ML19353B014
| ML19353B014 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/01/1989 |
| From: | Weiss S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19353B011 | List: |
| References | |
| NUDOCS 8912110054 | |
| Download: ML19353B014 (6) | |
Text
_ _ _ _
'e UNITED STATES E'
i NUCLEAR RE2ULATORY COMMISSION WASHINGTON, D, C. 20056 e
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P_UBLIC SERVICE COMPANY OF COLORADO DOCKET NO. 50-267 FORT ST. VRAIN NUCLEAR GENERATING STATION j
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. DPR-34 1
1 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Public Service Company of Colorado (the licensee) dated September 14 and revised on October 13, IS89, l
complies with the standards and requirements of the Atomic Energy Act l
of 1954, as amended (the Act), and the Comission's regulations set l
forth in 10 CFR Chapter I; l'
E.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Comission; r
C.
There is reasonable assurance:
(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; P
D.
The issuance of this license amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 i
of the Comission's regulations and all applicable requirements have been satisfied, i
P
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~2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendnent, and paragraph 2.D.(2) of Facility Operating License No. DPR-34 is hereby ar. ended to read as follows:
(2) Technical Specification The Technical Specifications contained in Appendices A and B, as rvvised through Amendment No. 74, are hereby incorporated in the ifcense. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION W
Seymour H. Weiss, Director Non-Power Reactor, Decommissioning and Environmental Project Directorate Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical l
Specifications Date of Issuance:
December 1, 1989 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 74 j
TO FACILITY OPERATING LICENSE NO. DPR-34 l
DOCKET NO. 50-267 Replace the following meges of the Appendix A Technical Specifications with
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the attached pages.
Tie revised pages are idtntified by amendment. number and i
contain vertical lines indicating the areas of change.
Remove Insert 6.1-5 6.1-5 6.1-Sa i
6.1-5b t
Fort St. Vrain #1 Technical Specifications Amendment No. 74 Page 6.1-5 within the spacer blocks.
The middle layer of lower reflector elements, excluding the central element in each I
core region, contains 25 weight percent boronated graphite pellets enclosed in hasta 11oy-X cans.
The top layer of reflector above the hexagonal columns contains I weight percent crushed boronated graphite.
The top layer of reflector above the permanent side reflector blocks contains I weight percent boronated graphite _ enclosed in steel cans.
I Defueling l
Defueling elements containing no fuel and made from grade l
H-091 graphite, which is equivalent to the grade HLM l
graphite used for permanent side reflector blocks, will be l
used to replace fuel elements during defueling.
The i
defueling elements have the same outer dimensions as the i
fuel elements that they will be replacing.
All defueling I
elements are of the same design, regardless of whether they l
are used to replace standard fuel elements or control I
column fuel elements from the central column of the region.
l Defueling elements do not have control rod channels or i
reserve shutdown holes.
(The control rods will net be l
l reinstalled in a defueled region.) Each defueling element i
has ninety (90) coolant channels which will align with the I
coolant channels of the elements above and below.
In j
addition, twelve (12) holes are blind crilled and loaced I
with boronated graphite lumped poison pins (LPD), with a
c Fort St. Vrain #1 Technical Specifications Amendment No. 74
[,
Page 6.1-Sa
., s
_l horon loading equivalent to about 350 ppm of homogeneous l-natural coron to replace the negative reactivity of the l
L l[
The essentially right circular cylindrical-geometry of the ll core that is modeled in the GAUGE code will be retained lj during defueling.
The outer ring of fuel elements (Regions.
ll 20-37) will be replaced with defueling elements first. The ll next most outer ring of fuel elements (Regions 8-19) will ll be replaced second. The central regions of fuel ' elements ll (Regions 2-7 and 1) will be replaced last.
l The top layer reflectors may be replaced with reflectors
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not containing boronated graphite in selected regions-l during defueling to maintain neutron count rate. Neutron l-source material, Californium 252, will also be added as l
necessary to. maintain a neutron count rate above the
-l required minimum until a demonstration of sube ri ticali t'y j;
has been performed in which all control rods have been I
withdrawn with a calculated k(ef f) not greater than 0.95
'l
_ assuming the conditions specified by SR 4.1.4 for SHUTDOWN l
MARGIN determination. After the demonstration, the startup l
chanrel low count rate red withdrawal prohibit may be l
bypassed and the minimum count rate need not be maintained.
l However, the startup channel high count rate scram will l
remain in service to respond to any approach to I
criticality.
J'[4.
viiirt Tt, Ura ta et
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Technical. Specifications b,
Amendment No. 74
- J Page 6.1-Sb Basp for Specification OF 6.1 The above specifications form the general design bases and criteria for the overall design features of the reactor core which were used to evaluate its general performance.
l Further details concerning these design features are given
_ ll in Section III of the FSAR, the Safety Analysis Report for Fort St. Vrain Reload 1 Test Elements FTE-1 through FTE-8,'
l General Atomic Document GLP-5494, June 30, 1977, and the Ll Safety Analysir Report For ' Reactor Oefueling, General l
Atomics Document GA-C19694.
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