ML20044B866
| ML20044B866 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/08/1993 |
| From: | Hagan R WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20044B867 | List: |
| References | |
| NA-93-0054, NA-93-54, NUDOCS 9303110181 | |
| Download: ML20044B866 (3) | |
Text
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WipLF CREEK NUCLEAR OPERATING CORPORATION 3
Robert C. Hagan Vice PresvJerit Nuclear Assurance March 8, 1993 i
NA 93-0054 i
U. S. Nuclear Regulatory Commission ATTN Document Control Desk Mail Station P1-137 Washington, D. C.
20555 References 1)
Letter NA 93-0013 dated January 28, 1993 j
from R. Hagan, WCNOC, to USNRC 2)
Letter NA 92-0073 dated October 28, 1992 from R. Hagan, WCNOC, to USNRC l
Subject:
Docket No. 50-482:
Change to Proposed Revision to Technical Specifications for Cycle 7 I
Gentlemen:
}
6 This letter transmits. a revision to the Reference 2 application for j
amendment to Facility Operating License No.
NPF-42 for Wolf Creek Generating Station (WCGS), Unit 1.
These changes are being proposed as a j
result of Nuclear Regulatory Commission (NRC) review and comments received in a February 19 and February 24, 1993, telephone conference.
These proposed changes are requested concurrently with the ' Reference l' proposed changes.
The Attachment to this letter provides marked up pages to supersede those submitted in Reference 2, as' appropriate.
Pages III, 2-1 and B 2-5 that t
were provided in Reference 2 with changes, no longer-require any revision and should be removed from the proposed change package.
The attached change pages raflect the addition of a revised Figure 2.1-1,
" Reactor Core-
-i Safety Limit - Four Loops in Operation". back into the-technical I
specifications versus the Core Operating Limits Report (COLR): maintaining the positive moderator temperature coefficient in Technical Specification 3/4.1.1.3'and adding Figure - 3.1-1, "BOL Moderator Temperature - Coefficient-versus Power Level" to the technical specifications versus putting this i
in mation in the COLR; addition of the Reactor Coolant System flow rate va..e to Table 3.2-1, "DNB parameters" versus locating this information in j
the COLR: maintaining the new and spent fuel enrichment values in Technical Specification 5.6 versus putting the information in the.COLR; and adding pertinent references used to support the COLR in the revised Technical-Specification 6.9.1.9.
The related table of contents and Bases changes are-also included in the Attachment.
as well as a typographical error correction.
These additional changes do not affect the basis - for the safety evaluation, no significant hazards consideration determination, or environmental impact determination provided in Reference 2.
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9303110181 930308 PDR ADOCK 05000482 P
PDR PO Box 411/ Burbngton, KS 66829 / Phone: (316) 364-8831 w
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An Equal Opponunity Employer M TMC/ VET i
3
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i NA 93-0054 l
Page 2 of 2 l
l requested in the February 19. 1993 telephone f
Additional,information was conference in two areas.
The first request was to ensure that the proposed i
change in Reference 2 concerning Table 3.7-2, " Steam Line Safety Valves per l
Loop
- to go from a lift setting of +12 to a value of +3% would not conflict i
with WCGS ASME Code compliance.
Ihe setpoint value is not specified in ASME Section XI and is determined by the Owner.
Therefore, this change does not conflict with ASME Section XI requirements.
The second request was to describe the process - used to ensure analysis results are reflected in plant surveillances.
Design parameters utilized l
in a reload design are controlled through several internal documents.
The i
Reload Design Initialization Checklist (RDIC) includes parameters ranging l
from enrichment to pump performance.
The RDIC is a controlled document which is reviewed each cycle by Operations, Licensing Engineering, Safety Analysis, Core Thermal-Hydraulic Analysis, and Core Design personnel.
Extensive reviews, documented in the RDIC, are performed to insure that I
correct plant input parameters are utilized in the design calculations.
The Reload Safety Analysis Checklist (RSAC) is another interface document which provides the mechanism for cycle-specific verification that the current safety analysis remains bounding.
Kinetics parameters and accident i
specific parameters are checked for each reload cycle.
Should the analysis require a plant setpoint change or other plant modification to ensure the i
plant configuration remains within the bounding analysis, the plant item is j
changed through the Plant Modification Request (PMR) process.
The PMR requires extensive reviews throughout Engineering and Operations prior to implementation, and includes items such as engineering studies, component modification orders, and affected procedures.
In accordance with 10 CFR 50.91, a copy of this submittal is being-provided to the designated Kansas State Official.
)
If you have any questions concerning this matter, please contact me at l
(316) 364-8831 extension 4553 or Mr. Kevin J. Moles of my staff at i
extension 4565.
l Very truly yours,
(~%
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l
&W obert C. Hagan Vice President Nuclear Assurance RCH/jad Attachment cc:
G. W. Allen (KDHE), w/a W.
D. Johnson (NRC), w/a J. L. Milhoan (NRC), w/a G. A. Pick (NRC), w/a V. D. Reckley (NRC), w/a 1
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i STATE OF KANStdi
)) ss COUNTY OF C0FFEY
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i Robert C. Hagan, of lawful age, being first duly sworn upon oath says that he is Vice President. Nuclear Assurance of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the content thereof; that he has executed that same for and on behalf of_ said Corporation with full power and authority to do so; and that the facts
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therein stated are true and correct to the 'best of his knowledge, information and belief.
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j day of N 9 kCkA, 1993.
SUBSCRIBED and sworn to before me this i
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Notary Tublic N.Sb Expiration Date i
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