ML20246K640

Forwards Safety Evaluation Supporting Temporary Increase in P-9 Permissive Setpoint During Preoperational Testing of New Digital Feedwater Control Sys

ML20246K640
Person / Time
Site:	Prairie Island
Issue date:	04/28/1989
From:	Musolf D NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.10, TASK-TM NUDOCS 8905180110
Download: ML20246K640 (8)

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Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401 '927 Telephone (612) 330-5500 April 28, 1989 Director of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Resetting P-9 to 20% During Feedwater Control System Testing The purpose of this letter is to provide, for the information of the NRC Staff, a copy of our safety evaluation supporting a temporary increase in the P-9 permissi- e setpoint during preoperational testing of the new digital feedwater control system in Unit No. 2.

1"he attached evaluation concludes that an increase in the P-9 setpoint from 10% to 20% for the duration of the test is consistent with the NRC Staff position expressed in NUREG-0737, Item II.K.3.10, and with the Prairie Island Updated Safety Analysis Report. The Prairie Isisnd Technical Specifications permit a P-9 setpoint of up to 50%

This issue was discussed with the NRC Project Manager for Prairie Island on April 24, 1989. At that time we provided our commitment to provide a copy of our evaluation.

Please contact us if you have any questions related to our evaluation.

O ; - G Mw s . O David Musolf \

Manager Nuclear Support Services c: Regional Administrator III, NRC Sr Resident Inspector, NRC Sr Project Manager, NRC fN \

8905180110 890428 \ \g PDR ADOCK 05000282 P PDC l

PR'-27-O% THU 14 203 P t- a i e- W e- Is 1 and NSP P. O3

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SAFETY EVALUATION No. 264 (NON-MODIFICATION) l l

The purpose of this safety Evaluation is to address the resetting of the P-9 permissive to twenty percent of rated power during the performance of operational testing on the Digital Feedwater Control System installed by modification for Units 1 and 2.

The Technical Specifications presently allow P-9 to be set at or below fifty percent of rated power. The P-9 permissive is normally set at ten percent of rated power in accordance with the commitment to II.K,3.10 of NUREG 0737 (attached). This commitment requires that a individual plant analysis be performed to ensure the probability of a small break LOCA resulting from a stuck open pressurizer PORV is substantially unaffected by increasing the P-9 setting. If the setting for P-9 were to be raised the analysis must be performed and supplied to the NRC prior to any change in the setpoint.

This subject was discussed in a telecon with the NRR project manager and NSP Licensing on April 24, 1989. The re'sult of that discussion pas that NRR saw no problem with raising the P-9 setpoint for the duration of the testing of the Feedwater Control System. It was requer. that NSF perform a review pursuant to 10CFR50.59 and forward a copy of the review to NRR prior to increasing the setpoint.

The reason it is desirable to increase the P-9 setpoint during the performance of the Feedwater Control System testing at about ten percent of rated power is to minimize the potential for a reactor trip during this period. The feedwater control system _

testing perturbs the feed flow and steam generator level in the positive direction. The intent is to verify the system's tuning constants. If a problem occurs the operator is expected to terminate the test at 55% narrow range steam generator level.

If this is not successful a turbine trip and feedwater isolation will occur at 67% narrow range level. In this case the reactor would trip if power is greater than 10% for the existing P-9 setpoint. Resetting P-9 to 20% would prevent a reactor trip in this instance.

The resetting of P-9 has been previously addressed as a part of Revision 32 to the plants Technical specifications. The basis for the 50% maximum setpoint for P-9 is to ensure that a reactor trip occurs when power is greater than the load rejection capacity of steam dump valves. Resetting P-9 to 20% power is clearly within this basis.

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,APf,-27-G9 THU 14a03 Pra ie ie 8sIand NSP P. 04 Page 2 of 3 USAR Section 7.2.3.4 addresses loss of load with turbine bypass. 1 This section states that the pressurizer relief valves might be actuated for the most adverse condition. Section 7.2.3.5 addresses the current P-9 setpoint. Section 14.4.9 addresses loss of external load. That section states that the plant is evaluated for a complete loss of load from one hundred two percent of rated power with out direct reactor trip and that this transient meets  ;

all acceptance criteria. This is also true for the current reload i pattern. Therefore resetting of P-9 to twenty percent for the j test period does not affect the USAR accident analysis.

The requirement to not increase P-9 that is in NUREG-0737 II.K.3.10 is based upon questions raised by the THI2 event regarding the probability of a pressurizer PORV opening and ,

failing to reset following a pressure transient or a loss of  !

load event without reactor trip. In this regard the following statements can be made:

1) The increase from 10% to 20% rated power for P-9 does not significantly change the probability of the event occurring due to the fact that the time to be spent at this power level is minimal.

2) The transient from 20% reactor power should not challenge the FORV's with proper steam dump operation.

3) In the c se of a PORV not reseating other aspects of i NUREG 07a7, such as valve position monitors and operator training, provide adequate assurance that any transient could be quickly terminated by shutting the appropriate PORV block valve.

I Based upon the above 1.nformation it is concluded that P-9 can be _

safely reset to 20% of rated power for the duration of Feedwater Control System testing for Unit 1 and Unit 2.

And that the setpoint change:

1) Does not create a possibility for an accident or

! malfunction of a different type than evaluated l

previously in the USAR/FSAR or subsequent l

commitments.

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,AP3-07-89 THU 1 c4 : 04 P t- a i t- ie Is I a rs d NSP Pa 05 l l

Page 3 of 3 l 4

l l 2) Does not increase the probability of occurrence of a l accident or malfunction of equipment important to safety previously analyzed in the USAR/FSAR on subsequent commitments.

3) Does not increase ~the consequences of an accident or malfunction important to safety previously analyzed in ,

the USAR/FSAR or subsequent commitments.

4) Does not reduce the margin of safety defined in the basis for the Technical Specifications.

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NORTHERN STATES POWER COMPANY l u t NN z A pou s. u t sw a s ora so4ci December 30, 1980 Director of Nuclear Reactor Regulation U S Nuclear Regulatory Commission

• Washington, DC 20555 '

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket No. 50-282 License No. DPR-42 50-306 DPR-60 Post TMI Requirements - RUREG-073 7 All licensees of operating plants were mailed a letter dated October 31, 1980 from Mr D G Eisenhut, Director, Division of Licensing, Office of Nuclear Reactor Regulation, which contained revised Post TMI Requirements in a document identified as NUREG-0737. The licensees were requested to furnish confirmation that the implementation dates indicated in Enclosure 1 to NUREG-0737 will be met or to furnish justification for delays.

In the attachment to this letter, Northern States Power Company provides commitments to those hardware, procedural and organizational implementation requirements that will be met or furnishes justification and requests for exemption commitments forare those dependent implementation requirements that may be delayed. Our on equipment availability and assume no changes in regulatory 31, 1980 letter. positions ,or interpretations beyond those stated in the October _

)

Mr Eisenhuc 's October 31, 1980 letter requires a large number of design descriptions, January 1 or 2,evaluations, and information transmittals to be submitted by 1981. Much of this material has already been submitted for NRC Staff review and other items will be addressed in Owners Group correspon-dence.

For the remaining submittals we find we will be unable to meet the required January 1 or 2,1981 date because of the lateness of the NRC's clarification information, the holiday vacation schedule, and the already heavy burden placed on our technical staff by other NRC requirements such as fire protection, environmental qualification, and on going TMI modification work.

We are directing our efforts to providing the written submittals for these items by February 1,1981.

NUREG-0737 items are involved:

Information submittals for the following

1. , . .

1 4

II.K.3.9 PROPORTIONAL INTEGRAL DERIVATIVE (PID) CONTROLLER - ,.

t' This modification was completed at Prairie Island when the two-out-of three low pressurizer pressure safety injection actuation logic change was made. The setpoint of the PORV interlock bistables was changed to 2335 psig. This in effect raised the permissive to the same setting as the trip setpoint.

Our derivative time conse. ant in the PID controller for . the PORV is set at zero which, in effect, removes the derivative action from the controller. Removal' of the derivative action will decrease the likelihood of opening the PORV since the actuation signal for the valve is then no longer sensitive to the ' rate of change of pressurizer pressure.

II.K.3.10 PROPOSED ANTICIPATORY TRIP MODIFICATIONS Prairie Island received NRC permission to include a P-9 permissive to omit reactor trip on turbine trip below a preset power level (50% was the original power level selected for P-9).

. II.K.3.10 requires delays in this type of modification until the small break LOCA probability analysis resultint; from a stuck open PORV is completed and shows there is little effect from the addition of this modification. The setpoint on Unit I has been lowered to approximately - 10% which is the present setpoint for the existing P-10 permissive . Unic 2 will be lowered from 30% to 10% during the next shutdown. This item 'is considered complete.

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II.K.3.12 CONFIRM EXISTENCE OF ANTICIPATORY TRIP UPON TURBINE TRIP Prairie Island has an anticipatory reactor trip on turbine trip. ~

This item is c^onsidered cemplete.

II.K.3.17 ECCS SYSTEM OUTAGES l

Prairie Island commits to submit a report by February 1,1981 detailing outage dates, length of outages and cause of each outage on ECCS

systems for the last five years of operation. The corrective I action necessary to prevent or minimize outages will be specified.

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.Mr. L. 0.~Mayer, Manager Nuclear Support Services . ,

l . Northern States Power Company -

414 Nicollet Mall-8th Floor Minneapolis, Minnesota . 55401

SUBJECT:

TMI TASK PLAN ITEMS II .K 3.9, II .K.3.10 AND II .K.3.12 FOR FRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 Your letter of December 30, 1980 indicates that you have taken the actions requested under Item II.K.3.9 (PID Controller) of NUREG-0737. This has been verified by the Prairie Island Resident Inspector. Therefore, this item is' resolved for your facilities.

Your letter also indicates that no change in the setpoint for the reactor trip on turbine trip other .than that which the NRC previously found acceptable is contemplated, as this is part of the original plant design, and therefore, Item II.K.3.10 is resolved for your facilities. In the future, shculd you request authorization to raise this setting the infonnation requested in Item II.K.3.10 should be supplied.

Your letter further confirms that Prairie Island Units 1 and 2 have an ~

anticipatory reactor trfp ori turbine trip, in conformance with Item II.K.3.12 of NUREG-0737. Therefore, this item is resolved for your facilities.

Sincerely,

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Robert A. Clark, Chief '-

Operating Reactors Branch #3 Division of Licensing cc: See next page

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II.K.3.10 PROPOSED ANTICIPATORY TRIP MODIFICATION ,

Position The anticipatory trip modification proposed by some licensees to confine the range of use to high power levels should not be made until it has been shown on a plant-by-plant basis that the probability of a small-break loss-of-coolant accident (LOCA) resulting from a stuck-open power-operated relief . valve (PORV) is substantially unaffected by the modification.

Changes to Previous Requirements and Guidance ,

There are no changes to the previous requirements.

Clarification This evaluation is required for only those licensees / applicants who propose the modification. .

Acolicability This requirement applies to selected Westinghouse operating reactors and operating license applicants.

Implementation Operating Reactors--Completion date for meeting requirements will be dictated by plant schedule for proposed modification.

Operating License Applicants--All applicants for operating license should submit documentation 4 months prior to the expected issuance of the staff safety evaluation report for an operating license or 4 months prior to the listed implementation date, whichever is later.

Type of Review

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A preimplementation review will be performed.

Documentation Recuired (1) The licensee is to submit the required analysis and document-proposed change for staff approval prior to implementation. Documentation is to be submitted as proposed by the licensee.

(2) Modification schedule is to be determined on a plant-specific basis.

Technical Specification Changes Recuired Changes to technical specifications will be required.

3-148 II.K.3.10-1

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