ML20032B812
| ML20032B812 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/03/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Werner R NORTHEAST UTILITIES |
| References | |
| TASK-15-13, TASK-15-16, TASK-15-18, TASK-15-19, TASK-15-20, TASK-RR NUDOCS 8111060373 | |
| Download: ML20032B812 (24) | |
Text
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i Docket No.
50-245 lg!
q Y C.
Ngb Hr. R. P. Werner 1-M Senior Vice President of Generation, i> gD N
Engineering and Construction
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Northeast Utilities Har or Connecticut 06101 Q$3
Dear fir. Werner:
SUBJQT: REVIEW OF SEP TOPICS XV-13, SPECTRUM OF ROD DROP ACCIDENTS:
XV-16 RADIOLOGICAL CONSEQUENCE 5 a FAILURE OF SMALL LINES CARRIflG PRIMARY COOLANT OUTSIDE.. % AINMEflT; XV-18, PADIO-LOGICAL CONSEQUENCES OF MAIN STEAM LINE FAILURE OUTSIDE C0!iTAIf3!EllT X7-19, LOSS OF C00LAflT ACCIDENTS RESULTING FROM SPECTRUM OF-POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE B00nJARY; XV-20, RADIOLOGICAL CONSEQUENCES OF FUEL DAMAGING ACCIDENTS - MILLSTONE 1 Enclosed are the staff's draft safety evaluations of Topics XV-13, XV-16, XV-18, XV-19 and XV-20 for Millstone 1.
These assessments com-pare your facility as described in Docket No. 50-245 with criteria currently used for licensing new plants.
Please inform us within 30 3g days if your as-built facility differs from the licensing bases assumed 3
in our assessments.
If no comments are received within 30 days we will
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assume the topic is complete.
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The results of our topic evaluation conclude that all of the above topics tgu d except lapic XV-18 are acceptable to current licensing criteria. Topic XV-18 Radiological Consequences of a Main Steam Line Failure Outside ADD
- Containment results in an exclusion area boundary t)yroid dose of nearly y*5/rs four times larger than allowed by the Standard Review Plan when using the present technical specification limits for iodine-131.
It is re-comended that Millstones technical specifications for fodiae concentration in reactor coolant be lowered to 0.2p.13 and 4.0pel/g dose equivalent iodine-131 for the equilibrium and iodine spike concentrations respectfully.
The computed thyroid doses for the steam line break accident are acceptable at these lower concentrations.
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4 These topic evaluations will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built condition at. tour facility. The assessments may be revised in the future if your facility design is changed or if NRC criteria relating to thess subjects are modified before the integrated assessment is completed.
j Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing l
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s SEP REVIEW 0F MILLST0nE 1 XV-13 SPECTRUM OF R03 [, ROP ACCIDENTS 1.
INTRODUCTION 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 environement via the turbine and conderser. SEP Topic XV-13 is intended to review the plant response and evaluate the radiological consequences of this accident.
II.
REVIEW CRITERIA Section 50.34 cf 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 to public health and safety result _ing from operation of the facility. The control 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 Part 100.11 provides guidelines concerning the general approach to calculations of the consequences of postulated accidents involv-ing a fission product release.
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1 III.
RELATED SAFETY TOPICS Topic II-2.C, " Atmospheric Transp. ort 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
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leak testing and ESF systems.
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. IV.
REVIEW GUIDELINES
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8 The review of the radiological consequences of a control rod drop accident was conducted in accordance with the Appendix to Standard Revie -
i Plan 15.4.0.
The plant is considered adequately designed against a control rod drop accident and the consequences acceptable if the resulting doses et the exclusion area and low population zone boundaries are well within the guiceline values of 10 CFR Part 100.
V.
EVALUATIOj The staff has completed the review of the Millstone submittal on the control rod drop accident. The applicant estimated that 880 fuel rods would perfor-ate out of a total 36540 rods. This fuel failure is higher than the usual 770 failed rods in a slightly larger core for a BWR. However, the staff analyzed this accident conservatively by using the licensee's fuel failure,
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estimate, the assumptions in SRP Section 15.4.9, Appendix A, Rev.
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and the assumptions in the Regulatory Guide 1.77.
A sumary of these assumptions is provided in the attached Table XV-13.1. The staff finds that the radio-logical consequences satisfy the acceptance criteria in SRP Section 15.4.9.
Appendix A.
VI.
CONCLUSIONS Using the assumptions outlined above and sumarized in Table XV-13-1 the resultant doses at the nearest exclusion area boundary are 0.51 rem to the thyroid and 0.21 run to the whole body. The resultant doses at the outer boundary of the LPZ are 0.06 rea to the thyroid and 0.03 rem to the whole body. For both instances the resultant raciological consequences are less than the acceptance criteria given in SRP Section 15.4.9 Appendix A, Rev. 1 i
and are well within the guideline values of 10 CFR Part 100.11.
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. k'e therefore conclude that the Millstone Unit 1 Nuclear Station design is acceptable for controlling or initigating the radiological consequences i
frem. the postulated control rod drop accicer.t.
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TABLE XV-13.1 Assumptions for tt Calculation of Radiological Consequences Following a Control Rod Drop Accident i
Amount of Fuel Failures = 880 rods Peaking Factor 1.5 Power 2011 MWt Activity release from failed fuel = 10% icdine 10% noble gases r
Amount of activity transported to the condenser prior to MSIV closure
= 10% iodine 100% noble gases Decontaminatic-f actor in the condenser = 10 for iodine i 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 /> Breathing rate (0-8 hr)
= 3.47 x 10-4 m /sec 3
(8-24 hr)
= 1.8 x 10-5 m /sec 3
X Exclusion Area /Q (0-2 hr)
= 1.3 x 10-5 X
Low Population Zone /Q (0-8 hr)
= 5.5 x 10-6 X
i Low Population Zone /Q (8-24 hr)
= 1.7 x 10-6 Total Rods in the core
= 36540 i
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j SEP REVIEW OF MILLSTONE 1 1
XV-16 RADIOLO3ICAL CONSEQUENCES OF FAILURE OF S"ALL LINES CARRYING PRIMARY COOLANT DUTSIDE CONTAINMENT I.
INTRODUCTION t
Rupture of IP es carrying primary coolant outside' containment can aller:
primary coolcr,t and the radioactivity contained therein to escape to the envi ror.m:r.t.
SEP Topic XV-16 is intended to review the radiological consequences of such failures. The review of this topic encompassed those lines which carry primary coolant outside containment, The scope included those lines that are not normally expected to be open to the primary system but can be opened during power operation (i.e., reactor coolant sample lines, instrument lines, etc.).
II.
REVIEW CRITERION All'small lines carrying primary coolant outside containment, should be-reviewed to ensure that any release of radioactivity from their postulated failure I(
is a small fraction of the 10 CFR Part 100 exposure guidelines. Small fraction is defined in the SRP to be no more than 101 of the 10 CFR Part 100 exposure guidelines.
III. RELATED SAFETY TOPICS AND INTERFACE 5 t
Lines which were excluded from this review included lines for which failure j
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outside containment is not postulated, such as lines with isolation valves inside containment, or lines for which interlocks prevent opening during power operation. The review also did not consider the release of radioisotopes i
from large pipes carrying primary system fluid prior to automatic isolation of such lines, (e.g. the main steam and feedwater lines).
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The consequences from failure: in the:c liner crc considered in SEP 1
Topic XV-18, " Radiological Consequences of Main Steam Line Failurc Outside Containment."
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t IV. PE"IEW GUIDELI CS The rev2ew aas conducted in accordance with SP.P 15.6.2.
The licensee was requested to provide plant specific information such as the identification f
of line; covered by this Topic, the size of these lines, break locctions and I
flow, etc. The licensee responded to this request in a letter dated August 18, 1980.
V.
EVALUATION We reviewed the applicant's submittal of the analysis of the radiological consequences of Failure of Small Lines Outside the Primary Containment ~~
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(Topic XV-16). The licensee analyzed the release from a break of the re-1 circulation loop sample line and estimated that the leak rates are 1380 t
lbs/hr. of water and 1015 lbs/hr of steam, and that the' break could be l
j solated from the control room within 30 minutes. Licensee has not pre-sented any infonnation on other small lines that could break outside the containment. For example,a break in the Sram Discharge Volume System could result in consequences which are more severe than a break in se-l t
lected sampling line. (Since a generic evaluation of the consequences of
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a break in Scram Discharge System is being performed separately, it has not been evaluated here).
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3-We have used the applicant's assumptions of primary coolant leakage and per-(.
. formed an independent calculation of the offsite dose consequences. Althcugh the applicant assumed that all the activity was released from a 375-foot i
stack withcut treatment, we have conservatively assumed that the air treat-ment syster nay not be operable and the release of activity could be at arcur.d level. We have further assumed that an iodine spike occurs during react" shutdown and the iodine release rate from the fuel is increased by a factor of 500. The assumptions used in our analysis are summarized'in Table XV-16.1.
The resulting doses are calculated to be 10.8 rems to the thyroid at the exclusion boundary and 0.3 rem to the thyroid at the low population zone.
The whole body dose from a Standardized Technical Specification gross activity limit of 100/E pCi/gm are negligible.
VI.
CONCLUSIONS It is concluded that the radiological consequences of small line failures outside containment are a small fraction of the 10 CFR 100 guidelines. The plant is therefore considered adequately designed against the radiological consequences of the sample line failure accident.
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TABLE XV - 16.1
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ASSUMPTIC1 USED IN THE RADIOLOGICAL CONSEQUENCE OFFAILUREOFSi;ALLLIrgS CARRYING PRIMARY COC'L!'!T l.
Iodine Decontamination Factor 1.0 in steam
- 10. in water 1
2.
Coolant Release Rate 1380 lb/hr water 1015 lb/hr steam 3.
Duration of Release 30 min.
4.
Cleanup System 530,000 lbs/hr 6
5.
Pri-ary Coalant Inventory 1.4 x 10 lbs.
I 6.
Maximum Coolant Iodine Activity 20 uCi gm 7.
Gross Gamma Activity Concentration 100 p Y,Ci/gm f
8.
X/Q Atmospheric Dispersion Coefficient EAB (0-2 hr)=6.1x10 sec/m
-4 3
LPZ (0-4 hr)=1.9x10-5 sec/m3
SEP REVIEW OF MILLST0iE 1 XV-18 MDIOLOGICAL C0!!SEQUE!!CES OF A MA!!! STEA" LItiE FAILUI'.E C"TSICE CC:! TAI::" Erit -
I. I!; TROD'f*TIO!;
Rupture of a steam line outside containnnt will allow radioactivity contained in the coolant to escape to tne environment. SEP Topic XV-18 is intevad to review the radiologicM 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.he released coolant.
II. REVIEW CRITERIA Section E0.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluat, ion
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of the design and perfomance of structures,. systems, and components of the facility witn the ob,jective of assessing the 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 I
health and safety.
In addition,10 CFR Part 100.11 provides an acceptable dose consequence limit for reactor siting.
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. Topic III-5.B. " Pipe Break Outside Containment" will cover the dynamic effects of the postulated pipe failure.
. IV.
REVIEW GUIDELINES The revie.. of the radiological consequences was conducted in accordance with Regulatory Guide 1.5, with the exception of meteorology, and SRP Section 15.6.4, which rec;uires that the review be performed for two cases, one using maximum iodine primary coolant concentrations permitted by the plant Technical 1
Spccifi:cticns (allowing for an iodine spike prior to the accident), and the other usine plant equilibrium Technical Specification limits for iodine con-centration. Because Millstone 1 has only one iodine concentration Technical Specification, only the case for equilibrium iodine concentration was reviewed to check that the Technical Specificatien limit is sufficient to assure that the calculated radiological consequences will not exceed a small fraction (10%) of the exposure guidelines given in 10 CFR Part 100.11.
j V.
EVALUATION l
The staff performed an independent review of the radiological consequences following a postulated main steam line break outside containment. The primary coolant iodine concentration was assumed to be the maximun concen-tration permitted by the licensee's Technical Specifications. The dose calculation was based on the assumption that all this iodine was I-131. The licensee's evaluation states that 39,000 lbs of reactor coolant is released L
to the environment prior to isolation of the break.
The amount of released coolant that the licensee calculated is much less l
than that recommended by SRP Section 15.6.4 for GE plants with the same Technical Specification for main steam isolation valve closure time, but with 80 to 100% higher power levels (SRP 15.6.4 recommends 100,000 lbs and 140,000 lbs for specified cases). The staff did not perform an independent l'(A l
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analysis of the mass of coolant released; we udged the value calcu!ated by tha licensee to be acceptable for the purposes of this dose calcul:ti:n; The release is assumed to occur near the ground level. The as:umptiens used for dose calculations are summarized in Table XV-18.1. Since the 0-2 hour Exclusion Area Boundary dose, calculated by assuming maximum Iodine-131 concentration permitted by Millstone 1 technical specifications, was greater that 10% of 10 CFR Part 100 guidelines, we performed another calculatica, using the GE Standard Techn1 cal Specifications for primary coolant equilibrium iodine concentration of 0.! pCi/g Dose Equivalent I-131. The standard tech-nical specification 5also include a limit on the maximum iodine concentration permitted in the primary c'olant for a short time (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />), which is 4 pC1/g Dose Equivalent I-131. A cas elation of the dose using this coolant concen!
tration was also performed, to check that the dose was less than the higher limit of 100% of 10 CFR Part 100 dose guidelines.
VI.
CONCLUSION Using the assumptions outlined above and Millstone's iodine concentration Technical Specifications, the resultant thyroid dose is 112 rems (for the limiting 0-2 hour EAB dose) for the case of no preaccident iodine spike.
This is larger than 10% of the exposure guidelines of 10 CFR Part 100, and, therefore, does not comply with the SRP Section 15.6.4 acceptance criterion.
However, if the GE Standard Technical Specifications (STS) for coolant iodine concentration is assumed, the resultant thyroid dose is 1.1 rems, and com-plies with the acceptance criterion.
The dose calculated with the higher STS for maximum short-term coolant iodine concentratioh (pre-accident iodine spike) is 22 rems thyroid, which also complies with the acceptance criterion of 100% of 10 CFR Part 100 dose guidelines for this case.
4-t Thcrefore, the doses from this acci i..t ca;inot meet the S.R.P 15.6.4 accept-cace criteria unless the GE STS fcr :ced o( coolant iodine concentratica are adopted for this plant.
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TABLE X" - 18.1 Assumptions for Calculatica of Dose from Main i
Steam Line Break Outside Centainment' Dur: tis.. of coolant release from break 5.5 sec.
4 Mass of coolant released 3.9 X 10 lbs.
Icdine Concentration in coolant (i)
Licensee 20 pCi/g maximum (ii) Staff (STS) 0.2 pCi/g D.E.1-131 (Equilibrium) 4.0 pCi/g D.E. 1-131 (Spike)
Dose Conversion Factor for thyroid dose 1.49 X 106 reg C1 Fraction of coolant released from break that is released to the environment 100%
Atmospheric dispersion coefficients 6.1X10-4-ff 0-2 hr EAB m
0-8 hr LPZ 3.0 X 10-5,jg m3
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itP nEitlEU OF !;1LLST0 tie 1 XV-19 LOSS OF CCCLA!;T ACCIDEtiTS P.ESULTit:0 FROM SPECTRU" 0F POSTULATED PIPIT!G BREM.5 6.ilHifd ihE REALIOR C00txNi PP.E55uRE o0dhDARY
.. IriTRODUCTI0ij Loss-of-cociant accidents (LOCA's) are panulated 12reaks in, the reactor coolant pressure boundary resulting in a loss of reactor coolant at a rate in excess of the capability of the reactor coolant makeup system.
LOCA's result in excessive fuel damage er m:lt unless coolant is rcpicnished. Excessive fuel d uage can result in significant radiological consequences to the environment via leakage from the containment.
SEP Topic XV-19 is intended to assure that the radiological consequences of a design basis LOCA from containment leakage and leakage from engineered safety features outside containment are within the exposure guideline values of 10 CFR Part 100.
II. REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a constructiorL
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2rmitoroperatinglicenseproYideananalysisandevaluationofthedesign 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. The LOCA 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 Part 100.11 provides dose guidelines for reactor siting against chich calculated accident dose consequences may be compared.
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.
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_2 Tcpic III-5.A, " Effects of Pipe Breaks on Structures Systems and Components Inside Centainment" ensures that the ability to c:hia.e safe shut-do-n or mitigate the consecuences of an accident are maintained. Various other topics examine such areas as containment integrity e-" 4*eletion, post accident chemistry, ESF systems, combustible gas control and control roca habitability.
IV. REVIEW GUIDELINES The review of the radiological consequences of a LOCA was conducted in accordance with the Appendices A, B, and D to Standard Revie. Plan 15.6.5 and Regulatory Guide 1.4 with the exception noted below. The plant is adequately designed against a LUCA and the dose mitigating features are acceptable if the resulting doses at the exclusion area and low population zone boundaries are within the guideline values of 10 CFR Part 100.
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V. EVALUATION
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Staff reviewed the licensee's submittal for evaluation of loss-of-coolant accident (LOCA). The licensee assumed a single failure criterion involving coincident loss of offsite power. Licensee states that the MSIV leakage and the bypass will be released into the condenser prior to venting to the atmosphere via the condenser vent line. Licensee's analysis assumes that MSIV leakage and the bypass leakage will not escape from the condenser during the first two hours and therefore will not contribute to exclusion area boundary (EAB) doses. Staff believes the licensee's assumption of two hour holdup of MSIV ano bypass leakage is overly optimistic. The li-censee has not submitted any information which uculd permit us to verify that i
.there are no MSIV ano Dypass leakage transport paths which release directly to the atmosphere.
1' 3-We performed an independent analysis of the dose consequences resulting from a LOCA assuming that the MSIV leakage and the bypass leakage were isolated intheconaenss[anowerereleasedtotheatmospnereviatheconcenser vent to the atnesobere after mixing with 50% of the condenser volume.
The staff used the :: 0:pheric dispersion factors oeveloped as a result of its i
review of tcpic II-2.C " Atmospheric Transport :nd Diffusion Characteristics for Accident Analytis". The important assumption used in the staff analysis are summari td in table XV-19-1.
I The calculated LOCA doses are summarized in Table XV-19-2 as Thyroid and Whole Body doses at the EAB and LPZ.
VI. CONCLUSION The staff concludes that the dose mitigation features provided to mitigate the consequences of a LOCA are such that the calculated doses are wiihin the 10 CFR 100 guidelines.
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TA3LE XV-19-1 Assu,T.ptions Caed In. Analysis Of Ti.e Of fsite Ra~diological Doses Following l,
A Design Basis LOCA.
- 1. Reactor Stretch Power:
2051 Mwt.
- 2. Fission Froducts Release Fractions:
25% Iodines 100% Noble Gases 3
- 3. Containmer.t Volume:
255,800 ft
- 4. Contain: rant Leak Rate:
0 - 100 Seconds 1.0% / day 100 Seconds - 30 days 0.762% / day i
- 5. SGTS Filter Efficiency:
90% for all forms of iodines (ReactorBuilding)
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- 6. MSIV Leakage into the Condenser:
0 - 100 Seconds 0.21% / day After 100 Seconds 0.16% / day
- 7. Bypass Leakage into the Condenser:
i 0 - 100 Seconds 0.045% / day After 100 Seconds 0.034% / day
- 8. ESF Leakage into Reactor Building:
1.0 GPM (Based on Total Coolant Volume =
5 3
1.2 x 10 f t )
- 9. Atmosphere Dispersion: As per Staff Reviews of Topic II-2,C.
(a) Containment and ESF Leakage: Elevated Release (b) MSIV Leakage and Bypass Leakage: Ground Level Release
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TABLE XV-19-2 LOCA DOSES II Exclusion Area Boundary
- Low Population Zone (0-2 hours)
(0-720 hours)
Thyroid Whole Body Thyroid Whole body rem rem rem re-l Containment Leakage 81.7 2.907 35.72
.664 f
MSIV Leakage 0.46 0.001 9.3
.006 Bypass Leakage 0.20 1.5
.002 ESF Leakage 7.0 5.8 52.3 0.67 TOTAL LOCA DOSE 89.4 2.91 1
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SEP REVIEW OF MILLSTONE 1 XV-20 RADIOLOGICAL CONSEQUENCE 5 GF FUEL DAMAGING ACCIDENTS I.
INTRODUCTION The safety objective of this topic is to assure that the offsite doses from fuel damaging accidents as a result cf fuel handling inside and out-side 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 operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the f acility with the objective of assessing the risk to public health and safety resulting from,
I, operation of the facility. A fuel handling accident in the fuel handling and storage facility resulting in damage to fuel cladding and subsequent j
release of radioactive material 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 Part 100 provides dose guidelines for reactor siting against which calculated accident dose consequences may be compared.
III. RELATED SAFETY TOPICS l
Topic II-2.C, " Atmospheric Transport and Diffusion Characteristics for Accident Analysis" provides the meteorological data used for calculating the offsite dose consequences, ik l
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- i The review of the fuel damaging accidents did not consider fucl d:msge :s a result cf crops of the spent fuel cask cr 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 objects is covered in two SEP Topics, IX-2,
" Overhead Handling Systems-Cranes" and XV-21, " Spent Fuel Cask Drop Accidents."
IV.
REVIEW GS:DELINES Accidents resulting from the movement of fuel inside and outside containment were rcvia.cd following the assumptions and procedures outlined in Standard Review Plest (SRP) Section 15.7.4 and Regulatory Guide 1.25.
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 event relative to other events which are evaluated against 10 CFR Part 100 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 accident, the methodology used by the staff is based on the fuel handling system described in the Northeast Nuclear 1
l Energy Company [[letter::A01012, Forwards Responses to NUREG-0654 Criteria & Facilities Emergency Plans Per NRC 800512 & s.Encls Available in Central Files Only.Emergency Plans Withheld (Ref 10CFR2.790)|letter dated June 30, 1980]], from W. G. Counsil, Senior Vice President, NNECO, to Director of NRR, NRC. The analysis was performed using
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i the guidelin2s and requirements of 5?P 15.7.4 and Regulatory Guide 1.25.
The list of assumptions and parameters used in the analysis is given in Table XV-70-1.
VI.
C0 'CLUSION j
The offsite thyroid and whole body doses for the postulated fuel handling accident cre 9 rem and 0.5 rem. respectively; these doses are well within the guideline values given in 10 CFR Part 100. The staff concludes that this systc.4 is ar.ceptable in mitigating the consequences of the fuel handling accidents.
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TABLE XV-20-1 FUEL HAfiDLIriG ACCIDEfC DOSE AS5.UMPTIONS Perer '_evel, MVT 2011 r,adici Peaking Factor 1.5 DC:ay time, hours 24 humber of fuel cssemblies affected 2
Nu.ber of fuel assemblies in core 580 Filter efficiencies for Iodine (SGTS) 90%
Iodine deccntamination factor 100 Activity release period 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> A/Q EAB (0-2 hr) 5.3 X 10-4 m3 sec.
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