A05364, Forwards Addl Response to Generic Ltr 83-28,Item 1.2, Post- Trip Review:Data & Info Capabilities, Per 860415 Telcon. Status of Responses to Generic Ltr Also Encl.All Action Items Either Complete or Closed

Forwards Addl Response to Generic Ltr 83-28,Item 1.2, Post- Trip Review:Data & Info Capabilities, Per 860415 Telcon. Status of Responses to Generic Ltr Also Encl.All Action Items Either Complete or Closed

ML20198B666
Person / Time
Site:	Millstone
Issue date:	05/13/1986
From:	Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	Noonan V Office of Nuclear Reactor Regulation
References
A05364, A5364, B12063, GL-83-28, NUDOCS 8605220083
Download: ML20198B666 (9)

Text

v o 4 N UTILITIES v e cue,ncw w semes= cwm-o.n.,.i On,c.. . soie.n s1,.. . e.,iin. Conn cticut 1 .m m.c.es i.ic c *** P.O. BOX 270 o e-a .. ia m . cw-- H ARTFORD, CONNECTICUT 06141-0270 k k J C 3 E,7 ,$'d, (203) 665-5000 May 13,1986 Docket No. 50-423 B12063 A05364 Office of Nuclear Reactor Regulation Atto: Mr. V. S. Noonan, Director PWR Project Directorate #5 Division of PWR Licensing - A U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Reference:

(1) 3. A. Zwolinski, letter to 3. F. Opeka, Draft Technical Evaluation Report (TER) for ATWS item 1.2 (Generic Letter 83-28), dated November 6,1985.

(2) W. G. Counsil letter to D. G. Eisenhut, Response to Generic Letter 83-28, dated November 8,1983.

(3) T. M. Novak letter to 3. F. Opeka, Issuance of Facility Operating License NPF-49-Millstone Unit No. 3, dated January 31, 1986.

Dear Mr. Noonan:

Millstone Nuclear Power Station, Unit No. 3 Request for Additional Information - Generic Letter 83-28, item 1.2 Ir. Reference (1), the NRC provided a draft Technical Evaluation Report (TER) representing the Staff's initial judgement of our response (Reference 2) to Generic Letter 83-28, item 1.2 for Millstone Unit Nos.1,2 and 3. The Staf f also requested a telephone conference to review the open items outlined in the draft TER.

In an April 15, 1986 telephone call, representatives of Northeast Nuclear Energy Company (NNECO) provided the Staff with the additional information on the open items from the draft TER for Millstone Unit No. 3. Based on this additional information, the Staff concluded that NNECO's verbal responses adequately addressed the Staff's concerns. The Staff and NNECO agreed that upon receipt of NNECO's formal documentation of the additional information, the Staf f would publish a final Safety Evaluation Report (SER) satisfactorily closing out this issue for Millstone Unit No. 3. Accordingly, NNECO hereby submits the j additional information (Attachment 1) provided in the April 15, 1986 telephone I call for Millstone Unit No. 3. l l

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The Millstone Unit No. 3 Operating License (Reference 3) contains a license condition that requires NNECO to submit responses to and implement the requirements of Generic Letter 83-28 Salem ATWS Events on a schedule which is consistent with our previous NRC submittals. Attachment 11 provides the status of the Millstone Unit No. 3 responses to Generic Letter 83-28. All the action items covered in Generic Letter 83-28 are either complete or closed for Millstone Unit No. 3. Items identified as " Complete" are those for which responses have been provided and no confirmation of status has yet been received from the staf f. Items identified as " Closed" are based upon formal communication or agreements with the staff. Please inform us of any disagreement of status described in Attachment II.

If you have any questions, please contact our licensing representative directly.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY F. 6wA.*

3. F. O'peka Senior Vice President

Attachment I Millstone Unit No. 3 Additional Information on Generic Letter 83-28, Item 1.2 In the NRC's draf t Technical Evaluation Report (TER) of NNECO's response to Generic Letter 83-28, item 1.2 (Post-Trip Review: Data and Information Capabilities) for Millstone Unit No. 3, the Staff reported NNECO's response was unsatisfactory in the following area:

o Parameters Recorded o Time history recorders performance characteristics o Data Retention Capability In an April 15, 1986 telephone call, NNECO provided the Staff with additional information on these items. This information is documented below as part of NNECO's formal response to Generic Letter 83-28, item 1.2 Parameters Recorded In Table 2, the NRC provided a list of desirable PWR parameters for Post-Trip review with the circled parameters designating parameters the NRC required further information on. This information is provided below.

a) Containment Isolation - Containment Isolation signals and the status of Containment Isolation valves are on the sequence of events (SOE) recorder.

b) Control Rod Position - Control Rod Position prior to the trip is recorded on a strip chart recorder. Reactor trip breaker status, which indicates if control rods should be inserted is monitored on the SOE recorder.

c) The following Trip Bistable / Alarm Status is recorded on the SOE recorder:

1. Neutron Flux

2. Containment Pressure

3. RCS Pressure

4. RCS Temperature

5. Pressurizer Level

6. RCS Flow

7. Steam Generator Pressure

8. Steam Generator Level

9. Feedwater Flow (alarm only - this does not provide a trip)

10. Steam Flow (alarm only - this does not provide a trip)

11. AC and DC Bus Voltage d) Containment Pressure is recorded on the Analog Pre-Post Trip Report at the interval described under Time History Recorder Performance Characteristics.

e) Containment Sump Level is recorded on a Strip Chart Recorder.

f) Containment Radiation is recorded on a Strip Chart Recorder.

g) Safety Injection pump and valve status is recorded on the SOE recorder.

h) MSIV position is recorded on the SOE recorder.

i) Auxiliary Feedwater system pump and valve status is recorded on the SOE recorder.

j) Diesel Generator status is recorded on the SOE recorder.

k) PORV position is recorded on the SOE recorder.

Time History Recorders Performance Characteristics The plant process computer for Millstone Unit No. 3 has the capability to record parameter values at the following sample intervals.

a) For the period 15 minutes pre-trip to 1 minute pre-trip the sample interval will be 15 seconds.

b) For the period 1 minute pre-trip to 1 minute post-trip, the sample interval will be 5 seconds.

c) For the period 1 minute post-trip to 15 minutes post-trip, the sample interval will be 15 seconds.

Data Retention Capability Pre-Post Trip Analog printouts and Strip Chart Recorder information is retained for the life of the plant.

L

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% F Attachment 11 Millstone Unit No. 3 Status of Response to Generic Letter 83-28

< " Required Actions Based on Generic Implication of Salem ATWS Events" On July 8,1983, the NRC issued Generic Letter 83-28 to all licensees to request information related to reactor trip system reliability and general management capability. The actions covered in the letter dealt with: 1) post-trip review.

, 2) equipment classification, 3) post-maintenance testing, and 4) reactor trip system reliability improvements.

The following discussion provides the status \of the Millstone Unit No. 3 responses

.! to Generic Letter 83-28.

Action item 1.1 Post-Trip Review (program description and procedure)

Discussion

- All positions (1.1.1 - 1.1.7) are resolved. The NRC has issued a safety evaluation

i. report.

Status Closed.

References 1, 8, 9, 10, 11 Action Item 1.2 Post-Trip Review (data and information capability)

Discussion

(- '

Reference 14 provided draf t technical ( c/u- . n report (TER) representing the Staff's initial judgement of NNECO% rcspm,a (Reference 4). Reference 17 provided the staff with the additional information on the open items from the TER.

Status '

Complete.

W . - _ _ _ _ _ _ _ _ _ l

Attachment 11 References 1, 4, 14, 17 Action item 2.1 Equipment. Classification and Vendor Interface (reactor trip system components)

Discussion Reference 4 and 7 provided a description of the vendor interface programs for Millstone _ Unit Nos.1, 2, 3, and Haddam Neck Plant. No safety evaluation has been issued by the NRC.

Status Complete.

References 1,2,3,4,6,7 Action item 2.2 Equipment Classification and Vendor Interface (programs for all safety-related components)

Discussion References 4 and 7 provided a description of the vendor interface programs for Millstone Unit Nos.1, 2, 3, and Haddam Neck Plant. No safety

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~*

evaluation has been issued by the NRC.

Status Complete.

References 1,2,3,4,6,7 Action item 3-1 Post-Maintenance Testing (reactor trip system components)

Discussion All positions (3.1.1 - 3.1.3) are considered resolved pending issuance of a safety evaluation by the NRC.

Status Complete.

~

s Attachment II References 1, 4, 6, 7 Action item 3.2 Post-Maintenance Testing (all other safety-related components)

Discussion All positions (3.2.1 - 3.2.3) are considered resolved pending issuance of a safety evaluation by the NRC.

Status Complete.

References 1,' 4, 6, 7 Action Item 4.1 Reactor Trip System Reliability (vendor-related modifications)

Discussion The modification for Millstone Unit No. 3 reactor trip breakers (Madels DS-416) recommended by the Westinghouse in their letter dated April 21, 1983 has been implemented (Westinghouse field change Nos. NEUM-10563, NEUM-10564 and NEUM-10565).

Status Closed.

References 1, 4

- Action Item 4.2 Reactor Trip Reliability (preventive maintenance and surveillance program for reactor trip breakers)

Discussion Reference 7 provided response to positions 4.2.1 and 4.2.2. In Reference 4, NNECO committed to address breaker / trip attachment life cycle testing and replacement issue (positions 4.2.3 and 4.2.4) via participation in the WOG effort.

The WOG program has been completed and the report has been submitted to the NRC (Reference 12).

Attachment 11 '

Status-Complete.

References 1,4,5,12,13 Action Item 4.3 Reactor Trip System Reliability (automatic actuation of shunt trip attachment for Westinghouse and B&W plants)

Discussion Action item 4.3 is resolved. The NRC has issued a safety evaluation report.

Status Closed.

References 1,4,15,16 Action item 4.4 Reactor Trip System Reliability (improvements lo maintenance and test procedures for B&W plants)

Discussion Not applicable for Millstone Unit No. 3 Status Closed.

References 1, 4 Action Item 4.5

' Reactor Trip System Reliability (system functional testing)

Discussion Millstone Unit No. 3 Technical Specification Table 4.3-1 includes the required surveillance requirement. Action item 4.5-2 is not applicable to Millstone Unit No. 3 since on-line functional testing will be performed. Reference 7 provided a response to Action. Item '4.5.3. These items are considered resolved pending issuance of a safety evaluation by the NRC.

4

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Attachment II Status Complete.

References 1, 4, 6, 7 References

1. Generic Letter 83-28, dated July 8,1983.

2. W. G. Counsil letter to D. G. Eisenhut, dated September 6,1983.

3. D. M. Crutchfield letter to W. G. Counsil, dated October 26, 1983.

4. W. G. Counsil letter to D. G. Eisenhut, dated November 8,1983.

5. W. G. Counsil letter to D. G. Eisenhut, dated March 16, 1984.

6. B. 3. Youngblood letter to 3. F. Opeka, dated May 20, 1985.

7. 3. F. Opeka letter to B. 3. Youngblood, dated September 5,1985.

8. B. 3. Youngblood letter to 3. F. Opeka, dated September 10, 1985.

9. 3. F. Opeka letter to B. 3. Youngblood, dated October 8,1985.

10. 3. F. Opeka letter to B. 3. Youngblood, dated November 7,1985.

11. Supplement 4 to the Millstone Unit No. 3 SER, dated November,1985.

12. WOG letter to R. Wright of NRC, dated December 5,1985.

13. WOG letter to the NRC, dated July 16, 1985.

14. 3. A. Zwolinski letter to 3. F. Opeka, dated November 6,1985.

15. 3. F. Opeka letter to B. 3. Youngblood, dated May 14, 1985.

16. Supplement 2 to the Millstone Unit No. 3 SER, dated September,1985.

17, 3. F. Opeka letter to V. S. Noonan, dated May 13, 1986.

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