ML20137Q101

Forwards Comments on AP600 ERG Passive Safety System Termination Guidance

ML20137Q101
Person / Time
Site:	05200003
Issue date:	04/09/1997
From:	Huffman W NRC (Affiliation Not Assigned)
To:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 9704100125
Download: ML20137Q101 (8)

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Mr.,Ni holas'J/ Liparulo, Manager.3 Nuclear.; Safety ind Regulatory' Analysis [t '! .

~Nuclea'randAdvanced; Technology l Division "

i WestinghouseiElectric; Corporation W>

P.O. Box'355 4;

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'SUBJECJ v COMMENTS ON AP600' EMERGENCY RESPONSE GUIDELINES (ERGS) PASSIVE

. SAFETY SYSTEM TERMINATION GUIDANCE

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Dear Mr. Liparulo:

TheNuclear'RegulatoryCommissionkNRC)ReactorSystemsBranch,withassis-  !

tance' from' a' contractor from Brookhaven National Laboratory, has reviewed the i AP600 ERGS to determine if they provide clear guidance on the termination of.

. passive safety. systems by the plant operators. Based on '...is review, the ,

istaff has generated comments.that may require the ERGS to be corrected and  ;

revised.. These comments are attached as an enclosure to this letter. l

~ Westinghouse is requested to review these comments'and arrange for a meeting 1

-or telephone conference to discuss what actions are necessary for resolution.- j We.also request that these comments be included in the open item tracking i

system so that the status and disposition of these-items can be tracked. j

.If you have any questions regarding this matter, you can contact me at 1

[ (301) 415-1141.-

Sincerely, 5 original signed by:

< William C. Huffman, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation I

. Docket No.52-003

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Enclosure:

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Mr'. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Mr. Ronald Simard, Director Advanced Plant Safety & Licensing Advanced Reactor Programs Westinghouse Electric Corporation Nuclear Energy Institute

, Energy Systems Business Unit 1776 Eye Street, N.W.

1 P.O. Box 355 Suite 300

[ Pittsburgh, PA 15230 Washington, DC 20006-3706 Ms. Cindy L. Haag Ms. Lynn Connor Advanced Plant Safety & Licensing Doc-Search Associates Westinghouse Electric Corporation Post Office Box 34 Energy Systems Business Unit Cabin John, MD 20818 Box 355

Pittsburgh, PA 15230 Mr. James E. Quinn, Projects Manager

! LMR and SBWR Programs

Mr. M. D. Beaumont. GE Nuclear Energy  ;

Nuclear and Advanced Technology Division 175 Curtner Avenue, M/C 165 Westinghouse Electric Corporation San Jose, CA 95125 One Montrose Metro i 11921 Rockville Pike Mr. Robert H. Buchholz '

Suite 350 GE Nuclear Energy Rockville, MD 20852 175 Curtner Avenue, MC-781 San Jose, CA 95125

- Mr. Sterling Franks U.S. Department of Energy Barton Z. Cowan, Esq.

NE-50 Eckert Seamans Cherin & Mellott 19901 Germantown Road 600 Grant Street 42nd Floor '

Germantown, MD 20874 Pittsburgh, PA 15219 Mr. S. M. Modro Mr. Ed Rodwell, Manager

- Nuclear Systems Analysis Technologies PWR Design Certification

> Lockheed Idaho Technologies Company Electric Power Research Institute  !

Post Office Box 1625 3412 Hillview Avenue '

Idaho Falls, ID 83415 Palo Alto, CA 94303 Mr. Frank A. Ross Mr. Charles Thompson, Nuclear Engineer )

U.S. Department of Energy, NE-42 AP600 Certification 1 Office of LWR Safety and Technology NE-50 l 19901 Germantown Road 19901 Germantown Road Germantown, MD 20674 Germantown, MD 20874 I

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Staff Comments on i AP600 Emergency Response Guidelines Termination Criteria Introduction

-The AP600 Emergency Response Guidelines (ERGS) were reviewed to determine if .

they provide sufficiently clear guidance concerning the termination by the operators of the passive safety systems. ,

The original version, Rev; 0, of these ERGS was reviewed before Rev. 2 was ,

received. Then after Rev. 2 was received and inserted into the ERG Volumes, the review was rechecked against Rev. 2. Some changes to the findings resulted from the Rev. 2 changes.  ;

The AP600 passive safety systems considered for this review task are:  ;

Core Makeup Tanks (CMT) i Passive RHR System (PRHR)

Automatic RCS Depressurization (ADS)

Accumulators Passive Containment Cooling (PCS)

In Containment RWST (IRWST)

It should be noted that the work for this portion of the review of the ERGS was not' global and was limited in scope to that described above.

Comments AE-0. " Reactor Trin or Safety In_iection" i

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1. 1 Step 25 This step requires the operator to check if the passive safety systems should be terminated. The operator checks the following CSFs: RCS subcooling, RCS heat sink, RCS pressure, & pressurizer level against criteria. If the criteria are satisfied, the operator leaves this guideline and transfers to AES-1,1, " Passive Safety Systems Termination." In general, for these critical safety functions (CSFs), the availability / status of the non-safety systems do not appear to be verified. In one case (RCS heat sink) there is a check of the non-safety systems, by checking total feed flow to SGs or narrow range SG level in one SG. However, there is also an "or" for PRHR operating. The rcd decay heat removal portion of 1 Step 25 is thus worded such that it can be satisfied with only PRHR operating. (Perhaps the "or" was intended to be an "and.") Yet the purpose of the step is partially to determine if PRHR should be  !

terminated. Thus, depending on the philosophy of when/whether to -l check for the availability of non-safety ba kup systems, the Enclosure

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transition to AES 1.1, Passive Safety Systems Termination, may be premature. Note: the steps after Step 25 and 26 would not be  :

performed, including:

Step 27 Initiate monitoring of CSF status trees Step 30 Check if ADS should be actuated Step 36 Check if PRHR should be isolated (checks for SFW operation and NR level in at least one SG) 1 Step'37 A return to Step 12.

- Transfers to other event based procedures in several steps.

In AES 1.1, " Passive Safety Systems Termination," Step 6 checks the following to see if PRHR should be isolated: (1) SFW in operation and (2) NR level in at least one SG. However, if these two criteria are agi met, one still proceeds to step 7 without securing PRHR at this time. Thus, given the criteria of AE-0 Step 25, you could enter AES 1.1 without SFW. If that's the case you continue in procedure AES-1.1 with PRHR operating and no SFW.

Further instructions do not appear to be given. Per the background document (knowledge section) step 6 of AES 1.1 is not a continuous action step. Thus, ,

while the intent of AES 1.1 is to terminate the safety systems, you enter it really not ready to terminate PRHR and the procedure does nothing further to solve that problem.

AES-0.1. " Reactor Trin ResDonse" d

#2 Page 2 of the guideline has a generic caution statement that if SI actuation

[ occurs during this guideline - go back to AE-0.

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Step 1 Requires the operator to take various decay heat removal actions to control RCS temperature. One of these actions is to initiate PRHR.

Step 1 also contains actions to terminate PRHR.

l- Step 3 Requires the operator to manually actuate PRHR if SFW is not

! available and SG narrow range level criteria is not satisfied.

l These steps do not explicitly direct the operator to another procedure (such

. as AE-0 or AES-1.1) for PRHR termination. The criteria in this ERG for PRHR termination are different from other ERGS. Moreover, subsequent steps in AES-0.1 do not provide PRHR termination guidance. It does not appear that the L caution statement, noted above, encompasses PRHR.

1. 3 Step 6 This step. checks CMT status. The CMT system auto actuates at a 4

certain pressurizer level. If. PZR level is less, the operator verifies that the CMT injection valves are open. If not, the i

valves are manually opened. The operator is directed to continue 2

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with Step 7 of the procedure, rather than being explicitly directed back to AE-0. This appears inconsistent with the purpose and entry conditions of the ERG.

Note: Step 7 looks at PZR pressure. If it is less than (Pil) psig, SI actuation is verified or manually actuated, and the operator is directed to AE-0.-

With regard to the CMT, Step 6 should probably have an independent statement to direct the operator to AE-0 instead of relying on Step 7. Procedure AES-0.1 appears to assume that CMT actuation on level (Step 6) will always occur simultaneously with a low RCS pressure actuation (Step 7) and therefore separate transfers are not necessary. That may not be true for spurious actuations.

1. 4 Step 14 The last step of this procedure directs the operator to AES-0.2 Natural Circulation Cooldown. AES-0.2 appears to assume PRHR and/or CMT are not actuated, but they may have been actuated in this procedure.

1. 5 In summary, the procedure has several steps that initiate PRHR. It does not appear that the actuation of PRHR transfers the operator to AE-0. Further, one step (Step 1) in AES-0.1 can isolate PRHR (and the termination criteria are not the same as in AES-1.1). It appears that PRHR could remain in operation as you transition out of this procedure into natural circulation .

cooldown which does not appear to address PRHR.

AES-0.2. " Natural Circulation Cooldown"

1. 6 The use of PRHR support systems (i.e., IRWST cooling) are not addressed for using PRHR in a natural circulation cooldown. The IRWST may eventually heat to saturation. It would seem that these support systems should be addressed in the ERGS.

t AE-1. " Loss of Reactor or Secondary Coolant"

1. 7 Steps S&6 These steps check if passive safety systems should be terminated and directs the operator to AES-1.1, " Passive Safety Systems Ter-4 mination." The same comment applies as discussed for AE-0 Step 25 with respect to the "or" for PRHR and feed flow.  ;

1. 9 Step 8 This step checks if passive containment cooling (PCS) should be l stopped. If the containment pressure is less than the criteria, l 1 the operator is instructed to stop PCS.

The termination criteria do not consider the status / availability of other containment heat removal systems prior to stopping PCS. It is not clear if the containment pressure criterion inherently j- considers this.

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1. 10 General Comments on AE-1:

(a) Unlike AE-0, this guideline does not direct the operator back to the beginning of the procedure. If the transfer to AES-1.1, " Passive Safety Systems Termination," at Steps 5 and 6 isn't satisfied, one gets to the end of AE-1 (Step 17-Evaluate Long Term Plant Status) apparently without formally satisfying the passive safety system termination criteria.

Thus, the ERG does not appear to provide sufficient coverage in this case.

l (b) It is unclear why the ERG does not take advantage of the Reactor Vessel Level (hot leg level) as a parameter to assist the operators in this procedure.

AES-1.1. " Passive Safety Systems Terminatign" Note that the termination criteria in AE-0 and AE-1 are not the same as in AES-1.1.

1. 11 Step 3 This step directs the operator to close the CMT isolation valves, )

apparently relying on the termination criteria of AE-0 and AE-1. I There are no checks of availability of non-safety backup systems. 1 (see comments on AE-0, Step 25 and AE-1, Step 6).

1. 12 Steps 4&22 These steps require the operator to verify CMT injection is not re-quired. Although there is a transfer back to AE-1, there is none for AE-0, to address a premature isolation of the CMT for reasons other than a loss of coolant (Is this needed?).

1. 13 Step 6 This step checks if PRHR should be isolated - requires SFW in I operation and NR level in at least one SG . l Given the criteria of AE-0 Step 25, you could enter AES-1.1 without SFW. If that's the case, you continue in AES-1.1 with only PRHR operating (no SFW). Further instructions do not appear to be given. Per the background document this is not a continuous action step (i.e., it is not so noted in the knowledge section).

1. 14 Step 8 This step checks if passive containment cooling should be stopped.

Terminatior criteria for PCS was not provided in AE-0 or AE-1, in contrast to the Entry Condition Statement of AES-1.1. The only check here is containment pressure less than the criterion. There is no check of availability of non-safety containment heat removal systems. l l

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AES-1.2. " Post LOCA Cooldown and Denressur12ation"

1. 15 Step 8' This step checks if ADS should be actuated. If ADS is actuated, plant conditions no longer satisfy the entry conditions to this guideline and a transfer back to AE-l~would appear to be ap-propriate. If not, the basis for remaining in this procedure should be discussed in the background document.

Beginning in AE-1, it appears that the operator may or may not transition to AES-1.2. Both have steps for actuating ADS. The final end states (the last few steps) appear different. This should be clarified?

1. 16 Step 9 This step checks-if CMT injection should be isolated. The availability of non-safety systems to perform RCS injection does not appear to be verified prior to CMT injection termination. Note that the AE-1.2 background document states "The CMTs can be isolated if RCS inventory is being maintained by the CVS makeup pups." This should be a step in the ERG itself, as is typically done elsewhere in the ERGS by checking SFW before securing PRHR.

1. 17 Step 17 This step checks if Passive Containment Cooling should be stopped without checking for the availability of non-safety backup systems.

This is the same comment as given on AE-1 Step 8.

AE-3. " Steam Generator Tube Ruoture" l Step 14 Checks for CMT isolation.

l Step 21 Checks for stopping PCS, Step 25 Checks to isolate SI Accumulators.

1. 18 Steps 14, 21, and 25 of AE-3 all terminate use of passive safety systems without checking on the availability of a defense-in-depth or other non-safety 1- related system which can back up the safety function of the system being

! secured. If a backup system is not necessary for this procedure, the back-ground document should clarify how the safety functions provided by these j systems are maintained after they are isolated.

General Comment on ADS Termination  !

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1. 19 All of the ERGS were reviewed to determine whether there was adequate guidance for terminating ADS if that should become necessary, for example due to ,

spurious actuation of ADS. The guidelines reviewed addressed ADS only within  !

the context of " verification of actuation / manual actuation backup." Unlike other " passive" safety systems (specifically CMT) where the background

documents make reference to spurious operation, the information on ADS did not  ;

address isolation at all, let alone due to spurious operation. It may  !

eventually be addressed in the more detailed E0P, to be developed by a COL 4

after design certification, for ERG AE-1, " Loss of Reactor or Secondary Coolant," step 11, Initiate Evaluation of Plant Statur. In the-Westinghouse l response to NRC's comunent #48 on the Adverse Systems Interaction document, Westinghouse mentions several actions associated with both the E0Ps and with operator response to spurious ADS actuations. However, these activities do not appear to be outlined in the ERGS, such that a COL could use them to ,

develop appropriate E0Ps.  !

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