ML20206U885

Forwards Response to NRC 860409 Request for Addl Info & Clarifications of Util Concerning Auxiliary Feedwater Design Basis Analysis & Circumstances of Wiring Problem During Automatic Initiation Mods

ML20206U885
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	06/30/1986
From:	Opeka J, Sears C CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Charemagne Grimes Office of Nuclear Reactor Regulation
References
A05679, A5679, NUDOCS 8607110183
Download: ML20206U885 (9)

Text

• e CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N. CONNECTICUT P.o BOX 270 HARTFORD. CONNECTICUT 06141-0270 TELEPHONE l 03-665-5000 June 30,1986 Docket No. 50-213 A05679 a

Office of Nuclear Reactor Regulation Attn: Mr. Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20555

References:

(1) Fs M. Akstulewicz letter to 3. F. Opeka, Request for Additional Information Concerning Auxiliary Feedwater System Status, dated April 9,1986.

(2) W. G. Counsil letter to D. M. Crutchfield, Haddam Neck Plant Auxiliary Feedwater Systems, dated May 19, 1980.

(3) D. M. Crutchfield letter to W. G. Counsil, dated October 5, 1932.

(4) D. M. Crutchfield letter to W. G. Counsil, dated April 28, 1983.

(5) 3. F. Opeka letter to C. I. Grimes, dated March 31, 1986.

(6) 3. F. Opeka letter to 3. A. Zwolinski, dated September 20, 1985.

Gentlemen:

Haddam Neck Plant Response to Request for AdditionalInformation Concerning the Auxiliary Feedwater System Re ference (1) requested additional information and clarifications to our September 20,1985 letter (Reference (6)) which provided information on the auxiliary feedwater design basis analysis and circumstances surrounding a wiring problem that occurred to the auxiliary feedwater system when automatic initiation modifications were made. Enclosure 1 provides Connecticut Yankee Atomic Power Company's response to these requests.

8607110183 DR 860630 pl ADOCK 05000213 PDR [I l

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Should you have any further questions regarding this matter, feel free to contact us.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY.

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3. F. Opeka '

Senior Vice President a

By: C. F. Sears

Vice President 1

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l Docket No. 50-213 A05679 Enclosure 1 Additional Information Concerning Auxiliary Feedwater System June 30,1986 i

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I Enclosure 1 i

Question 1  ;

It is unclear from the licensee's September 20, 1985 letter what has been [

assumed as the design basis for auxiliary feedwater (AFW) flow following a i' system demand. Provide a discussion of this issue considering both the current plant configuration and the configuration used in the revised accident analyses  !

currently under way. This discussion should address minimum AFW flow required  ;

to meet all design basis accident decay heat removal requirements including the t assumption of a loss of off-site power concurrent with the most limiting single  ;

active failure.

Response ,

t The design basis requirements'for the AFW were provided in Reference (2). This f document discussed the AFW flow required to meet all design basis accident .

decay heat removal requirements. The loss of main feedwater event was '

identified as the limiting design basis event for AFW and as such defines the minimum AFW flow requirements. This was found acceptable as documented in  ;

the NRC's SER (Reference (3)). The current plant configuration is the  ;

configuration being used in the revised accident analyses currently under way. l These analyses address the adequacy of the minimum AFW flow for both four-and three-loop plant operation including the assumption of a loss of off-site power concurrent with the most limiting single active failure.

Question 2 .

Provide a discussion of the effects of any changes in your assumptions on AFW system safety function and on system reliability for a loss of main feedwater '

transient and loss of off-site power. This discussion should include justification for the assumed availability of the motor-driven AFW pump given that it is not  :

included in current plant technical specifications.  ;

Response

l The analyses performed (References (6) and (2)) did not take credit for the  !

motor-driven AFW pump. In addition the analyses did not take credit for the '

steam dump to condenser system. This results in the steam generators (SGs) relieving steam through the SG safety valves. Therefore, there would not be any difference in the AFW system assumptions for a loss of main feedwater transient regardless of the availability of off-site power. The analyses currently being ,

finalized also do not take credit for the motor-driven AFW pump or the steam dump to condenser system. l Question 3 Explain how the current AFW system automatic initiation capability satisfies the requirements of NUREG-0737, Item II.E.1.2, for a safety grade system given that  !

it relies on non-safety-related control air.

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Response

The AFW automatic initiation system does not rely on the control air system since the control valves fail in the safe direction. If control air were lost (e.g.,

compressors failing to maintain control air pressure or a seismic event rupturing a control air line causing rapid depressurization of the control air system) the AFW system would be automatically initiated. The reasons are as follows:

o The air operated valves required to start the AFW system include the steam admission valves (MS-PICV-1206A and B) to the AFW Terry turbines and the feedwater bypass valves (FW-FCV-1301-1, 2,3, and 4).

All of these valves fail open on a loss of control air.

o Automatic initiation of the above valves is dependent upon safety-related instrumentation and actuation logic signals which activate solenoid-operated valves in the control air system which are powered by safety-related DC power sources.(I) Loss of DC power results in fail-safe opening of steam admission valves and feedwater bypass valves.

One concern that was identified (2) during a review of selected plant design changes was that on loss of control air, resulting in rapid depressurization of the control air supplying the operators of the steam admission valves, overspeed of the steam-driven turbines may result. This could occur if the Terry turbine steam admission valves opened too rapidly such that the response time of the turbine govemor would be too slow to prevent turbine overspeed. In evaluating this situation, design modification alternatives have been identified which wculd control the rate of opening of the steam admission valves and therefore prevent turbine overspeed. A project is currently under way to determine if modifications are appropriate. Even if overspeed results, ample time is available to take manual action to restore the auxiliary feedwater pumps.

Question 4 In the discussion of four-loop operation, the licensee talks about random failures and two SGs lost. It is not clear that these assumptions are valid because the licensee would deliberately remove two SGs from service by tripping two reactor coolant pumps. In particular, the last sentence of this section beginning with

" detailed analyses of this scenario..." makes absolutely no sense. Provide additional justification for the assumptions used in the analyses for transients and accidents during four-loop operation.

(1) The D. M. Crutchfield letter to W. G. Counsil, dated October 5,1982, includes the Safety Evaluation Report which evaluated these instruments, actuation logic circuits and power sources and found them to be acceptable. We find that this conclusion is still valid.

(2) The 3. F. Opeka letter to 3. A. Zwolinski, dated September 20, 1985, discusses the status of this concern. With this letter an update is provided. l 2 of 6 L

Response

Reference (2) stated that one AFW pump flow of 318.2 gpm is sufficient to remove decay heat and pump heat 15 minutes after shutdown. Based upon a comparison with Table 1, it was concluded that only two SGs were required.

With the revised flow rates as shown in Table 2, it is seen that now 3 SGs are required. The statement in Reference (6) clarified this statement in that it was the result of a hand calculation and not detailed analyses performed. The RETRAN analysis assumed that all four generators are fed with the RC pumps-tripped in two generators. it did not assume that the two generators were isolated. Each of the two turbine-driven AFW pumps feed all four SGs. A single active failure can not result in AFW flow being delivered to only two SGs during four-loop operation. To have two SGs isolated requires failure of two feedwater control valves (bypass valves) in addition to the failure of one AFW pump.

Therefore, loss of main feedwater during four-loop operation with AFW flow to only two steam generators was not analyzed.

4 Question 5 Provide an evaluation of how much the AFW flow was reduced by a " design change on steam-driven pumps." Further the analysis takes credit for use of l PORVs in mitigating transients. Provide a description of how the PORV design satisfies all safety grade criteria such as emergency power, separation of power supplies, seismic, etc.

Response

Minimum expected AFW flow rates for the original design bases prior to the design change on the steam-driven AFW pumps were presented in Reference (2).

Attached is Table I which includes the data presented in Reference (2) and a comparable Table 2 which presents the reduced AFW flow rates utilized in Reference (6).

The limiting case for four loop operation in the design basis reanalysis utilizes 288 gpm from one AFW pump to four steam generators at 1000 psig. The limiting case for three loop operation utilizes 396 gpm from both AFW pumps to two steam generators at 1000 psig. A calculation is currently being verified that involves further refinements to the auxiliary feedwater flowrates. These results will be provided as a revised Table 2 on or before July 14, 1986.

The analyses discussed in Reference (6) looked at peak pressure with and without credit for the use of PORVs. The peak RCS pressure remains below the acceptance criterion of 110% of design pressure even without crediting PORVs.

Therefore, credit for the PORVs is not necessary and was not taken for

mitigation of the loss of feedwater event.

Question 6 In order to meet the diversity requirements of Branch Technical Position ASB RSB 10-1, the motor-driven AFW pump should be powered frorn an emergency bus. Provide an evaluation of how the current configuration of the AFW pumps j and associated power supplies satisfies the criteria stated above.

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Response

1 The Haddam Neck Plant AFW design is based on two 100 percent capacity 4

turbine-driven pumps and their associated valves, instrumentation, circuitry, and piping. This configuration does not meet all of the guidelines presented in Branch Technical Position ASB RSB 10-1, however, evaluations have been performed by both the NRC Staff and CYAPCO, which have concluded the-original design basis configuration bas a high degree of reliability and does not represent a significant - contributor to the overall core melt frequency.

Specifically:

o Reference (3) provided a safety evaluation of TMI Action Plan j ltem II.E.1.2. This concluded that the AFW system has an acceptable

means to automatically initiate and that the related flow indication systems are acceptable.

) o Reference (4) provided a safety evaluation of TMI Action Plan 4

Item II.E.1.1. ' This evaluation reviewed the case of a main steam line break with continued feedwater addition and in doing so evaluated the reliability of the AFW system. The AFW system was found to perform its function satisfactorily and a safety grade motor-driven pump was i

determined not to be required.

l o Reference (5) forwarded, among other evaluations, a probabilistic risk

assessment of the AFW system. This assessment did not credit the availability of the electrically-driven AFW pump (used primarily for start-up). The availability of the AFW system was found acceptable

except in the case of a main steam line break upstream of the non-return valves and outside containment. For this particular break, feed and bleed operations are credited for decay heat removal. CYAPCO will be providing documentation by July 13, 1986 which will describe under what circumstances feed and bleed operations are credited and discuss the need for increased surveillan.es and inspections of that portion of the main steam line where, if a break occurred, AFW unavailability may result. Overall, the failure of the AFW system was found to be an insignificant contributor to the core melt frequency.

In summary, both CYAPCO and the NRC have previously concluded that the

absence of a safety grade motor-driven auxiliary feedwater pump is acceptable.

In our view, nothing has occurred to invalidate that conclusion.

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4 Table 1 Minimum Expected AFW Flow Rates Prior to AFW System Design Changes I Total flow to all intact Condition steam generators (SGs) (GPM)

IPump 2 Pumps I ruptured SG and 3 intact SGs at 1,000 psig 0 0 4 intact SGs at 1,000 psig 350 545 3 intact SGs at 1,000 psig 340 515 with the other 1 isolated 2 intact SGs at 1,000 psig 320 450 with the other 2 isolated 1 intact SG at 1,000 psig 245 300 with the other 3 isolated 4

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Table 2 Minimum Expected AFW Flow Rates Current Flow Rate (Reduced)

Total flow to all intact Condition steam generators (SGs) (GPM) 1 Pump 2 Pumps I ruptured SG and 3 intact SGs at 1,000 psig 0 0 4 intact SGs at 1,000 psig 335 525 3 intact SGs at 1,000 psig 325 490 with the other 1 isolated 2 intact SGs at 1,000 psig 310 430 with the other 2 isolated I intact SG at 1,000 psig 235 290 with the other 3 isolated ,

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