ML19320B256
| ML19320B256 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/13/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19320B251 | List: |
| References | |
| NUDOCS 8007100110 | |
| Download: ML19320B256 (6) | |
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/,p SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 30 TO FACILITY OPERATING LICENSE NO. DPR-64 POWER AUTHORITY OF THE STATE OF NEW YORK INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET N0. 50-286 Introduction By letter dated January 17, 1977, the staff requesteo ne Pc-er Authority of the State of New York, Inc. (the licensee) to evaluate tne p eviously unevalu-ated potential consequences of a postulated Fuel Handling Accident Insice Containment (FHAIC) at Indian Point Unit 3 (Indian Pt. 3).
Tne licensee submi t'ted, in a letter dated March 21, 1977, an evaluation of the FHAIC.
The staff reviewed this submittal and requested in a letter dated May 5, 1977, that the licensee provide a basis for his model for mixing and for isolating the containment before a complete release of activity occurs.
The staff also requested an analysis including the worst single faile. curing tnis accident.
The licensee stated, by letter dated June 15, 1977, that the potential conse-quences for the worst single failure in the accident are 277.8 Rem Thyroid and 1.24 Rem Whole Body.
The licensee cescribed the assumptions used for containment mixing and locations of monitors which will automatically isolate the containment.
The staff reviewed the licensee's June 15, 1977, submittal and concluded additional actions were needed to provide adequate assurance that the potential consequences of this accident were less than the guidelines of 10 CFR Part 100.
By ' letter dated January 12, 1978, the staff proposed possible means to provide adequate assurance that the consequences of the FHAIC are within tne guidelines of 10 CFR Part 100 for Indian Point 3.
These proposais being:
(1) increase the minimum time after shutdown before refueling, (2) reduncant radiation monitors on the operating floor which will automatically isolate the containment, (3) a safety grade duct and charcoal filter on the purge exhaust from the containment, (4) smoke tests or other experiments or analysis which will demonstrate that the radioactivity released from the damaged fuel assembly woulc be mixed in the containment, or (5) conservative analysis which demonstrates that the containment would be isolated in a timely manner by the existing monitors assuming a single failure.
The licensee, in a letter dated February 14, 1978, agreed te increase the minimum time between shutdown and fuel handlir.g from ~3C 5ct. s to 12C hours in the Inc:an Doirt 3 Te:hnical Speci'ications.
T-i s 7:--ics' heci:ation chance as evalustad and arproved in the safety evata t':,-
catec ar:r. 22, 1978, fur Indian ?oint 3.
Tnis chance recucas the agrituce Of saicactivity in the spent fuel assemblies available for release durin; nis accicent and orcvides additional assurance tnat tn potent al cc se:cences are within the i
Part 100 guidelines.
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. In addition, in our review of Unresolved Safety Issues at Indian Point 3, we noted that a change in the Technical Specifications was needed in Section 3.8 to add protection against potential heavy load drops on spent fuel.
(See p. 27 of Enclosure 1 of ota- '.pril 9,1980 letter.)
By letter dated May 6,1980, the licensee proposed changes to the Technical Specifications in response to our February 27, 1980 letter.
These changes add surveillance requirements for the containment vent and purge system and add precautions for the control of heavy loads over the reactor when the vessel head is removed.
Evaluation We have completed our review of the licensee's March 21,1977, June 15, 1977, February 14, 1978 and May 6, 1980 submittals, which address the potential consequences of an accident involving spent fuel handling inside containment. We have performed an independent analysis of the FHAIC. Our assumptions and the resulting potential consecuences at the Exclusion Area Boundary are given in Table 1.
We concluae that testing the Purge System charcoal fil.ers esery 723 nours of system operation for a 90% methyl iodine removal ef ficiency at 9% re'ative humidity will provide adequate assurances that the cnarccal adsorcer iodine i
removal efficiency is at least that which we have given in Tacle 1 and have assumed in our ovaluation of the FHAIC.
Based on the acave ciscussec technical specifications on the charcoal adsorbers, degradation of tne adsorcers during operation of the Containment Purge System (CPS) and a margin of safety to assure the charcoal radiofodine removal efficiencies are at least the efficiencies assumed in our evaluation of the FHAIC, we have assigned a 70% charcoal radiciodine removal efficiency for the CPS.
We conclude that tne implementation of these Technical S ecificaticas into the Indian Point 3 Technical Specifications will provide acequate assurance tnat the potential consequences of a postulated FHAIC are appropriately within the guidelines of 10 CFR Part 100.
Appropriately within the guicelines of 10 CFR Part 100 has been defined as less than 100 Rem to the thyroid.
This is based on the probability of this event relative to other events which are evaluated against 10 CFR Part 100 exposure guidelines.
Whole body cosas were also examined, but they are not controlling due to decay c' tne snort-lived radio-isotopes prior to fuel handling.
In cur review, we did not require that the CPS be safety g ace anc cid not consider the Single Failure Criter'ia, IEEE Standards, seismic design and equipment quality group classification.
The CPS is not safety grade. We conclude that this is acceptable because the potential consequences of the postulated FHAIC are within the exposure guidelines of 10 CFR Part 100 with no
. credit given for operation of the CPS.
In addition, the surveillance requirements for the CPS filters discussed above are less than the require.
maats on safety grade ventilation filter svstens ce:a;se
- ave e :ctential
- sea;ences
- ? :nisacciuentaper: aria:elyw'tr!-
- e::s.re : ':i inis Of Fr Part 100, mc re stringent surveill ance req.f ra e-.s :r tri ----'afety grace CPS filters are not needed.
A cecent stuoy has incicated that drooping a scent
'.e' asse : '
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ccre during refueling ooerations may notentially ca.se :s a;5
- i fue' rins tnan hc5 ceen assumed for evaluating the FhAIC.
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as "':icated nat up to all of the fuel pins in two spent fuel asse.f es. t e ens crcpped and the one hit, may be daniaged because of the embr'tt:s e,t c' fuel :' adding material from radiation in the core.
The radiation e.--
'ti.sre t -o;:t occur within the fuel's first few mcaths of operations.
The probability of the postulated fuel handling accident ins'ce containment is small.
Net only have there been several hundred rea:t:r years of plant operating experience with only a few accidents involving spent f.e' reing crepted into the core, but ncne cf these accidents has resulted in.essura:'e releases of activity.
The potential camage to spent fuel estimated :y : e 3 ucy was based on the assumption that a spent fuel assembly falls acc;; il 'est dirs:'tly onto one other assembly in the cc.re; an impact which results 'n e greatest energy available for crushing the fuel pins in both assemblies.
'n's ype c' impact is unlikely because the falling assembly would be sucjec ed o c ag forces in the water which shculd cause the assembly to skew out of a var-ical fall path.
Based on the above, we have concluded that the likelihocd of a spent fuel assembly falling into the core and damaging all the fuel pins 'n two assemblies is sufficiently small that refueling operations inside containment are not a safety concern which requires immeciate remedial action.
Mc-e'ce, because there is a chance that more than one spent fuel assemoly may be camaged during refueling, we are reviewing the study and the probability ant consequences of dropping a spent fuel assembly in the core and damaging more fuel pins than the equivalent of one assembly.
The objective of this revie-is to cetermine if any additional restrictions on fuel handling operatiens ca plant coerating procedures are needed.
Any conclusions of this review wnich are applicable to this plant will be implemented.
We have calculated the potential radiological consecuences of a fuel assembly drop onto the core assuming all the fuel pins ir two spent f.ei assem li es are ruptured.
If, for both assemblies, we use the assumptions given in Regulatory Guide 1.25, and taking no credit for the non-ESF charcoal filters, the cotential consequences of this accident are greater than the guidelines of 10 CFR Part 100.
However, the source term defined in Regulatory Guide 1.25 is ccnservative Decause-(l) these two assemblies are unlikely to botn na/e tre nigh power 5" gn. "Fue; Asserst hanc!' g Ac:' cent Ar.a' ! 5 E2L3
- s : -e:--ical
- o;rt RE-A-Ti-227, Octcce-1975.
peaking factor and clad gap activi'ty used in Regulatory Guice 1.25, and (2) the pool decontamination factor for inorganic iodine may well be greater than that used in Regulatory Guide 1.25.
Taking into account more realistic values for
- power peaking factor, clad gap activity and pool decontamina-icn, we conclude that potential consequences of tnis postulated accicent shou:c not be greater than the exoosure guidelines of 10 CFR Part 100.
Because t ese note tial
- ensecuences a"e less than the guicelines of 10 CFR 21-t ' :. -e
. Eve concluded
- 1
- n: sc:iti:nal restricticr.s an fuel nanc!;ng c; era;':r.1 c: :
1--
- erating
- ' cecares are r.setec wrile e c eview is uncerwa;.
The results of this analysis warranted an investigation of a similar accident in the spant fuel pool even thouok it is located outside containment.
For this, a drop of 2-12 feet was postulat.
and the analysis performed in the same manner as previously described. Rasults indicate that in this scenario damage to the missile or target is minimal. No fuel pins in either fuel assembly were calculated to be ruptured.
4 After performing an independent analysis of the radiological consequences of an FHAIC to any individual located at the nearest exclusion coundary, the staff concludes that the doses for one assembly failure are appropriately within the guideline values of 10 CFR Part 100 and for failure of two assemblies are within the guideline values of 10 CFR Part 100 and are, therefore, acceptable.
We also conclude that the proposed change to the Technical Specifications which adds precautions for the control of heavy lcads over the reactor when the vessel head is removed is acceptable. With this change, we find that Indian Point 3 has substantial protection against potential heavy load drops on spent fuel in the reactor vessel and that there is reasonable assurance that the health and safety of the public is protected while the generic task
" Control of Heavy Loads Nea_r Spent Fuel" is completed and its results imple-mented.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental. impact and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
. Conclusion T
We have concluded, based on the considerations discussed above, that:
(1)- because the amendment does not involve a significant increase in the probability or consequences of accidents previously cor.sidered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: June 13,1980
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6-Tacle 1 ASSUMPTIONS FOR AND FOTENTIAL CONSEOUENCE5 :~ T-E ::STULATED FUEL HANDLING ACCIDENTS AT TnE ExCLUS;;' ARE-5:J ;A?-
FOR IhDIAN POINT STATION L W 3 Assumotions:
Guidance in Regulatory Guice 1.25 Power Level 3C25 Vat Fuel Exocsure Time 3 years Power Peaking Factor 1.55 4
Equivalent Number of Assemblies Danaged 1
Number of Assemblies in Core 193 Charcoal Filter Efficiency Elemental and Organic 7C percent Decay time before mcving fuel 120 hcars 0-2 hours, X/Q Value, Exclusion Area Boundary 2
(ground level release) 1.1 x 10
- se:/m*
Coses. Re.
1 Thyrcic Whole Body Exclusion Area Boundary (EAB)
Consequences from Accidents Inside Containment 73
.83 4
J Amendment No. 30 l
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