ML20245A376
| ML20245A376 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 04/13/1989 |
| From: | Withers B WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20245A378 | List: |
| References | |
| WM-89-0114, WM-89-114, NUDOCS 8904250174 | |
| Download: ML20245A376 (8) | |
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NUCLEAR OPERATING CORPORATION Bart D. Withers President and Chief Executive Offcer April 13, 1989 WM 89-0114 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk i
Mail Station F1-137 Washington, D. C. 20555 Subj ect:
Docket No. 50-482: Revision to Technical Specifications 3.1.3.4 and Figure 3.1 Fully Withdrawn Position l
for Rod Cluster Control Assemblies l
Gentlemen:
The purpose of this letter is to transmit an application for amendment to Facility Operating License No.
NPF-42 for Wolf Creek Generating Station (WCGS), Unit No.
1.
This license amendment request proposes revising Technical Specifications 3.1.3.4 and Figure 3.1-1 to change the fully withdrawn position of the Rod Cluster Control Assemblies (RCCA) for WCGS to a range of 222 to 231 steps, inclusive.
A complete Safety Evaluation and No Significant Hazards Consideration determination are provided as Attachments 1 and II respectively.
The proposed changes to the Technical Specifications are provided as Attachment III.
In accordance with 10 CFR 50.91, a copy of this application, with attachments is being provided to the designated Kansas State official. This proposed revision to the WCGS Technical Specifications will be fully implemented within 30 days of formal Nuclear Regulatory Commission approval.
8904250174 890413 PDR ADOCK 05000482
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P FDC 1
P.O. Box 411/ Burlington, KC 66839 / Phone: (316> 364-8831 An Equal Opport iruty Employer MMfCNET
WM 89-0114 Page 2 of 2 April 13, 1989 If you have any questions concerning this matter, please contact me or Mr. O. L. Maynard of my staff.
Very truly yours, Bart D. Withers President and Chief Executive Officer BDW/j ad Attachments I - Safety Evaluation II - Addressing The Standards In 10 CFR 50.92 III - Proposed Technical Specification Changes cc:
G. W. Allen (KDRE), w/a B. L. Bartlett (NRC), w/a E. J. Holler (NRC), w/a R. D. Martin (NRC), w/a D. V. Pickett (NRC), w/a
STATE OP KANSAS
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) SS COUNTY OF COPPEY
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l Bart D. Withers, of lawful age, being first duly sworn upon oath says that he is President and Chief Executive Officer 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 l
therein stated are true and correct to the best of his knowledge, information and belief.
-_-.s Bart D. Withers President and Chief Executive Officer SUBSCRIBED and sworn to before me this /)
day of 1989.
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ATTACHMENT I SAFETY EVALUATION
Attachment I to WM 89-0114 Page 1 of 2 SAFETY EVALUATION Description of Amendment Request This license amendment request proposes to revise Technical Specifications 3.1.3.4 and Figure 3.1-1 to change the fully withdraw position of the Rod Cluster Control Assemblies (RCCA's) for Wolf Creek Generating Station to a range of 222 to 231 steps, inclusive.
Evaluation Wolf Creek Generating Station (WCGS) Technical Specifications require all i
shutdown rods to be fully withdrawn and all control rod banks to be withdrawn in accordance with Figure 3.1-1.
Past operational history at WCGS has shown that long periods of operation with the control rods withdrawn 'to 228 steps has lead to control rod wear by fretting against the upper internals guide surface due to flow induced vibration. In order to minimize the effect of this control rod wear, axial repositioning of the control rods can be used to eliminate further degradation at locations where control rod wear has been observed.
This Technical Specification change would allow l
axial repositioning between 222 steps and 231 steps withdrawn.
With the RCCAs positioned at 222 steps withdrawn, the tips of the RCCAs will be approximately 3 steps (1.875 inches) into the active fuel region.
Based on the Cycle 4 design, at beginning-of-life (BOL),
the boron concentration will be approximately 2 ppm lower (vs.
rods at 228 steps), the axial peak power (Fz) will increase by approximately 0.6% and axial offset (A.O.)
will be about 1% more negative.
Due to the burnup effects discussed above, the reactivity impact will be larger at end-of-life (EOL).
At EOL, the boron concentration will be approximately 6 ppm lower (vs.
rods at 228 steps).
The axial peak power (Fz) and A.O. will be only minimally impacted at E0L by depleting the core with the RCCAs positioned at 222 steps withdrawn and after pulling the RCCAs out to any position between 222 and 231 steps withdrawn at EOL.
Based on a review of non-LOCA safety analyses, it is concluded that there is no impact on the non-LOCA rafety analyses due to the proposed Technical Specification changes for the fully withdrawn RCCA position.
The position of the control rods relative to each other will not change, and thus the Updated Safety Analysis Report (USAR) conclusion that the DNBR design basis acceptance criteria is met for Condition II events remains valid.
The RCCA repositioning does not invalidate the Technical Specification on control rod drop time limit of 2.2
- seconds, and thus the maximum rod drop time assumption used in non-LOCA safety analyses remains valid.
The proposed change also does not invalidate any tripped rod characteristics used in non-LOCA safety analyses, and thus modeling of transient reponse following reactor trip remains valid. Based on the above information, it is concluded that the proposed RCCA repositioning does not affect any non-LOCA transient behavior and the results and conclusions presented in the Wolf Creek USAR remain valid.
. Attachment I to WM 89-0114 Page 2 of 2 The small and large break LOCA analyses were evaluated to determine any possible impact by this Technical Specification change. For the small break
- analysis, it is assumed that the core is brought to suberitical by the trip reactivity insertion of the control rods.
Since the maximum rod drop time has not changed due to the RCCA repositioning, there will be no effect on i
the USAR small break LOCA analysis.
The large break analysis does not take l
credit for negative reactivity introduced by the control rods.
During a large break LOCA, the reactor is assumed to be brought to subcritical by the presence of voids in the core caused by the rapid depressurization of the RCS.
Siner credit is not taken for the control rods, there will be no effect on ti e USAR large break analysis for RCCA repositioning.
I A' review et the impact of the RCCA repositioning on the RCCL design and l
performance requirements has been performed.
Changing the fuily withdrawn definition to a raage of 222 to 231 steps will not pose a challenge to Control Rod Drive Mechanism (CRDM) operation.
Mechanically, the repos!.tioning will not effect the operation of the CRDMs and does not represent a design or operational issue. At 231 steps, the RCCAs are still engaged in the top of the fuel assembly which will allow for a smooth rod drop.
The above safety evaluation has determined that the current safety analysis design basis continues to be met for repositioning the RCCAs to fully withdrawn in the range of 222 to 231 steps withdrawn.
Based on the above discussions and the considerations presented in Attachment II, the proposed revision to the WCGS Technical Specifications does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report; or create a possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report; or reduce the margin of safety as defined in the basis for any technical specification.
Therefore, the proposed revision does not adversely affect or endanger the health or safety of the general public or involve a significant safety hazard.
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ATTACHMENT II ADDRESSING THE STANDARDS IN 10 CFR 50.92 l
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Attachment II to WM 89-0114 Page 1 of 1 Addressing The Standards In 10 CFR 50.92 This license amendment request proposes to revise Technical Specifications 3.1.3.4 and Figure 3.1-1 to change the fully withdrawn position of the Rod Cluster Control Assemblies (RCCA's) for Wolf Creek Generating Station (WCGS) to a range of 222 to 231 steps, inclusive.
The following sections discuss proposed changes under the three 10 CFR 50.92 standards:
Standard 1
Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated This license amendment request will allow axial repositioning of the control rods to minimize the effect of control rod wear by eliminating futher degradation at locations at which wear has been observed.
The probability and consequences of accidents and transients previously evaluated in the Updated Safety Analysis Report, including LOCA and non-LOCA events, have been evaluated and/or reanalyzed and are not affected by the proposed changes.
Therefore, the proposed changes would not involve an increase in the probability of occurrence of an accident previously evaluated.
Standard 2 - Create the Possibility of a New or Different Kind of Accident from Any Accident Previously Evaluated i
The proposed Technical Specification changes do not create any new failure modes from those assumed in the accident analyses. The accidents assumed to occur at the previous fully withdrawn position are the same as those for the prosposed position range of 222 to 231 steps.
No changes to the operation of the control rod d rive mechanism are associated with the change.
The re fo re,
the possibility of a new or dif ferent kind of accident would not be_ created by the proposed changes.
Standard 3 - Involve Significant Reduction in a Margin of Safety The proposed Technical Specification changes do not affect the rod drop time limit or the control rod bank and rod insertion limits. The affected safety analyses have been evaluated, and it has been determined that all applicable safety criteria are met with no significant adverse affects on analyses results.
Therefore, the margin of safety as defined by the USAR, safety
- analyses, and the Technical Specification Bases would not be significantly reduced.
Based on the above discussions and those presented in Attachment I, it has been determinded that the requested Technical Specification revisions do not involve a significant increase in the probability or consequences of an accident or other adverse condition over previous evaluations; or create the possibility of a new or different kind of accident over previous evaluations; or involve a significant reduction in a margin of safety.
Therefore, the requested changes do not involve a significant hazards consideration.