ML17332A683
| ML17332A683 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/15/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17332A682 | List: |
| References | |
| NUDOCS 9503200317 | |
| Download: ML17332A683 (7) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
193 TO FACILITY OPERATING LICENSE NO.
DPR-58 AND AMENDMENT NO.
179 TO FACILITY OPERATING LICENSE NO.
DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NOS.
1 AND 2 DOCKET NOS.
50-315 AND 50-316
- 1. 0 INTRODUCTION By letters dated January 17, 1994 (Ref.
1) and February 10, 1995 (Ref. 3), the Indiana Michigan Power Company (the licensee) requested amendments to the Technical Specifications (TS) appended to Facility Operating License Nos.
DPR-58 and DPR-74 for the Donald C.
Cook Nuclear Plant, Unit Nos.
1 and 2.
The proposed amendments would increase the TS limit for control rod misalignment below, at, and above 85% rated thermal power.
The February 10, 1995 submittal clarified a potential inconsistency between ACTION statement
- 3. 1.3. I.c.2.d, which requires a power reduction to 75%,
and the possibility that the proposed amendment may allow a greater misalignment below 85% power than that allowed above 85% power.
The licensee indicated in its submittal that experience at D.
C.
Cook with the Analog Rod Position Indication (ARPI) System has shown that indicated rod misalignment could be greater than 12 steps.
When this occurs, as per TS action item 3.1.3.2, an incore flux map must be taken every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The licensee pointed out that at D.
C.
Cook, to ensure compliance with the action statement, flux maps are taken every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to verify the actual location of the rod In all cases the flux maps have shown that there was no actual rod misalignment.
In addition, the licensee stated that misaligned rods can be detected during normal operation (via changes in power, quadrant tilt or flux) and are realigned using established procedures.
It is the licensee's contention that this excessive use of flux mapping will lead to increased maintenance and possibly an ALARA [as low as is reasonably achievable]
concern Consequently, changing the TS to allow 18 steps misalignment will reduce t'e usage of the flux mapping system.
- 2. 0 EVALUATION 2.1 Anal sis Method The licensee performed neutronics analyses to evaluate the impact of rod cluster control assembly (RCCA) misalignment on 'steady-state power distributions and normal operational transients, such as load-follow 95032003i7 9503i5 PDR ADOCK 050003i5 P
PDR operations.
The principal tool used in these calculations is the Westinghouse Advanced Nodal Computer Code (ANC) (Ref. 2).
All the analyses utilizing ANC were performed in a three-dimensional mode, while full-core and quarter-core models were used for the analyses.
All calculations were performed by Westinghouse using approved methodologies.
Discrete pin power and pin burnup information was obtained from )NC calculations.
The licensee provided tabulated data regarding Feo " and fsso'",
where it was pointed out that it is the changes in Fe values that are op major concern rather than absolute values of F,.
The licensee used the Unit 2 cycle 9 ANC model for the rod position indicator
- analysis, since the Unit 2 core has axial blankets and therefore larger peaking factors can be expected.
'To verify the ANC code, the licensee used the Unit 2 cycle 9
ANC model to deplete the core and compare the results of the power distribution and boron letdown calculations with those of measured values.
2.2 Misali nment Calculations 2.2. 1 85X Rated Thermal Power (RTP)
To determine power levels at which peaking factors increase due to RCCA misalignment, the licensee assumed misalignment of 30 steps (18 step indicated
+ 12 step uncertainty) for calculations from the power dependent insertion limit (PDIL).
The licensee analyzed misalignment of groups of RCCAs in the control bank (groups 1
and 2 in control bank D) since it is more probable that the RCCAs in one group would misstep rather than different RCCAs from different groups would misstep.
However, single RCCA misalignment calculations analysis was also performed.
In particular, the misalignment of H-8 was investigated since this RCCA is in the middle of the core and.is physically very difficult to monitor with an excore detector.
Analysis conducted by the licensee showed that peaking factors do not change substantially with power change as long as the reactor operation is on Constant Axial Offset Control.
Data provided by the licensee for power at or below 85X power showed that at BOL [beginning of life] for single rod misalignment and for group misalignment, the maximum increase in F~ is 4.7X and the maximum increase in F, is 2.8X.
For the same power range, but at HOL
[middle of life], similar calculations'ave a maximum increase of 6.6X F~ and a Oe7X increase in F~.
EOL [end of life] calculations showed an increase of 7.4X and 0.9X in F, and F~, respectively.
The licensee carried out other sensitivity studies for similar RCCA misalignments from different D-bank positions and other than PDIL.
These analyses showed an increase of 7.3%%d in F~ and an increase of 2.7X in F~.
Full core analysis of increases in the peaking factors were conducted.
The rods H-12 and D-12 were misaligned 30 steps from ARO [all rods out] and PDIL resulting in an increase of 1.7X in F, and an increase of 0.9X in F,.
These increases in the peaking factors are well within the 15X at 85Xo RTP; consequently, an 18-step misalignment up to 85X RTP can be tolerated.
2.2.2 100X RTP The licensee also performed sensitivity misalignment studies of control bank D
at 100X RTP.
Data for 24 steps (12 step indicated
+ 12 step uncertainty),
27 steps (15 step indicated
+ 12 step uncertainty),
and 30 steps (18 step indicated + 12 step uncertainty),
misalignment shows a maximum increase in F~
of 8.9X and a maximum increase in F~ of 4.1X at EOL.
A similar misalignment case was performed from a D-bank position of 215 steps rather than from PDIL.
The result was a very minor change in the peaking factors, similar to those changes observed due to misalignment for PDIL.
The result of the analysis indicated that the increase in F~ due to an additional misalignment of three steps over the existing 12 step (indicated) misalignment was found to be 0.9X.
The increase in F
due to an additional six steps (indicated) misalignment was found to be I.LX.
Finally, load-follow calculations showed an increase in F, of 3.4X.
To be more conservative, and since the analysis was performed for a typical cycle, the licensee has chosen to increase the margin on F~ to 6X.
The implication is that provided that a
6% mar in in F exists, then an additional misalignment of six steps (that is a total of 18 steps indicated) is allowed at 100X RTP.
Similar analysis for F+ showed an increase of 0.&X.
Review of the analyses conducted by the licensee indicates that if the requirements given in the TS 3.1.3.1 are satisfied, TS limits on F
and F~ will not be violated even with an 1& step indicated misalignment.
consequently, the staff concurs with the licensee, that if TS 3. 1.3. 1 limiting conditions for operation are not violated, the initial conditions assumed in the accident analyses for Cook 1 and 2 are not
- violated, and the proposed change from 12 to 18 step (indicated) is therefore acceptable.
- 2. 3
~Summar The NRC staff has reviewed the reports submitted by the licensee for the operation of Cook Units 1 and 2 and finds the proposed changes to TS 3/4. 1.3 to be acceptable.
2.4 Administrative Chan e
The staff noted that the words "in the core" were added to Unit 1 TS 4.1.3.1.2 after the words "Each full length rod not fully inserted".
This change is consistent with the wording for Unit 2, more accurately reflects the intent of the specification and is therefore acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Hichigan State official was notified of the proposed issuance of the amendments.
The State official had no comments.
- 4. 0 ENVIRONMENTAL CONSIDERATION The amendments change the requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements.
The staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released
- offsite, and that there is no significant increase in individual or, cumulative occupational radiation exposure.
The Commission has previously issued a
proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding (59 FR 10008).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(I) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
6.0 REFERENCES
1.
E.E. Fitzpatrick, Indiana Michigan Power Company, letter to U.S. Nuclear Regulatory Commission, Document control desk, Washington, DC, 20555, January 17, 1994.
2.
Westinghouse Electric Company, "ANC A Westinghouse Advanced Nodal Computer Code",
WCAP-10965-P-A, December 1985. [Proprietary information.
Not publicly available.]
3.
E.E. Fitzpatrick, Indiana Michigan Power Company, letter to U.S. Nuclear Regulatory Commission, Document control desk, Washington, DC, 20555, February 10, 1995.
Principal Contributor:
A. Attard Date:
March 15,
- 1995,
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