (performed 5/12/86, 9/2/86,
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12/30/86 and 4/25/87) and 1R43*EDG-102 (performed 7/15/86, 2/26/87 and 5/2/87) diesel generators.
The inspector, through a review of the engineering support in dispositioning diesel generator cam saddle indications and discussions with QA personnel, determined that the routine engineering support to QA was adequate and timely.
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, C. Engineering Support To' Maintenance L
.The Nuclear Engineering Department (NED) routinely interfaces with the
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maintenance organization through Engineering Change Request disposition-l.
ing, Engineering Evaluation and Assistance Request-(EEAR) dispositioning and memos. Potential modifications (EEARs), Maintenance Work Requests (MWRs) and safety significant problems are addressed on a priority basis.
.The NED organization typically gets involved with maintenance.related problems through the plan of the day meeting and interfaces with the systems engineers. These items are then tracked on the Plan of the Day tracking system.
.The inspector reviewed-the memos, listed in Attachment A, to assess the
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type of.information exchanged, its adequacy and timeliness.
Based on the I
reviewed sample and discussions with maintenance personnel the inspector determined that the engineering orga.,ization is providing adequate and timely support to the maintenance deparw?ent.
D. Engineering Support to Operations The inspector interviewed licensee representatives from-the Operating Shift, NED System Engineering, NED Fuels, Plant Staff Reactor. Engineering, and Plant Staff System Engineering to assess the adequacy of routine support received from engineering.
. Based on discussions with operating shift personnel the inspector-determined tilat the engineering organization interfaces primarily with the system engineers of the' plant staff. The operating shift is knowledgeable of the NED organization but they rely upon the system engineers to obtain engineering resolution for plant problems.
They did acknowledge that the NED organization assisted the plant staff system engineers in solving plant problems.
From a' review of selected examples, the inspector determined that the system engineers interface well with the on shift personnel. Operations personnel provide input for modifications and serve on the Modification Review Committee.
They also respond to NED requests for assistance especially in the area of installing test equipment to monitor plant, parameters. Monitoring of the service water flows during extended diesel generator operation is an example of such interface. A systems engineer is assigned a system or systems to follow on a day to day basis. Backup CSEs are'also assigned for contingencies. A system engineer is also assigned at.the NED organization for each system.
Key NED personnel are available on call to provide round the clock coverage at the plant. A few of the system engineer positions are currently being filled by contractor personnel. The licensee is actively pursuing filling these positions with permanent employees.
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The request for engineering assistanca can be made through daily plant meetings, memos, EEAR and verbal request. The inspector witnessed one such daily meeting on July 15, 1987, and observed the request for assistance being made.
The licensee also follows problems identified at other plants through several systems and takes action as deemed appropriate. Monitoring of drywell temperature conditions at BWR facilities during the initial plant startup program and the monitoring of feedwate* line for potential thermal stratification are examples of such licensee followup. The NED responds to special requests from the operations staff. An example was the analysis to justify use of the reactor building standby ventilation system to maintain the reactor i
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building negative pressure if the normel building ventilation is not
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available for short periods of time.
The shift technical advisors (STAS) serve as the reactor engineering presence on a daily basis. They have been trained as station nuclear engineers and may perform certain simple tasks.
They are required to consult with the Reactor Engineer or the Nuclear Engineer prior to undertaking complicated tasks. The Reactor Engineer also interfaces with the NED Fuels Division personnel, especially the Nuclear Core Analysis Section. The Reactor Engineer f requently requests estimated critical position for each reactor startup from NED.
When the plant is operating, Fuel Division personnel monitor the core performance and establish the control rod patterns when required.
The licensee is utilizing the 3D code SIMULATE to perform most of their core calculations.
They have been routinely requested to predict the estimated critical positions when the plant is being started up.
The NED Fuets Division is developing its own capability to do portions of the reload analysis. The licensee has completed many of their benchmark calculations and is in the process of preparing a report to submit to the NRC to support their plans.
The licensee can also use the code as a backup to the plant computer to calculate the thermal limits. This is in addition to the BUCLE capability that the licensee is able to use. The licensee is also training several of the plant staff reactor engineers in the use of the SIMULATE code in the event personnel from NED are unavailable. The level of experience of the personnel in the Core Analysis Section is quite high. Four of the six positions available are currently filled with persons with advanced degrees, t'-ae with PhD's with over 47 years of combined nuclear experience.
The so remaining positions are soon to be filled.
Based on the above, it is concluded that the engineering organization is involved with the day to day operation of the power station.
The individuals interviewed are aware of the interfaces with other departments. The working relationships appear to provide outputs that are acceptable to the receiving party.
Communication between work groups appears to be effective. Many of the actions taken by the work groups are based on requests received in day to day communication.
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6.0 Facility Tours
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The inspector toured the Control Room and the facility to assess plant and
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l equipment conditions including cleanliness, apparent physical condition,
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fire hazards and work in progress. The inspector noted that the areas appeared to be in very good condition with no discrepancies observed.
7.0 Exit Interview At the conclusion of the site inspection, on July 17, 1987, an exit interview was conducted with the licensee's senior site representatives (listed in Attachment B). The findings were identified and inspection items were discussed.
At no time during this inspection was written material provided to the licensee by the inspector.
Based on the NRC Region I review of this report and discussions held with licensee representatives during this inspection, it was determined that this report does not contain information subject to 10 CFR 2.790 restrictions.
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ATTACHMENT A
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L OTHER DOCUMENTS REVIEWED TA.
'LILCo Deficiency Report 86-361.
Indication in Valve. Intervals
'86-367 Indication in Valve Disk Assembly-8G-374'
Structural Steel Gouged 87-019 Incorrect Weld Rod Used - Training 87-093 Incorrect Wald Rod Used - Retraining 86-408 Turbine Lube Oil Check Valve Not Installed 86-408 Dynamic Qualification of Limitoque SMC-04 Opreators86-417 SKV Circuit Breaker Damaged 86-015 TDI -Instruction Manual Incorrect Identification Of Filter 87-082 Turbine Building Sprinkler System Valve Not Shown On Print '
87-10 Linear Indications Discovered in 2"'90 Elbow B.
Memos ME 87-096 Maintenance... Requests Extension of LDRs - Dedication of disk for TCVs.
ME'87-060 Maintenance recommending possible disposition of LDR 87-055 and-87-072 concerning SW Snubbers.
MEi86-273 NSD 86-2871 IE Information Notice 86-61 NSD 86-2907'. Engineering response to Maintenance regarding T46*A0V035B Failure to pass stroke time.
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NSD 86-2970 IE Information Notice 86-82 Failure of Scram Discharge' Volume Cent and Drain Valves Information.
NSD 86-3174' Disassembly / Inspection of RCIC Turbine Exhaust Vacuum Break Check Valves.
NED 86-3179 Drywell Cooling System Performance - Followup Action Items
'C.
Calculations Reviewed.
.11.600.02-NP(S)-1E41-PSR-004-3-85185-1 Pipe Support 1E41-PSR-004 11.600.02-NS(B)-192 Analysis of Class 2 Sleeved Penetration x-13 11.600.02-AX-118-2-85185-1 Pipe Stress Analysis Sumcary 85-036-EED-2 TDI Loading Calculations D.
Documents Reviewed SP 12.004.02 Safety Evaluation Revision 3.
SP 12. 010.02 Station Modification Activities Revision 9.
~ PD-NE-01 Nuclear Organization Interim Management Control Program for Station Modification, Revision 4.
NE-01-02 Preparation, Review and Approval of Design Input Packages, Design Output Packages and Station Modification Packages, Revision 4 NE-01-04 Station Modification Document Control, Revision 4 i
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Attachment A
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NE-01-03 Preparation, Review and Approval of E&DCRs and ECRs, Revision 4 NE-01-05 Modification Review Committee Charter, Revision 3 QAP 15.1 Control of Nonconformances, Revision 6 MEGADS 2.7 CSE Training Revision 0 Quality Assurance, Safety and Compliance Monthly Report June 1987 MDE-86-037 Memo on 1986 Station Modifications Progress Year to Date/Look Ahead QA Audit QC 86-11 Interim Station Modification Program EED-2 TDI Generator Load T racking EED Report dated April 7, 1987, Plar.t Design Modification as they affect Emergency Diesel Generator Loads EED Report dated May 11, 1987, Plant Design Modification as they affect Emergency Oiesel Generator Loads EED Report June 5, 1987, Plant Design Modificat!on as they affect Emergency Diesel Generator Loads W.0. 10-45200-000 Engineering Report on Flooding of Safety Related Equipment by Backflow through the Equipment and Floer Drain System in No.
83-44 i
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ATTACHMENT B PERSONS CONTACTED 1.
Long Island Lighting Company A. Arena, Nuclear Systems Engineer M. Branco, Core Analysis Section Head
- L. Britt, Manager Licensing and Regulatory Affairs
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- J. Carey, Licensing Engineer H. Chau, Engineering Assurance Division Manager M. Giannattasio, Electrical /I&C Section Head
- R. Grunseich, Operational Compliance Engineer
- R. Kascak, Nuclear Systems Engineering Division Manager J. Keneally, Nuclear Systems Engineering Section Head
- J. Leonard, Vice President Nuclear Operations C. Losnedahl, Nuclear Engineer
- E. Montgomery, Nuclear Project Engineering Division Manager
- J. Notaro, Manager Quality Assurance
- J. Novarro, Manager Power _ Engineering Department T. Pappas, Electrical Systems and Controls Section Sr. Engineer
- M. Potkin, Modification Engineer
- R. Purcell, Outage /Modifiaction Manager J. Rigert, Nuclear Analysis Division Manager
- C. Seaman, Manager Quality Control
- S. Skorupski, Assistant Vice President Nuclear Operations
- D. Smith, Compliance Engineer
- W. Steiger, Plant Manager J. Swanson, Systems Engineer T. Travis, Reliability Engineer W. Tunney, Nuclear Fuel Division Manager
- E. Youngling, Manager Nuclear Engineering Division 2.
US NUCLEAR REGULATORY COMMISSION
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- F. Crescenzo, Resident Inspector
- C, Warren, Senior Resident Inspector
- Denotes those present at the exit meeting held at the conclusion of the inspection on July 17, 1987.
The inspectors also contacted other licensee personnel in the technical, operational and quality assurance departments.
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