ML19341C559

Responds to Co Pirg Questions Re Facility.Util Has Not Experienced Problems W/Fuel Particle Coatings

ML19341C559
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/20/1981
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Kuzmycz G Office of Nuclear Reactor Regulation
References
P-81061, NUDOCS 8103030744
Download: ML19341C559 (5)

Text

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public senice company of Cdende February 20, 1981 0 $b ,

Mr. George Kuzmyc:

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U. S. Nuclear Regulatory Commission Division Project Management

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(.'y Q Washington, D. C. 20555 4 y ocket No. 50-267

SUBJECT:

Response to Questions Submitted by John Loges

Dear Mr. Kuzmyc:

We have reviewed the questions submitted by Mr. John Loges, Colorado Public Interest Research Group. Because some of the questions are not clear we may not have addressed the question, but we have attempted to provide some interpretation of questions to the best of our ability.

1. If the reactor is pushed toward its rated capacity, will the efficiency change comparably?

PSC Response:

The overall thermal efficiency of the unit increases with increasing power levels. It should be noted, however, that the unit operates at an overall thermal efficiency of approximately 38?; when the unit is at 70?; power. Increasing the power level to 100% will only serve to increase the overall thermal efficiency by approximately 1% (i .e. , 39?; overall thermall efficiency).

2. Since increased efficiency would apparently require

. increased temperature, do you perceive a prcblem with

, continuing shrinkage?

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PSC Response:

We do not understand the intent of the question with reference to

" shrinkage" versus increased temperature. If the question refers to normally expected fuel element shrinkage it should be pointed out that the shrinkage phenomenon is not related to temperature, but rather is a phenomenon resulting from neutron flux.

Shrinkage of fuel element , which is accounted for in the design

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I will occur whether the plant is operated at 70*; power or 100?;

power.

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3. Has the proposed Emergency Safety Tower been cited, and, if so, on what considerations has the location been chosen?

PSC Response:

Again we cannot directly determine what is meant by " emergency safety tower" as we have no such terminology in our emergency plans. If the reference is directed toward emergency response facilities, PSC has located the response facilities, PSC has located the on site emergency response facility in an area adjacent to the plant. This facility is nearing completion and should be available for use by mid March 1981. In the interim a tempora ry on site emergency response facility har been established and is operational.

If the question refers to the near site emergency response facility, PSC has located this facility in Fort Lupton.

4. Will the Tower be constructed and operative before the plant is pushed beyond 70%?

PSC Response:

As indicated in question 3 above the emergency response facilities are operative at the present time. We expect further guidance to be issued by the NRC on the various emergency response facilities, but to date such guidance has not been issued. The existing facilities, however, have been reviewed by

the NRC and do meet the immediate action guidelines resulting from TMI-2.

5. Have PSCo and other public officials and authorities complied fully with the provisions of the latest NRC regulations?

PSC Response: i PSC has complied wi t h the various regulations and guidance issued by the NRC to the extent that such guidance which was developed for light water reactor technology is applicable to gas cooled reactor technology.

6. When will there be evacuation drills to determine whether new safety plans are in fact practical?

PSC Response:

A combined PSC-State of Colorado drill was conducted on February ,

28, 1980 to test the feasibility of the emergency response plans. l While it is impractical to conduct actual evacuation of people within the 5 mile emergency planning zone for Fort St. Vrain the

i drill was conducted in sufficient detail to determine that evacuation could be successfully accomplished. We did evacuate one of the schools in the area, on a cooperative basis with the School District to ensure that our planning was adequate. Drills will continue to be conducted in the future to ensure that plans are adequate, but again it is not our intent during such drills to actually evacuate the public.

It thould ~eo noted that because of its characteristics we have no postulated accident at Fort St. Vrain that would result in exceeding the protective action guidelines for evacuation of people living near the plant. We have, however, done the necessasry planning to ensure such an evacuation could be

. accomplished.

7. Have you resolved the issue of having an expert emergency technician on 30 minute call?

PSC Response:

Again because of the inherent safety features at Fort St. Vrain slightly different criteria has beec applied with reference to our Technical Advisor (expert :mcrg:ncy technician). Our Technical Advisor has a response time of up to one (1) hour.

This issue has been reviewed and accepted by the NRC.

8. What are the results of the recert computer analysis of the performance of the core of the reactor?

PSC Response:

Wit 5 cut more specifics it is difficult to determine which computer analysis is being questioned. On an overall basis core performance is well within expected design criteria.

9. Is there still a problem with the carbon cadding on the fuel rods? Are there still changes in configuration?

PSC Response:

We have not experienced any problems with the fuel particle coatings. Performance of the coatings to dat: has been excellent. I

10. If the rods do change configuration, does the change affect safety considerations?

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4 PSC Response:

If this question is directed at core configuration changes resulting from temperature fluctuations or temperature redistribution the answer is no. Although slight changes in core configuration: do occur these changes have no overall safety implications.

11. Will the capacity increases be accompanied by changes in the kind or amounts of radiation generated by the reactor?

PSC Response:

Circulating activity within the core of Fort St. Vrain is directly related to fuel performance and the integrity of fuel particle coatings. Increasing the unit capacity, while such capacity increases result in temperature increases, does not directly result in increase in circulating activity. We would expect to see some slight increase in circulating activity at higher power levels, but this increase is not directly proportional to power level.

It should be noted that the circulating activity utilized for the design basis accident is based on 30,000 C1, and that the present circulating activity is less than 285 C1. This is directly attributable to the excellent fuel performance experienced to date as well as design conservatism.

12. Of the radio-isotopes generated by the plant, how is it decided which ones to monitor? Which ones are monitored?

PSC Response:

The radio-nuclides monitored represent the total activity within the core. While we do not perform isotopic analyses for each and every nuclide the total activity circulating in the core is monitored and equivalent activities are expressed in terms of the predominate nuclides. The predominate nuclides that are utilized in cetermining dose conversion factors for accident situations are the noble gases (Kryptons and Xenons) and the ladines.

13. Given the exceptionally long time the plant has required to become commercially operational, will the aesigned life span of the plant have to be re-appraised to consider the deterioration of materials caused by aging?

PSC Response: )

The plant is designed for an overall life of 30 years based on i 1

full power operation and various design cycles. Aging of materials is far over shadowed by design cycles. The plant is

under a system of constant surveillance testing that is utilized to dete.mine component or material degradation. Based on this testing program to date we see no indications that would be cause for reducing the expected design life of the plant.

14. Wnen will the transcript of the NRCACRS meeting on Monday, January 26 become available?

PSC Response:

We have no control over this matter. This question should more appropriately be directed to the NRC.

Very truly yours, 7Y WW Don W. Warembourg Manager, Nuclear Production Fort St. Vrain Nuclear Generating Station DWW/dkm