ML20053C861
| ML20053C861 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/28/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Fiedler P, Fielder P GENERAL PUBLIC UTILITIES CORP. |
| References | |
| TASK-02-02.C, TASK-03-08.B, TASK-15-13, TASK-15-18, TASK-2-2.C, TASK-3-8.B, TASK-RR LSO5-82-05-072, LSO5-82-5-72, NUDOCS 8206030068 | |
| Download: ML20053C861 (10) | |
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l May 28,1982 Docket No. 50-219 LS05-82-05-072 Mr. P.B. Fiedler Vice President and Director - Oyster Creek Oyster Creek Nuclear Cenerating Station Post Office Box 388 Forked River, New Jersey 08731
Dear Mr. Fiedler:
SUBJECT:
SEP TOPICS XV-13. RADIOLOGICAL CONSEOUENCES OF CONTROL R0D DROP ACCIDENTS AND XV-18, RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE BREAK Enclosed is our draft evaluation of the subject SEP Topics XV-13 and XV-18.
In sumary, the staff concludes that for Topic XV-13, the Oyster Creek Nuclear Generating Station design is acceptable for controlling or mitigating the radiological consequences of a control rod drop accident without any modifications. However, for Topic XV-18, acceptable exclusion area boundary doses cannot be assured when considering iodine spiking coupled with the maximum iodine concentration allowed by present plant Technical Specifications.
Therefore, it is recomended that the plant Technical Specifications for controlling and limiting reactor coolant iodine concentrations be changed to conform with reactor coolant iodine limits contained in the GE Standard Technical Specifications for BWR's. With this change, the computed exclusion area boundary dose fmm a v.ain steam line break accident are acceptable.
I Please provide your comments and position on the recomended change regarding these topic dispositions within 30 days from receipt of this letter.
Did *g(sh These evaluations will be a basic input to the integrated safety assessment for your facility. These topic evaluations may be changed in the future if your facility design is changed or if NRC criteria relating to these topics is modified before the integrated assessment is complete.
ADDI Sincerely, b-Original nicacd by:
Dennis M. Crutchfield, Chief lgg6ggggg8!$@o$f9 Operating Reactors Branch #5 o
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Oyster Creek -
Docket No. 50-219 Mr. P. B. Fiedler Revised 3/30/82 CC G. F. Trowbridge, Esquire Resident Inspector Shaw, Pittman, Potts and Trowbridge c/o U. S. NRC 1800 M Street', N. W.
Post Office Box 445 Washington, D. C.
20036 Forked River, New Jersey 08731
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J. B. Lieberman, Esquire Commissioner Berlack, Israels & Lieberman New Jersey Department-of Energy 26 Broadway 101 Commerce Street New York, New York 10004 Newark, New Jersey 07102 Ronald C. Haynes, Regional Administr:: tor i
Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsyl.v:11a 19406 J.. Knubel BWR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, New Jersey 07054
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Deputy Attorney General' State of New Jersey
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Department of Law and Public Safety 36 West State Street - CN 112 Trenton,' New Jersey 08625
' Mayor Lacey Township 818 Lacey Road Fo.rked 3iver, New Jersey 08731 i
U. S. Environmental. Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Licensing Supervisor Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731 M
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0YSTER CREEK NUCLEAR POWER STATION XV-13 SPECTRUM 0F R0D DROP ACCIDENTS I.
INTRODUCTION i
An uncoupled control rod may hang up in the core when the control rod is withdrawn and drop later when the consequences are most severe.
As a result, radioactivity may be released from the core to the environment via the turbine and condensers.
SEP Topic XV-13 is intended to review the plant response and evaluate;the radiological consquences of thi's accident.
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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 the objective of assessing the risk of public health and safety resulting from operation of'the facility.
The cont'rol rod drop accident is 'one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety.
In addition, 10 CFR 100.11'provides guidelines concerning the general
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approach to calculations of the consequences of postulated accidents involv-ing a fission product release.
III.
RELATED SAFETY TOPICS Topic II-2.C, " Atmospheric Transport and Diffusion Characteristics for Accident Analysis" provides the meteorological data used to evalu' ate the offsite doses.
The atmospheric dispersion factors used in the staff evaluation are based upon a ground level releas_e for EAB and LPZ distances of 414 meters and 1208 meters respectively.
Topic III-8.B, " Control Ro'd' Drive Mechanism
~ Integrity" eva]uates the reliability and operability of control rod drives.
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VariousotherSEPtopicsevaluatesuchitemsasconta*inmentiN1ation, containment leak testing and ESF systems.
O IV.
REVIEW GUIDELIt!ES Therehiewoftheradiologicalconsequencesofacontrolroddrop
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accident was conducted in accordance with the Appendix to Standard Review Plan 15.4.9.
The plant is considered adequately designed.against a cootrol, rod drop accident and the consequences acceptable if-the resulting doses at,the exclusion area and low population zone. boundaries are well within theguidelinehaluesof10CFRPart100.
V.
EVALUATI0tl
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Thestaffperformedanindependentrehiewof'theradiologicalconsequences following a postulated control rod drop accident.
In the Safety EYaluation ReportSupplement#3datedtiohember,1971,thestaffestimatedthat330 fuel rods could perforate if a rod drop accident occurred at the licensed power of 1930 MW.
In that SER Supplement, the staff also assumed that, t
100% of the noble gas inhentory and 50% of the iodine inhentory are released i
from the perforated fuel rods to the primary coolant.
While the approach of assuming a fuel melt source term for perforated fuel rods is not consistent with the method suggested in SRP Section 15.4.9, Appendix A, Reh.1(i.e.,moreconscrhatihe),thestaffwillcontinuetomakethat assump. tion in the SEP ehaluation of the Oyster Creek plant.
All other assumptions used in the radiological consequence analysis are consistent with the assumptions gihe'n in the rehiew proce,d.ures section of
_5RPSection15.4.9,AppendixA,Reh'.1. A sumary of these assumptions,
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isprohidedbytheattachedtabletothisSEPevaluation.
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VI.
C0!;CLUSIONS Using the assumptior.s outlined above the resultant do.ses at the. nearest exclusion area boundary are 2.6 rem to the thyroid and 0.5 rem to the whole body. The resultant doses at the outer boundary of the LPZ are 1.2 rem to the thyroid and 0.1 rem to the whole body.
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both instances the resultant radiological consequences are less than the acceptance criteria given in SRP Section 15.4.9 Appendix A, Rev. I and are well within the guideline values of 10 CFR Part 100.11.
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,is acceptable for controlling or mitigating the radiological consequences from the postulated control rod drop accident.
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TABLE XV-13.1 Oyster Creek Assumptions for the Calculation of Radiological ~
Consequences Following a Control Rod Drop Accident 330 rods Amount of fuel Failures
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Peak'ing Factor 1.5
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1930 MW Power
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t Activity release from failed fuel = 50% iodine 100% noble gases Amount of activity transported to the 10% iodine condenser prior to'MSIV closure
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100% noble gases Partition factor in the condenser
= 10 for iodine 1 for noble gases 1% per day Leak rate from condenser
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Leak duration
= 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
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Breathing rate (0-8 hr)
= 3.47 x 10-4 3
m /sec m3 sec (8-24 hr)
= 1.75 x 10-4
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Exclusion Area /Q (0-2 hr)
= 7.6 x 10 X Low Population Zone /Q (0-8 hr) *
= 6.5 x 10-5
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Low Population Zone
/Q (8-24 hr)
= 4.3 x 10-5 --
X Total' Rods in.the core
= 27440 (FSAR Fig XIII-2-4)
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XV-18 RADIOLOGICAL C0!;5EQUEf4CES OF A FAlf; STEAM LINE FAILURE OUTSIDE CONTAllMENT I.
INTRODUCTION Failure of a steam line outside containment will allow radioactivity contained in the coolant to escape to the environment.
SEP Topic XV-18 is intended to review the radiological consequences of such failures.
The review will encompass those design features which limit the release of radioactivity including technical specifications which limit the amount of radioactivity in the released coolant.
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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
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of the design and performance of structures, systems, and components of the facility with the objective of assessing 4he risk to public health and safety resulting from operation of the facility.
The steam line break accident is one of the postulated accidents used to evaluate the adequacy of these structures, systems,.and components with respect to the public health and sqfety.
In addition, 10 CFR Part 100.11 provides an acceptable dose consequence limit for reactor siting.
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III.
RELATED SAFETY TOPICS Topic II-2.C, " Atmospheric Transport and Diffusion Characteristics for Accident Analysis" provides the meteorological data used to evaluate the offsite doses.
However, the atmospheric dispersion factors for the main steamline break accident (i.e., assuming an elevated release at 30' meters) were not calculated for this topic. Therefore, we have taken the atmospheric dispersion factors for a 30m elevated release from Figure 1 of Regul'atory Guide 1.5 for an EAB distance of 414
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meters and an LP7 distance of 1208 meters. These values are reported in the
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REVIEW GUIDELINES
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The review of the radiological consequences was conducted in accordance with Regulatory Guide 1.5, with the exception of meteorology, and SRP Section 15.6.4.
As suggested in SRP Section 15.6.4, the r$ view was performed for two specific cases. In the first case the radiological consequences of a main steam line failure are computed with the assumption of maximum primary coolant concentrations o,f jodine permitted by the plant Techncial Specifications under (temporary) iodine spiking conditions, and are then compared to the exposure guidelines given in 10 CFR Part 100.11.
In the second case the plant equilibrium Techni. cal Specification limit for iodine concentration is assumed, and the calculated radiological consequences are compared to a small fraction, (10%) of the dose guidelines of 10 CFR Part 100.
V.
EVALUATION The staf.f performed an independen't assessment of the radiological
' consequences following a postulated main steam line failure outside j _.
. In FSAR Amendment 68 (March, 1972), the licensee' containment.
4 estimated that based upon a MSIV isolation time of 10.5 seconds (per p unds of steam and reactor f
plant technical specificati6ns) about 103,000 o
coolant would be released to the environment prior to isolation of the break.
The staff believes this release estimate to be consistent with th'e release estimates given for large plants in SRP Section 15.6.4 Appendix A,'Rev. 1 and has used the licensee's estimate in its evaluation.
Consistent with Regulatory Guide 1.5, all of the' iodine contained in the reactor coolan't and all the noble gases are assumed to be released from the coolant to the environment.
The maximum
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primary coolant iodine concentration permitted by the Standard'
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Technical Specifications were assumed for equilibrium and spiking conditions, in accordance with SRP 15.6.4.
Consistent with Regulatory Guide 1.5, the total activity is released in a two hour period at an assumed to occur height of 30 meters with fumigation condition. A summary of the assumptions used in esti-7 mating the radiological consequences is provided in Table 1.
VI. CONCLUSIONS
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Using the assumptions outlined above and the Standard TechrIfcal Specification
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values for iodine activity in the coolant, the resultant thyroid -doses at the exclu'sion area boundary are 72 rem for the case of a preaccident iodine spike and 4 rem for the case where the coolant iodine concentration is initially at the equilibrium technical specification limit-and, therefore, comply with the SRP Acceptance Criteria.
It is noted, however, that the licensee's technical specifications do not conform to the Standard Technical Speci.fications and do not make any allowance for the iodine spiking phenomenon.
If the plant's current equilibrium icdine technical specification value of 8 pCi/ gram of gross iodine were used, the calculated conse'quences would increase substantially, and would significantly exceed a small fraction of the Part 100 guidelines.
Based on the actual iodine concentration observed during the plant's operating period, we conclude that the 8 pCi/ gram value would be unduly conservative, and that actual plant behavior is more accurately reflected by the Standard Technical Specifica-tions for BWR plants.
Therefore, the Standard Technical Specifications for reactor water fission product concentrations should be incorporated into the plant's tech-nical specifications, thereby ensuring that the consequences of a main steam line failure are within the guidelines of 100CFR100.
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c TABLE XV-18.1 Oyster Creek Assumptions For The Calculation of Radiological Consequences Following a Postulated Main Steam Line Failure Outside Containment O
. Mass of Primary Coo ant Re eased before MSIV l
l closure, pounds (FSAR Amendment 68) 103,000 Maximum closure time for MSIV's per plan't technical specifications, seconds 10.5 Maximum equilibrium specification iodine 0.2 concentration, pCi/ gram DEI-131 Short-term (spiking) iodine concentration, pCi/gm DEI-131 4.0 Fraction of iodine released, percent 100.
Fraction of noble gases released, percent 100.
3 Atmospheric diffusion factors, sec/m 0-2 hour Exclusion' Area Boundary 7,4_x.10-4.;_
0-8 hour low Population Zone Boundary 3.0 x 10-4 Other assumptions are as in Regulatory Guide 1.5, " Assumptions Used for Evaluating the Potential Radiological Consequences o,f a Steam Line Break Accident for Boiling Water Reactors.
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