ML17258A988

SEP Review of NRC Safety Topic V-11.B Associated W/Electrical,Instrumentation & Control Portions of RHR Sys for Ginna Nuclear Power Plant.

ML17258A988
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Site:	Ginna
Issue date:	04/21/1981
From:	EG&G, INC.
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ML17258A987	List: ML17258A988
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EGG-1183-4154, NUDOCS 8104290492
Download: ML17258A988 (17)

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EGG 1183-4154 21 April 198l Energy Measurements Group SYSTKMAYIC EYALUAi/ON PRQGPAM REY)FW OF NRC SWFETV ToeIC V->>.a eSSOCI@YED WITH rHS SL<CTaICat.,

Ir.'STRUiYaKNTATIQN, AND COP~TRQL PQRTIOF~S QF THE RESIDUAL HEAT- Ri~h"<OVAL S'f STEM FOR THE GINhfA P4UCLEAR FOYER'F.R PLANT I l I ~

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8 10.4 2'90 'V~~ SAN RAMON OPERATIONS BSC1 QI Q CROW CANYCN RQAQ SAN RAMQN, CAI IFQRNIA 94583

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EGG ll83-4154 SYSTEMATIC EYALQATION PROGRAM REVIEW OF NRC SAFETY TOPIC V-ll.B ASSOCIATED V/ITH THE ELECTRICAL, INSTRUMENTATIONIAND CONTROL PORTIONS OF THE RESIDUAL HEAT REMOVAL SYSTEM FOR THE GINNA NUCLEAR POVfER PLANT

ABSTRACT This report documen.s the technical evaluation'nd review of NRC safety topic V-. 11.B, associated with the electrical, instrumentation, and control portions of the residual heat removal (RHR) system for the Ginna nuclear power plant. Current licensing criteria are used to evaluate..the overpr essure protection and independence of the RHR system.

111

FOREMGRD This report is supplied as part of the Systematic Evaluation Program being conducted for the U.S. Nuclear Regul atory Cori ssi on by Lawrence Livermore National Laboratory. The work was performed by, EGEG, Energy Measurements Group, San Ramon Opera ti ons for Lawrence Li vermore National Labor atory under U.S. Department of Energy contract number DE-AC08-76NV01183.

TABLE OF CONTENTS Page

1. INTRODUCTION......... ~ ~ ~ 1 CRITERIA....... '..

2.

3.

4.

CURRENT REYIEM SYSTEM

! ICENSING GUiDE'NES;. -..;.......

DESCRIPTION...........

3 5

7

5. EVALUATION AHD CONCLUSIONS.......... 9

6. SUM'CRY. . . . . , . . . . , . . . 11 REFERENCES . . . . . . . . . . . . . . 13 APPENDIX A HRC SAFETY TOPICS RELATED TO THiS REPORT . . . A-1

SYSTEMATIC EVALUATION PROGRAM REVIEW OF NRC SAFETY TOPIC V-11.B ASSOCIATED WITH THE ELECTRICAL, INSTRUMENTATION, AND CONTROL PORTIONS .

OF THE RES'IDUAL HEAT REMOVAL SYSTEM FOR THE GINNA NUCLEAR POWER PLANT

1. INTRODUCTION A number of plants have residual heat removal (PHR) systems in which the design pressuie:.".rating is lower than the reactor coolant system (RCS) pressure boundary to which the system is connected. The RHR system normally is located outside of primary containment and has motor-operated valves (MOVs),which'solate it from the RCS. There is, therefore, a po-tential that these systems would be subjected to pressure stresses in excess of their design -rating if the isolation MOVs were opened inadvert-ently'while the RCS was above the RHR system design pressure rating. This could result in a LOCA outside containment and a loss of reflood capability since the coolant inventory could be lost. Generally, interlocks are provided to prevent isolation MOVs from opening under high RCS pressure conditions .

It is important to incorporate features into the system design which will prevent overpressurizing the low pressure-rated RHR systems which interface with the reactor coolant pr ssure boundary. The current licensing criteria requires redundant, diverse interlocks to prevent opening of the isolation MOVs when RCS pressure exceeds PAR pressure design

limits. The current licensing criteria also requires automatic closure of the isolation MOYs when RCS exceeds RHR pressure design limits.

r The objective of this review is to ensure that the plant has adequate measures to protect a low pressure-rated RHR system that inter-faces with the RCS from failures due to excessive pressure and that such protection is suitably redundant and diverse.

This review applies to the interlocks associated with the isola-tion MOYs of the RHR system. O.her protection schemes such as double-testable check valves are discussed in reports on other NRC Safety Topics.'2-

2 CURRENT LICENSING CRITERIA Branch Technical Position ICSB-3 [Ref. 13, entitled "Isolation of Low Pressure Systems from the High Pressure RCS," states that:

The isolation NOYs should have independent and diverse rl ocks to prevent opening unl ess the primary system 'nt pressure is below the subsystem design pressure. 'lso, the isolation NOV operators should receive a signal to close the valves automatically when the primary system pressure ex-ceeds the subsystem design pressure.

Branch Technical Posi ion RSB 5-1 [Ref. 2], entitled "Design Re-quirements for the Residual Heat Removal System," states that:

Isolation shall be provided by at leas. two power-operated valves .in series, and the valves snail have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR system design pres-sure. The valves shall have independent, diverse interlocks to protect against one or both valves being open during an increase above RHR system design 'pressure. If the RHR system discharge line is used for an emergency core cooling system (ECCS) function, the power-operated valv is to be opened upon receipt of a safety injection signal once the reactor coolant pressure has decreased below the ECCS design pressure.

3. REYIEM GU:OELINES The NRC guidelines used in this revie~ are as follows:

( 1) Identify the valves which isolate the RHR system from the reactor coolant pressure boundry; (Refer to NRC memorandum from 8. L. Siegel, RSB, to P.

A. Di Benedetto, SEP; which is enclosure 3 of a letter from Crutchfield NRC, SEPB, to Oittmore, LLNL, dated 6'-10-80 [Ref. 3]).

Eval uate the design features which provide protec-tion against ihe overpressurization cf the RHR system.

(3) Identify the related topic reviews in an appendix to this report.

(4) Compile a list of the major EI8C systems that are necessary for DBE and for safe shutdown of the plant. Submit the compilation of necessary items for safe shutdown as an appendix to NRC Safety Topic YII-3, entitled'Systems Required for Safe Shutdown."

(5) If power is locked-out to the RHR isolation l10Ys, review to determine if any functions of the inter-locks or permissives are adversely affected. (The report on NRC Safety Topic YI-7.C, among others, states which values have power locked out).

4. SYSTEM OESCRIPTION The RHR loop consists of two pumps, two heat exchangers, and the necessary val ves, piping, and instrumentation. Ouring plant cool down, coolant flows frcm the RCS to the RHR pumps, through the tube side of the RHR hea exchangers and back to the RCS. The single inlet line to the RHR loop comnences at the hot leg of reactor coolant loop A, through two re-dundant pumps 'nd their associated heat exchangers, and back to I the cold leg of reactor coolant loop 8 via a single header.

The RHR pumps and hea. exchangers serve dual functions. Al.hough the normal duty of the RHR pumps and heat exchangers is performed during periods of reac.or shutdown, this equipment is aligned during the injection phase after a loss-of-coolant-accident (LOCA) to perform the 1 ow-head safe y injection (LPSI) function. In addition, during . the recirculation phase of a LOCA the capability may be divided between the. core-cooling function and the containment-cooling function as a part of the containment spray system.

5. EVALUATION AND CONCLUSIONS The suction line of the RHR system is isolated from the loop A hot leg of the RCS by MOV-700 and MOV-701 in series. The discharge line of the RHR system is isolated from the loop 8 cold leg of the RCS by MOV-720 and MOV-721 in series. [Ref. 4, drawing 33013-436-A].

All permissive interlocks associated with the RHR sys..em isola-tion MOVs are designed to open the valves; there are no permissive inter-locks associated with isolation MOY closure.

Section 4.1 of- the SEP review of Safe Shutdown Systems [Ref; 5i states that the permissive intei locks required o open .he four RHR sys.em isolation valves are as listed below:

MOV-700....RCS pressure must be less than 410 psig.

RHR suction valves MOV-850A and MOY-8508 fran the containment sump must be closed.

MOY-701....The valve .is operated by a key switch.

RHR suc ti on val ves MOV-850A and MOY-8508 from the containment sump must be closed.

MOV-720....No interlocks exist; valve operated by key switch. 1 MOV-721....RCS pressure must be less than 410 psig.

The RHR system discharge line is not used for an ECCS function that would require MOY-720 or MOV-721 .o open; however, a branch of the RHR discharge line provides low pressure safety injecto'n (LPSI) to the reactor vessel via parallel lines . Isolation between the RHR system and LPSI injection into the reactor vessel is provided by two separate paths from the RHR discharge line, ~th each path containing an MOY and check valve, MOV-852A and check valve 853A provide isolation in one path, while MOY-852B and check valve 8538 provide isolation in the other path [Ref. 4, drawing 33013-436-A; Ref. 6, drawing 33013-432-A]. The LPSI isolation MOVs open on a SI signal regardless of RCS pressure; .here are no interlocks associated

~ith closure of the'LPS'I isolation MOVs, although key switch closure cap-ability is provided.

Section -".1 of the SEP review of Safe Shutdown System [Ref. 53 states in part that:

A branch of the RHR discharge line provides low pressure safety injection (LPSI) to the reac .or vessel via parallel lines with one normally closed motor-operated valve (MOV) and one check valve in each line . The MOY position indi-cation is provided in the control room an< these valves receive an open signal coincident with the safety injection (SI) signal. The MOVs in the LPSI lines open ,on an SI signal before RCS pr ssure drops below RHR design pressure.

The plant complies to all EIhC aspects of the "RHR Interlock Requirements" review cri teria listed in Section 2 of this repor. except for the followiag:

(1) The plant RHR system does not satisfy BTP ICSB 3 [Ref.

13 and BTP RSB 5-1 [Ref. 2] because the RHR discharge and suction i sol ation MOYs do not have independent diverse interlocks to prevent opening the valves until RCS pressure is below 4:0 psig. Only the inboard valves MOY-700 and MOY-721 have thi s in erl ock .. The outboard valves MOY-701 and MOV-720 are manually control ed wi th key- 1 ocked swi tches.

1 By procedure, MOV-701 and MOY-720 are not opened until RCS pressure is less than 410 psig.

(2) The plant RHR system does not satisfy BTP ICSB 3 [Ref.

13 and BTP RSB 5-1 [Ref. 2j because all RHR isolation MOVs lack an interlock feature to close them when RCS pressure increases above the RHR design pressure.

(3) The plant RHR system does not satisfy BTP ICSB 3 [Ref.

13 and BTP RSB 5-1 [Ref. 23 because the isolation MOYs in the LPSI lines (MOV-852A and MOV-852B) open on an SI signal before RCS pressure drops below RHR design pressure.

6.

SUMMARY

The plant RHR interlock system fails to satisfy current. licensing criteria for the following reasons:

(1) The RHR suction and discharge isolation MOVs do not have independent diverse interlocks to prevent opening the isolation MOVs until RCS pressure is below 410 pslgo (2) All RHR isolation MOVs lack an interlock feature to close them when RCS pressure increases above RHR design pressure.

(3) The isolation MOVs in the LPSI lines open on an SI .

signal regardless of RCS pressure.

The resolution of items 1, 2 and 3 are presented in Sections 3.l and 3.2 of SEP Topic V-ll.A.

REFERENCES

1. U.S. Nuclear Regulatory Comnission, Branch Technical Position ICSB 3, "Isolation of Low Pressure Systems from the High Pressure Reactor Coolant System."

2. U.S. Nuclear Regulatory Commission, Brancn Technical Position RSB 5'-1, "Design Requirements of the Residual Heat Removal System."

3. NRC (D. M. Crutchfield) letter to LLNL (N. H. Di tmore), dated June 10, 1980.

4. Ginna drawing, 33013-436-A, "Auxiliary Coolant System". ~

SEP Review of Safe Shutdown Systems for .he R,E. Ginna Nuclear Power Plant, Revision 1, .undated.

6. Ginna drawing, 33013-432-A, "Safety inject;on System."

APPENOIX A HRC SAFETY TOPICS RELATED TO THIS REPORT 1 ~ III-1, "Classifica.ion of Structures, Systems and Components."

2. III-10.A "Thermal Overload of MOYs."

3. Y-10-.B, "RHR System Rel iabil i.y."

4. V-ll.A, "Requirements for Isolation of High and Low Pres'sure Systems."'.

Y1-7.C "ECCS Single Failure Criterion and Requiremen;s for Locking Out Power to Valves Including Independence of Interlocks on'ECCS Valves..",

6. YIII-3, "Systems Requi.red .or Safe Shutdown."

7. XYI, "Technical Specj.fications".

RE(UEST FOR ADDITIONAL INFORMATION ON SEP TOPIC VI-7.C.l FOR R. E. GINNA

1. For each of the seven automatic transfers from one dc train to the other, provide the short circuit analyses and the protective device coordination curves. Short circuit analyses should be provided for each of the following initial conditions:

a. Full battery charge with equilizing charge in progress

b. Battery near discharge with chargers not available.

2. Describe the methods used to assure that fault interrupting devices remain within the curves provided in response to question l. Your answer to this question should address breaker test frequency used vice that recommended by the breaker manufacturer and production lot verifica-tion of fuse characteristics.

3. Provide electrical schemhtiW from load to dc bus for each transfer and the other drawings given in References 4 and 5 of Enclosure 2 to your letter of March 27, 1981.