ML17326B223
| ML17326B223 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/04/1986 |
| From: | Youngblood B Office of Nuclear Reactor Regulation |
| To: | Dolan J AMERICAN ELECTRIC POWER SERVICE CORP., INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| References | |
| NUDOCS 8604160083 | |
| Download: ML17326B223 (5) | |
Text
Docket No.:
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Mr. John Dolan, Vice President Indiana and Michigan Electric Company c/o American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43216
Dear fir. Dolan:
Subject:
Cycle 6 Reload for Donald C.
Cook Nuclear Plant, Unit No.
2 In our review of the IMEC safety analysis for the Cycle 6 reload for the Donald C.
Cook Nuclear Plant, Unit No. 2, we have identified additional information that we need to complete the review.
The questions attached identify this information.
Please let us know if there are any questions on this matter.
Sincerely,
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Enclosure:
As stated cc:
See next page
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COOK CYCLE 6 RELOAD 1.
Section 15.4.1 of XN-NF-85-64, Rev.
1:
The analysis of uncontrolled control rod assembly withdrawal from subcritical or low power was revised.
In Section 1.0 the reason for the revision was stated to be conservative biasing of the effective delayed neutron fraction.
Provide the effectiye delayed neutron fraction used including the amount of biasing and justification for the amount of biasing that was utilized.
2.
XN-NF-85-64, Rev.
1:
The reanalysis of uncontrolled control rod assembly withdrawal from low power indicated that use of new more conservative assumptions for the delayed neutron fraction have a significant effect on the calculated results.
Provide analyses showing the effect of the more conservative delayed neutron fraction on other anticipated transients including increased steam generator cooling, loss of load, reactor coolant pump shaft failure, control rod withdrawal events during power operation and control rod drop.
3.
Table 3.2-2 of Proposed Technical Specification Changes:
This specification was added to be consistent with Section 15.4. 1 of XN-NF-85-64, Rev.
1.
Section 15.4. 1 contained an analysis of inadvertent control rod assembly withdrawal with the reactor just critical.
The initial pressurizer pressure was 2175 psia.
The initial reactor system temperature was indicated to be approximately 550'F in Figure 15.4.1-5.
The proposed technical specification Table 3.2-2 requires that the reactor coolant system temperature be less than 576.3'F.
Provide a
technical specification that is consistent with the safety analysis.
4.
Section 15.4.3 of XN-NF-85-64, Rev.
1:
This section describes the analysis of a single control rod withdrawal accident.
The analysis was performed from an initial powei level of "rated plus tolerance".
Other power levels and initial conditions were not examined.
The Exxon methodology report XN-NF-84-73 provides for a sensitivity study to determine the most severe initial conditions for the analysis of uncontrolled rod bank withdrawal at power (Section 15.4.2) and states that single control rod withdrawal at power will be calculated for the most limiting case of Section 15.4.2.
The most limiting cases of rod bank withdrawal at power in Section 15.4.2 were at 60% power and 11%
power.
Provide analyses of single rod withdrawal demonstrating that the most severe initial conditions have been selected.
5.
Section 15.4.3 of XN-NF-85-64, Rev.
1:
Withdrawal of a single control rod was calculated to damage up to 3.7K of the core.
Provide a description of the methodology used to calculate axial and radial power distribution during the transient.
Justify that the power distribution in the vicinity of the affected rod is conservatively calculated.
6.
Section 3.4.2 of Proposed Technical Specifications:
Action state (a) provides for the placement of an RHR loop in service, if no pressurizer code safety valve is operable since the RHR system is equipped with pressure relief.
This requirement also appears in the Standard Technical Specifications for Westinghouse Plants (NUREG-0452).
A second action (b) in the proposed technical specifications does not appear in NUREG-0452.
Action (b) requires that power be removed from the motor breakers of the SI pumps.
The SI pumps cannot discharge against reactor system pressure at the safety valve setpoint and would be unaffected by failure of the pressurizer safety valves to open.
The disabling of the SI pumps appears unnecessary and a source of possible operator error and loss of ECCS pumps when required.
Justify that Action (6) improves plant safety.