ML19323B229
| ML19323B229 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/09/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323B225 | List: |
| References | |
| NUDOCS 8005120104 | |
| Download: ML19323B229 (5) | |
Text
8 0 051 e 610'/
[) hm
. UNITED STATES
[O s
s_,
ed
! 9, _. J,/ %
NUCLEAR REGULATORY COMMISSION
< n m
=
a,. \\'s-C. d,.
C WASHING TON. O. C. 20555 f
5 b ~ &a%+s ! j:'e m"s*
s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATICN REGARDING THE FUEL HANDLING ACCIDENT INSIDE CONTAINUENT JOSEPH M.
FARLEY *;UCLEAR PLANT UNIT NO. 1 ALABAMA POWER COMPANY C0CKET No. 50-3?S Introduction Ey letter dated March 11, 1977, the staff requested the Alabama Power Company
- APCa) to evaluate the potential consequences of a costulated fuel handling accident inside containment (FHAIC) at Farley Nuclear Plant Unit NO. 1 Tne licensee sutnitted the evaluatien of the FHAIC in Amendment No. 67 to the Final Safety Analysis P.ecort (FS AR) cated September 9,1977.
In addition, by letter dated June S,1978, the staff requested APCo to modify the charcoal filter for the containment purge system to meet seismic Category I requirements.
These modifications were required to assure that the consequences of a FHAIC at Farley were well within the guidelines of 10 CFR Part 100. APCo's response dated August 7,1973 stated that the modifications associated with the charcoal adsorbers in the con-tainment purge exhaust system would be completed prior to startuo following the first refueling outage in accordance with NRC reouirements.
The licensee further stated that the analysis for the FHAIC, which includes credit for the charcoal adsorbers in the containment ource exhaust system, is provided in FSAR Section 15.4.5.4 (Paragraph i5.a.5.4.c, Case 1).
APCo concluded that, since this analysis demonstrates that the calculated offsite doses for the Farley plant are within the NRC determined limit of 250 of 10 CFR Part 100 guideline values, no revised analyses are required.
The following is our evaluation of APCo's subnittals.
Evaluation The staff has reviewed the FHAIC analysis crovided by APCo in FSAR Arendment No. 63. We determined that the calculated offsite dose due to radioactive ~ cas released from the costulated damaced fuel aculd exceed the staff's acceptance criteria (255 of 10 CFR Part 500) for the postulated l
l FHAIC accident.
In making this determination, we did nct alicw credit for the reduction of iodine activity by the charcoal adsorbers because l
these units were not seismically qualified. Thus, APCo, in rescense to
- "e staff's recuest by letter dated June 3,1973, pro:csed to rodify the cnarcoal filters for tne containment ourga system to eet seismic Categor, I recuire ents.
F5 AR Amendment No. 71, dated December 29, 197-3, irdicates tnat the filters.v0uld be ide seismically qualified ;rior to
- crt:.; fcilowing the first re#ueli ; oucage for Unit i and prior to
'nitial fuel ic3dil; for Unit E.
The licensee has orill; statet that the
.:e on jnit i vas compie:e; daring t"e "arca H79 cutage.
~be staf#
.: cludes that :his is a 5:tisfactory e3ns of crovid'n; assurirce that
. the consequences of a FHAIC will be well within the dose guidelines of 10 CFR Part 100 and are, therefore, acceptable. Assumptions used in the analysis are shown in Table 1 and the potential offsite doses are shown in Table 2.
A recent studyl has indicated that dropping a spent fuel assembly into the core during refueling operations may potentially cause damage to -more fuel pins than has been assumed for evaluating the FHAIC. This study has indicated that up to all the fuel pins in two spent fuel assemblies, the one dropped ano the one hit, may be damaged because of the embrittlement o.f fuel claading ma-terial from radiation in the core.
The probability of the postulated fuel handling accident inside containment is small. Not only have there been several hundred reactor-years of plant operat-ing experience with only a few accidents involving spnt fuel being cropped into the core, but none of these accidents has resulted in measurable releases of activity. The potential damage to spent fuel estimated by the study was baseo on the assumption tnat a spent fuel assembly falls about 14 feet directly onto one other assemoly in the core; an impact wnicn results in the greatest-energy available for crushing the fuel pins in both assemblies. This type of impact is un nKeiy oecause the falling assembly woulo be subjected to drag forces in the water which should cause tne assembly to skew out of a vertical fall path.
Based on the above, we have concluded that the likelihooo of a spent fuel as-sembly falling into the core and camaging all the-fuel pins in two assemblies is sufficiently small that refueling inside contai.nment is not a safety con-cern which requires immediate remedial action.
We have, howt.
r, conseFvatively calculated the potential radiological conse-quences of a fuel assembly drop onto the reactor core with the rupture of all tne fuel pins in two fuel assemblies.
We have also assumed for this postulated accident that the source term for both spent fuel assemblies is that given in Regulatory Guice 1.25.
This is conservative oecause (1) these two assemolies snould not have the power peaking factor and clad gap activity recommended in Regulatory Guice 1.25 and (2) the pool decontamination factor for inorganic io-cine snould be greater than that recommended in Regulatory Guice 1.25.
The cal-culated potential radiological consequences at the exclusion area boundary anc iow population zone for the complete rupture of fuel pins in two assemblies are twice the values given in Table 1.
Because tnese potential consequences are within the guidelines of 10 CFR Part 100 using the conservative assumptions of Regulatory Guice 1.25, we have concluded that the potential consequences of :nis postulated accident are acceptable and no accitional restrictions on fuel hand-ling operations and plant operating procecures are neeced.
The results of this analysis warranted an investigation of a similar accident in the spent fuel pool.
For this, a drop of 2-1/2 feet was postulated and tne,
analysis performed in the same manner as previously descriced.
Results inoicate nat in this scenario camage to tne missile cr target is minimal.
No fuel pins in either fuel assemoly were calculated to be rupturea.
I J N. Singh, "Fael Assembly Handling Accident Analysis," EG&G Idaho Technical Report RE-A-78-227, October 1978.
- Environmental Considerations The environmental impacts of an accident involving the handling of spent fuel inside containment have been addressed in Section VI. A of the Final Environ-mental Statement (FES) dated June,1972, for the operation of Farley Nuclear Plant Unit No.1.
Conclusion The staff has evaluated APCo's analysis of the postulated FHAIC.
Af ter performing an independent analysis of the radiological consequences of a FHAIC to any individual locatec at the nearest exclusion area boundary, the staff conciudes tnat tne cases for one assembly failure are well within the guideline values of 10 CFR Part 100 and for failure of two assemblies within the guideline values of 10 CFR Part 100 and are, therefore.. acceptable.
The staff has also concluced, based on the considerations discussed above, that: (1) because this action does not involve a significant increase in the probability or consequences of accidents previousl.y considered and does not involve a significant decrease in a safety margin, it does not involve a sig-nificant 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 De conducted in compliance with the Commission's regulations and the issuance of this amendment will not De inimical to the common defense and security or to the health ano safety of the putlic.
Date:
April 9,1980
(
l l
4_
TABLE 1 FUEL HANDLING ACCIDENT ASSUMPTIONS Shutdown Time, hours 100 Total Number of Fuel Rods in the Core 32,028 Number of Fuel Rods Involved in the Refueling Accident 204 Power Peaking Factor 1.65 Iodine Fractions Released from Pool Elemental 0.75 Organic 0.25 ffective Filter Efficiency, %
Elemental 90 Organic 30 3
X/Q Values, Sec/n
-4 0-2 hours @ 1235 meters 3.9 x 10
-5 0-8 hours 0 3218 meters 8.3 x 10
- Memorandun from L. Hulnan to G. Knighton, dated September 4,1979.
t i
4 i l
TABLE 2 POTENTIAL OFFSITE DOSES DUE TO DESIGN BASIS ACCIDENTS Two-Hour Course of Accident Exclusion Boundary Low Population Zone (1260 Meters)
(3200 Meters)
Thyroid Whole Body Thyroid Whole Body
)
Accident (rem)
(ren)
(rem)
(rem)
Fuel Handling 27
<1 6
<1 i
v 4
4 t
l-i
--m-
_