ML20236E326

Forwards Assessment & Tech Spec Clarification Re 881212 Loss of 5-volt Dc Power Supply & Loss of Rod Display & Operator Control Modules.Restoration of Normal Rod Position Indication,Rather than rod-by-rod Indication,Proper

ML20236E326
Person / Time
Site:	Perry
Issue date:	03/20/1989
From:	Virgilio M Office of Nuclear Reactor Regulation
To:	Greenman E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 8903240098
Download: ML20236E326 (6)

Text

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Clf o UNITED STATES

' g E' o NUCLEAR REGULATORY COMMISSION g

• g l WASHINGTON, D. C. 20555 March 20, 1989

...../ l I

i l l MEMORANDUM T0: Edward G. Greenman, Director l

Division of Reactor Projects, Region III i i

FROM: Martin J. Virgilio l Acting Assistant Director l for Regions III and V Division of Reactor Projects III, IV, Y, and Special Projects

SUBJECT:

REQUEST FOR TECHNICAL ASSISTANCE - TECHNICAL SPECIFICATION CLARIFICATION PERRY NUCLEAR POWERPLANT,UNITNO.1(AIT0060)

On December 12, 1988, a loss of power from a 5 volt DC power supply (PS-52) resulted in loss of both the rod display module (RDM) and the operator control module (OCM) at the Perry Nuclear Power Plant, Unit No. 1. This resulted in inoperability of the rod position indication system (RPIS) and the rod control and informationsystem(RCIS),respectively.

By memorandum dated January 24, 1988, you requested clarification

.astowhethertheTechnicalSpecifications(TS)shouldindicate (see TS 3.1.3.5 and 3.1.3.1.b of the Perry Nuclear Power Plant TS) ,

that control rod insertion in lieu of taking no action to alter rod position is the appropriate action when accompanied by a loss of all rod position indication. You noted the potential for i stuck rods and out of sequence rod movement as concerns. You also requested clarification as to whether obtaining rod position  !

indication by back panel readings on a rod-by-rod basis would be sufficient as a means of avoiding entry into or allowing exit from the action statement of TS 3.1.5. '

Enclosed is our staff's detailed assessment and TS clarification as requested. In summary, the staff believes that shutdown of the reactor (by scramming) is proper rather than non-movement of the control rods if the above-described losses of RPIS and RCIS occur. Additionally, the staff believes that restoration of normal rod position indication rather than obtaining rod-by-rod position indication from the back panels is proper. The latter would not be an acceptable alternative for obtaining rod position indication unless evaluated with procedures and training developed

Contact:

T. Colburn 49-21369 pf8I

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'e' 2 o and appropriate TS in place. Should you have any questions concerning our response, please contact Timothy Colburn (492-1369),

the Perry Project Manager.

1

/s/

Martin J. Virgilio i Acting Assistant Director. l

'for Regions III and V ,

Division of Reactor Projects III, IV, V, and Special Projects-  !

Enclosure:

i As stated j l

cc w/ enclosure: i W. Hodges, SRyd l G. Holahan, NRR ,%y. Distribution.,n.

8sDockettF11e g ijiOGC  !

M. Miller, RIII Local 's~'PDRs EJordan- 'l G. O'Dwyer, R1 Perry PDIII-3 r/f BGrimes ')

P. Hiland, SRI Perry MVirgilio ACRS(10)

M. Caruso, NRR JHannon TColburn PKreutzer i

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• SEE PREVIOUS CONCURRENCE Vki  !

Office: LA/PDIII-3 PM/PDIII-3 PD/PDIII-3 (A)AD:tRSP35 I

Surname: *PKreutzer *TColburn/mr *JHannon MVirgilio Date: 03/10/89 03/13/89 03/13/89 J /90/89 <

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__m___.____

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and appropriate TS in pla e. Should you have any questions ..

concerning our response, please. contact Timothy Colburn (492-1369),  ;

' the Perry. Project Manager. q l

J Martin J. Virgilio Acting Assistant Director for Regions III and V  ;

Division of Reactor Projects III, l IV, V, and Special Projects j

Enclosure:

As stated j i

cc w/ enclosure:

W. Hodges, SRXB i G. Holahan, NRR T. Colburn, NRR .

M. Miller, RTII .!

G. O'Durjer, RI. Perry P. Hiland,' SRI Perry '!

M. Caruso, NRR ,

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N' NC- k Office: LA/PDIII-3 PM/PDIII-3 PD/PDIII-3 )AD:DRSP35 i Surname: PXgeg TColburn/mr JHannon M rgilio j/pzer/89 Date: 3 // 3 /89 3 / /3 /89 / 89 ]

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, 1 TECHNICAL SPECIFICATION CLARIFICATION FOR THE PERRY NUCLEAR PLANT (AIT 0060)

Region III requested NRR assistance concerning the interpretation of certain Technical Specifications that were pertinent to an event at the Perry Nuclear 1 Plant,' Unit 1(Perry)onDecember 12,1988(Ref.1). The. event, which l apparently began with a loss of power from a 5 volt DC power su  ;

resulted in a loss of both the control rod display module (RDM)pply and the (PS-52),

operatorcontrolmodule(OCM). This in turn, resulted in inoperability of the

-rodpositionindicationsystem(RPIS)andtherodcontrolandinformation ,

system (RCIS). This left all of the control rods inoperabic but apparently  !

trippable with the reactor at 100% power. Additional details concerning this event may be found in the Region III Daily Report provided as Enclosure 1 to  !

Reference 1.

Two Technical Specifications (Ref. 2) are of primary interest for this event.

The first is Specification 3.1.3.1 which states that all control rods shall be operable. The other Specification is 3.1.3.5 which states that at least one l RPIS shall be operable. With the luss of the RDM and the OCM, the operator was not able to make control rod movements in the normal manner. Thus, none of the control rods were Operable and Action Statement c of Specification 3.1.3.1 should have been entered by the licensee. This requires that the plant must be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more than eight control rods inoperable. Action Statement b of Specification 3.1.3.1 could have also been entered and would have led, though less directly, to the same requirement as Action Statement c. - Even though Specification 3.1.3.5 was not met because of the loss of the RPIS, the requirements of Action Statement a cannot be met because there were no Operable control rods. This also requires that the reactor then be in at least Hot Shutdown within the next twelve hours. During the event, the licensee did not attempt any actions other than to replace the  :

defective power supply. With a good power supply, the licensee evidently exited Action Statement c (or equivalently Action Statement b) of Specification 3.1.3.1 before the time limit of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> had expired. That is, the OCM, the i RDM, the RPIS, and the RCIS were all operable again and Specifications 3.1.3.1  !

and 3.1.3.5 were met.

Region III requested that NRR provide clarification to two concerns with the Specifications. These are:

(1) Should the Technical Specifications indicate (as they apparently do) that control rod insertion in lieu of taking no action to alter rod position is the appropriate action when accompanied by a loss of al'1 rod position ,

indication? Region III is concerned that scramming the rods, given the loss of insertion capability, would be an unconservative action unless the operators noted increasing reactor power. Regien III is also concerned about subsequent potential for stuck rods and out-of-sequence rod movement that could result in hot spots and potential core damage. i c

.' 2 (2) Whether or not obtaining rod position indication by back panel readings on a rod-by-rod basis is: sufficient for not ever entering the Action

, Statement of Technical Specification 3.1.3.5 or exiting the Action

} Statement once operators are setup to monitor back panel rod indication.

l or whether restoration of the normal control rod position indication is  ;

required.  !

Before responding te Region III's concerns, a brief discussion of the intent of the two applicable Specifications is pertinent. The intent of Specifica-tion 3.1.3.1 is that all (or nearly all) control rods should be operable. The >

intent of Specification 3.1.3.5 is that at least one RPIS should be operable or i that a control rod with an inoperable position indicator should be moved to a position with an operable position indicator. The Sasis for these two Specifications states that the occurrence of eight (or more) inoperable control rods could be indicative of a generic problem and the reactor must be shut down (emphasis added) for investigation and resolution of the problem. The Perry event was an unusual event that, among other things, clearly resulted in more than eight inoperable control rods. In fact, all of the control rods were inoperable. Therefore, entering Action Statement c of Specification 3.1.3.1 i would have led to the reactor being placed in at least Hot Shutdown within the '

next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This result would meet, therefore, both the scope and intent of the two applicable Specifications.

For the Perry event, it is not obvious that control rods could not or did not receive erroneous insert or withdrawal signals from the failed systems (the

' RDM, the OCM, the RPIS and the RCIS). Thus, the exact state of the reactor was not absolutely known during the event. On this basis, NRR cannot recommend a no-action position, that is, leaving the reactor in its assumed pre-event position. In fact, if control rods had moved because of erroneous signals, the no-action position could cause fuel thermal limits to be exceeded so that fuel failures could possibly result. The Technical Specification position is clearly the preferred position because it places the reactor in a known, safe condition. Here we have no evidence that the failed systems would affect the trip function of the Reactor Protection System or the scrammability of the control rods. If they oid, this event would be an ATWS precursor. The disadvantage of tripping the reactor is, of course, the challenge to safety systems. All things considered, we advise against the no-action position and .

prefer the Technical Specification position. In addition, we conclude that a l reactor trip is not unconservative and that the subsequent )otential for stuck rods and out-of-sequence rod movement that could result in lot spots and potential core damage should not be of concern. However, for a Perry type of event, we suggest a possible modification to the Technical Specification position if the action statement has been entered to bring the reactor to the Hot i Shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This possible modification is to reduce power by reducing recirculation flow, consistent with the applicable flow control line and stability restrictions, if a portion of the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is used to perform trouble-shooting before bringing the reactor to a Hot Shutdown condition. This reduction in power without resorting to control rod movement has the advantage of ensuring ample margin to the thermal limits.

Our response to the second of Region III's two concerns is straightforward. It ,

is not permissible, as the Specification is now written, to obtain rod position  !

indication by back panel readings on a rod-by-rod basis to prevent entering the

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Action Statement of Specification 3.1.3.5 or to allow exiting the Action" j Statement once operators are setup to monitor back panel rod indication. In this instance, restoration of the normal rod position indication is required. .j However, credit could be taken for using the back panel as an alternative l method for obtaining rod position indication if a licensee followed the usual )

procedures for modifying Technical Specifications. This means that the  !

licensee would have to (1) evaluate the back panel as an alternate means of rod .j position indication including establishing'a time interval for performing the , j rod position readings and the development of procedures, and (2) submit the '!

evaluation and a proposed Technical Specification modification to the NRC. If ar evaluation led to a favorable Safety Evaluation, the NRC would issue

,a license amendment authorizing the use of the back panel as an alternative .;

'means for determining rod position for Specification 3.1.3.5. It should be= j noted that for the loss of the'S volt DC power supply, this would only permit j exiting TS 3.1.3.5, the LCO and associated ACTION STATEMENTS of TS 3.1.3.1 {

would still apply. j

. Based on the considerations discussed above for Region III's two concerns, we  !

do not conclude that any changes are required to the GE BWR Standard Technical Specifications. ,

i REFERENCES l

1. Memorandum from Edward G. Greenman of Region III to Martin J. Virgilio of l NRR on " Request for Tecnnical Assistance . Technical Specification  !

Clarification Perry Nuclear Plant, Unit No. 1 (AIT 0600)," dated  !

January 24, 1989, i

2. Perry Technical Specifications.

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