ML20207F685

Application for Amend to License DPR-61,changing Tech Spec Figure 2.2-2 to Reflect Results of Three Loop Flow Measurement Conducted During Cycle 14 Startup & Modifiying Tech Spec 3.20, RCS Flow,Temp & Pressure. Fee Paid

ML20207F685
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	12/31/1986
From:	Mroczka E, Sears C CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Charemagne Grimes Office of Nuclear Reactor Regulation
Shared Package
ML20207F691	List: B12389, Proposed Tech Spec,Modifying Tech Spec 3.20, RCS Flow,Temp & Pressure to Include Three Loop Operation Flow Requirement & Tech Spec Figure2.2-2,reflecting Results of Three Loop Flow Measurement ML20207F685
References
B12389, NUDOCS 8701060189
Download: ML20207F685 (5)

Text

CONNECTICUT YANKEE ATOMIC POWER COMPANY BERLIN, CO N N ECTIC uT P.O. BOX 270 H ARTFORD. CONNECTICUT 06101 Tskarmons 203-666-6911 December 31,1986 Docket No. 50-213 B12389 Office of Nuclear Reactor Regulation Attn: Mr. Christopher 1. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:

Haddam Neck Plant Proposed Revision to Technical Specifications Reactor Coolant System Pursuant to 10CFR50.90, Connecticut Yankee Atomic Power Company (CYAPCO) hereby proposes to amend its Operating License, No. DPR-61, by incorporating the changes identified in Attachment 1 into the Technical Specifications of the Haddam Neck Plant.

The proposed Technical Specification changes are required due to the repgits of the three loop flow measurement conducted during the Cycle 14 startup.w The Technical Specification affected by the measurement result is the three loop safety limit curve (Specification 2.2, Figure 2.2-2). Additionally, Technical Specification 3.20, Reactor Coolant System Flow, Temperature and Pressure, is being modified to include a three loop operation flow rate requirement.

The current four and th(ee loop safety limits were provided in Amendment 3 to the operating license.(2i These safety limits were based on design hot channel factors and core flow rates. The Cycle 6 reload analysis and safety limit evaluation performed by Yankee Atomic Electric Co. assumed the three loop core flow rate was 198,100 gpm or 81% of the four loop core flow rate (244,600 The corresponding three loop vessel flow rate, assuming 4.5% bypass gpm)(.3) flow is 207,500 gpm. The measured three loop vessel flow rate was 202,520 gpm and identified a nonconservatism in the flow rate assumed for the safety (1) See 3. F. Opeka letter to T. E. Murley dated August 7,1986, Core 14 Startup Physics Test Report.

(2) R. A. Purple letter to D. C. Switzer dated June 20.1975 on New Thermal Hydraulic Safety Limit Curves for Section 2.2, SER included.

(3) See F. M. Akstulewicz letter to 3. F. Opeka dated April 14,1986, Cycle 14 Reload Technical Specifications. o}

8701060189 861231 Ok D PDR ADOCK 05000213 P PDR k\ g0 0000

limits and safety analysis. Issuance of the subject amendment would serve to

' formalize the resolution of this discrepancy.

The three loop safety limits were adjusted to account for a lower core flow rate.

The adjustment was based on sensitivities of the DNBR to changes in inputs (e.g.,

flow, pressure, temperature and power) to the approved COBRA III-C model.

The adjustment was determined by calculating a lower core inlet temperature to compensate for a reduced core flow to maintain the same DNBR and core exit void fraction. The revised three loop safety limits are shown in the revised Figure 2.2-2.

These results show that the allowed inlet temperature for 65% power and 2000 psia is reduced from 587.20F to 581.20F. This value is well above the Technical Specification value of 540.60F and 5130F operating point. The revised three loop vessel flow rate of 197,200 gpm (188,300 gpm core flow rate) was conservatively based on the minimum flow rate required to preserve the current Variable Low Pressure Trip for three loop operation. This required flow rate yields 3% margin to the measured flow rate value.

The revised vessel flow rate of 197,200 gpm was also evaluated for the impact on the three loop safety analysis. The three loop safety analysis was performed at a power level of 86% and an inlet temperature of 5480F. The DNBR sensitivities to core flow rate and power level, discussed above, were used to show that the conservative power level of 86% versus 67% more than compensates for the reduction in core flow rate. The reduction in DNBR due to the flow reduction was 4%, while the increase in DNBR due to the drop in power-level was 36%

Therefore, the three loop analyses remain conservative for the current operating conditions.

The three loop vessel flow rate used above is proposed to be incorporated in Technical Specification 3.20. The surveillance requirement was revised to allow confirmation of the three loop flow rate based on the four loop result. The Cycle 14 test results demonstrated that the three loop flow rate is 78% of the four loop flow rate with measurement uncertainties included. The proposed three loop value of 197,200 gpm is 77% of the four loop Technical Specification value.

CYAPCO has reviewed the attached proposed changes pursuant to 10CFR50.59 and has determined that they do not constitute an unreviewed safety question.

The probability of occurrence or the consequences of a previously analyzed accident have not been increased and the possibility for a new type of accident has not been created. There are no failure modes associated with the change that could be an initiating event for a LOCA or non-LOCA event.

CYAPCO has reviewed the proposed changes, in accordance with 10CFR50.92, and has concluded that they do not involve a significant hazards consideration in that these changes would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated, in that there is no increase in the

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. probability of occurrence of a LOCA or non-LOCA event as a result of this change. There are no safety systems affected by the change.

The variable low pressure trip was evaluated based on the proposed three loop safety limits. It has been determined that the changes in the three loop safety limits do not result in a reduction of protection provided by this trip.

2. Create the possibility of a new or different kind of accident from any previously analyzed. The plant response for LOCA and non-LOCA

- events was not significantly affected by the reduction in the '

~ -allowable three loop flow rate and there were no new failure modes and no safety systems adversely affected due to this change.

Therefore, no new unanalyzed accidents are created.

3. Invol_ve a significant impact on the margin of safety, in that the protective boundaries are not affected since the peak cladding temperature for the design basis LOCA remains below the regulatory

limit of 23000F. Also, the safety limits were modified to account for the decrease -in minimum allowable RCS flow. These safety limits

are still based on the MDNBR and maximum void fraction criteria of 1.3. and' 32%, respectively.- Therefore, the margin of safety has not 4

been affected and there is no impact on the protective boundaries.

The proposed reduction in the three loop RCS flow rate does not change the. basis of Technical Specification 2.2. The basis continues to assure fuel rod integrity and the prevention of fission product release. The reactor core safety limits have been modified to account for the reduction in the flow rate consistent with the value specified in Technical Specification 3.20. '

. The Commission has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (51FR7750, March 6,1986). The changes proposed herein are enveloped by example (ii), a change that constitutes an additional limitation, restriction or control not presently included in the

technical specifications. The previously assumed 3-loop flow rate used in current safety limits is being replaced by a 3-loop flow technical specification requirement and 3-loop safety limits which are consistent with this additional requirement.

The Haddam Neck Nuclear Review Board has reviewed and approved the attached proposed revisions and has concurred with the above determinations.

In accordance with 10CFR50.91(b), we are providing the State of Connecticut

.with a copy of the proposed amendment.

i Pursuant to the requirements of 10CFR170.12(c), enclosed with this amendment is the application fee of $150.00.

With respect to the status of 3-loop operation at Haddam Neck the following informatior is provided. Our letter dated August 7,1936 informed the NRC that j 3-loop operation was administratively prohibited. In a subsequent letter dated

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September . 24; 1986 we. provided the Staff with a sequence of events for resolving this issue. Resolution of the. lower than required flow rate has been attained through the development of new 3-loop safety limit curves consistent with the additional 3-loop ilow requirement. Therefore, the 3-loop operation capability will .be reinstated with the administrative implementation. of the attached changes within the next several weeks. Issuance of this amendment request will serve to formally resolve this issue.-

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY

$.D N E. 3. Mroczka 3 Senior Vice President ,

By: C. F. Sears Vice President cc: Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protection Hartford, CT 06116

' STATE OF CONNECTICUT )

) ss. B erlin COUNTY OF HARTFORD )

Then personally appeared before me C. F. Sears, who being duly sworn, did state that he is Vice President of Connecticut Yankee Atomic Power Company, a .

Licensee herein, that he is authorized to execute and file _the foregoing. '

information in the name and on behalf of the Licensees herein and that the statements contained in said information are true and correct to the best of his -

knowledge and belief.

us, Y, Y mi4Ao tary P lic f/ Comm: sin DM U"ch 31. M33

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Docket No. 50-213 i B12389

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Attachment 1 Proposed Revision to Technical Specifications Reactor Coolant System December,1986

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