ML18045A452

Forwards Draft of SEP Review of NRC Safety Topic VI-7.B Associated W/Electrical,Instrumentation & Control Portion of Engineered Safety Feature Switchover from Injection to Recirculation Mode for Palisades Nuclear Power Plant.

ML18045A452
Person / Time
Site:	Palisades
Issue date:	07/28/1980
From:	Nishimura M EG&G, INC.
To:	Scholl R Office of Nuclear Reactor Regulation
References
TASK-06-07.B, TASK-6-7.B, TASK-RR ESD-6897, NUDOCS 8008060329
Download: ML18045A452 (9)

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Energy Measurements Group

• Ramon Operations 2801 OLD. CROW CANYON ROAD. SAN RAMON. CA *TEL. (415) 837-5381 *MAIL: BOX 204. SAN RAMON. CA 94583 28 July 1980 ESD# 6897 Mr. R. F. Scholl, Jr. US Nuclear Regulatory Commission Division of Operating Reactors SEP Branch 7920 Norfolk Avenue Bethesda, MD 20014

SUBJECT:

PALISADES NUCLEAR POWER PLANT (DOCKET NO. 50-255)

Ray: Attached is a rough draft .report on the Palisades plant detailing the NRC safety topic VI-7.B (ESF Swithoven From Injection to Recirculation Mode). Please review the report and forward your comments and recommendations

• to me. Sincerely, NISHIMURA GROUP LEADER MWN/ss cc: EG&G*> D. *Laudenbach B. Mayn . 8 0080 BO 3J 'f LLNL

• NRC M. Dittmore D. Allison St.

i* /' 1 :. ; " I

.. SYSTEMATIC EVALUATION PROGRAM REVIEW OF NRC SAFETY VI-7.B ASSOCIATED WITH THE ELECTRICAL, INSTRUMENTATION, AND CONTROL PORTIONi , OF THE ESF SWITCHOVER FROM INJECTION TO RECIRCULATION MODE FOR THE PALISADES NUCLEAR POWER PLANT By M. W. Nishimura*

July 1980 ROUGH DRAFT *EG&G, Inc., Energy Measurements Group, San Ramon Operations I '!'

. ... e* . TABLE OF CONTENTS 1. INTRODUCTION

• 2. SYSTEM DESCRIPTION

3. EVALUATION AND CONCLUSIONS.

REFERENCES.

APPENDIX A NRC SAFETY TOPICS RELATED TO THIS REPORT. 1 2 4 5 6 ii ' I . ' ... . : t ' I I I I i !

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.. r .l SYSTEMATIC EVALUATION PROGRAM REVIEW OF NRC SAFETY TOPIC VI-7. B ASSOCIATED WITH THE ELECTRICAL, INSTRUMENTATION, AND CONTROL PORTION OF THE ESF SWITCHOVER FROM INJECTION TO RECIRCULATION MODE FOR THE NUCLEAR POWER PLANT M. W. Nishimura EG&G, Inc., Energy Measurements Group San Ramon Operations

1. INTRODUCTION Most pressurized water reactors (PWRs) require operator action to realign the emergency core cooling system (ECCS) for the recirculation mode following a loss-of-coolant accident (LOCA). The NRC staff has been quiring (on a case-by-case basis) the use of some automatic features to realign the ECCS from the injectiori to the recirculation mode of operation.

The safety objective of this requirement is to increase the reliability of 1 ong-term core cooling by requiring no operator action to change system realignment to the recirculation mode. This report reviews the ECCS control system and operator action required to align the ECCS from injection mode to recirculation mode following a LOCA. -1 t , i* I ! I. i l ! I i *! r i !

2. SYSTEM DESCRIPTION The safety injection and refueling water (SIRW) tank low-level control system is designed to transfer the suction of the safety injection (SI) and containment spray pumps to the containment sump when the SIRW tank is essentially empty. During post accident cooling of the core, the system performs the functions necessary for recirculating and cooling water which has accumulated in the containment building In the recirculation mode, the system automatically provides component cooling water to the shell side of the shutdown cooling heat exchangers by opening reci rcul ati on valves CV3030 and CV3029 and at the same time closing injection valves CV3057 and CV3031. The circuit is designed on a two-channel concept with each channel i_nitiating the tion of separate and redundant hydraulic loops. The SIRW tank is provided with four level switches (LS0327 through LS0330) to detect a low. level in the tank. Each switch is ed to an auxiliary relay from separate preferred a-c supplies.

Consistent with the two-channel concept, a separate circuit is provided for control of the injection/recirculation valves (CV3029 and CV3057 or CV3030 and CV3031) that are associated with one injection/recirculation loop. Each circuit controls the operation of one of the two redundant component cooling water valves (CV0945 or CV0946) to* the shutdown heat exchangers via the component cooling water heat exchangers, as well as the service water valve (CV0823 or CV0826) to provide service water from one of the component cooling water heat exchangers.

The low-level control circuits have no normal or shutdown cooling operating functions, and ate only after the SIRW tank has been emptied. f'\' l' r 1: ! l: /!

1 * .1 Coincident two-*out-of-four low-level signals initiate the culation actuation signal (RAS), which opens the containment sump valves (CV3029 and CV3030), closes _the SIRW tank valves (CV3031 and CV3057), stops the low-pressure pumps, and closes the valves in the pump minimum-flow

,,. lines. A manual bypass is provided s; that the low-pressure*

injection pumps may be restarted if the operator deems this necessary for 1 ong-term core co*o 1 i ng. The system has redundant 1 ow-1 evel control circuits, each of which controls*

a redundant recirculation loop and* the cooling system valves. Each of the redundant control circuits is supplied from a separate preferred a-c source.* Failure of the power source in any one of the level switch . circuits will cause the circuit to fail in a mode that i ni ti ates recirculation.

The control circuit may be tested while the plant is in tion. This test will initiate the operation of the valves and the trip signal of the 1ow pressure (LP) injection pump. The test may be initiated by the test switches provided in the control room or by actuating the level switches mounted at the SIRW tank. Operation of one of the two redundant test switches on the control panel wi 11 deenergi ze t1'lo 1 evel switch auxi 1-i ary relay circuits and provide a two-out-of-four low-level signal which will initiate operation of the valves. Releasing the test switch will conclude the test, and valve operators will return to the normal positions.

In addition, individual valve operation may be tested manually using the control switches.

• Failure in any one a-c source of the four switch circuits will cause that channel to be in a tripped state, thereby changing the 1 ogi c from two-out-of-four to one-out-of-three logic. i 1'r

\.;.. 3. EVALUATION AND CONCLUSIONS The injection and recirculation paths have two emergency

' water loops that receive water from the SIRW tank during the injection mode and from the containment sump during the recirculation mode. The d-c power supply to the control logjc relays {as shown in Palisades drawing E-246 [Ref. l]) have two separate d-c power supply sources for each of the two cooling 1 oops. The two d-c power supplies .come from d-c panels Dll and 021. If one d-c power supply fails. durfog the injection phase, the cooling loop for which the power supply has failed will hot complete the automatic over to the recirculation mode. The injection/recirculation pump in the failed loop wi 11 continue to operate after the SIRW tank water has been depleted.

Continued operation may result in pump failure because when d-c power to the control logic relays is lost, the containment sump valve will fail in the closed position and the SIRW tank outlet valve will fail in the open position. A 1 though damage to one emergency cooling path may occur, single failure criterion is not jeopardized.

The second redundant cooling path (independent of the first cooling path) wil 1 provide adequate emergency cooling water to the core *. ,, I 1 " ! i I i i: i' I' I

• 1...9', .. REFERENCES

1. Consumer Power Company drawing numbers M-204 (Piping and Instrumented Diagram-Safety Injection, Containment Spray and Shutdown Cooling Sy!?tem);

E-246 (Schematic Diagram SIRvl Tank and Containment Sump Valves); E-207 (Schematic Diagram Containment High Pressure, High Radiation, and SIRW Tank Low Level). 2. Consumer Power Company, Palisades Final Safety Analysis Report (filmed June 1978). i* /.

APPENDIX A NRC SAFETY TOPICS RELATED TO THIS REPORT 1. Safety Topic VI-3 2. Safety Topic Vl-4 3. Safety Topic VI-7 4. Safety Topic VI-7.C 5. Safety Topic VI-9 6. Safety Topic Vl-10 Cathy #4/#10/CEB/amr 11 Containment Pressure and Heat Removal Capabi 1ity.11 11 Containment

!Solation System.11 11 Emergency Core Cooling System. 11 11 ECCS Single Failure Criterion and Requirements for Locking Out Power to Valves Including Independence of Interlocks on ECCS Valves.11 11 Main Steam Isolation.

11 11 Selected ESF Aspects.11 I 11 .. -..