ML19323B469

Summarizes Background,Status & Future Plans Re Changes in Main Steam Line Rupture Matrix Sys,Safety Task Force Priority Item 51.Includes Description & Justification of short-term Change & Conclusions

ML19323B469
Person / Time
Site:	Crystal River
Issue date:	05/06/1980
From:	Moore G FLORIDA POWER CORP.
To:	Office of Nuclear Reactor Regulation
References
3--1-A, 3-0-1-A, TAC-12961, NUDOCS 8005120408
Download: ML19323B469 (4)

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8005120 %

i Florida l

Power May 6, 1980 File:

3-0-1-a Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Contaission Washington, D.C. 20555 S ubj ect :

Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Nuclear Safety Task Force Priority Items and Confirmatory Order for Crystal River Unit 3, April 14,1980

Dear Sir:

This letter provides a suninary of the background, status, and future plans related to changes in the Main Steam Line Rupture Matrix System, Safety Task Force Priority Item No. 51.

Background

One of the principal objectives of the CR-3 Nuclear Safety Task Force was to identify design or operational features which could lead to loss of all secondary cooling, thus requi ring operator action and increasing the potential for water relief through the pressurizer relief or safety valves.

A number of simplified event tree analyses were performed by the Task Force.

Several of these event tree analyses, including those for steam line break, loss of main feedwater, cxcessive main feedwater, loss of AC power, and loss of ICS power, include sequences which lead to initiation of the steam line rupture matrix.

This, in turn, causes isolation of both main and auxiliary feedwater, and without subsequent operator action, prevents heat removal on the secondary system and leads to the HPI heat removal mode.

Because of the comparatively low prob-dbilities assigned to Correct operator action, the event trees thus show an undesirably high likelihood of ending up in this less desirable cooling code.

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General Office 3201 ininy-founn street scuin. P O Box 14o42, St Petersburg. Florda 33733 813 - 866 5151

Director Page Two May 6, 1980 Office of Nuclear Reactor Regulation In light of these conclusions and similar conclusions reached by the Task Force in reviewing the interaction of the Steam Line Rup-ture Matrix System and the EFW system, the Task Force recommended, subject to NRC staff approval, that the rupture matrix signals be deleted from emergency feedwater valves FWV-161 and FWV-162, in the short-term (as soon as possible), and that longer term changes be incorporated in the rupture matrix to preclude feeding the affected generator in the event of a steam line rupture without inhibiting EFW for all other events.

These conclusions are fully consistent with Staff's Reconnendation 4 of draft NUREG-0667 which states:

"4.

Steam Line Break Detection and Mitigation System A.

Eliminate Adverse Interaction with AFW System B.

Reevaluate and Modify Such That System is Capable of Differentiating Between Steam Line Break and Over-cooling and Undercooling Transients."

Elimination of rupture matrix signals to emergency feedwater valves FWV-161 and -162 satisfies the NRC Staff Recommendation 4.A.

The longer tenn changes contemplated are expected to satisfy Recommen-dation 4.B.

Description of Short-Term Change Valves FWV-161 and -162 are preset at 22% open to pass a minimum of 500 gpm.

The Steam Line Rupture Matrix System isolates valves FWV-161 and -162, FWV-33 through -36 as well as main feedwater and main steam if low pressure is sensed in both steam generators.

The sroposed short-term change would eliminate the rupture matrix sig-nals to FWV-161 and -162, thus allowing EFW to be delivered to both steam generators, even if the rupture matrix actuates on both steam generators.

Justification for Short-Term Change The adverse effect of the short-term change is the potential for increased mass and energy release to the containment and associated increase in peak building pressure for an SLB inside containment

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when compared to the analysis described in Section 14.2.2.1.5 of the FSAR. Analysis is now being perfonned with the intent of show-l ing that adequate operator action time is available to preclude l

overpressurizing the containment in this event.

The results of l

this analysis and additional details including basis of the analy-l sis, risk assessment of short-term operation with the change, com-l parison with other operating plants, etc, will be completed by mid-l May and submitted to the NRC as soon as possible after that.

Cirector Page Three May 6,1980 Office of Nuclear Reactor Regulation Conclusions Significant improvement in the availability and reliability of the EFW system can be achieved by elimination of the Steam Line Rupture Matrix signals to emergency feedwater val ves FW V-161 and -162.

This is consistent with NRC Recommendation 4 of NUREG-0667.

How-ever, since this requires a deviation from the present FSAR Steam Line Break Analysis in Section 14.2.2.1.5, NRC Staff approval is requi red.

If you have any questions about this submittal, please contact this office.

Sincerely, FLOR DA POWER CORPORATION I

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g' George /C. Moore Vice President Power Production GCMemhT0706 cc:

Director Of fice of Inspection and Enforcement U.S. Nuclear Regulatory Commission Suite 3100 101 Marietta Street Atlanta, GA 30303

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STATE OF FLORIDA COUNTY OF PINELLAS G. C. Moore states that he is the Vice President, Power Production, of Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information and belief.

0 liM vGr. sCT Moore Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 6th day of May,1980.

% At Notary Public Notary Public, State of Florida at Large, My Commission Expires: August 8, 1983 CameronNotary 3(D12)