ML20212E758
| ML20212E758 | |
| Person / Time | |
|---|---|
| Issue date: | 02/18/1987 |
| From: | Correia R, Jocelyn Craig NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20212E751 | List: |
| References | |
| REF-PT21-87, REF-QA-99900507 NUDOCS 8703040478 | |
| Download: ML20212E758 (26) | |
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ORGANIZATION: SARGENT AND LUNDY ENGINEERS CHICAGO, ILLIN0IS REPORT INSPECTION INSPECTION NO.: 99900507/86-01 DATE:
12/1-5/86 ON-SITE HOURS:
125 CORRESPONDENCE ADDRESS: Sergent and Lundy Engineers ATTN: Mr. L. E. Ackmann Senior Partner 55 East Monroe Street Chicago, Illinois 60603 ORGANIZATIONAL CONTACT: Mr. H. S. Taylor, Associate and Head, QA Division TELEPHONE NUMBER:
(312) 269-6371 NUCLEAR INDUSTRY ACTIVITY: Design and engineering services for nuclear facilities under construction ar.d operating f acilities requesting modification and reanalyses of existing systems, comperents and structures.
2 S 7 ASSIGNED INSPECTOR:
Richard P. Correia, Special Projects Inspection Date Section (SPIS)
OTHER INSPECTOR (S):
K. C. Leu, SPIS V. Venczel, Conspitant
[. 4 2//5/M
[JohnW.Craig, Chief.SPIS,VendorProgramBranch APPROVED BY:
' Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21, 10 CFR Part 50, Appendix B and Sargent and Lundy I M T Topical Report (TR) No. SL-TR-1A.
B.
SCOPE:
(1) Status of previous findings, (2) Dedication of commercial grade items overview, arc' (3) 10 CFR Part 21 evaluation and reporting review.
PLANT SITE APPLICABILITY: Byron 1/? (50-454/455); Braidwood 1/2(50-456/457);
Clinton 1(50-461); LaSalle 1/2(50-373/374); Zimer I (50-358).
8703040478 870224 PDR GA999 EECSALE 99900507 PDR
s ORGANIZATION:
SARGENT AND LUNDY ENGINEERS CHICAG0, ILLIN0IS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 2 of 16 A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
None.
C.
UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVI0lls INSPECTION FINDINGS:
1.
(Closed) Unresolved Iten (64-02)
During(the previous NRC inspection, it was learned that Sargent Lundy S&L) had perforred a review of 70 structural pipe support anchors which had been issued for field installation without proper qualification. The Clinton Structural Archor Review Summary listed the 70 anchors and to what extent their qualifications affected the safety of the piping subsystems they were e part of.
Durirg this inspection, the NPC inspectors reviewed the status of the 70 anchors, a representative exanple of the previous ard current cualifying calculations and S&L's corrective actions to prevent such a reoccurrence. The inspectors deternined that this issue had been adequately addressed and corrective actions had been implenented to prevent future renccurrences. Details of this review are provided in Section E.1 of this report. This item is censidered closed.
?.
(Closedi Unresolved Item (84-02)
During the last NRC inspection, a review of an STL internal audit was performed. This audit had been committed to beiro performed by S&L as a result of nonconformance A.1 of Inspection Report Nc.
99900507/83-03. The audit had identified three (3) nonconformances.
two of which had been satisfactorily resolved when the last NPC inspection tock place. The third open nonconformance has since been closed satisfactorily by S&L QA personnel.
It pertained to the Byron /Braidwood plents and involved a review of all project proceduret and guidelines to determine which needed +o be converted into pro.iect instructions.
During this NPC inspection, the NPC inspectors reviewed the close-out of this nonconformance and audit and determined that these issues are resolved.
Details of this review are provided ir Section E.2 cf this report. This item is considered closed.
ORGANIZATION: SARGENT AND LUNDY ENGINEERS CHICAGO, ILLIN0IS INSPECTION REPORT PAGE 3 of 16 NO.: 99900507/86-01 RESULTS:
3.
(Closed) Unresolved Item (84-01)
During the previous NRC inspection, it was determined that the analysis of the fire protection / suppression systen at the Byron /
Braidwood plants had a deficiency similar to one found at the The deficiency was identified as a reduced system LaSalle plants.
capebility to supply water to certain areas after piping has aged.
During this inspection, the NRC inspector reviewed the latest Byron /
Braidwood plant's fire protectien/ suppression system analyses and Details determined that the problem had beer addressed and resolved. This item of this review are provided in Section E.3 of this report.
is considered closed.
E.
OTHER FINDINGS AND COMMENTS:
1.
The Clirton Structural Anchor Review
Background
In February 1962, S&L initiated a general review with respect to design qualificatien of 70 structural anchcrs at Clinton Unit 1.
It was speculated by S&L engineers that the review ection resulted fron Safety Relief Valve (SRV) load reanalysis of anchors on systems Consequently, Sll issued a report near the suppression pool.
"Clinton Structural Archor Review Sumary," EMD No. 04749? dated The report concluded that 13 anchort required redesign May 4, 1984 of welds and/or components in order to r:eet applicable ASME code The other 57 anchort, their designs and analyses requirements.
meetirC the code requirements, were acceptable.
Procedure Review The NRC inspecter reviewed S&L's " Procedure for Evaluation of 049437, dated September 4, 1984 for Anchors," Rev. O, EMD No.
utilization of proper techniques for analy.ing ASME class 2 & 3 i
and ANSI B31.1 piping systems, ard basic types of interrediate structural anchors used to restrain the pipe. Also, a review of
" Project Instructinr for Piping Analysis Corponent Support Design and Structural Load Verification," Rev. 4, No. PI-CP-016, dated was conducted. This instructicn established the March 19, 1984, assignment of responsibility for design and detailing of all items and welds for various types of supports between applicable engineering and design organizations within S&L. The NRC inspector found ther appropriate and acceptable.
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SARGENT AND LUNDY ENGINEERS CHICAG0, ILLIN0IS PEPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 4 c' 16 Anchor Review Engineering Mechanics Division (EMD) Anchors The NRC inspecter selected the fo110 win 9 four EMD anchor stress packages from the 13 redesigned anchors for review. EMD stress calculations, in general, included any component end associated welding which interface directly with the piping.
The following lists the four packages which were reviewed:
ISX46009A, EMD No. 034479, Rev. 01, dated November 14, 1985 ICC13029A, EMD No. 027556 Rev. 01, dated February 20, 1985 1CC13069A, EMD No. 028029, Rev. 01, dated July 29, 1981 1RT0902aA, EMD No. 025062, Rev. 02, dated November 7, 1980 Analyses showed that there were overstresses in pipe welds and/or stanchions in the original designs.
S&L performed reanalyses of these anchors, provided an alternate design for anchor ICC13Cf9A, and recommended specific changes in weld sizes and plate dimensions for the other three anchors as listed below.
ISX46009A, cberge the partial-penetration 1" weld between the:
pipe and plate, and the 5/16" partial-penetration weld between plate halves to full peretration welds 1CC13029A, change stanchion from the existir9 2"O SCH1f0 to a 2.50"0 SCH80 and make the weld between pipe and stanchion a full penetration 1RT09024A, replace the single 3/4" x 8" x 17"2 with two, 1-3/8" x 10" x 10"2 nake til the welds between pipe and plates and between plate halves full penetration.
The NRC inspector found the proposed design changes and alternate design for the above four anchors acceptable.
The remaining group of 57 anchors, S&L determined that their design calculations and analyses were acceptable, and therefore, oc redesign was recessary. The NRC inspector selected the fc11 ewing four EMD anchor packages from different piping systems for review.
ICC1306SA, EMD No. 028029, Rev. 01, dated July 29, 1981 ICC13067A, EMD No. 028029, Rev. 01, dated July 29, 1981 1MS24014A, EMD No. 028024, Fev. 01, dated June 11, 1984 1RH2!001A, EMD No. 022407, Rev. 03, dated August 12, 1980 l
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SAPGENT AND LUNDY ENGINEERS CHICAG0, ILLIN0IS I
INSPECTION REPORT PAGE 5 of 16 NO.: 99900507/86-01 RESULTS:
In order to confirm the adequacy of the above anchor designs, applicable design loads such as dead load, temperature, ss+cvic, and displacement, were included for review. Different load combinations reflecting operational, upset, emergency and faulted conditions were used in developing the final design loads. Three directional moments and forces were included in designing and verifying stanchions and their welds, and the stresses were compered to the allowables provided by equations 8, 9, 10, and 11 of ASME Section III, Subsection NC, 1983 edition. The analyses and calculations were adequate and acceptable.
Structure 1 Engineering Division (SED) Anchors Anchors which included stanchions and corponents and associated welding interfacing between the piping end the supporting building structures were designed by SED. The NRC inspector selected the following 6 SED anchor packages for review:
M-1CC13067A, " Calculations for Mechanical Component Support No. M-1CC13067A," Rev. 1, dated August 10, 1981 and Pev. 01, dated May 17, 1985.
M-ISXa6009A, "Calculatir.ns for Mechanical Component Support No. M-1SX46009A," Rev. 2. dated August 27, 1981 and Rev. 02, dated June 21, 1983.
M-1RT09024A, " Calculations for Mechanical Component Support No. M-1RT09024A, Rev. 1, dated October 15, 1981 and Rev. 82, dated August 19, 1985.
M-1CC13029A, " Calculations fo-Mechanical Component Sunport No. M-1CC13029A," Rev. E2, dated January 21, 1985.
M-1RH21001A, " Calculations for Mechanical Compnnent Support No M-1RH21001A, Rev. G2, deted October 12, 1983.
M-1MS24014A, " Calculations for Mechanical Component Support No. M-1MS2401aA," Rev. El, dated April 23, 1985.
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ORGANIZATION: SARGENT AND LUNDY ENGINEERS CHICAG0, ILLIN0IS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 6 of 10 Design load types considered in the above structural anchor designs were comprised of design, emergency and faulted conditions. The faulted condition, when applicable, usually governed the design.
Structural members were checked for stress and deflection. Supports' welded connections at the expansion anchor plate (s) were designed for bending and torsional stress effects. Exnansion anchor plate checking included shear, torsion, deflection, bolt size and web crippling.
Embedded plates and studs were checked for stress and pull out force.
Applicable criteria from the AISC and ASME Codes were used. Weld procedures were consistent with Baldwin Associated (BA) QC Data Sheet (BA was S&L's piping and pipe support contractor), and weld examination procedures were in accordance with ASME ND and NF requirements.
in the review of anchor 1RT09024A, the NRC inspector found that anchor loadings in the final design were greatly increased for the emergency and faulted conditions, but were unchanged for design condition.
It was later found that the loading changes in the emergency and faulted conditions were due to combinirg loads at seismic /nonseismic interface which included loads fron both sides of the interface. The design condition irvolved only dead and thermal loads, and therefore was not affected. A transmittal letter, " Design Information Transmittal,"
t DIT-CP-EMD-1503-1, dated July 25, 1985 confirmed the loadire changes.
In reviewing the anchor drawing #or M-1RHE1001A, the NPC inspector found that the EMD review / concurrence was not indicated on the drawing.
S&L obtained by telex fren the Clir. ton site a copy o' transmittal letter No. F-5143. This letter was sert from SED to EMD on July 13, 1904 for review and conraent of the r#er drawing, and was approved l
and signed by EMD the same day.
The inspector found the SED ar.9.or 6.,gns acceptable.
Cortactive Actions S&L has taken the followirg corrective actions ccocerning welded anchors and welded attachments in response to the Clinton 1 anchors reevaluation program.
1.
PI-CP-016 was revised requiring EMD to prepare, review and approve welded attachments and/or ancher calculations prior to the release of design in'crmation for a piping subsystem analysis.
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a ORGANIZATION: SARGENT AND LUFDY ENGINEERS CHICAGO, ILLIN0IS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 7 of 16 2.
PI-CP-016 was revised reouiring EMD to review pipe support drawires which contain welded anchors, shear lugs, and integral attachment designs.
3.
The Division of Responsibility (00R) established assignment of responsibility for all items and welds for the various types of supports.
The inspector observed that the corrective actions have'been imple-mented in S&L's current anchor designs.
As a result of the review work of the selected anchor designs regarding procedures, code reouirements and regulations, engineering calculations, and the corrective actions and implementatiers, the NRC inspector believes that the current designs and analyses of the 70 structural anchors at Clinton Unit 1 are acceptable. This open item is censidered closed. See Section 0.1 of this report.
2.
S&L Internal Audit Review An internal audit was committed to be performed by St.L as a result of NRC inspection nonconformance 99900507/83-03, A.1, for the use of and control of piping line lists on all major S&L projects.
S&L's audit identified three nonconformances, two of which were resolved at the tire of NRC inspection 99900507/84-03. During this NRC inspection, a re-review of the S&L internal audit was conducted.
Three S&L audit reports were reviewed; the first G-190, dated May 3, 1984, addressed the NRC nonformance identified in inspection report 99900507/83-03 as part of the reason for the S&L internal audit and included the results of the audit, in which the three previously mentiered nonconformances were cited. The second audit report reviewed, G-190.1, dated March 25, 1985, eddressed the reaudit of the findings of internal audit identified ir Report G-190.
The reaudit report addressed two of the three ncnconformances which were reviewed by S&L for corrective actions and close-out.
This reaudit report identified that two of the ncnconformances were considered closed by combined corrective actinns upon the issuance of a project l
instruction. The project instruction, PI-BB-61 effectively corrected the issues identified with the status and current revision cf piping lire lists on the Byron /Braidwood plants.
ORGANIZATION: SARGENT AND LUNDY ENGINEERS CHICAGO, ILLIN0IS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE E cf 16 The third report, G-190.2, dated August 16, 1085, addressed the third remaining nonconforrance identified in S&L audit report G-190. This portion of the S&L audit was addressed as the open item 84-02 in NRC inspection report 99900507/04-03. The nonconformance identified that S&L project procedures and guidelines were being used in safety-related design activities. S&L's OA program, Section 3.01, requires in part that safety-related desigr activities be planned and controlled by department standards and procedures, project instructions and QA procedures. Existing Byron /Braidwood project procedures were reviewed by S&L project staff tn determine if any should be reissued as project instructions. Subsequent profeet procedures were either voided as they were no longer recuired or charged to project instructions to comply with S&L 0A requirements.
The nonconformance was closed tc the satisfaction of the internal S&L reaudit. The NRC inspector reviewed the applicable audit documentation and concurs that the nonconformances corrective actions should prevent recccurrences. This iten is considered closed.
See Secticn D.2 of this report.
3.
Byron /Braidwond Fire Protection / Suppression System A concern addressed during the previous NRC inspection involved LeSalle Urits 1 & 2 fire protection /suppressian system not beirg capable of supplyinc water to certain areas after the piping has aged and that a similar problem existed at the Byron /Braidwood plants.
Durino tH s NRC inspection, the NRC inspector discussed the LaSalle issue with S8L persorrel in more detail in crder to evaluate the Byron /Braidword fire protectior/suppressicn analysis.
A Sf.L LaSalle project engineer told the NRC inspectcr that the problen at LaSalle was addressed during an NRC site inspectier when it was noted that the as-built confinuretion of the system differed from the as-analyzed systen. S&L project analysts performed computer analyses of the as-built configuration and included aged-pipe factors ir areas which had rot previcusly used this criterion. The results of this analyses showed that very i
minor modifications were recuired. The LaSalle fire protection issue was addressed during an SFL project meeting. At this tine l
the Byron /Braidwood fire protection / suppression system analysis was still in prccess and the as-built and aging considerations were included. The NRC ir.spector reviewed the Byron /Braidwood fire protection calculation files and found that the correct factors were considered end the results showed that the system would be capable of meeting its intended functiont es designed.
This item is considered closed.
See Section D.3 of this inspectior.
ORGANIZATION: SARGENT AND LUNDY ENGINEERS CHICAG0, ILLINDIS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 9 c' 16 4.
Computer Software Error Notification The previous NRC inspection cited S&L with a norconformance for not establishing and implementing a procedure for the handling of computer software error reports.
During that inspection, S&L Procedure CSD-10.4.3, " Computer Software Error Notification" was in draft and was te he reviewed during a future NRC inspection.
The current revision of CSD-10.4.3 Rev. 2, dated October -16, 1086 was reviewed by the NRC inspector. The procedure outlines how i
computer software errors are to be received, documented, respont -
bilities determired for reporting, evaluation, corrective actions te be made and problem resolution recorded. The procedure provides 'or both inhouse S&L software errors and software procured from outside sources which includes liaisoning and evaluation of reported errors.
The NRC irspector asked S&L QA personnel what procedures er processes were in place which will assure that once an error har.
been reported, that all affected designs which utilized the computer ccde would be identified and evaluated. Section 4.1 9 of Procedure CSD-10.4.3 reouires that the division head who is the recipient of an Error Notification Report is responsible for calculations, decurenting any recessary corrective actions in accordance with applicable departrent standards, maintaining files pertaining te any corrective acticns and reperting closure of the corrective action to the Documentation Librariar, Ccmputer Services Division (CSD).
The NRC inspector reviewed computer error LPGEOD (09.5.220-3.5) 86-1 which addrassed an error in the "Largebore" computer progran concern-ing the calculation of an eccentricity used to transforn structural menber reacticns to the center-of-oravity (c.g.) of the structural weld.
During the review, it was noted that reouired procedural steps were followed including return receipt of user evaluatien results. Only the Byron /Braidwood pro.iect was affected by this particular code error.
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A review of the error notification evaluation pe-formed by the affected structural groups was performed. Upon receipt of the error rctificaticn the Structural. Analytical Division (SAD) (the group respersible for the "Largebore" code) performed a data searcF of all designs which could have been affected by tha arror.
A list of all affected designs was iteni ed by proiect and was sent to all 9
ORGANIZATION: SARGENT AND LUNDY Et!GINEERS CHICAGO, ILLIN0IS
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REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 10 of if responsible project managers for further evaluation.
Each affected design was then evaluated fer possible effect due to the code error.
Corrective actions (evaluations and calculation revisions) were then performed and results documented. The sumary of the results of the evaluations were then compiled and included in the Progran Error Evaluaticn Report (per procedure CSD-10.4.3) for close out of corrective action activities. No nonconfomances were found during this part of the inspection.
5.
10 CFR Part 21 Review S&L procedure, GOP B-14, rev. 1, " Reporting of safety-related defects ard nonconformances for nuclear plants in accordarce with 10 CFR 21" was reviewed by the NRC inspector. The structure of the General Operating Procedure (GOP) is outlined as follows:
1.
Intro-Policy 2.
Definition of 10 CFR 21 Terms 3.
Reporting Criteria 4
Evaluation of Defect /Nonccmpliance 5.
Notification to NRC 6.
Records The S&L reporting criteria cutlined in GOP B-14 is as follows:
1.
Is the potential defect in a USA facility?
2.
Is it within the S&L scope of responsibility?
3.
Has the basic component been " offered for acceptance" 4.
Could the deviation result in a substantial safety hazard?
"Cffered to the Purchaser for Acceptance" is defined as follows: A document shall be considered " offered for acceptance" when it has beer communicated to another party that will use it in design, in manu-facturing or in preparing a dccument for the manufacturing of any basic component.
If a reported devietion meets the reportinc criteria, form GOP B-14.1 is filled out and sent to the Head, QA Division. The Head, QA Division then forwards the fom to the Head, Nuclear Safeguards and Licensine Division who is also the Chaiman of the Nuclear Safety Review Committee (NSRC). The NSRC u nsists of rembers chosen by the chaiman from appropriate divisions and departments depending upon the nature of the deviation. The chairman establishes the schedule for the completion of the required evaluation.
If
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ORGANIZATION:
SARGENT AND LUNDY ENGINEERS CHICAG0, ILLINOIS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 11 of 16 applicable, the licensee or client is notified that a possible defect has been reported and is under review.
If a defect has been evaluated as being reportable, the chairman documents the information, provides a suggested corrective action and notifies the Director of Engineerina.
The Director of Engineering documents receipt of the notification and returns one copy to the Head, Nuclear Safeguards and Licensing Division.
Copies are also sent to the Director of Services, Project Director, the reporting employees' Department Manager (as applicable) and Division Head and Head, QA Division. The Director of-Engineering is responsible for notifying the NRC and all affected clients and projects. All compiled documentation for each evaluation shall be maintained in QA Records whether the evaluation results in a report to the NRC or not.
As an example, the NRC inspector reviewed the QA file on a non-reported potential defect. The evaluation was of a potential defect at the Zimmer Unit I station.
It concerned sedimentation in the intake structure. The S&L evaluation judged that the problem was
,not a defect reportable under 10 CFR 21.
The initiating incident was sedimentation measurements taken inside the intake flure of the Zimmer station en April 24, 1979. Soundings indicated a sedimentetion profile along the entire length of the I
intake flune ranging from 5 to 12 feet in thickness. An essential service water pump located in the river pump house that had been used during preoperational testing started to cavitate and a pump bearing
- ailure occurred.
It was felt that sedimentation may have contributed to the cavitction and bearing failure but could not be conclusively determined.
The higher levels of sedinentation found was determined to have occurred during the 1976 to 1979 time period when river flooding over flowed the bulkhead at the entrance to the river. During most of this period, the service water pumps were not in operation i
and flow velocity in the intake fiume was essentially zero.
It was S&L's understanding that sedimentation would occur and that routine sedimentation monitoring and removal would be performed by Cincinnatti Gas & Electric Company. There was no indication that the water intake structure had any defects or flaws in desian or construction which would have contributed to the sedimentatior. build up in the flume.
Cincinatti Gas & Electric Ccmpany had reported the cccurrence to the NPC, Regicn III, under the requirements of 10 CFR Part 50.55(e).
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ORGANIZATION: SARGENT AND LUNDY ENGINEERS CHICAGO, ILLINDIS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 12 cf 16 The S&L evaluation resulted in a non-reportable defect in the S&L design under 10 CFR Part 21. The NRC inspector reviewed S&L's adherence to 10 CFR Part 71 and GOP B-14's requirements and found no violations or nonconformances with this part of the inspection.
6.
Dedication of Commercial Grade Items The objective of this area of the inspection was to review what measures S&L has established for the dedication of commercial grade The NRC components for use in safety-related applications.
inspectors found that S&L does not have a generic progran for dedication. However, on a case-ty-case basis, S&L has developed client-specific dedication processes conforming to their customers requirements.
It should be noted that S&L considers the spare parts procurement program developed for clients as proprietary. As such, the follewing inforration is given in general terms. The inspection report is based on:
personal interviews; review of procedures governing work activities; examination of work performed using gcverning precedures; inspection of turrover documentation; and S&L technical reports presented at the American Power and ASME PVP conferences.
Durino this inspection, the dedication program developed for Ccmmenwealth Edison Company's (CECO) Byron and Braidwood plants was reviewed. The program ercompasses replacemert/ spare parts for both operational plar.ts, and for plants under construction, a)
CECO's Dedication Prooram CEC 0 requested S&L te develop a Spare Parts Precurenent Program for replacement, maintenance and initial provisioning which will satisfy NRC requirements, increase plant availability, and achieve cost effectiveness. The program was to be structured to provide CECO with, among other things, the ability to select between using qualified vendors to procure safety-related component parts or to upgrade (dedicate) equivalent commercial-grade component for safety-related applications. To accorplish these objectives, S&L perforned five tasks:
- 1) component classification; E) part classification; 3) parts procurement requirements identification, based on classifice-tions; 4) computer data base development for tasks 1 thru 3; and
- 5) provide procedures and instructions for the utility to perform continued work recuired by the program.
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ORGANIZATION: SARGENT AND LUNDY ENGINEERS CHICAGO, ILLIN0IS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 13 of 16 The program is expected to be turned over to the utility for implementation in calender year 1987. These tasks are in various stages of completion at the time of the inspection and are described below.
1)
Component Classification - Component classification is based on the comoonent's system function and its contribution to safety of the plant. Each component is classified with respect to its nost critical operatirg mode during plant normal and postulated accident condition as outlined in the American National Standards, ANSI /ANS-51.1 1983 (Nuclear Safety Criteria for the Design of Stationery Pressurized Water Reactor Plar.ts). Safety function is also assigned to each component to describe the attributes or preperties the component must possess to permit its function during the most demanding operating mode.
Consideration is given to the effect on the system when loss of safety function occurs.
Safety furetion and compenent safety-related status is identified csing information contained in the plant's Safety Related Components List (SRCL), piping and instrurent diagrams, control and instrumentation diagrams, electrical schematics, FSAR and CEC 0's commitrents to the NRC. With the identification of the components safety function, re-classification is feasible in some cases based on engine-ering evaluation and justificatior.
For example, a passive valve motor operator is considered safety related because of its association with a Class 1E circuit. However, with the installatien of a simple isolatier device, such as a fuse, the operator may be reclassified as non-safety related.
2)
Parts Classification - Each part of a safety-related compe-nent is assigned a Parts Classification (PC) code using Failure Mode and Effects Analysis (FMEA) with cor.sideratier given to previously determined safety function. Then, a parts ranking is established depending on the part's con-l tribution to the component's safety function. The four l
classifications for parts ranking are:
PR-3 Parts governed by the ASME Boiler ard Pressure Vessel codesSection III.
PR-2 Parts essential to the component's specific safe +y functional performance, bu+ nre not PR-3 (failure effect cccurs).
ORGANIZATION: SAPGENT AND LUNDY ENGINEERS CHICAGO, ILLIN0IS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 14 of 16 PR-1 Parts whose failure may degrade sore of the component's functions, but not the safety functions; continued availability of the component / plant is challenged.
PR-0 Parts which do ret contribute to the components safety function, performance or normal plant availability.
The purpose of parts classification is to inforr CEC 0's operations, technical and maintenance organizations the relationship of various components.
Ranking of PR-1 or higher are considered important te plant availebility and should be subject to suitable procurement, surveillance, and maintenance efforts.
3)
Parts Procurement - Procurement requirements consistent with Regulatory requirements are established including the use of commercial grade itens for safety-related applica-tions through dedication. The establishment of procurement requirements is an iterative process. A deternination is made whether or not an item satisfies the NRC definition for "Cormercial Grade Item".
Section 21.2 ef 10 CFR 21 states that e "comercial grade item" must:
"nnt be subject to design or specificatien requirements that are unique to facilities or activities licensed by the NRC; be used in applications cther than facilities or activities licensed by the NRC; and be ordered from manufacturers / suppliers on the basis of specifications set forth in the manufacturer's published product description (a catalog or specialized brochure item)."
For those parts categorized as safety-related, established procurement requirements dictate procurement from qualified nuclear vendors under the provisions of 10 CFR 21 trd 10 CFR 50, Appendix B.
Each original vendor is contacted for availability of the original components under 10 CFR 21 ard 10 CFR 50, Appendix B procurements. Those parts unavailable through oualified nuclear vendors are identified for procurement as commercial crade and subject to dedica-u
ORGANIZATION:
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PAGE 15 of 16 tion to safety-related applicatiens by the utility.
Such dedication is to be acccmplished by establishing a documented link between the original and replacement component. Other components established as commercial grade items, based on safety-functions, are also rubject to dedication.
After a conponent is classified as a comercial grade item, procedures are established for each item to be purchased (based on the vendors published data) and upgraded to a safety-related application.
For example, a safety-related instrument operating in a mild environment must meet seismic requirements to be qualified.
The licensee could buy the instrument as a commercial grade prcduct and dedicate it for safety-related application through testing and inspection, in order to document the adenuacy of the conponents for their intended safety function.
If the design or fabrication process for the replacement iten differs from the original procured item, the licensee may perform a similarity analysis.
4)
Computer Data Base - A computerized data base information system was developed to facilitate component classificttion, parts classification, and the establishment of procurement requirements. After the Spere Parts Procurement Program is completed and turned over to the client (CECO), the data base becomes the utility's basis for inventory contrcl and spare parts procurenent activities.
5)
Precedures and Instructions - Documentation is being developed to provic'e CEC 0 with a users manual to imple-ment the Spare Parts Procurement Program.
b.
Review of Perfonned Work The NRC inspector performed a review of S&L's ccmpliance to procedures governing work activities ard quality ren.uirements of the client-specific dedication pregram invoked by S&L's Quality Manual.
It was found that procedures governine work ectivity were written, approved, and controlled in accordance with prescribed measures. Training was civen to personnel for i
the use and implementation of developed procedures. Several
ORGANIZATION: SARGENT AND LUNDY ENGINEERS CHICAGO. ILLIN0IS REPORT INSPECTION NO.: 99900507/86-01 RESULTS:
PAGE 16 of 16 different homogeneous component matrix reports were reviewed:
the classification of components and parts; establishment of parts procurement requirements (including the mechanism for commercial grade item dedicatier); and the development of computer database entries appeared to be consistent with governing working instructions and S&L's Quality Manual.
In reviewing work performed by S&L for CEC 0's Spare Parts Procurement Program, the NRC inspector found that work was performed in accordance with certrolled anc prescribed procedures.
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