ML20042B771
| ML20042B771 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 03/23/1982 |
| From: | Crutchfiled D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-15-12, TASK-RR LSO5-82-03-095, LSO5-82-3-95, NUDOCS 8203260051 | |
| Download: ML20042B771 (8) | |
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[' ff l I Mr. W. G. Counsil Vice President Nuclear Engineering and Operations y
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Dear Mr. Counsf1:
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SUBJECT:
HADDAM NECK - SEP TOPIC XV-12, SPECTRUM OF CONTROL ROD EJECTION ACCIDENTS (SYSTEMS)
By letter dated September 30, 1981, you submitted a safety assessment report for the above topic. Additional information was provided in your letter of
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February 19, 1982.
The staff has reviewed this assessment and our conclusions are presented in the enclosed safety evaluation report, which completes the systems review of this topic for Haddam Neck plant.
The potential radiological consequences of these events will be addressed in a separate evaluation.
This evaluation will be a basic input to the integrated assessment for your j
facility.
is changed or if NRC criteria relating to this topic are modified be integrated assessment is completed, ji60g i
8 Sincerely, i
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/ o*b 9, # v Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 p '5k N Division of Licensing g,
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Enclosure:
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Mr. W. G. Counsil CC Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Superintendent Haddam Neck Plant RFD #1 Post Office Box 127E Easc Hampton, Connecticut 06424 Mr. Richard R. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Russell Library 119 Broad Street Middletown, Connecticut 06457 Board of Selectmen Town Hall Haddam, Connecticut 06103 State of Connecticut 0Ffice of Policy and Management ATTN: Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06115 U. S. Environmental Protection Agency Region 1 Office ATTN: Regional Radiation Representative JFK Federal Building Bosten, Massachusetts 02203 Resident Inspector Haddam Neck Nuclear Power Station c/o U. S. NRC East Haddam Post Office East Haddam, Connecticut 06423 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region 1 Office of Inspr.ction and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406
SYSTEMATIC EVALUATION PROGRAM TOPIC XV-12 HADDAM NECK 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 ar:d 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 nomal 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 perfomance 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 detemination of the margins of safety during nomal 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
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2-O 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 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 greater 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
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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.
Deviation, if any, from the criteria specified in the Standard Review Plan are identified.
The potential radiological consequences are assessed in a separate evaluation.
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Evaluation For this event the analysis used assumptions including positive moderator coefficients for beginning of life cases and negative for end of life cases, maximum values of the total peaking factor, minimum values of the delayed neutrum fraction, a conservatively fast time for rod ejection, large ejected rod worth, the most reactive control rod stuck out of the core and a 3% uncertainty added to the power level.
Four cases were analyzed: zero power and full power, both at begining of life and end of life. No credit was taken for pressurizer spray or the actuation of the PORV's in the peak pressure calculation. Peak pressure was not calculated for the full power cases due to the lower additional heat transfer to the reactor coolant than the zero power cases. The following table shows the results of this analysis:
CASE CONDITION INITIAL POWER MAX CENTERLINE PEAK PRESSURE TEMPERATURE REACHED 0
1 BOL 1880 4230 F 2
E0L 1880 4330 F 3
BOL 0
2930 F 2470 psia U
4 E0L 0
3740 F 2450 psia Thus rod ejection was calculated not to raise primary system pressure above the design pressure or cause any fuel melting.
Fuel average temperature for the hottett pellet was predicted to be lower than 3400 F for all cases. This temperature corresponds to a fuel enthalpy of less than 200 cal /gm. Thus for this event the acceptance criterion of radially averaged enthalpy less than 280 cal /gm and pressures lower than the value to cause stress are satisfied.
The licensee has not performed a calculation to find the number of fuel rods in DNB, since they claim fuel enthalpy of less than 200 cal /gm constitutes no clad damage. We do net necessarily agree with this evaluation,
4-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 will use 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.
The staff now believes that a fuel energy deposition criterion in terms of 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.1). 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 enthalpf es, 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 cri.2rion.
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. 2) 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 were typical of the conservative analyses performed for these REAs.
CE also reported results from more realistic three-dimensional analyses.
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.
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.
6-VI Conclusion The spectrum of rod ejection accidents have been analyzed and the acceptance criteria have been met. These results have been obtained using methods which are more conservative than present day state of-the-art methods.
The staff concludes that the analysis performed for the spectrum of rod ejection accidents meets current requirements and the results are acceptable.
This completes Topic XV-12.
VII References _
1.
P.E. MacDonald et al., " Assessment of Light-Water-Reactor Fuel Damage During a Reactivity-Initiated Accident," Nucl. Safety 21, September -
October 1980, P 582.
2.
R.E. Uhrig (FPL) letter to D.G. Eisenhut (NRC) dated August 25, 1981.