ML20041F709
| ML20041F709 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 03/05/1982 |
| From: | Caruso R Office of Nuclear Reactor Regulation |
| To: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| References | |
| TASK-15-08, TASK-15-8, TASK-RR LSO5-82-03-042, LSO5-82-3-42, NUDOCS 8203170321 | |
| Download: ML20041F709 (7) | |
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March 5,1982 Docket No. 50-29 LS05-82-03-042 4
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Mr. James A. Kay 2-17Jggg 7
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Dear Mr. Kay:
SUBJECT:
YANKEE - SEP TOPIC XV-8, CONTROL ROD MISOPERATION By letter dated January 4,1982, you submitted a safety assessment report for the above topic. The staff has reviewed this assessment and our conclusion is presented in the enclosed safety evaluation report, which completes this topic for the Yankee Nuclear Power Station.
This evaluation will be a basic input to the integrated assessment for your facility. The evaluation may be revised in the future if your facility design is changed or if NRC criteria relating to this topic is modified before the integrated assessment is completed.
Sincerely, Ralph Caruso, Project Manager Igoy Operating Reactors Branch No. 5
,5 Division of Licensing
Enclosure:
ud As stated DSLA SDOI cc w/ enclosure:
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Docket No. 50-29 Mr. Janes A. Kay cc Mr. Janes E. Tribble, President Yankee Atomic Electric Company 25 Research Drive Westborough, Massachusetts 01581 Greenfield Conmunity College 1 College Drive Greenfield, Massachusetts 01301 Chai'rma n Board of Selectmen Town of Rowe Rowe, Massachusetts 01367 Energy Facilities Siting Council 14th Floor One Ashburton Place Boston, Massachusetts 02108 U. S. Environmental Protection Agency Region I Office ATTN: EIS COORDINATOR JFK Federal Building Boston, Massachusetts 02203 Resident Inspector Yankee Rowe Nuclear Power Station c/o U.S. NRC Post Office Box 28 Monroe Bridge, Massachusetts 01350 l'
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n YANKEE NUCLEAR POWER STATION TOPIC XV-8 CONTROL ROD MIS 0pERATION Introduction Control Rod Misoperation events include (1) inadvertently withdrawing one or several rods: (2) leaving one or several rods behind during cank withdrawal and (3) inserting one or several rods inadvertently. The Standard Review Plan also states that " single failures in equipment or errors in operation" should be considered. The withdrawal events add reactivity to the core causing an increase in core power, heat flux and coolant temperature and pressure and thus reduction in margin to fuel design limits on DNBR and fuel centerline temperature. The dropped rod events cause an initial decrease in core power, pressure and temperature.
Returning the reactor to normal power results in a distorted core power distribution with a reduction in margin to DNBR.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license prnvide an analysis and evalua-tion of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility.
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Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10 " Reactor Design" requires that the core, and associated coolant, control and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrences.
GDC 15 " Reactor Coolant Systems Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurences.
GDC 20 " Protection System Functions" requires that the protection system be designed to initiate automatically the operation of reactivity control systems to assure that specified acceptable fuel, design limits are not exceeded as a result of anticipated operational occurrences.
GDC 25 " Protection System Requirements for Reactivity Control Malfunctions" requires that specified acceptable fuel design limits not be exceeded for any single malfunction of the reactivity control systems such as accidental withdrawal of control rods.
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b, III.
RELATED SAFETY TOPICS
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Topic IV-2 describes the reactivity control system and any failure modes that could lead to control rod misoperation.
Other SEP topies addrat: :uch items as the reactor protection systam.
IV.
REVIEW GUIDELINES
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The review is conducted in accordance with SRP 15.4.1, 15.4.2 and 15.4.3.
The acceptance criteria for Control Rod Misoperation ' events are:
- 1) The thermal margin limit of DNBR >1.30 is met, and-2)
Fuel centerline temperature does not exceed the melting point.
In actuality the first limit is approached before the second','so the review considers the first criteria only.
i The evaluation includes review of the analysis for the event and identifi-cation of the features, in the plant that mitigate the consequences of the event as well as the ability of these systems to function as required. The extent to which operator action is required is also evaluated.
Deviations from the criteria specified in the Standard Review Plan are identified.
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V.
Evaluation Safety evaluation of reload cores are performed for the: uncontrolled group rod withdrawal and the single rod drop events.
Other events are shown to be boun,ded by these or other more limiting events.
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1 CONTROL ROD WITHDRAWAL The group rod withdrawal event analysis for the present (Core XV) core was bounded by the Cycle XI analysis, which used conservative assumptions. The t
minimum DHBR calculated for the Core XI event was 2.30 for a reactivity insertion of 0.02X10 AP/sec. The minimum DNBR for the Core XV event was determined to be greater than 2.30 because of more favorable peaking factors even when the lower steady-state design value DNBR (3.13 vs 3.24) and the higher core inlet temperature (515"F vs 511'F) were taken into account. Thus j
for this event the acceptance criterion of DNBR greater than 1.30 is satisfied.
j CONTROL R0D DROP Using conservative assumptions and approved methods the minimum DNBR predicted for a single control rod drop was 2.18 for Core XI. The lower peaking factors and linear heat generation rates for Core XV lead to even more favorable results for Core XV. Thus, for this event the acceptance criterion for DNBR j
greater than 1.30 is satisfied.
CONTROL R0D MALPOSITIONING The Technical Specifications limit mal' positioning of control rods to i 8 inches.
l If all control rods in the most reactive group were each misaligned by 18 inches, the reactivity and power peaking effects would be less limiting than the single rod drop.
Therefore, the acceptance criteria for the control rod malpositioning event is satisfied.
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SINGLE R0D WITHDRAWAL
.1 Since single rod withdrawal produces more localized power peaking than does group rod withdrawal, the DNB ratios are lower.
Scoping calculations for Core XI showed DNBR greater than 2.0.
Shutdown requirements, power-dependent-i nsertion-limits, and allowable peak-red LHGR limits have all changed favorably
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5 since the Core XI analysis. Thus single rod withdrawal events for Core XV are bounded by the Core XI analysis and therefore satisfy the acceptance criteria.
VI.
CONCLUSION The possibilities for control rod misoperation events, including operator error and single failures have been reviewed and analyzed.
In all cases the acceptance criterion of DNBR greater than 1.30 has been met.
The staff concludes that the analyses perforced for the control rod misopera-tion events meet current requirements and are acceptable. There are no deviations from the SRP. This completes Topic XV-8.
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