ML20040A084
| ML20040A084 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/12/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-15-13, TASK-15-20, TASK-RR LSO5-81-01-029, LSO5-81-1-29, NUDOCS 8201200320 | |
| Download: ML20040A084 (10) | |
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Dear Mr. Iloffman:
SUBJECT:
BIG ROCK POINT SEP TOPICS XV-13, SPECTRUM 0F R0D DROP ACCIDENTS AND XV-20, RADIOLOGICAL CONSEQUENCES OF FUEL DAMAING ACCIDENTS Enclosed are the staff's review of SEP Topics XV-13 and XV-20. The results of the review indicate that the Big Rock Point Plant is acceptably designed for controlling or mitigating the radiological consequences of the Fuel Handling Accident (XV-20). for the Rod Drop Accident (XV-13), the design is acceptable provided there is no significient reactor coolant leakage between the pressurized reactor steam side and vented shell side of the isolation condenser.
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, gof Original cim ' y f:
s lll Dennis M. Crutchfield, Chief ga %E b8)
Operating Reactors Branch No. 5 Division of Licensing
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Enclosure:
As stated cc w/ enclosure:
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SEP REVIEW 0F BIG ROCK POINT i
XV-13 SPECTRUM 0F R0D DROP ACCIDENTS I.
INTRODUCTION 4
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An uncoupled control rod may hang up in the core when the control rod is i
withdrawn and drop later when the consequences are most severe. As a result, j
radioactivity may be released from the. core to the environment via the turbine and condenser.
SEP Topic XV-13 is intended to review the plant i
response and evaluate the radiological consequences of this accident.
II.
R,EVIEW CRITERIA i
Section 50.34 of 10 CFR Part ~50 ' requires that each applicant for a construction a
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permit or operating license provide an analysis and evalua' tion of the design
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and performance of structures, systems, and components of the facility with I
the objective of assessing the risk to public health and safety resulting from operation of.the facility. The control rod drop accident is one of the postulated accidents used to eval,uate the adequacy of these st'ructures, systems, and components with respect to the public health and safety.
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In addition, 10 CFR Part 100.11 provides guidelines concerning the general approach'to calculations of the consequences of postulated accidents involv-ing a fission product release.
II.
RELATED SAFETY TOPICS j
Topic II-2.C, " Atmospheric Transport and Diffusion Characteristics for Accident Analysis" provides the meteorological data used to evaluate the offsite doses. Topic III-8.B " Control Rod Drive Mechanism Integrity" evaluates the reliability and operability of control rod drives. Various other SEP topics evaluate such items as containment isolation, containment leak testing and ESF systems..
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IV.
REVIEW GUIDELINES
. The review of the radiological consequences of a control rod drop accident was conducted in accordance with the Appendix to Standard Review Plan.15.4.9. The' plant is considered adequate,1y designed against a control rod drop a'ccident and the consequences acceptable if the resulting doses at the exclusion area and low population zone boundaries are well within the guideline v'alues of 10 CFR Part 100.
V.
EVALUATION The staff has completed the review of the Big Rock Point submittal on the control rod drop accident. The applicant estimated that 464 fuel rods would perforate this being all fuel bundles adjacent to the dropped con-trol rod, and about 5% of the core. The staff analyzed this accident conservatively by using the licensee's fuel failure estimate, the assump-tions in SRP Section 15.4.9, App'endix A, Rev.
l., and the assumptions in the Regulatory Guide 1.77.
A summary of these assumpticn; is provided in the attached Table XV-13.1. The staff finds that the radiological con-sequences satisfy the acceptance criteria in SRP Section 15.4.9. Appendix A..
VI.
CONCLUSIONS Using the assumptions outlined above and summarized in Table XV-13-1 the resultant doses at the nearest exclusion area boundary are 0.54 rem to the thyroid and 0.12 rem to the whole body. These radiological conse-quences are less than the acceptance criteria given in SRP Section 15.4.9 Appendix A, Rev. 1 and are well within the guideline values of 10 CFR Part 100.11.
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l We therefore conclude that the Big Rock Point design is. acceptable for controlling or mitigating the radiological consequences from the postulated control rod drop accident, provided t' hat there is no sig-nificant radiation leakage through the isolation condenser.
Unacceptable 1
radiation leakage through the isolation condenser can be detected by the radiat' ion _ monitor in the atmosphere vent of the isolation condenser.
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TABLE XV-13.1 Assumptions for the Calculation of Radiological Consequences Following a Control Rod Drop Accident Amount of Fuel Failures = 464 rods Peaking Factor 1.5 Power 240 MWt Activity release from failed fuel = 10% iodine 10%' noble gases m
Amount of activity transported to-the condenser ~ prior to MSIV closure
= 10% iodine 100% noble gases Decontamination factor in the condenser = 10 for iodine 1 for noble gases Leak rate from condenser 1% per day
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Leak duration 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
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Breathing rate (0-8 hr) 3.47 x 10 m /sec
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Exclusion Area /Q (0-2 hr)
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- value from Memoran'dum from Hulman to Knighton, July 31, 1979 9
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BIGROCKP0iNTNUCLEARGENIRATINGSTATION
-XV-20 RADIOLOGICAL CONSEQUENCES OF FUEL DAMAGING ACCIDENTS I.
INTRODUCTION Thesafetyobjectiheofthistopicistoassurethattheoffsitedoses from fuel damaging accidents as a result of fuel handling inside and outside containment are well within the guideline value of 10 CFR Part 100.
II. REVIEW CRITERIA Section 50.34 of 10 CFR Part 50, " Contents of Applications: Technical Information," requires that each applicant for a construction permit,or operatinglicenseprohideananalysisandehaluationof.thedesignand performance of structures, systems, and components of the facilit'y with theobjectiheofassessingthe risk to public health and safety resulting from operation of the facility. A fuel handling accident in the fuel handling and storage facility resulting in damage 'to fuel cladding and 4
subsequentreleaseofradioactihematerialisoneofthepostulatedaccidents usedtoehaluatetheadequacyofthesestructures, systems,andcomponents with respect to the public health and safety.
Inaddition,10CFRPart100prohidesdoseguidelinesforreactorsiting against which calculated accident dose consequences may be compared.
III RELATED SAFETY TOPICS Topic II-2.C, " Atmospheric Transport and Diffusion CharacterStics for AccidentAnalysis"prohidesthemeteorologicaldatausedforcalculating offsite dose consequences.
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The review of the fuel damaging accidents did not consider fuel damage as a result of drops of the spent fuel cask or other heavy objects which can be carried either over an open reactor vessel or the spent fuel pool. Review of the drops of casks and heavy object is covered in'two SEP Topics, IX-2,-
" Overhead Handling Systems-Cranes" and XV-21, " Spent Fuel Cask Drop Accidents."
IV. REVIEW GUIDELINES Aegidents resulting from the movement of fuel inside and outside containment were reviewed following the assumptions and procedures outlined in Standard Review Plan (SRP) Section 15.7.4 and Regulatory Guide l.'25.
The dose to an individual from a postulated fuel handling accident should be "well within" the exposure guidelines'of 10 CFR Part 100.
(Whole body doses are also examined but are not controlling due to the decay of the short-lived radio-isotopes prior to fuel handling). This is based on the probability of this eventrelatihetoothereventswhichareevaluatedagainst10.CFRPart100 exposure guidelines. 'The review considers single failure, seismic design and equipment qualification only when the potential consequences might exceed the guidelines of 10 CFR Part 100 in the absence of containment isolation and effluent filtration. The system design is considered to be acceptable if the limiting doses are well within the 10 CFR 100 guidelines.'
V. EVALUATION In the evaluation of the fuel handling acc'ident, the methodology used by the staff is based on the fuel handling system described in the Consumer Power Company letter dated July 3,1981, from Robert A. Vincent to Director, NRR, NRC. The analysis was performed using the guidelines and requirements of SRP'15.7.4 and Regulatory Guide 1.25. The list of assumptions and parameters used in the analysis is given in Table XV-20-1.
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VI. CONCLUSION The offsite thyroid and whole body doses for the postulated fuel hundling accidentare47 remand.7remrespectihely;'thesedosesarewellwithin l
theguidelineshaluesgihenin10CFRPart100. The staff concludes that l
l this system is acceptable in mitigating the consequences of the fuel handling i
accident.
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TABLE XV-20-1 Fuel Handling Accident Dose Assumptions PowerLehol,MWT 240 Radial Peaking Factor 1.5 Decay time,, hours 12 Number of fuel assemblies affected
.1 Number of fuel assemblies in core 84 Actihityreleaseperiod 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> No filtration Iodine decontamination factor 100 X/Q sec/m3 6.x 10-4*
j Value in Memorandum from Hulman to Knighton, July 31, 1979 1
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I TABLE XV-20-1 i
i Fuel Handling Accident Dose Assumptions l
PowerLehel,MWT 240 Radial Peaking Factor 1.5 1
Decay time, hours 12 Number of fuel assemblies affected
'l Number of fuel assemblies in core-84 Actihityreleaseperiod 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> No filtration Iodine decontamination i
factor 100 3
X/Q sec/m 6.x 10-4*
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