ML20205N948
| ML20205N948 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Haddam Neck, 05000000 |
| Issue date: | 01/31/1988 |
| From: | Udy A IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20205N941 | List: |
| References | |
| CON-FIN-D-6001 EGG-NTA-7428, GL-83-28, TAC-53690, NUDOCS 8811070120 | |
| Download: ML20205N948 (18) | |
Text
c-.
,'T' EGG-NTA-7428 TECHNICAL EVALUATION REPORT CONFORMANCE TO GENERIC LETTER 83-28. ITEM 2.2.1--
EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:
MILLSTONE-2 Occket No. 50-336 f
Alan C. Udy Published January 1988 P
Idaho National Engineering Laboratory EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
20555 t
Under DOE Contract No. DE-AC07-76!D01570 FIN No. 06001
,,, - - ~ -
~,
l unoac w&
Qe
ABSTRACT This EG&G Idaho, Inc., report provides a review of the submittals from Unit No. 2 of the Millstone Nuclear Power Station for conformance to Generic Letter 83-28, Item 2.2.1.
I l
l-l i
i l
Docket No. 50-336 TAC No. 53690 11
~
FOREWORD 8
This report is supplied as part of the program for evaluating licensee / applicant conformance to Generic Letter 83-28 "Required Actionh Based on Generic Implications of Salem ATWS Events." This work is being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Divt..on of Engineering and System Technology, by EG&G Idaho, Inc., Electrical, Instrumentation and Control Systems Evaluation Unit.
The U.S. Nuclear Regulatory Commission funded this work under the authorization B&R 20-19-10-11-3, FIN No. 06001.
1 l
l t
i i
i Docket No. 50-336 TAC Wo. 53690 r
111
CONTENTS ABSTRACT..............................................................
11 FOREWORD..............................................................
iii 1.
INTRODUCTION.....................................................
1 2.
REVIEW CONTENT AND FORMAT........................................
2 3.
I T EM 2. 2.1 - P R OG RAM.............................................
3 3.1 Guideline..................................................
3 3.2 Evaluation.................................................
3 3.3 Conclus?on.................................................
4 4
ITEM 2.2.1.1 - IDENTIFICATION CRITERIA...........................
5 i
4.1 Guideline..................................................
5 4.2 Evaluation.................................................
5
[
4.3 Conclusion.................................................
6 5.
ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM.......................
7 5.1 Guideline..................................................
7 5.2 Evaluation.................................................
7 5.3 Conclusion.................................................
7 6.
ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING...........
8 6.1 Guideline...
8 6.2 Evaluation.................................................
8 6.3 Conclusion.................................................
8 7.
IT EM 2. 2.1. 4 - MAN AGEMENT CONTROLS...............................
9 7.1 Guideline..................................................
9 7.2 Evaluation................................................
9 l
7.3 Conclusion.................................................
9 l
E.
ITEM 2.2.1.5 - DESIGN VERIFICATION AND PROCUREMENT...............
10 i
r a
1 8.1 Guideline..................................................
10 4
]
8.2 Evaluation.................................................
10 i
i.
8.3 Conclusion.................................................
10 r
I' 9.
ITEM 2.2.1.6
"!MPORTANT TO SAFETY" COMPONENTS..................
11 i
9.1 Guideline..................................................
11
- 10. CONCLUSION.......................................................
12 11.
REFERENCES.......................................................
13
[
l f
I iv i
l I
t i
CONFORMANCE TO GENERIC LETTER 83-28, ITEM 2.2.1--
EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED CCMp0NENTS:
MILLSTONE-2 1.
INTRODUCTION On February 25. 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident was terminated manually by the operator about 30 seconds af te,' the initiation of the 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 incicent, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated based on steam generator low-low level during plant startup.
In this gase, the reactor was trippec manually oy the operator almost coincidentally with the I
automatic trip.
Following these incidents, on February 28, 1933, the NRC Executive Director for Operations (EDO), directed the NRC staff to investigate and I
report on the generic implications of these occurrences at Unit 1 of the i
Salem Nuclear Power Plart.
The results of tLs staff's inquiry into the generic implications of the Salem incidents are recor.ed in NUREG-1000, "Generic Implications of the ATVS Events at the Salem Nuclear Power i
}
Plant." As a result of this investigation, the Commission (NRC) requested I
(by Generic Letter 83-28 dated July 8, 1983 ) all licensees of operating 4
reactors, applicants for an operating license, and holders of construction permits to respond to the generic issues raised by the analyses of these two ATWS events.
I This report is an evaluation of the responses submitted by the Northeast Utilities Company, the licensee for Unit 2 of sne Millstone Nuclear Powet-Station, for Item 2.2.1 of Generic Letter 83-28. The documents reviewed as a part of this evaluation are listed in the references at the end of this recort, j
I l
l
2.
REVIEW CONTENT AND FORMAT t
Item 2.2.1 of Generic Letter 83-28 requests the licensee / applicant t
submit, for staff review, a description of their programs for safety-related equipment classification including supporting information, in considerable detail, as indicated in the guideline section for each item within this report.
L As previously stated, each of the six items of Item 2.2.1 is evaluated in a separate section in which the guideline is presented; an evaluation of 4
+
the licensee's/ applicant's response is made; and conclusions about the licensee's nr applicant's program for safety-related equipment classification are drawn.
3 e
2
7_________--__________________________
3.
Item 2.2.1 - PROGRAM 3.1 Guideline l
Licensees and applicants should confirm that an equipment classification program exists that provides assurance that all safety-related components are designated as safety-related on plant documentation and in the information handling system that controls safety-related activities.
The purpose of this program is to ensure that personnel performing activities that aff6ct safety-related components are-aware that they are working on safety-related components and are guided by safety-related procedures and constraints.
Features of this program are f
evaluated in the remainder of this report.
3.2 Evaluation The licensee for Unit 2 of the Millstone Nuclear Power Station responded to these requirements with submittals dated November 8, 1983 and March 13, 1987.3 These submittals include information that describes the Millstene-2 safety-related equipment classification program.
In the review of the licensee's rssponse to this item, it was assumed that the information and documentation supporting this program is available for audit upon recuest.
The licensee's computer-based Production Maintenance Management System (pMMS) and the Material. Equi: ment and Parts List (MEPL) make up the equipment classification program.
Both the PMMS and the MEPL use Nuclear Engineering and Operations Procedures (6.10 and 6.01) for control.
- Thus, the Nuclear Engineering and Operation's Department has the responsibility for both the PMMS and the MEPL.
3 m
+
Work orders are generated.by the PM S and include a designation for safety-related work activities. The plant procedures for surveillance testing, administrative control procedures, drawings, and purchase documents also designate safety-related equipment.
3.3 Conclusittn We have reviewed the licensee's submittals and find that the licensee's response is adequate.
l l
l l
I l
i l
l 1
l l
l l
l 4
4.
ITEM 2.2.1.1 - IDENTIFICATION CRITERIA 4.1 Guideline The applicant or licensee should confirm that the program used for equipment classification includes criterid used for identifying components as safety-related.
l 4.2 Evaluation i
r i
The licensee's response orovided the criteria for the original and ongoing identificatiSn of systems, structures, and comoonents as safety-related.
The criteria consist of a listing of safety-related system',, structures, and comoonents, and includes foundations and supports.
The listing consists of 19 items, most of which contain suo-items. 7ne licensee states thst these criteria are currently used to identify safety-related components in accordance with cuality assurance procedures.
Generic Letter 83-28 identifies as safety-related those structures, systems, and comoonents that assure (following a design basis event)
I (1) the integrity of tne reactor coolant boundary. (2) the capability to shut down the reactor and to maintain it in a safe shutdown condition, and (3) the capability to prevent or to mitigate consequential offsite exposures.
i i
The licensee defines (but the definition is not limited to) as safety-related those systems, structures, and components, including associated foundations, supports, and auxiliary systems, that:
(1) are a portion of the reactor coolant pressure boundary: (2) are used for emergency core cooling, reactor shutdown, residual heat removal and cooling water systems which support the previous systems; and (3) reactivity control systems, primary to secondary containment, portions of radioactive i
waste control systems and systems that might contain radioactive materials and whose postulated failure could result in offsite exposures.
l 5
[
l
tY 4.3 Conclusion L
We find that the criteria used to identify safety-related components encompasses the definition for safety-related that is part of the generic letter. Therefore the identification criteria meets the requirements of f
Item 2.2.1.1.
i l
l f
(
i i
e I
e t
k 6
-,, - - - -I
l S.
ITEM 2.2.1.2 - INFORMATION HANDLING SYSTEM l
5.1 Guideline I
The licensee or applicant should confirm that the program for equipment classification incledes an information handling system that is used to identify safety-related components. The response should confirtn that this information handling system includes a list of safety-related t
equipment and that procedures exist which govern its development and validation.
5.2 Evaluation The Itcensee's submittals identify the hord-copy MEPL as the information handling system that lists safety-related structures, systems, ccmperents, and parts. The PMS is a computerized data base that will I
eventually replace the MEPL. Currently the two systems co-exist, and the MEDL is the governing document.
The iicensee's cescription of the systems included the methods used for the develcoment of these systems, the process by which new safety-related items are entered, how changes in thte classification of listed items are mace, how listed items are verified, how unauthorizec cnanges to tne listing are prevented, and how the listing is maintainec and distributed to online comouter terminal users.
Revisions to both the MEPL and the PMMS are controlled by the MEPL engineer.
5.3 Cenelusion f
We find that the information contained in the license 0's submitt'41s is l
sufficient for us to conclud;t that the licenste's information handling system fo.* equipment classification meets the guideline requirements.
Therefore, the information provided by the licensee for this item is acceptable.
7
~
6.
ITEM 2.2.1.3 - USE OF EQUIPMENT CLASSIFICATION LISTING 6.1 Guideline The licensee's or applicant's description should confirm that the program for equipment classification includes criteria and procedures that govern how station personnel use the equipment classification information handling system to determine that an activity is safety-rslated.
The description should also include the procedures for maintunance, surveillance, parts replacement. and other activities defined in the introduction to 10 CFR 50, Appendix B, that apply to safety-related comconents.
6.2 Evaluation l
The licensee states that either the H *w or the PMMS is consulted to l
determine the safety-related status of the above work activities. Work orders control H.ase activities. These work orcers are generated by the PMMS.
The licensee has also provided a listing of Administrative Control I
Procedures that address these activities. The Work 0*das.viignates a scocific activity as safety-related or not and designates the procedures t'o be used in the above activities.
i 6.3 Conclusion We find that the licensee'? dsseription of plant administrative
[
controls and procedures meets the requirements of this item and is, f
therefore, acceptable, i
i b
6 i
i F
0 i
f
7.
ITEM 2.2.1.4 - MANAGEMENT CONTROLS 7.1 Guideline The applicant or licensee should briefly describe the management controls that are used to verify that the procedures for preparation, validation, and *outine utilization of the information. handling system have
(
been and are being followed.
7.2 Evaluation The licensee's response states that their Quality Assurance Program Tc;1 cal Report serves as the method of managerial control. QAP 18.0 (audits), as imolemented by NQA 1.14, is used to verify the preparation, validation, and routine use of the information handling system.
Quality assurance reviews and audits occur on a scheduled easis and assure that the programs and their implementation are correct.
I 7.3 Conclusion We find that the management controls used by the licensee assure that the information handling system is maintained, is current, and is used as intended.
Therefore, tne licensee's response for this item is acceptable.
t I
i l
l t
i
[
9 i
I
,,c_
.n_.
-,---wne,,
, -,, =
8.
ITEM 2.2.1.5 - DESIGN VERIFICATION AND PROCUREMENT 8.1 Guideline The applicant's or licensee's submittal should document that past osage demonstrates that appropriate design verification and qualification testing are specified for the procurement of safety-related components and parts. The specifications should include qualification testing for expected safety service conditions and should provide support for the applicant's/ licensee's receipt of testing documentation to support the limits of life recommended by the supplier.
If such documentation is not available, confirmation that the present program meets these requirements should be provided.
8.2 Evaluation Nuclear Engineering and Operations (NEO) Procedure 5.04 is tne centrolling procecure fer procurement.
NEO Procedure 6.02 outlines the appropriate documentation reouirements.
The originator of the purchase requistion is responsible for specifying the technical, testing, and cccumentation recuirements of the purchase.
Responsibility for ensuring that qualification testing and testing documentation is specified are also I
delineated in Procecure 6.02.
l 8.3 Cene1usion The li:ensee's response for this item is considered to be complete.
The information provided addresses the concerns of this item and is acceptable.
10
9.
ITEM 2.2.1.6 "IMPORTANT TO SAFETY" COMPONENTS 9.1 Guideline Generic Letter 83-28 states that the licersee's equipment classification program should include (in addition to the safety-related components) a broader class of components designated as "Important to
(
Safety." However, since the generic letter do:t not recuire the licensee to furnish this information as patt of theio response, this item will not t.e reviewed.
s i
1 l
,i i
h 11
.y.
l r
- 10. CONCLUSION
(
Based on'our review of the licensee's response to the specific requirements of Item 2.2.1, we find that the information provided by the licensee to resolve the concerns of Items 2.2.1.1, 2.2.1.2, 2.2.1.3, l
2.2.1.4, and 2.2.1.5 meets the requirements of Generic Letter 83-28 and is i
acceptable.
Item 2.2.1.6 was not reviewed, as noted in Section 9.1.
.i t
[
h l
L t
b f
l i
(
r h
i 6
l 1
t 12
11.
REFERENCES 1.
Letter, NRC (O. G. Eisenhut) to all Licensees of Operating Reactors, Applicants for Operating License, and Holders of Construction Permits, "Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8,1983.
2.
Lette*, Northeast Utilities Company (W. G. Counsil) to NRC (O. G. Eisenhut), NRC, "Resconse to Generic Letter 83-28 Generic Implications of Salem ATWS Events," November 6, 1983, A03381, Attachments 3 and 5.
3.
Letter, Northeast Utilities Company (E. J. Mroczka) to NRC, "Generic Ltster 83-28, Item 2.2
'Eauipment Classification' "
l March 13, 1987, A06176, A06384, A06385, B12375.
l l
[
l I
e 13
i
.. ~...,.
7
""S
BlSUOGRAPHIC DATA SHEET EG3-NTA-7428
)
i
= i.w,.ve,
- o.,..
...n
,,,,....u.....
CONFORFANCE TO GENERIC LETTER 83-28. ITEM 2.2.1--
EQUIPMENT CLASSIFICATION FOR ALL OTHER SAFETY-RELATED COMPONENTS:
MILLSTONE-2
- a"*"'<**'a January 1988 Alan C. Udy
,...ri...o...
January 1988 u.,4.
..,- i. 4
..=m,.c,,,.m w i EG&G Idaho. Inc.
P. O. Box 1625 tra.,,=
,=a===
Idaho Falls. 10 83415-2409 06001
......4..,...................,,...
Division of Engineering and System Technology Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission
.ho. e..o.-
WashD.gton. DC 20555
- i.. p s.. eat....ctag
, a
? t. 41.J yog. y e This EG&G Idaho. Inc.. report provides a review of the submittals from Unit No. 2 of the Millstone Nuclear Power Station regarding conformance to Generic Letter 83-28 Item 2.2.1.
i
..u,..,........
.....w......
Unlimited Distribution
,. n c...,. c......u,.
....... o... e,.
p., - 1.,, 4 < i. d s...,
Uncitssified
...si n g...m h
i