ML20009A718
| ML20009A718 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 07/08/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Loges J AFFILIATION NOT ASSIGNED |
| References | |
| NUDOCS 8107140006 | |
| Download: ML20009A718 (7) | |
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. J y-JUL 8 1981 DISTRIBUTION:
MJ J 33
__ Docket.f o Mt p:NRC?PDR a S ocal PDR t
Docket No. 50-267 SSPB Reading JRMill er GKuzmycz LTremper Mr. John Loges OELD Colorado Public Interest 01&E(3)
Research Group ACRS(16) 1530 Tenth Avenue TERA Greeley, Colorado 80531 NSIC G. McKinley, ACRS
Dear Mr. Loges:
R. Black, OELD
'In response to your questions dated January-27,1981 dealing with the Fort i
St.:Vrain Nuclear Generating Station, please find two enclosures. The first provides answers to the questions which you submitted at the ACRS meeting.
The second is Amendment No. 23.'to the Fort. St. Vrain Technical Specifications.
Amendment No. 23 delas with some of the topics that you were interested in and provides background information on the rise to power of Fort St. Vrain.
t I~ apologize for the delay which was due to pmcessing both the Amendment and answering your questions. Should you require any further infomation, do not hesitate to write or call the project manager, George Kuzmycz, Division of Licensing Special Projects, Washington, D.C.
20555, (301) 492-8198.
Sincerely, Original signed by
& L. Todeso)
Robert L. Tedesco, Assistant Director for Licensing Division of Licensing N
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Docket No.: 50-267 Mr. John Losas Colorado Public Interest Research Group
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1530 Tenth Avenue Greeley, Colorado 80631 Dea \\ Hr. Loges:
r In re onse to your questions dealing with the Fort St. Vrain nuclear generat g station, please find two enclosures. The first 'atteupts to answer th questions which you submitted at' the ACRS meeting, while the second is As idment No. 23 to the Fort St. Vrain Technical Specificationt..
Amendment No.. deals with some of thejlopics that you were interested i
and provides kground information/on the rise to power of Fort St.
Vra i n.
I apologize for the delay ut you must realize that issuing the Amendment took some time awa frony' answering your questions. Should you require any further inforr ion, do not hesitate to write or call the project manager, George /
sqycz, Division of Licensing.
Special Projects, Washington, D.C. ' rSS.
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Robert L. Tedesc Assistant Di ector for Licensing Divisior, of Licensin '
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Mr. O. R. Lee, Vice President Senior Vice President Electric Production Public Service Company Public Service Company of of Colorado Colorado P. O. Box 840 P. O. Box 840 Denver, Colorado 80201 Ocnver, Colorado 80201 James B. Graham, Manager Licensing and Regulation East Coast Office General Atomic Company 2021 K Street, N.W.
Suite 709 Washington, D. C.
20006 Mr. J. K. Fuller, Vice President Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201 Mr. W. Dickerson NRC Resident Inspecter 16805 Weld County Road 191/2 Platteville, Colorado 80651 Director, Division of Planning Department of Local Affairs 615 Columbine Building 1845 Sherman Street Denver, Colortdo 80203 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative, Radiation Programs Environmental Protection Agency 1860 Lincoln Street Denver, Colorado 80203 Mr. Don Warembourg Nuclear Production Manager Public Service Company of Colorad9 16805 Weld County Road 191/2 Platteville, Colorado 80651 1
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ENCLOSURE 1 If the reactor is pushed toward its rated ' capacity, will the efficiency 1.
change comparably?
Response
The overall thermal efficiency of the unit increases with increasing power It should be noted, however, that the unit operates at an overall levels.
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 thennal efficiency by approximately 1% (i.e., 39% overall thennal efficiency).
2.
Since increased efficiency would apparently require increased temperature, do you perceive a problem with continuing shrinkage?
Response
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 elements will occur whether the plant is operated at 70%
swer or 100% power, and has been accounted for in the design.
3.
Has the proposed Emergency Safety Tower been sited, and, if so, on what considerations has the location been chosen?
Response
If the reference is directed toward emergency response facilities, PSC has located the response facilities in an area adjacent to the plant. This facility building is complete and should be available for use by fall of 1981.
In the interim a temporary onsite emergency response facility has been established and is operational.
If the question refers to the near site emergem:y response facility, PSCo has located this new facility in Fort Lupton.
The locations for either facility have been detennined using the guidelines provided by the NRC for emergency planning.
4.
Will the tower be constructed and operative before the plant is pushed beyond 70%?
Response
The emergency response facility building is complete and the ventilation equipment, communication equipment, various drawings and documents are installed.
Instrumentation is in the process of being installed and should be complete in the Fall of 1981. PSCo plans to operate Fort St.
Vrain at 70% power until the testing up to 100% is finished and the NRC reviews the results and approves steady state 100% operation. We expect this review to be completed sometime in December of 1981.
4 5.
Have PSCo and other public officials and authorities complied fully with the provisions of the latest NRC regulations?
Response
PSCo enmplied with 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.
The State of Colorado, in relation to their dealings with PSCo and Fort St.
Vrain, has also complied with NRC regulations on an acceptable Emergency Plan.
6.
When will there be evacuation drills to deten-..t whether new safety plans are in fact practical?
Retponse:
A combined PSC-State of Colorado drill was conducted on February 28, 1980 to test the feasibility of the emergency response plans. While it is im-prtctical to conduct actual evacuation of people within the 5 mile emergency planning zone for Fort St. Vrain, the drill was conducted in sufficient detail to determine that evacuation could be successfully accomplished.
One of the schools in the area was evacuated on a cooperative basis with the School District to ensure that planning was adequate. Drills will continue to be conducted in the future to ensure that plans are adequate, but again it is not the intent of such drills to actually evacuate the j
J public.
7.
Have you resolved the issue of having an expert emergency technician on 30 minute call?
Response
Because 'f the inherent safety features of Fort St. Vrain, slightly different criteria have been applied with reference to the Technical Advisor (expert emergency technician). The PSCo Technical Advisor j
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 recent computer analysis of the performance of the core of the reactor?
Response
The results of the computer analysis are presented in Section 4.0 of the enclosed amendment.
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9.
Is there still a problem with the carbon cladding on the fuel rods?
Are there still changes in configuration?
10.
If the rods do change configuration, does the change affect safety considerations?
Response
The reference, we assume, is to the coatings on the fissile particles that were manufactured slightly thicker than specified. Due to this increase in overall particle voluma, a problem arose in squeezing enough particles into a fuel rod to yield a thorium to uranium ratio of 4.25 to 1.
The NRC staff has reviewed the above situation and presented this analysis as Section 8.0 " Design Features" of Amendment No. 22. This amendment presents the staff's position:
" Based on our evaluation of the information supplied, we concluded that there is reasonable assurance that a decrease in the nominal Th:U ratio from 4.25:1 will result in negligible changes in fuel particle performance during normal operation and design basis accident conditions and that the technical specifications change is, therefore, acceptable."
- 11. Will the capacity increase be acecmpanied by changes in the kind or amounts of radiation generated by the reactor?
Response
Circulating activity within the core of the Fort St. Vrain reactor 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 an 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 Ci, and that the present circulating activity is less than 285 Cf. This is directly attributable to the 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 a e monitored?
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Response
The radio-nuclides monitored represent the total activity within the core.
1 While PSCo ooes not perform isotopic analyses for each and every nuclide, the total activity circu? sting in the core is monitored and equivalent activities are expressed in terms of predominate nuclides.
The predominate nuclides that are utilized in determining dose conversion factors for accident situations are the noble gases (Kryptons and Xenons) and the Iodines.
- 13. Given the exceptionally long time the plant has required to become,
commercially operational, will the designed life span of the plant 4
-have to be re-appraised to consider the deterioration of materials caused by aging?
Response
The Fort St. Vrain reactor is designed for an overall life of 30 years based on full power operation and various design cycles. The plant personnel follow a predetermined system of surveillance testing that is used to monitor plant, conponent and material performance. Various components in the plant do not have a 30 year life and are replaced as their useful life approaches; for example, valve packing is replaced periodically.
- 14. When will the transcript of the NRC-ACRS meeting on Monday, January 26, becone available?
Response
The transcript of the ACRS meeting is available at the local Pubite Document Room in Greeley, Colorado.
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