ML20011D103
| ML20011D103 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 09/30/1989 |
| From: | Udy A EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20011D104 | List: |
| References | |
| CON-FIN-D-6002 EGG-NTA-7249, GL-83-28, TAC-64022, TAC-64049, NUDOCS 8910310022 | |
| Download: ML20011D103 (16) | |
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EGG-NTA 7249 ENCLOSURE 3 TECWICAL' EVALUATION REPORT CONFORMANCE TO GENERIC LETTER 83 20. ITEM t.2.1-EQUIPMENT CLA551FICAT!0N FOR ALL 0THER SAFETY RELATED COMP 0NENTS-
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Docket Nos. 50 456/50 457 l
Alan C. Udy l
Pubitshed September 1989 t
Idaho National Engineering Laboratory EG&G Idaho, Inc.
Idaho Falls Idaho 83415 Prepared for the U.S. Nuclear Regulatory Comission Washington D.C.
20555 Under DOE Contract No. DE AC07-761001570 FIN No. D6002 TAC Nos. 64022/64049
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SL4 MARY This EG48 Idaho, Inc., report provides a review of the licensee's submittals for Unit Nos. I and 2 of the Braidwood Station for conformance to Generic Letter 83-28, Ites 2.2.1.
Itse 2.2.1 of Generic Letter 83 29 requires licensees and applicants to submit a detailed description of their programs for safety related equipment classification for staff review. It also describes guidelines that the licensee's or applicant's programs should encompass. The review concludes that the' licensee complies with the requirements of this ites.
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FIN No. D6002 B&R 20-19-40-41-3 Docket Nos. 50-456/50-457 TAC Nos. 64022/64049 s
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PREFACE s
This report is supplied as part of the program for evaluating 1
l licensee / applicant conformance to Generic Letter 83 28 ' Required Actions Based on Generic laplications of Sales ATW$ Events.' This work is being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear i
Reactor Regulation, Division of Systems Technology, by EG&G Idaho, Inc.,
Regulatory and Technical Assistance Unit.,
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SUMMARY
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PREFACE...............................................................
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INTRODUCTION.....................................................
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REVIEW CONTENT AND FORMAT........................................
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ITEM t.t.1 - PROCRAM.............................................
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3.1 Guide 11oe..................................................
3 3.2 Evaluation.......................
,3 3.3 Conclusion................................................
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ITEM t.2.1.1 - IDENTIFICATION CRITERIA...........................
4 4.1 Guideline..................................................
4 4.2 Eval uat i on...................
4 4.3 C o nc l u s i on................................................
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ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM.......................
5 5.1 Guideline..................................................
5 5.2 Ev al uat i on..........................
5 l.3 Conclusion.................................................
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ITEM 2.2.1.3 - USE OF THE EQUIPMENT CLASSIFICATION LISTING.......
6 6.1 Guideline..................................................
6 6.2 E val u a t i on............................
6 6.3 Conclusion................................................
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ITEM 2.2.1.4 - MANAGEMENT CONTROLS...............................
7 7.1 G u i d e l i ne..................................................
7 7.2 Ev al u a t i on..............................
7 7.3 Conclusion.................................................
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ITEM 2.2.1.5 - DESIGN VERIFICATION AND PROCUREMENT...............
9 8.1 Guideline..................................................
9 8.2 Ev al ua t i on...................................
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8.3 Conclusion............................e....................
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ITEM 2.2.1.6 "IMPORTANT T0 SAFETY" COMPONENTS..................
10 9.1 Guideline..................................................
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- 10. CONCLUSION.......................................................
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- 11. REFERENCES.......................................................
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CONFORMANCE TO RENERIC LETTER R3 28. ITEM 2.2.1-.
EDUlfMEE_ILASSIEP.ATION FOR ALL OTHER 1AFETY RELATED COMPONENT $r RRAIDW000 1/ 2 1.
INTRODUCTION On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salen Generating Station failed to open upon an automatic reactor trip signal from the reactor protection system. This incident was terminated j
manually by the operator about 30 seconds after the initiation of the j
automatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior to this incident, on February 22, 1983, at Unit 1 of the Salen Generating f
Station, an automatic trip signal was generated based on steam generator low-low level during plant startup.
In this case, the reactor was tripped manuelly by the operator almost coincidentally with the automatic trip.
Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (E00) directed the NRC staff to investigate and report on the generic imp 1tcations of these occurrences at Unit 1.
The l-results of the staff's inquiry into the generic implications of the Sales-1 incidents are reported in NUREG 1000, " Generic Implications of the ATWS Events at the Salen Nuclear Power Plant.' As a result of this investigation,theCommission(NRC) requested (byGenericLetter83-28 dated I
July 8,1983 ) that all licensees of operating reactors, applicants for an operating license, and holders of construction permits respond to the generic issues raised by the analyses of these two ATWS events.
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This report is an evaluation of the responses submitted by Connonwealth l
Edison, the licensee for the Braidwood Station, for Item 2.2.1 of Generic L
Letter 83 28. The documents reviewed as a part of this evaluation are listed in the References (Section 11) at the end of this report.
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KVIN CONTDIT AND F0IMAT l
Item t.2.1 of Generic Letter 83 28 requests the licensee to submit a description of their programs for safety related equipment classification for staff review. Detailed supporting information should aise be included i
in the description, as indicated in the guideline section for each ites within this report.
As previously indicated, each of the six items of Item t.2.1 is evaluated in a separate section in which the guideline is presented; an evaluation of the licensee's response is made and conclusions about the t
programs of the licensee for safety related equipment classification are drawn.
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ITDI 2.2.1 PROGRAM 3.1 Guideline Licensees should confire that an equipment classification program is in place that will provide assurance that safety related components are designated as safety related on plant documentation. The program should provide assurance that the equipment classification information handling system is used so that activities that may affect safety related components are designated safety related. By using the information handling system, personnel are made aware that they are working on safety related components and are directed to, and are guided by, safety related procedures and constraints. Licensee responses that address the features of this program are eval'uated in the remainder of this report.
3.2 Evaluation The applicant (now licensee) for the Braidwood Station responded to these requirements with submittals dated November 5, 19838 and February tg. 1984.3 The licensee submitted additional information on April 21, 198g.4 These submittals describe the licensee's safety-related equipment classification program.
In the review of the licensee's responses to this ites, it was assumed that the information and documentation supporting this program is available for audit upon request.
3.3 cenclusion i
We have reviewed licensee's submittals and find that, in general, the licensee's responses are acceptable.
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ITEM t.2.1.1 IDENTIFICATION CRITERIA 4.1 Gyldelina The licensee should confirm that their program used for equipment classification includes the criteria used for identifying components as safety related.
t 4.t Evaluatina The licensee states that the criteria used to classify structures, systems, components, and parts as safety related are contained in the Station Nuclear Engineering Department's quality procedure Q.11, exhibits 8 t
and C, in station administrative procedures, and in engineering proa.edure manuals. Procedures that are based on the requirements of the corporate quality assurance manual control the class,1fication of safety-related systems and components. These procedures were not included in the response.
4.3 tonclusion The licensee's responses to this item are complete and address the i
staff's concerns. We find the licensee's responses for this item acceptable.
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ITEM 2.2.1.2 - INFORMAT!061 HANDLING SYSTEM 5.1 Guidelina i
The licensee should confirm that the program for equipment classification includes an information handling system that is used to identify safety related components. The response should confirm that this information handling system includes a list of safety-related equipment and that procedures exist to govern its development and validation.
5.2 Evaluation The licensee states that the information handling system consists of listings made using computer programs and techniques.
It is identified as the Safety-Related Component List, and consists of the Q List, valve list, instrument index list, mechanical equipment list, and as built piping and i
instrument diagrams. The licensee states'that vendor input is used i
extensively, and that detailed parts lists are being generated using the same method. This list is maintained current by Sargent and Lundy in a -
program under the control of the PWR Systems Design Section of the Nuclear Engineering Department, using quality assurance procedures Q.P. 3 3,
' Classification of Systems, Components, Parts and Material,' Q.P. 3 51,
" Design Control for Operations - Plant Modification,' Q.P. 4 51,
' Procurement Document Control for Operations - Processing Purchase l
Documents," and Engineering Procedure Q.lt.4, " Control and Maintenance of the Safety-Related Component List (SRCL) for Byron and Braidwood Stations."
The licensee's Quality Assurance Manual is the basis for the procedures.
5.3 conclusion The licensee's responses describe a system that meets the recomendations of this ites. Therefore, we find the licensee's responses for this item ac'ceptable.
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ITEM 2.2.1.3 USE OF THE EQUlpMENT CLAS$1FICATION LISTING l
6.1 Guideline l_
The licensee's description should confire that the program for
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equipment classification includes criteria and procedures that govern how 1
station personnel use the equipment classification information handling system to determine that an activity is safety related. The description should also include the procedures for maintenance, surveillance, parts i
replacement, and other activities applicable to safety related components.
as defined in the introduction to 10 CFR 50, Appendix 8.
6.2 Evaluation Administrative Procedure AP 1600 1, ' Initiating and Processing a Nuclear Work Request (NWR),' requires the operating engineer to determine the safety related classification of compo'nents and activities. This procedure al,so requires the verification of this classification'by various personnel, including work analysts, Quality Control, Quality Assurance, and In service Inspection technical staff. Ap 1600 1 is said to assure that i
safety related procedures, cautions, and constraints are utilized on activities that are designated safety-related.
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Quality Procedure t-53, ' Quality Assurance Program for Operations Classification of structures, Systems and components,' gives directions and instructions on the use of the Safety-Related Component List and assigns the responsibility of detemining the safety-related classification of l
components and activities to the operating engineer.
6.3 tonclusion We find that the licensee's description of plant administrative controls and procedures meets the requirements of this item. Therefore, we fin ( the licensee's responses for this item acceptable.
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ITEM t.2.1.4 - MANAGEMENT CONTROLS j
7.1 Ruideline 1
The licensee should briefly describe the management controls that are used to verify that the procedures for preparation, validation, and the routine use of the information handling system have been, and are being, followed.
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7.2 Evaluation The licensee describes the following procedures as providing the management controls intended.
Q.P. 3-3
' Classification of Systems, Components, Parts and Material' Q.P. 3 51
' Design Control for Operations - Plant Modification' Q.P. 4-51
" Procurement Document Control for Operations -
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L Processing Purchase Documents
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- Control and Maintenance of the Safety Related Component List (SRCL) for Byron and Braidwood Stations."
These procedures provide for audits and inspections whose results inform management of the status and performance of the equipment classification l
program.
PWR Engineering has the responsibility for the preparation and maintenance of the SRCL in accordance with procedure Q.12.4 for changes involving the addition, deletion, non-identical replacement, or relocation of components. These components can be safety related or non safety related but associated with safety related systems.
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e Thelicensee'sarchitectengineer,largentandLundy($&L),isstated to have the delegated authority to control, issue, and maintain the SRCL in accordance with S&L Project Instruction p!-ts.48. The NR Engineering manager is responsible for forwarding any non 5&L designed modifications to S&L to assure that those changes are incorporated into the SRCL.
7.3 Conclusian We find that the management controls used by the Itcensee assure that the infursation handling system is maintained, is current, and is used as intended. Therefers, we find the Itcensee's responses for this item acceptable.
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ITEM t.t.1.5 DESIGN VERIFICATION AND PRDCUREMENT 4.1 Guida11na The licensee's submittals should document that past usage demonstrates that the appropriate design verification and qualification testing are specified for the procurement of safety related components and parts.
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specification should include qualification testing for expected safety service conditions and provide support for the licensee's receipt of testing documentation to support the Ilmits of life recommended by the supplier.
If such documentation is not available, confinnation that the present program meets these requirements should be provided.
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The licensee states that the design criteria of the station Nuclear i
Engineering Department assure that design verification and qualification testing is specified for the safety related equipment and components procured. The licensee states that service conditions and the requirement to identify maintenance schedules to achieve the expected component or part itfe are specified.
8.3 Conclusion We conclude that the licensee has addressed the concerns of this ites.
Therefore, we find the licensee's responses for this ites acceptable.
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ITDI 2.2.1.6
'IMPORTANT TO SAFETV' COMPONENTS i
g.1 Guida11am Generic Letter 83 28 states that the licensee's equipment f
classification progree should include (in addition to the safety-related components) a broader class of components designated as 'leportant f.e j
Safety." However, since the generic letter does not require the licensee to furnish this information as part of their response, this ites will not be reviewed.
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- 10. CONCLU5!0N Sased en our review of the licensee's response to the specific requirements of itse 2.2.1, we find that the inferentien provided by the licensee to resolve these concerns meets the requirements of Generic Letter 83 28 and is acceptable.
Ites 2.2.1.8 was not reviewed as noted in Section 9.1.
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- 11. REFERENCES i
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Letter, NRC (O. 8. Eisenhut) to all Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Permits,
' Required Actions Based on Seneric !ap11 cations of $41em ATWS Events s
(8eneric Letter 83 28)," July 8, 1983.
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Letter, Consonwealth Edison (P. L. Barnes) to NRC (H. R. Denton),
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' Response to teneric Letter No. 83 28,' November 5, 1983.
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Letter, Commonwealth Edison (P. L. Barnes)bruary toNRC(H.R.Denton),
' Response to teneric Letter No. 83 28," Fe 29, 1984.
4.
Letter, Commonwealth Edison (R. A. Chrzarowski) 4.2.3 and 4.2 to NRC (T. E. Nurley),
' Generic Letter No. 83 28,' Items 2.2 (Part I),
April 21, 1989.
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