ML20039D431
| ML20039D431 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/29/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| References | |
| TASK-15-12, TASK-RR LSO5-81-12-099, LSO5-81-12-99, NUDOCS 8201040082 | |
| Download: ML20039D431 (8) | |
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December 29, 1981 D
Docket No.50-029 f,
LS05-81-12-099 ftECEWED -
5 DECSIG81 Mr. James A. Kay
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Senior Engineer - Licensing a
Yankee Atomic Electric Company s#
3 1671 Worcester Road g
Framingham, Massachusetts 01701
Dear Mr. Kay:
SUBJECT.: YANKEE R0WE, SEP TOPIC XV-12, RADIOLOGICAL CONSEQUENCES OF SPECTRUM OF R0D EJECTION ACCIDENTS Enclosed is our evaluation of the SEP Topic XV-12, Radiological Consequences for Spectrum of Rod Ejection Accidents. The results of the review indicate that the Yankee Rowe plant is adequately designed for controlling or miti-gating the radiological consequences for this accident.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions of your plant. This assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this subject are modified before the integrated assessment is completed.
Sincerely, Dennis M Crutchfield, Chief l
Operating Reactors Branch No. 5 Division of Licensirig I
Enclosure:
As stated cc w/ enclosure:
s's ypQ See next page
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Docket No.50-029 j
LS05-81-12-099 Mr. James A. Kay Senior Engineer - Licensing Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701
Dear Mr. Kay:
SUBJECT:
YANKEE R0WE, SEP TOPIC XV-12 RADIOLOGICAL CONSEQUENCES OF SPECTRUM 0F ROD EJECTION ACCIDENTS Enclosed is our evaluation of the SEP Topic XV-12, Radiological Consequences for Spectrum of Rod Ejection Accidents. The results of the review indicate that the Yankee Rowe plant is adequately designed for controlling or miti-gating the radiological consequences for this accident.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions of your plant. This assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this subject are modified before the integrated assessment is completed.
Sincerely, h $ k-Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing i
Enclosure:
As stated cc w/ enclosure:
See next page
Mr. James A. Kay CC Mr. James E. Tribble, President Yankee Atomic Electric Company 25 Research Drive Westborough, Massachusetts 01581 Greenfield Connunity College 1 College Drive Greenfield, Massachusetts 01301 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
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YANKEE R0WE XV-12 SPECTRUM 0F ROD EJECTION ACCIDENTS 1.
INTRODUCTION Ejection of a control element assembly from the core can occur if the control element drive mechanism housing or the nozzle on the reactor vessel head breaks off circumferentially. The ejection of a control element assembly by the reactor coolant system pressure can cause a severe reactivity excursion.
This accident may result in radioactivity being released to the environment through the steam generator and containment leaks.
SEP Topic XV-12 is intended to evaluate the radiological consequences of this accident. The review will encompass those plant design features which limit the release including the plant technical specifications on primary to secondary system leakage.
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 evaluation of the design-and performance of structures, systems, and components of the facility with f
the objective of assessing the risk to public health and safety resulting l
from operation of the facility. The control rod ejection accident is one f
of the postulated accidents used to evaluate the adequacy of these structures, l
systems, and components with respect to the public health and safety, i!~
u General Design Criterion (GDC) 28, " Reactivity Limits", of Appendix A to 10 l
CFR Part 50, requires the reactivity control system to be designed with appro-i l
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.priate limits on the potential amount and rate of reactivity increase. GDC 28
,also requires that these postulated reactivity accidents include consideration i
/ of;the rod ejection unless such an accident is prevented by positive means.
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In addition, 10 CFR Part 100,11 provides dose guidelines for reactor siting against which calculated accident dose consequences may be compared.
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E III. Rldi[ED SAFETY TOPICS q,
Topic,II-2.C, " Atmospheric Transport and Diffusion Characteristics for i
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1AccidentAnalysis"providesthemeteorologicaldatauscdtoevaluatethe
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offs,ite doses. Various other'SEP topics evaluate such items as containment
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isola'tio), cont'ainment leak testing, ESF systems, and steam generator integrity, b'
IV. REVIEW G3IDELINES l
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The review of the radiological consequences of a control rod ejection accident
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was conducted in accordance with the Appendix to Standard Review Plan 15.4.8
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,n<Cl and Regulatory Guio'd 1.77.
Existing plant technical specifications will be igT taken into account in calculating the radiological consequences. The plant is considered adequately designed against a control rod ejection accident if the resulting doses at the exclusion area and low population zone boundaries tare well within (defined as 25% of 'the limits) the guideline values of 10 CFR s
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) Part 103,
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s Tne staff-perf.ormed 'an independent < assessment of the consequences of the
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accident, since the evaluation provided by the licensee was at variance with the provi'sions.of Regulatory Guide 1.77.
Q, The parameters assumed'in the analysis are given in Table XV-12-1. Two
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cases were considered: First, the gap content of all failed rods plus 100% of the noble gases, and 25% of the radioiodines in melted fuel are assumed to be instantaneously available for leakage from the cqntainment;
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Second, all of the gap cont'ent' of f ailed rods plus 100% of the noble gases and 50%
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of the radioiodines in melted fuel are instantaneously mixed in the primary water, which leaks to the secondary system at the technical specification
. leak rate. The results of the calculations are given for'both the Exclusion Area Boundary and the Low Population Zone for both release path-ways in Table XV-12-2.
VI. CONCLUSIONS
.The plant is adequately designed against a control rod ejection accident.
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TABLE XV-12-1 Assumptions MMe in Analysis of the Radiological Consequences of a Control Rod Ejection Accident 1.
Reactor Power = 600 MWth, 2.
Loss of Offsite Power following the accident.
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3.
Reactor Coolant System (RCS) volume = 2940 ft,
4.
Time to equalize the primary and secondary pressures 8 hrs.
5.
Gap Activity 10% of contained radioiodines and noble gases "from 10%
fuel rod failures."
6.
Iodine decontamination factor of 10 between water and steam.
7.
0.25% fuel melting as a result of the accident.
8.
Peaking factor 1.0.
9.
Primary to secondary leak rate 1 gpm per steam generator.
- 10. Containment leak rate 0.2% per day for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 0.1% thereafter.
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Meteorological conditions: EAB 0-2 hr. 2.84 x 10-4 LPZ 0-8 hr. 2.84 x 10-5 8-24 hr. 1.92 x 10-5 24-96 hr. 1.62 x 10-5 96-720 hr. 1.04 x 10-5
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TABLE XV-12-2 Calculated Doses for the Control Rod Ejection Accident 4
Case Location Thyroid dose (Rem)
Whole Body dose (Rem)
Containment Exclusion Area 7.0 0.03 leakage Boundary Low Population 10.6 0.01 Zone Steam Generator Exclusion Area 25.1 0.09 l
release Boundary
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Low Population Zone 9.2 0.02 4
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