IR 05000302/1990004
| ML20012C327 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/02/1990 |
| From: | Jape F, Casey Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20012C323 | List: |
| References | |
| 50-302-90-04, 50-302-90-4, NUDOCS 9003210037 | |
| Download: ML20012C327 (18) | |
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UNIT E3 STAT ES
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o NUCLEAR RE!ULATORY COMMisslON
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REGION il
- 4,i.
101 MARIETTA STREET.N.W.
ATL. ANT A, GEORGI A 30323 k.....,/
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Report No.:
50-302/90-04 Licensee:
Florida Power Corporation 3201 34th Street, South
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St. Petersburg, FL 33733 Docket No.: 50-302 License No.:. DPR-72
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Facility Name: Crystal River 3
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Inspection Conducted:
February 5-9, 1990
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Inspector:
.3 + f6 C. Smith, Team Leader
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Date Signed Team Members:
K. Poertner
. Wright Approved by:
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(144 OeM#1 Y2 MO F. Jape, Sec~ tion Chief
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Date Signed Quality Performance Section Operations Branch Division of Reactor Safety l
SUMr.ARY
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Scope:
This routine, announced inspectior. was condected in de areas of des 191 design change, and plant modifications an et,ee3seent of the licensee s activities in the functional areas of DBD, and cerfiguration llanagement Control was also pwformed.
Results:
In the areas inspected, violations or deviations were not identified.
An' enhanced DBD Program is presently being implemented with on-site system walkdowns and reviews of surveillance and operating procedures to verify the quality of the DBDs.
A CMIS program has been developed to provide current and
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pertinent design basis information including cross references.
Management's involvement in assuring quality is demonstrated by their support to Nuclear Engineering by providing adequate staffing level, training of engineering personnel, and budgeting of design engineering activities.
Review of MAR packages reveals that Nuclear Engineering personnel are intimately involved in ensuring the quality and technical adequacy of plant modifications.
Design reviews performed by Nuclear Engineering personnel of MAR packages which are prepared by outside engineering organizations are thorough and technically adequate.
9003210037 900306 4.
PDR ADOCK 05000302
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... itechnically correct;:and were1 controlled in' accordance with the requirements'~
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i REPORT DETAILS
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. Persons Contacted
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' Licensee Employees j
'*K. Baker, Manager, Nuclear Configuration Management
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- T. Catchpole, Senior Nuclesr Engineering Assurance Specialist
- C. Doyel, Manager, Nuclear Mechanical / Structural Engineering
- A. Friend, Nuclear Principal Licensing Engineer
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- E. ' Froats, Supervisor, Nuclear Licensing
- J. Maseda, Supervisor, Nuclear Engineering
- W. Nisula, Manager, Nuclear Engineering Projects
- R. Pinney, Manager, Nuclear Engineering Assurance
- D. Shook,- Manager, Nuclear Electrical /I&C Engineering
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- N. Sulouff, Senior Nuclear Engineer i
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- P. Tanguay. Director, Nuclear Operations Engineering and Projects
'*R. Widell, Director, Nuclear Operations Site' Support i
Other licensee-employees contacted during this inspection included engineers, and administrative personnel.
- Attended exit interview 2.
-CR-3 SALP Report Period September 1, 1987-December 31, 1988 J
The-inspector examined the corrective actions the licentee has
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implemeated 4 4mpeow the fullowMg Engir,eering/ Technical Support SALP
ceficient'anas identifiw6 by the subject report:
- Priorit'.2ation of Wor a! ceded Inprovement-
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Discussions with the MNDE and revia.w of procedures NEP-201, Preparation i
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and Frncessing of REIs,. Sh-and Engineerir.g Studies; AF410, i
"Freparation and uandling of Field Problem Report, disclosed the
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currently used oriuriti:atton technique being ~used by -the CRT and MBPC
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(multi-departmental committee's) that decide what modification' will be
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implemented and when.
Frloritization of proposed modifications =is
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accomplished by a computerized six year ILS based on criteria ratings
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determined by the MBPC _ This ILS is based on a long range planning
function that developes' schedules for three consecutive fuel cycles
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(six years on a rolling basis) and introduces the concept of a
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"mid-cycle" maintenance outage to be' conducted midway between reactor refuelings. Work scope included in the ILS includes major capitol and s,
0&K modifications, functional maintenance (corrective, preventive,
inspection, and tests) and interfacing program (FPC, NRC, INPO)
commitments.
The ILS is updated as changes occur and issued after management's semiannual review.
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SNES has raised the threshold on REIs, consequently more FPRs and
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L Engineering Questions are being resolved at the site level rather than
being upgraded to REIs which had inundated NOE in the past.
Review of r
the SNES monthly reports for December 1989 and January 1990 disclosed no
new REIs were issued and although site engineering is processing more FPRs and Engineering Questions, the backlog of each has remained
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relatively stable.
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Examination of the January 15, 1990, " List of open FPRs" sorted by
priority revealed all but four 1987. FPRs are of 1988 vintage or younger. This sorting and a separate listing of the " top 20 FPRs"
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reported, get distributed to responsible site management every two weeks.
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j ItLappears the licensee has developed and is implementing and adequate
work prioritization program that covers both urgent and long term -
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projects.
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- Inadequate consideration of EQ i
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FPC committed to providing an EQ refresher training course by the end
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of 1989 to all nuclear engineering personnel who act in the _'apacity of j
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L Design Engineer.
In addition, the overall EQ training was to be
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evaluated to determine future needs.
The ir.soector verified that the
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above mentioned formal classroom training was conducted during the
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period August 23 to October 2, 1989, and the inspector examined the Nuclear Awareness Training Lesson Plan Modules 1 thrL 3 that were i
presented.
The scope of Module-1 included a review of the basic
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elements of the CR-3 EQ progra covering applicable regulations, i
ic program implementation and resrensiblitie).
Mudule 2 pruvides the
engineering staff with the Krawled,e cf how to integrate EQ program i
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considerstions into their day to day tisks such as M% reviews,
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evaluatior - of - specifications for ecuipment, and discusses the up-keep of FQ documentation. Module 3 diccusses the casic limitations
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'2 a plicable to the installation and maintenance of EQ equipment, j
_ The inspector examined the work done to enhance NOTD Lesson Plan No.
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ST-03, "The How G Why of Plant Procedures". EQ Training Modrie which is
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included as part of the NOTD special training program.
.The above in-depth, formal training provided in-house by FPC should
- provide the design engineer with the expertise necessary to properly
. handle EQ related matters.
- Inadequate Evaluation of Incoming VTI
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i To enhance and tighten controls in this area the licensee revised, reissued, and provided formalized training to all SNES personnel personnel on procedures NOD-06, " Technical Information Program";
AI-404, " Review of Technical Information"; and NEP-221, " Engineering Software and Vendor Technical Information Review.
This training was performed on May 2, 4, and 5, 1989.
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FPC has established Technical Information Submital Agreements with CR-3
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prime component vendors. They are B&W,-NSSS; Colt' Industries,
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diesel generators; and-Asea Brown Beveri, the prime switchgear vendor.
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Letters have also been issued to all vendors on the ANSL to establish the MNEA as the single point of contact within FPC for VTI.
These letters also requested the vendors to identify a responsible person
.within their organization for a point of ';ontact and to provide a list of VT1 they have previously issued so that FPC can determin? if this information should be requested for evaluation.
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Enclosure 6 to Al-404A, (Revision 0 June 27.1989) depicts the
Engineering / Responsible Department'Flowpath for all VTI reviews.
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intermediate steps involve Nuclear Compliance Department action which
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tracts the status of the document review in the NOTES computerizec'
program.
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The above licensee efforts should enhance the receiving, handling, and
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processing'of VII to ensure c.onsistent, thorough reviews.
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' Inadequate Resolution of Issues from a Safety Standpoint All permanant and tempuary modifications, both major and minor, require
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safety evaluations.
FPC utilizes the guidant:e contained in Exhibit 2 of
NEP-211, Revision 3, " Modification Approval Records" to clari'y the type of considerations that should be used vnsn reviewing & proposed modification for an unreviewed safety question.
This guidance currently-in use was first implemented on Decembtr 16, 1986. Design engineers are
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reciuired to review NEP-211, and discuss any questions they may nave.
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concerning this 14% procedure (including refety evah ttions) with their
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supervisor. Any enoineer completing this self-study training and approved by his supervisor is con;idered qualified to make MAR 10 CFR 53.59
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safe ty - reviews.-
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FPC-has bew. actively involved in 10 CFR 50.69 guideline workshops, s d as _ '. hat hele on June 22-23, 1989, in Baltimore Md.
The NRC staff
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presentation at the subject workshop concluded that the NSAC-125 guidelines
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are in' general aggreement with the staffs views and recommended that utilities adopt NSAC-125 guidelines into their procedures, i
On June 1,1989, FPC instituted NEP-210 Minor Modifications, which incorporated a " screening" process consistent with that described in
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_NSAC-125, Guidelines for 10 CFR 50.59 Safety Evaluations.
The
screening process employed for minor modifications is detailed such that
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a 10 CFR 50.59 safety evaluation is in effect, being performed. The process is restrictive in that the change shall not be one that adversely affects a system, structure, or comporent either explicitly
or implicitly as described in the FSAR or impact the TSs.
FPC has steted that they plan to utilize other NSAC-125 guidelines (IOC dated December 20, 1989, N0EP 89-0854) at CR-3.
One of the 1990 goals for Nuclear Licensing is to " Develops and assure implementation of
' guidance for performing 10 CFR 50.59 evaluations based on NSAC-125".
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Corporate. Design Engineering e
FPCs Nuclear Engineering Department has basically two design engineering
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groups. _ NDE located _in St. Pertersburg, corporate offices and the SNES
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located a the CR'-3 site.
This inspection dealt with NOE.
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' Organization NOE is divided into three primary sections, Nuclear Engineering Projects, o
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h a-Mechanical / Structural discipline engineering section, and an Electrical /I&C discipline engineering section.
Nuclear Engineering
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Projects is the single point of accountabitity for the management of o
p" external engineering contracts including cost control and scheduling of -
work assigned to contracted organizations.
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Nuclear Engineering Projects is responsible for assuring that effective
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design control of cMtracted organizations is being maintained through the coordination of audits and resolution ~of non-conformances.
The two
discipline engineering sections within N0E are responsible for providing
engineering and design services to CR-3.
In addition, these sections provide engineering services and manpower as required to support nuclear
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generation related studies, research and development projects, licensing
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and other activities as assigned by the MNOE.
Dependent upo:' the priority of the problem, concern,or nuestion, the MBPC
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decides unich REls, Ses, or engineering studies will be undertaken unless
funds tn sceomplish the work tre contained and controlled within the-
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- Section Managars (Nv
- lesr Engi:*eering) budgst.
The MBPC is comprired of l
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appr W mately seven CR-3 Corporate Offue Managers of Engineering,
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Licensing and Nucler Operations who provide initial _ recommendations for
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pric.'f ty ano spproval of project requests. Final selection of these MBPC proritized projects is performec' at the VP: level.
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The preuarer of r. project and the MSPC give the project a priority rating j
e f ' O to 10 for each of seven ce.tagories _ with 10' being the highest
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priority.
Per NEP-201 the catergeries rated include; public relations
_ (INPO), regulatory issues, nuclear safety, personnel safety, ' personnel
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performance,. reliability, and support.
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Non-safety related MARS are handled in exactly the same way as safety
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related or important to safety projects except they do not get a l
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verification engineer review and signature.
Instead they receive a r
" design check" for accuracy performed by a second engineer, different from the design engineer.
This design check is not meant to meet the requirements of the " design verification" as described in ANSI N45.2.11-1974.
Discussions with the MNDE disclosed that " design veri fi cati on'_'
must be performed by a degreed engineer.
Design verification for safety related MARS is performed in accordance with ANSI N45.2.12-1974 requirements per NEP-261.
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The NCE organizations engineering' output is examined by two corporate QA
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groups.
NEA personnel review design packages and other design documents u
primarily for conformance to requirements. imposed by the QA program, NRC
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regulations, and industry codes - and standards.
This review is not a
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technical' evaluation of a design change. QA' audit personnel as a minimum r
perform an annual audit of NOES organization, activities, systems design,
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and maintenance or modification functions.
The inspector held a discussion with the lead auditor. and examined the audit checklists used for audit report 89-11, Raw Water System - Decay Heat Closed Cycle Cooling.. This audit went beyond a programatic. review in that a-- good technical review
was performed on the modification calculations involved. The audit group
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-was' composed of: one nuclear engineer, two mechanical engineers and four
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non-dejreed engineering personnel.
Discussions with the QA audit group L
supervisor revealed 'that if a particular expertise was needed, they may
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use contract personnel to complement the FPC auditing process.
For example,' Audit 89-12 which evaluated various NOE/SNES design activities L2 utilized a NVS Corp. technical advisor and two FPC NGRC laision personnel-to perform,the audit.
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' Staffing h
A survey of 19 NOE.personne1 revealed and average of 15.1 years engineering-
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_ experience of which 12.6 years was nuciear related. Discussions with the W40E revealed that only.cne of 35 FPC engineering personnel was non-depreed
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. Two :ontractor persunnel were also non-degceed. Sempt Position Question-
naires are written to establish the knowledge and skill requirements for i
these positions and the principal activities and responsib1Ms of tha job are ;specifically ident',fied in the EPQ and are centrclied by their supervisors.
ine turnover rate' within the NCE organization 56, baen minimal, lhe only position currently' vacant is fer a Nuclear Electricai Engineer' Supervisor; hovlever, this was a newly created position. Currently only. the Senior VP can approve the-hiring of an enqircer for a new NCE position or approve the hiring of contract help.
The MNOE can hire
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engineers to fill vacancies that occur within his department.
Since FPC has only one nuclear plant des'1gn engineer, contractor personnel
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are frequently used especially during times of heavy work activity.
H When contractor personnel are used NOE relies largely on A/E qualified project teams that were previously used by the utility who have already been through site indoctrination and trained to FPC procedures.
C When L contractor services are procured from a new A/E, the specifications
specify areas-of experience required and personnel supplied by the A/E
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are screened by interviews conducted by the MNOE ~and his supervisors.
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' Once hired, they are given FPC. procedure training and indoctrination.
Auditing of external organizations performing design or design related activities is performed by the Quality Programs Department using ANSI N45.2.11 requirements.
The management of design activities assigned to external engineering organizations is discussed in NEP-103, h
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Communications
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The NSSS and A/E, as well as other external engineering organizations are provided with copies of Design Basis Documents, the Safety Listing,
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Equipment Lists, aperture cards of all design drawings, the FSAR, and
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NEP-104, "Interf ace Design Control" describes the general flow of t
inf ormation to' and f rom the NSSS, A/E, and other external interf aces..
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To a' certain degree' most significant major modifications continue to be f
developed by the NSSS or by the A/E, FPC utilizes a " project management" L
concept with these organizations which requires the NSSS and A/E to-have
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b a' Project Management Manual. This manual serves as the interface document l
or. bridge between FPCs Design Engineering Procedures / Commitments and'
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those_of the external engineering organizations.
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NEP-103 requires the use of Work Authorizations to initiate specific tasks.
The Work Authorization identifies inputs, limitations,
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requirements, and specifies the required deliverables.
l With these clear interface instructions a modificaticn package, along with drawing changes, can be initiated by the NSSS or A/E using FPC format.
Analyses and calculatiens are developed by the NSSS and A/E uring their
i own formats with these documents which are reviewed by an FPC Pru.iect
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Engineer, ther entered into FFCs Fystsm as un acceptable design document,
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Coordination between the two FPC design groups NOE/SNES is, a:complished
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by. monthly cec?dinhtion meetings.
At these meetings simila discipline Managers and Suporthors discuss the scope of work each group is active'y
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working on.
These. meetings help prevent potential conflicts between MARS.
being plenned or implemented by the two groups.
NOE personnel mako-frequent visits to CR-3 to verify design conditions and may use SNES to
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do. Some work for them when travel is not justified. Managament receives s
. monthly.and quarterly MAF. status reports for tracking purposes.
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' Management Support CR-3 minor capital equipment purchases (<550,000) require MN0E approval level whereas major capital equipment purchases (>$50,000) requires
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various increasing levels of corporate management approval depending on the amount. The inspector concluded from' examination of actual SNES/NPSE O&M and other budget expenditures for CR-3 year 1989 and from discussions concerning CR-3s 1990 Capital Equipment Budget with responsible _
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management, that the utility is committed to supporting engineering needs
and enhancing the safety of the plant.
Management's policy is to encourage it's technical staff to become active f
in professional societies and' nuclear industry groups.
A FPC survey of 51 GOC technical personnel revealed 24 had memberships in professional organizations or : societies, two were active participates on society committees, and 31 were were active participates of industrial associations.
FPC industrial association participation includes EPRI,
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NUMARC, SQUG, SEE, EEI, and BWOG activities.
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Management support of engineering -is also evident in that the licensee pays c;ues for membership in professional societies such as ASME, ASCE, etc. for active employees and provides expenses to attend local and national meeti ng s.-
The licensee also pays Professional Engineering.
Licensing fees for its employees licensed in the State of Florida.
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' Reporting Procedures
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NEP-141, " Corrective ' Action" provides direction for the establishment of I
a system to identify -design or design related deficiencies or problems;
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to determine.the cause of. such deficiencies or problems; to identify and-
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p implement appropriate corrective / preventive. actions' to resolve the
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F deficiencies or problems; and a mechanism to document such actions. The L
MNEA gis responsible for evaluating and processing internally identified
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L potential. design deficiencies, for tracking / trending design deficiencies L
anc related corrective action.
The inspectors examined the Quality Programs Department, "Nonconformance
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Trend Report" for the.3rd Quarter of 1989 which contained the MNEA design
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deficiency trending input data.
This report analyses those
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non-conformances corsidered by the NN durine the %LP appraisals.
The
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J 198?. Engineering / Technical Support t.re a appears to now a slight
' improvenient of 18 total weighted deficiency pints accurec versus 21
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- paints f o* the first thrie quartees of the hst J7/83 NRC CALP period, y
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Configuration.Managemer.t t-The inspector reviewed the scope and status of the licensee's
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Configuration Management Project.
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The licenste's Configuration Management Pmject r onsists of the following seven programs:
Analysis Basis Document Field Validation Program l
Enhanced Design Basis Document Program EDBD Field Validation Program
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Safety List. System (SLS) Program Configuration Management Information System Program
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Configuration Control Process Evaluation / Enhancement Program y
Configuration Management Implementation Program
The Analysis Basis Documents address the analyzed transient events in Chapter 14 of the FSAR and two additional events, Small Break LOCA and Loss of Feedwater. The purpose of the program was to validate tnat CR-3 actually conforms to the FSAR Chapter 14 Assumptions, u=
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This effort has been completed and the' ABD's have gone through a-field
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validation effort to ensure that key items and assumptions in the ABD's
.are actually reflected in the as built plant hardware ~and software, i
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The licensee had previously initiated and completed a Design - Basis
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Document Program. The initial program was initiated in October 1983 and p
was completed in December 1987.
Due to weaknesses identified in the
current DBD's' the licensee ' has implemented an Enhanced Design Basis'
Document Program and an EDBD Field Validation Program. The licensee has presently issued four EDBD's and five more are scheduled to be; issued by March 1990.
Cive additional systems are scheduled to be completed by
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December 31,'1990, and the remaining systerns are scheduled to be-completed.
i by November 1992.
The licensae's program encompasses 37 systems. The priority for completion is determined if the system is required for safe
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. shutdown.
If = so, it will be completed first, then the remaining safety-related systems will be accomplished.
The licensee hrs presently p
. field validated one'EDBD. The licensee plans to i nue the EDBU's as they are completed then field validate the document and resise the EDBD as-
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requi*ed.
.l The EDBU's are prepared by one of three A/E's with FPC oversight and review.
The vdidation process is conducted onsite and includes a.
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walkdown of the system and review of surveillance 6nd operating procedures.
t The licensee's proprem is presently on schedule and spoears comprehensive.
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Yhe Safety. List bystem Program developed an Automated Saf ety List System, The Automated Safety Lis; is-intended to be the cornerstone for development of - the Configeration Management Information. System.
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l" Safety List is the FPC tern for "Q" list and the Automated Safety List System has been developed and implemented throughout FPC.
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The Configuration -Management. Information System Program will develop an information system to provide for the controlled' maintenance and use of-configuration data. This system is a computer based system tied into the FPC main-frame computer. The system consists of the Safety List System,_
' Hanger List System, Document Reference System - Instrument List System, Value List' System and Equipment List System.
The CMIS has been.irrplemented and the licensee is presently in_a data entry mode. The end result.of this program will be a computer.ized information system that will easily provide-
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current and pertinent design basis information, including cross reference information.
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5..
Engineering Calculations
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The SALP inspection report number 50-302/88-35 documented programmatic inadequacies related to the control of engineering calculations.
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Deficiencies related to engineering calculations included failure of the calculations to consistently identify assumptions and design inputs.
Additionally, the calculations were not consistently filed, neither were they readily retrievable.
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The inspectors reviewed objective evidence and interviewed licensee management-to verify the technical adequacy of corrective actions
implemented to resolve the above deficiencies.
Nuclear Engineering Procedure, NEP-213, Analysis / Calculation, Revision 1,
was approved
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September 30, 1988, to correct the deficiencies documented in the SALP report.
The-procedure specifies requirements for administrative control of' design analysis and calculations prepared f or CR-3.
lhe requirements
are applicable to design analysis or calculations prepared by Nuclear Engineering, and includes provision for FPC review of analysis / calculations,
which contains design basis information, prepared by external
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u l-engineering / vendor organizations.
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. administrative controls specified in NEP-213, Revision 3, along with
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The
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- those -available via the ' computer based " Records Management Seek System",
are designed to provide ready ' access to and retrieval of design. analyses i
p and calculations.
These controls allow for retrieval by design discipline,
system, tag number, A/E document identification number, date, MAR number,
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-or any combination of known information.
Requirements have 61so been established for Analysis / Calculation Logs to be maintained by the Site
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Nuclear Engineering Services and Fos:11 Engineering Document Centrol on a b
computer. based format.
The logs a t s-mai.nained un-to-d ne and are
. distributed on a quarterly bash to Lil FPC disciplines, and the Manager,
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Nuclear Engineering Projects, for further distribution to A/Es, hard copy
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file locations and the Manager, Ccnfiguration Management.
The administrative controls for aralysis/ calculation 4 propt.rt d t,y i
external engineering / vendor organizations are specified in procedures
NET-103, Management of Design Activities Assigned to Extarnal Engineering Organizations, and NEP-104, Interface Design Control.
Procedure NEP-103
requires that analysis /caiculations be identified in the work r
authorization scope of work as deliverables if these documents are a
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necessary part of the work activity.
This requirement ensures that FPC can maintain a - complete file of design basis information.
Deliverables received by Nuclear Engineering from external. engineering organizations
are required to be reviewed by a FPC Design Engineer to assure that it is complete and verified in accordance with Nuclear Engineering Procedures Manual.
The Design Engineer is also required to review the submitted
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work-for technical adequacy and administrative completeness.
Provisions
'have been made for FPC to retain DBD prepared by others via delivery of the ; magnetic disc in addition to paper copies of analysis / calculations.
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Additional requirements related to the review of engineering software
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. prepared by external engineering organizations are specified in procedure
- i NEP-221, Engineering Software and Vendor Technical Information Review, Revision 3.
This procedure delineates the review responsibilities of the FPC Design. Engineer which are required to be performed in accordance with
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the requi.rements of procedures NEP-103 and NEP-104.
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Based on review of the above procedures the inspectors determined that requirements for preparation, review, approval, and configuration control of instrument -set points had not been specifically addressed.
In response to the inspector's request for information concerning this subject, licensee management stated that G/C Inc. had been retained to develop a setpoint determination methodology for CR-3.
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The scope of the task, to be performed under contract number NPM007D also includes a methodology for setpoint change control and a policy for controlling loop configurations and related loop error calculations.
This task which has a scheduled. activity start date of December 5,1989, is scheduled for completion April 10, 1990.
(An obvious error concerning the completion date was made on contract NPM007AD, which shows a completion date of April 10, 1989).
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The inspectors performed a review of the following listed calculations and supporting documents to verify technical adequacy and procedural compliance.
Acceptance criteria used during this review included
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1) requirements of the calculations as defined by the stated purpose; validity and documentation of assumptions and engineering judgments; documentation and retrievability of references; validity and correctness of computation; i
technical adequacy of results and conclusions; and compliance with procedural requirements:
Analysis /Calcul.ation Loc dated February 5,1990 Document No, f.-83-0008, EDG Alarm Moaule Calculation for Setpolut, dated August ?S, 1989 Document No. E-89-1003, MCC Feeder Breaker Siring; MCC 3AZ BKR.5BL and 5BP; MCC SB1, SKR 3BL and 3BR, dated June 8, 1989 Document No. E-89-0039, Celculatior, of SB Battery Ampare-hours ar.d Number of Plates, dated September 1, 1989 No deficiencies were identified during review of the above documents.
Within this area no violations or deviations were identified.
6.
Design,DesignChangesandModifications(37700)
a MAR Nos. T87-10-19-01, & 87-10-19-01, EDG Cumulative Timers Installation Temporary plant modification, MAR No. T87-10-19-01, was developed and installed to provide a cumulative time that EDG-3A operated above its 30 minute 3000kw rating.
Provisions were made for alarms to be generated each time the EDG load goes above 3000kw for greater than 10 seconds.
Additional alarms were generated after five minutes, twenty-four minutes, and twenty-nine minutes of EDG operation above 3000kw.
The scope of the design change involved removal of an existing timing and alarm scheme and replacing it with a new timing scheme.
Hardware installed by the plant modification included an ETI, three mechanical rotary reset timers, a dual alarm module, associated reset pushbuttons and alarms, and a watt transducer having 4-20 MA output.
The above temporary plant modification was made permanent via development and installation of MAR No. 87-10-19-01.
Additional design changes included:
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' Removing EDG-3A five minute timer and its associated reset button and alarm
' Combining the twenty-four einute and twenty-nine minute alarms into a common annunciator window, while maintaining separate
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trouble contact inputs to the Sequence of Events Recorder
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' Changing the watts transducer installed via MAR T89-10-19-01
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with one meeting seismic and environmental requirements
' Rearranging annunciator windows related to EDG-3A alarms located on SSF A2.
The inspectors reviewed the plant modification packages and
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interviewed license engineering personnel to verify the technical adequacy of the design change.
The following interim drawings along with MAR related sketches were reviewed during this effort:
Drawing #B208-027 SH EG-22 (Sheet.1), Elem. Diag. EDG-3A Timers
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Drawing #E708-027 SH EG-22 (Sheet 2) Elem. Diag. EDG-3A Timers The inspectors verified that the hardware changes were (1) consistent with the design scope and (2) were bounded by the safety evaluation performed
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in accordar.ce with the requirements of 100FR50.59.
The procurement deficiency identified by the licensee related to the use of a nnn-semieslly qualified, non-EQ watts transducer was reviewed by the
inspectors.
The watts transducer procured commercial grade on Purchase Order #F9061464K was installed via TMAR-87-19-10-01.
This MAR package specified applicable codes and standards for seismic and EQ requirements for the enclosure containing the rotary timers, relay and alarm unit. None were specified for the watt transducer.
At the time of the inspection the licensee stated that the installed transmitter was identical to the one replaced, which was siesmically qualified.
A similar argument was presented to establish seismic qualification for the watts transducer.
However, during the inspection period the licensee was unable to provide objective evidence via appropriate documentation to corroborate this claim.
This required documentation was provided Ly the licensee on February 14, 1990.
An
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evaluation for seismic adequacy was subsequently performed.
Results of this review indicate that the transmitter has been seismically analysed in a full configuration.
The inspector has no futher questions regarding this matter, b
MAR No. 88-05-24-01, DHP/EFP Engineered Safeguard Start Hodification The above plant modification was developed to resolve limitation in the EDG loading capacity.
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The ' design scope involved modifications to the ES loading sequence to
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maintain -the 'EDG' load below the 3000 kw 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating. This was
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accomplished via rearranging the ES sequence block loads. Additionally,
[J logic changes were made which precluded simultaneous actuation of the
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EFP-l'and DHP-1 A based on a 500 psi Reactor Coolant' System pressure during a LOOP. Similar changes will be made for Train B equipment.
The_following represents the scope of the plant modification:
- The ES actuation _ matrix for starting the LPI pumps (DHP-1A and DHP-18) is changed from a 1500 psig Reactor Coolant
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pressure to a 500 psig RC pressure energize to trip.
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'The ES 4_ psig Reactor Building (P.B) pressure diverse actuation t_
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signal.for LP1 pump start remains when offsite power exists.
L However,.the 4 psig RB signal is blocked from starting the LPI
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C pumps when offsite power is unavailable.
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'The E5 actuation matrix _for sequence starting the Decay Heat E
Closed Cycle Cooling pumps (DCP-1A and DCP-1B) is moved from
Block 4 to Block 6.
Block 6 is' set to start at 25 seconds.
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- Reactor _ Build' ng $ pray pumps (BSP-1A'and IB) is moved from
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Block 6 is set to start at 25 seconds.
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'The Es actuation matrix for sequence starting the Raw Water i
pumps;(RW-03A and 03B) is moved from Block 4 to Block 5.
Block'5=is set to start at 20 seconds.
- The ES actuation permissive matrix for sequence starting the
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'LPI pumps (DHP-1A and 1B) will be moved from Block 1 to Block 4.
t Block 4 is set to start'at 15 seconds.
'The Emergency Feedwater motor driven pump (EFP-1) has a dedicated ES sequence start matrix to start five seconds after
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EDG loading on Loss of off site. power (Loop) without ES= actuation and-five secondt. after Block three loading with ES actuation.
If
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an ES condition occurs after LOOP the EFP-1 will be tripped and
-re-sequenced, If the ES condition is a LOCA that allows RC
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pressure to decrease below 500 psig with a LOOP event, EFP-1 will be tripped'due to LPI pumps starting.
There is no EFP-1-trip function during non-LOOP events.
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'The ES manual actuation of LPI will not cause automatic start of DHP-1A and DHP-1B pumps until HPI is either automatically or manually actuated to properly sequence load the pump by the ES
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sequence timing permissive matrix.
'The Es manual or automatic actuation of LPI during a LOOP will complete the energize to trip logic to trip EFP-1.
The turbine driven EFW pump EFP-2 is unaffected by this modification, i
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The. inspectors conducted-interviews with licensee management and
determined that discussions related to the development of the MAR l
package had been conducted with-the NRC.
The MAR package is
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scheduled to-be installed during RF07.
However, NRC approval and-a. TS -change.is required prior to making _ the. installed plant
modification operational.
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The safety evaluation performed in accordance with the requirements.
l of 10 CFR 50.59 was - reviewed by the inspectors to assess the y
technical ' adequacy of the evaluation.
No deficiencies were
identified.
The following interim electcical elementary drawings
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were also reviewed to verify that wiring and logic changes
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_ accurately implemented the design obiectives.
The scope of this
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review was limited to one channel that makes up the two out of three logic required to initiate Loperation of Train A equipment affected-by the plant modification.
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Drawing #B208-02B, SH. ES-A20, Elem. Diag., Engineered. Safeguard-H.'P.
' Injection and Load Sequence j
Drawing #B208-028. SH. ES-A37, Elem. Diag.
E.S. Act. "A" L.P.
Injection Channel RC6
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Drawing - ~ #B208-026 EF-01,-Elem. Diag., Motor Driven Emergency-b Feed Pump 3A (EFP-1)
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Drawing #B206-021 SH. DH-01, Elem. Diag. D.H. Removal Pump 3A i;
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Drawing #B208-028, SH. ES-A42 Elem. Diag. Engineered Safeguard-Drawing #B208-028, SH. ES-A66, Elem. Diag. Engineered Safeguard v
Drawing #B203-028, SH. ES-A24, Elem. Diag. Engineered Safeguard Drawing #B208-028, SH. ES-30, Elem. Diag. Engineered Safeguard-Drawing #B208-050, RW-04, Elem. Diag. Decay Heat Sea Water Pump 3A
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Drawing #B208-019, SH, DC-01, Elem. Diag. Decay Heat Closed
' Cycle Cooling Water Pump 3A-Drawing.
- B208-028 SH.
ES-A67, Elem.
Diag.
Engineered
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Safeguard-Reactor Building Spray Start Based -on review of the above drawings and discussions with the-responsible engineer, the inspectors oetermined that the elementary drawings accurately depicted Train A equipment operation as described in the MAR summary description.
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Within this area no violations or deviations were identified.
.c TMAR.8S-03-01-01, MUV-16 Rebuild.
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.This temporary MAR was initiated to replace the valve stem of MUV-16 with a replacement stem manufactured by the licensee. This TMAR will-be removed when a vendor supplied stem is delivered. No deficiencies were noted.
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d MAR 89-05-15-01, DHV-3,4 Pressure lock relief.
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This MAR was. initiated to drill a 0.125. inch hole in the upstream disc on' Decay Heat System wedge gate va'1ves DHV-3, and 4.
The purpose
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of the modification is to prevent pressure locking of. the valves as j
described-ia INPO SOER 84-7. A minor descrepancy was noted with the
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- safety evaluation in that the evaluation stated that the valves are not mentioned in tbc Technical Specifications.
This incorrect statament did not ef fect the safety evaluation conclusion and this ert$r wasLbrought to the attention of the licensee.
e MAR 88"SI-20-01, Replace RCV-139.
This MAR was initiated to replace valve RCV-139 with a 3/4",1500#
Valan globe valve. The existing valve is no longer manufactured and-has.been damaged beyond repair.
No discrepancies were noted.
. ithin this area no vislations or deviations were identified.
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Exit Interview The inspection scope and results were summarized on February 9,1990, with those ' persons indicated in paragraph'l.- The. inspectors described the areas _ inspected and discussed in detail the inspection results. listed below. ; Proprieta ry-information is. not. contained in this report.
' Dissenting comments were not-received from the licensee.
The use of non-seismically qualified ' transmitter in a seismic' application was discussed with; the licensee.
The licensee provided qualification-documentation on. February 14, 1990,.to. substantiate their claim that the-Linstalled watts. transducer is. identical to the one it replaced except for Ethe'4-20 MA output.
The replaced transmitter was seismically qualified.
8.-
Acronyms and Initialisms
- [ABD-Analysis Basis Documents
..A/E
. Architect Engineer Al-Administrative Instructor ANSI.
American National' Standards Institute ANSL Approved Nuclear Suppliers List 1ASCE:
American Society of Civil Engineers
_ASME American Society of Mechanical Engineers
_B&W Babcock & Wilcox BWOG B&W Owners Group CMIS-Configuration Management Information System
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G CRT Change Review Team CR-3-Crystal River 3 DBD Design Basis Document EDBD Enhanced Design Basis Document EDG Emergency Diesel Generator EFP Emergency Feedwater Pump EPQ Exempt Position Questionnaire EPRI Electrical Power Research Institute EPQ Exempt Position Questionnaire EQ Environmental Qua11fication ES Engineered Safeguard ETI Elapse Time Indicator FPC Florida Powar Cerporation FPR Field Problem Report FSAR Final Safety Analysis Report G/CI Gilbert Commonwealth. Inc.
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GOC General Office Corporate I&C Instrumentation & Controls ILS Integrated Living Schedule INPO Institute of Nuclear Power Operations IOC Inter-office Correspondence Kw-Kilowatt LOCA Loss of Coolant Accident LOOP Loss of Off-Site Power LPI Low Pressure Injection MAR
' Modification Approval Record MBPC Modification Budget Planning Committee MCC Motor Control Center MNEA Manager Nuclear Engineering Assurance MNOE Manager Nuclear Operations Engineering NEA Nuclear Engineering Assurance NEP Nuclear Engineering Procedure NGRC Nuclear General Review Committee N00 Nuclear Operations Department NOE Nuclear Operations Engineering NOEP Nuclear Operations Engineering Project NOTO Nuclear Operations Training Department NOTES Nuclear Operations Training Expiditing System NOE/SNES Nuclear Operations Engineering / Site Nuclear Engineering Services NSAC Nuclear Safety Analysis Center NSSS Nuclear Steam Supply System NUMARC Nuclear Utilities Management and Resources Council NUS Nuclear Utility Services O&M Operations & Maintenance QA Quality Assurance RC Reactor Coolant REI Request Engineering Informaion SALP Systematic Assessment of Licensee Performance
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SEE-Southeast Electric Exchange SLS Safety List System SNES.
Site-Nuclear Engineering Serv',ces
. SP Special Projects
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SQUG Seismic Qualifications Utilities Group TS
- Technical Specification
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VP Vice President VTI Vendor Technical Information
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