B12190, Forwards Reviews for Isap Topic 1.02, High/Low Pressure Valve Interlocks, Isap Topic 1.07, Vital Bus Feed Realignment Mods, Isap Topic 1.11, Primary Auxiliary Bldg Ventilation Sys Mods & Isap Topic 1.19, Control Room..

Forwards Reviews for Isap Topic 1.02, High/Low Pressure Valve Interlocks, Isap Topic 1.07, Vital Bus Feed Realignment Mods, Isap Topic 1.11, Primary Auxiliary Bldg Ventilation Sys Mods & Isap Topic 1.19, Control Room..

ML20214W107
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	09/18/1986
From:	Opeka J, Sears C CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Charemagne Grimes Office of Nuclear Reactor Regulation
References
B12190, NUDOCS 8610020498
Download: ML20214W107 (12)

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CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N, CONNECTICUT P o. BOX 270 HARTFORD CONNECTICUT 06141-0270 TELEPHONE 203-665-5000 September 18,1986 Docket No. 50-213 B12190 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) 3. F. Opeka letter to C. I. Grimes, dated May 17,1985.

(2) H. L. Thompson letter to 3. F. Opeka, dated July 31,1985.

Gentlemen:

Haddam Neck Plant Integrated Safety Assessment Program In Reference (1), Connecticut . Yankee Atomic Power Company (CYAPCO) provided a proposed scope for the Integrated Safety Assessment Program (ISAP) review of the Haddam Neck Plant. In R'eference (2), the Staff formally issued the results of the ISAP screening review process, establishing the scope of ISAP for Haddam Neck and initiating issue-specific evaluations. Reference (1) also indicated that for each issue or topic included in ISAP, CYAPCO would provide a discussion of the safety objective and an evaluation of the plant design with respect to the issue being addressed to identify specific items to be considered in the integrated assessment. In accordance with this commitment, reviews for the following ISAP topics are attached:

1) ISAP Topic No.1.02 "High/ Low Pressure Valve Interlocks"

2) ISAP Topic No.1.07 " Vital Bus Feed Realignment Modifications"

3) ISAP Topic No.1.11 "PAB Ventilation System Modifications"

4) ISAP Topic No.1.19 " Control Room Design Review" If you have any questions concerning the attached reviews, please contact us.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY d.k.D L J. F. Opeka' Senior Vice President 8610020498 860918 PDR ADOCK 05000213 b

P PDR By: C. F. Sears Vice President a

Docket No. 50-213 B12190 Haddam Neck ISAP Topic No.1.02 High/ Low Pressure Valve Interlocks September 1986 J

F Haddam Neck ISAP Topic No.1.02 High/ Low Pressure Valve Interlocks I. Introduction Several systems that have a relatively low design pressure are connected to the reactor coolant pressure boundary. The valves that form the interface between the high and low pressure systems must have sufficient redundancy and interlocks to assure that the low pressure systems are not subjected to coolant pressures that exceed design limits. However, during certain system operating modes (e.g., shutdown cooling and ECCS injection) these valves must open to assure adequate reactor safety.

Haddam Neck has three systems directly connected to the Reactor Coolant System (RCS) that have a lower design pressure rating than the RCS; the Residual Heat Removal (RHR) system, Safety Injection (SI) system and Chemical and Volume Control System (CVCS).

High/ Low Pressure Valve Interlocks was the subject of SEP Topic V-II.A (Reference 1). The RHR system and CVCS have been mechanically modified to demonstrate compliance with current !! censing criteria for isolation of high and low pressure systems. Hene, this ISAP topic addresses only the SI systems. These systems are the high pressure safety injection (HPSI) and the low pressure safety injection (LP31) systems. Both HPSI and LPSI are isolated from RCS pressure by moter-operated valves, which automatically open on a safety injection signal, and by check valves.

Over-pressurization of these systems could potentia 3y result in a LOCA outside containment.

11. Review Criteria

1) SEP Topic V-II.A " Requirements for Isolation of High and Low Pressure Systems"

2) Standard Review Plan, Section 7.3

3) 10 CFR 50, Appendix R III. Related Topics / Interfaces ISAP Topic No.1.03 " Containment Penetration Evaluations" IV. Evaluation, The proposed modification would result in the addition of pressure interlocks to prevent opening of the MOVs and hence prevent potential over-pressurization of the HPSI and LPSI piping. In addition, the design would include a feature for automatic closure of these valves when primary pressure is above the systems' design pressures.

This proposed design modification may have competing effects on pub!!c risk. Although the overall frequency of a LOCA outside containment may decrease, the unavailability of HPSI and LPSI could increase.

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V. Conclusions CYAPCO is currently evaluating the need for pressure interlocks on the HPSI and LPSI system - piping. The expected safety impact and optimization of design alternatives will be further defined and evaluated in the integrated assessment.

' VI. References

1. " Integrated Plant Safety Assessment, System Evaluation Program (SEP), Haddam Neck Plant," NUREG-0826, March 1983.

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Docket No. 50-213 B12190 Haddam Neck ISAP Topic No.1.07 Vital Bus Feed Realignment Modifications September 1986 1

Haddam Neck ISAP Topic No.1.07 Vital Bus Feed Realignment Modifications I. Introduction The Vital AC Power System at Haddam Neck consists of four Vital AC each fed by an associated inverter.

buses,f one o the two 125V DC buses. Should the Inverter associated with a VitalEach inverter i AC bus be unavailable for any reason, that bus can be manually realigned to take power from one of the other Inverters. The current configuration at Haddam Neck is such that the alternate feed for each Vital AC bus is from an inverter in the opposite power division. The principal concern with this Vital AC bus feed alignment is that it does not comply with the single failure criterion.

II. Review Criteria

1. 10 CFR 50, Appendix K "ECCS Evaluation Models"

2. 10 CFR 50, Appendix R " Fire Protection Program for Nuclear Power Facilities"

3. SEP Topic No. VI-7.C.! " Independence of Redundant Onsite Power Sources" III. Related Topic / interfaces

1. ISAP Topic No.1.01 "Switchgear Room Cooling Modifications"

2. ISAP Topic No.1.14 " Appendix R Modifications" IV. Evaluation The proposed project is to modify the Vital AC bus alternate feeds so that the alternate supply for a vital bus is from the other inverter in the same power division. These design changes were proposed as part of the planned fire protection modifications in the switchgear room. This modification will prevent paralleling of redundant sources and would prohibit placing more than two inverters on the same battery, or more than two vital buses on the same inverter.

V. Conclusions It is not expected that this modification will significantly decrease risk to the health and safety of the public. However, further evaluation will be presented in the integrated assessment.

VI. References

1. D. M. Crutchfield letter to W. G. Counsil, "SEP Topic VI-7.C.1,

" Independence of Redundant Onsite Power Sources - Haddam Neck",

dated January 11,1982.

2. W. G. Counsil letter to D. M. Crutchfield, "SEP Topic VI-7.C.1,

" Independence of Redundant Onsite Power Sources - Haddam Neck",

dated December 2,1981.

3. C.1. Grimes letter to 3. F. Opeka, "Haddam Neck Plant - IPSAR Section 4.24.2, Onsite Standby DC Power Systems," dated January 13, 1986.

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Docket No. 50-213 B12190 Haddam Neck ISAP Topic No.1.11 PAB Ventilation System Modifications September 1986 J

Haddam Neck ISAP Topic No.1.11 PAB Ventilation System Modifications I. Introduction The Primary Auxiliary Building (PAB) contains the high and low pressure safety injection pumps, charging pumps, residual heat removal pumps, and the component cooling water pumps. These pumps would generate a significant heat load in this building during a postulated LOCA. If a loss of normal power (LNP) occurs concurrent with the LOCA, the PAB supply and exhaust fans would trip and not auto-start when the emergency diesel generators came up to speed. In the case of a large break LOCA resulting in substantial fuel damage, the radiation level inside the PAB might prohibit an operator from entering the bouding and restarting the ventilation system (which can only be restarted locally). Elevated temperatures within the enclosure would result and could have an adverse impact on critical electrical equipment such as the pump motors.

SEP Topic IX-5 discussed potential reviews of the design and operation of the ve.ntilation systems to assure their capability for providing a safe environment for plant personnel and for engineered safety features.

11. Review Criteria

1) SEP Topic IX-5 " Ventilation Systems" III. Related Topics / Interfaces ISAP Topic No.1.01 "Switchgear Room Cooling Modifications" IV. Evaluation This project analyzes the necessity of providing a remote operation station located outside the PAB to make it possible to restart the PAB fans and open the dampers, in case of'a large break LOCA coincident with a LNP scenario.

It is expected that the temperatures in the PAB may level off, possibly at temperatures well below that required to substantiate electrical equipment qualification.

V. Conclusions it is expected that the benefit to public safety from installation of a remote station for PAB ventilation operation is minimal. Further evaluation of this topic will be performed in the integrated assessment.

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Docket No. 50-213 B12190 Haddam Neck ISAP Topic No.1.19 Control Room Design Review September 1986

r Haddam Neck ISAP Topic No.1.19 Control Room Design Review I. ~ Introduction It has been recognized that the control rooms in many nuclear power plants contain significant human engineering discrepancies. These discrepancies might compromise the ability of the control room to provide safe ~and effective facilities during emergency operations and can impair the emergency response capabilities of the control room operators. Such human engineering discrepancies have been identified as the root cause behind:

o unintentional plant shutdowns and transients caused by operation of the wrong device by a control room operator, o unintentional disabling of decay heat removal and engineered safeguards systems due to operator errors while manipulating controls, and o premature termination of engineered safeguards systems due to -

cognitive errors arising from incorrect interpretation of control board instruments.

II. Review Criteria

1. Regulatory Guide 1.47 " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems"

2. NUREG-0737, Supplement 1, " Clarification of TMI Action Plan Requirements - Requirements for Emergency Response Capability"

3. NUREG-0801, " Evaluation Criteria for Detailed Control Room Design Review"

4. NUREG-0700," Guidelines for Control Room Design Review" III. Related Topics / Interfaces

1. ISAP Topic No.1.20 " Safety Parameter Display System"

2. ISAP Topic No.1.21 " Regulatory Guide 1.97 Instrumentation"

3. ISAP Topic No. 1.22 -

" Emergency Response Facilities Instrumentation" IV. Evaluation The proposed project involves a systematic review of the Haddam Neck control of NUREGsroom desig(n.

0737 The CRDR Supplement willand 1),0700 encompass the criteria 0801 for existing designand andguidelines all the contemplated (present and future) modifications. The outcome of the review will be the identification of recommendations for possible control

• room design changes. The evaluation of this project is based on the implementation of these recommendations.

As specified in the CRDR Implementation Plan, the review will consist of the following:

1. Estab!!shment of a qualified multidisciplinary review team.

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2. Performance of task analysis to identify control room operator tasks and information and control requirements during emergency operations.

3. A comparison of the information and control requirements with the control room inventory to identify discrepancies.

4. A control room survey to identify deviations from accepted human engineering guidelines.

5. Assessment of human engineering discrepancies (HEDs) to determine which HEDs are significant and should be corrected.

6. Selection of design improvements and establishment of implementation schedules.

7. Verification that selected design improvements will provide the necessary correction.

8. Verification that improvements will not introduce new HEDs.

9. Coordination of control room improvements with other programs such as Safety Parameter Display System (SPDS), operator training, Regulatory Guide 1.97 instrumentation, and upgraded emergency operating procedures.

Human engineering discrepancies in the control room can be characterized by the corresponding quality of the man / machine interface. This characterization can be based on engineering analysis and historical data.

Information obtained from the Haddam Neck PSS Indicated that the man / machine interface in the Haddam Neck control room is generally considered to be fair. In addition, a review of the Haddam Neck operating experience showed that there have been four reactor trips, all of which were caused by operator errors that could be linked to the same control room design deficiency. They were all caused by manipulating a reactor coolant pump (RCP) switch and shutting the RCP off, in place of another switch used during normal operations. This particular item has been corrected (the RCP switches were placed inside a plastic lid) since the last occurrence. This is a good example of how plant improvements that are typically made based on actual plant operating experience can improve the man / machine interf ace.

V. Conclusions Based on the above analysis, impicmentation of a new control room design may result in other similar benefits. Further analysis of this topic, including the associated resource burden, will be undertaken in the Integrated assessment.

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VI. References ,

1. F. M. Akstulewicz letter to 3. F. Opeka, " Review of the Connecticut Yankee Atomic Power Company Detailed Control Room Design Review Program Plan Submittal," dated May 19,1986.

2. NUREG-0737, Supplement 1, " Requirements for Emergency Response Capability," dated December 17,1982.

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