ML20054B863
| ML20054B863 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 04/16/1982 |
| From: | Caruso R Office of Nuclear Reactor Regulation |
| To: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| References | |
| TASK-15-12, TASK-RR LSO5-82-04-046, LSO5-82-4-46, NUDOCS 8204190263 | |
| Download: ML20054B863 (9) | |
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April 16,1982 Docket No.50-029 LS05 04-046 ss G
/fggE/P Mr. James A. Kay g
Senior Engineer - Licensing 4
80 i2 Yankee Atomic Electric Company T
e 1671 Worcester Road 9-Framingham, Massachusetts 01701 7
4
Dear Mr. Kay:
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SUBJECT:
YANKEE NUCLEAR POWER STATION - SEP TOPIC XV-12, SPECTRUM OF CONTROL R0D EJECTION ACCIDENTS (SYSTEMS)
In your letter dated June 30,1981 (FYR 81-95), you submitted a safety assessment on the above topic. Your letter of March 9,1982 (FYR 82930),
provided additional information. The staff has reviewed your assessment and our conclusions are presented in the enclosed safety evaluation re-port. The potential radiological consequences of rod ejection accidents were addressed in our December 29, 1981 letter. The enclosed evaluation completes SEP Topic XV-12 for the Yankee Nuclear Power Station.
The enclosed safety evaluation will be a basic input to the integrated safety assessment for your facility. The assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely,
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O FFICI AL R E.t.OR D COPY usc+a no-anow r,nc roriu aia <ia aot tsicu o 4a
Yankee Docket No. 50-29 Revised 3/30/82 Mr. James A. Kay cc Mr. James E. Tribble, President Yankee Atomic Electric Company 25 Research Drive Westborough, Massachusetts 01581 Chairman 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:
Regional Radiation Representative 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 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 l
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SYSTEMATIC EVALUATIOS PROGRAM TOPIC XV-lq YANKEE NUCLEAR POWER STATION TOPIC: XV-12, Spectrum of Control Rod Ejection Accidents 1.
Introduction A control rod ejection accident is caused by the mechanical failure of a control rod mechanism pressure housing such that reactor coolant system pressure ejects the control rod and drive shaft.
Ejection of a control rod results in a rapid increase in reactivity, power production and a corresponding pressure increase. The increase in power is lowered by Doppler feedback.
Reactor trip occurs on high neutron flux. The potential for fuel damage is further reduced by the control rod insertion limits during nonnal operation.
II.
Review Criteria Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide 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 detennination of the margins of safety during nonnal operation and transient conditions anticipated during the life of the facility.
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 28 " Reactivity Limits" requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure
- boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the j
core.
III.
Related Safety Topics SEP Topic XV-19 considers the effects of rupture of the reactor coolant pressure boundary by the ejected rod.
IV.
Review Guidelines The review is performed in accordance with SRP 15.4.8 and Regulatory Guide 1.77 " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors."
The acceptance criteria for control rod ejection accidents are:
1.
Reactivity excursions should not result in a radially averaged enthalpy greatcr than 280 cal /g at any axial location in any fuel rod.
2.
The maximum reactor pressure during any portion of the assumed excursion should be less than the value that will cause stresses to exceed the " Service Limit C" as defined in the ASME Code.
3.
The fission product inventory in the fuel rods calculated to experience a departure from nucleate boiling (DNB) condition is an input to the radiological evaluation.
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 esaluated.
Deviation, if any, from the criteria specified in the Standard Review Plan are identified.
The potential radiological consequences were previously assessed in a separate evaluation.
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V. Evaluation Rod ejections are evaluated for each reload core using methods similar i
to the Core XI analysis submitted in Reference 1.
These methods have been found to be acceptable by the NRC in Reference 2.
Typically results vary from core to core because of changes in rod worth, moderator and fuel temperature reactivity coefficients, axial power distribution and radial two dimensional peaking factors.
Analysis is performed for both zero power and full power conditions. The following assumptions were used:
1.
The fuel temperature coefficient is the least negative value expected throughout core life during power operation.
2.
The ejected rod worth is the maximum throughout core life.
3.
The calculated maximum ejected rod peaking throughout core life is increased by 10 percent in order to account for calcu-lational uncertainities.
All current regt.irements with respect to initial conditions and reactor parameters are met.
Results for Core XV were that radially averaged enthalpy for both low power and full power cases was less than 200 cal /gm. Thus for this event the acceptance criteria of radially averaged enthalpy less than 280 cal /gm was satisfied.
The licensee has not performed a calculation to find the number of fuel rods in DNB, since they claim fuel entahalpy of less than 200 cal /gm results in no clad damage. We do not necessarily agree with this eval-uation, however, for the reasons discussed below, we believe that this plant would not have more than the standard Westinghouse number of 10%
of the fuel rods in DNB.
The radiological evaluation (Ref. 3) used this 10% value.
At the time of writing the SRP and the Regulatory Guide, the staff believed that the use of DNB to calculate fuel cladding failures provided suitably (and in fact overly) conservative results.
- i The staff now believes that a fuel energy deposition criterion in terms of r
enthalpy is a better measure of fuel cladding failures than DNB for a rod ejection accident (REA). Such an enthalpy criterion would be a function of burnup because of the change from an oxidation failure mechanism for fresh fuel to a PCI failure mechanism for exposed fuel. We believe that this failure criterion is on the order of 180 cal /g for fresh fuel and 140 cal /g for exposed fuel when the enthalpy value is expressed as the peak (as a function of time) radially averaged fuel pellet ehthalpy (See Ref. 4). We have previously considered a lower enthalpy value for high-burnup fuel be-cause of the existence of an 85 cal /g failure in one of the SPERT tests.
We now tend to disregard the single 85 cal /g SPERT test failure as perhaps being waterlogged in light of its prior high-power irradiation history, the similarity to waterlogged failure enthalpies, and the 176 cal /g failure of a similar rod.
Nevertheless, because of the scarcity of data, there is a lot of uncertainty in the 140 cal /g value and we therefore do not want to use it as a precise criterion.
To estimate the number of fuel failures during a PWR REA, we recently asked Combustion Engineering to perform an analysis using an enthalpy criterion.
In their response (Ref. 5) they described four cases for two plants (HFP and HZP for CESSAR and St. Lucie 2). Only one of the four cases studied resulted in rods exceeding 140 cal /g, and in that case only 2% of the rods exceeded that value. These CE analyses w re typical of the conservative analyses performed for these REAs.
CE also reported results from more realistic three-dimensional analyses.
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j 5-l In addition to the Combustion Engineering analyses, analyses have been per-formed by Westinghouse and Babcock & Wilcox. The Westinghouse analyses used realistic three-dimensional calculations and included full-power and zero-power cases. The zero-power cases and the full-power case, when adjus-ted to the proper rod worth, give results of less than 140 cal /g. The Babcock & Wilcox calculations give similar results.
I We conclude, therefore, that realistic REA calculations using appropriate three-dimensional models would predict very few cladding failures using a 140 cal /g criterion.
Consequently, the use at this time of an assumed 10%
amount of failed fuel in a radiological dose calculation for REAs is more realistic than the DNB criterion but is still suitably conservative from our point of view. This position would be reevaluated if we established any more limiting mechanism for fuel cladding failure resulting from reactivity initiated accidents.
VI Conclusion i
The spectrum of rod ejection accidents has been analyzed and the acceptance criteria have been met. These results have been obtained using currently approved methods.
The staff concludes that the analyses performed for the spectrum of rod ejection accidents meets current requirements and are acceptable.
This com-(
pletes Topic XV-12.
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VII References 1.
YAEC Letter to AEC, Proposed Change No.115 (Core XI Refueling),
March 29,1974.
2.
D0L/AEC Letter to YAEC, Amendment No. 9, July 30,1974.
3.
D. Crutchfield (NRC) letter to J. Kay (YAEC) dated December 29, 1981.
4.
P.E. Mcdonald et al., " Assessment of Light-Water-Reactor Fuel Damage -
During a Reactivity-Initiated Accident," Nuclear Safety 21, September -
1 October 1980 Page 582.
5.
R.E. Uhrig (FPL) letter to D.G. Eisenhut (NRC) dated August 25, 1981.
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