ML19273B267
| ML19273B267 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 03/02/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19273B263 | List: |
| References | |
| NUDOCS 7904060056 | |
| Download: ML19273B267 (7) | |
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UNITED STATES y'
/g NUCLEAR REGULATORY COMMISSION j
W ASHINGTON, D. C. 20665 g
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 43 TO FACILITY OPERATING LICENSE NO. DPR-39 SUPPORTING AMENDMENT N0. 40 TO FACILITY OPERATING LICENSE N0. DPR-48 COMMONWEALTH EDIS0N COMPANY ZION STATION UNITS 1 AND 2 DOCKET N05. 50-295 AND 50-304 Introduction By letter dated September 21, 1978, Commonwealth Edison Company (the licensee) requested a change to Technical Specifications appended to Facility Operating Licenses DPR-39 and DPR-48 for Zion Station Units 1 and 2, respectively.
The proposed amendments would allow the containment equipment hatch of each unit to remain open during refueling operations. Associated with this proposal is the staff's review of a postulated fuel handling accident inside containment (FHAIC) for which, by letter dated January 14, 1977, we requested that the licensee submit an analysis of that postulated accident. By letter dated March 14, 1977, the licensee provided that analysis.
Discussion The proposed changes to Technical Specification 3.13 would allow the equipment hatch between the containment and the fuel handling building to remain open during all refueling outages.
As discussed in the licensee's submittal, the proposed changes would reduce occupational exposures by eliminating the wait time for passing through the equipment hatch during refueling outages.
The licensee has estimated that 0.9 to 2.7 man-rems per outage will be saved by leaving the equipment hatch open.
In addition, removal of the hatch during refueling would reduce outage time by a minimum of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per outage by eliminating at least one removal and replacement of the hatch and reducing maintenar.ce on the hatch through less usage.
7904060056
. The only potential accidents affected by the proposal would be the fuel handling accidents both inside containment and in the fuel handling building.
The FHAIC was not evaluated during licensing of the Zion Units.
In response to our request, the licensee submitted an analysis of the FHAIC on March 14, 1977.
Based on containment air mixing and containment isolation, the licensee concluded that the potential consequences of an FHAIC would be well within the guidelines of 10 CFR Part 100.
The licensee's evaluation, however, did not include the effect of the equipment hatch remaining open after the accident. The licensee's letter of September 21, 1978 addressed the effect of keeping the equipment hatch open during refueling and therefore during an FHAIC.
The licensee concluded that, with the hatch open, the consequences of an FHAIC would not exceed those calculated for an accident in the fuel handling building.
Evaluation We have reviewed the licensee's basis for proposing to keep the equipment hatch open during refueling with respect to occupational exposure control. We agree with the licensee that the present Technical Specifications restrict the movement of workers into and out of containment.
This restriction on movement will result in a higher occupational exposure compared to that received with the equipment hatch open.
We conclude that keeping the equipment hatch open during refueling will help keep occupational exposures "as low as reasonably achievable" (ALARA).
We have reviewed the licensee's proposal to determine if the revised Technical Specifications would decrease the safety margins with respect to postulated accidents. The only postulated accident which would be affected is the fuel handling accident both in the containment (FHAIC) and in the fuel handling building.
We have independently evaluated the potential consequences of an FHAIC.
With the equipment hatch open, both the containment and the fuel handling building should be at the same pressure.
As the fuel handling accident would not release a significant amount of energy there should be no driving force to cause exfiltration of the activity from the containment other than through the containment purge system and the auxiliary building ventilation system. The purge inlet system supplies 40,000 cfm of air from 32 ducts surrounding the periphery of the refueling pool.
These ducts are angled upward at 45 degrees and should promote
. A manipulator crane mixing of the activity in the containment.
ventilation system directs air down from the crane toward the pool, The purge exhaust intake duct is thus further promoting mixing.
located 17 feet above the operating floor near the containment wall We approximately 45 feet from the nearest refueling pool surface.
have estimated that the released activity would uniformly mix in the volume of air between the refueling pool and the purge exhaust duct This only on the side of the containment where the duct is located.
represents approximately 4% of the containment free volume.
The air would be drawn into the duct and be directed past the containment An air sample would be drawn from isolation valves to the plant vent.
the purge d uct downstream of the purge isolation valve and directed Upon detection of high radiation levels the to a radiation detector.
detector would initiate closure of the containment purge supply and In its letter of March 14, 1977, the licensee states exhaust valves.
that it would take approximately 32 seconds from the time the initial activity reaches the outboard isolation valve until initiation of valvt The licensee assumed a purge valve closure time of 12 second:,
closure.
however, we assumed a more conservative purge valve closure time of 60 seconds, as specified in the plant Technical Specifications, resulting in a total release time of 92 seconds. The resulting doses to an individual at the site boundary would be 84 Rem to the thyroid and 0.3 Rem to the whole body.
We have concluded that the potential consequences of the postulated fuel handling accidents are appropriately within the guidelines of 10 CFR part 100 and are, therefore, acceptable.
Appropriately within the guidelines of 10 CFR Part 100 has been defined as less than 100 Rem to the thyroid. This is based against the 10 CFR Part 100 Whole body doses were also examined, but they exposure guidelines.
are not controlling due to decay of the short-lived radioisotopes prior to fuel handling.
We have reviewed the staff's Safety Evaluation Report (SER) of October 6, 1972, for the Zion Station. That SER surmiarizes the analysis of the postulated fuel handling accident inside the fuel handling building.
The proposed change does not affect the SER analysis since the activity would be exhausted through the ESF grade auxiliary building ventilation In the event that the activity released as system charcoal filters.
a result of a fuel handling accident in the fuel handling building were to pass through the equipment hatch to the containment purge system rather than through the auxiliary building charcoal filters, the resulting offsite doses would be less than those estimated for the This is because the additional air volume in the fuel handling FHAIC.
building would result in much greater mixing of the activity than that assumed for the FHAIC analysis. The purge isolation time would remain the same, therefore, the total activity released would be less.
. If all the activity released from the fuel handling accidents were released to the environment with no credit taken for containment isolation, mixing inside containment, or effluent filtration, the thyroid dose to an individual at the site boundary would be 138 Rem.
The whole body dose to the individual at the site boundary would be 0.5 Rem. As this limiting case is well within the guidelines of 10 CFR Part 100, and the expected accident scenarios are as previously described, we conclude that it is not necessary for the licensee to test the mixing within containment or the containment purge or auxiliary building ventilation system with the equipment hatch open.
The assumptions used in calculating the consequences are listed in Table 1.
The low population zone doses would be lower due to a lower atmospheric dispersion factor.
To assure that the auxiliary building charcoal filters will be in operation to mitigate the consequences of a fuel handling accident, we modified the licensee's proposed revision of Technical Specification 3.13.2 to read, "The auxiliary building ventilation system shall be operating in the charcoal filter mode whenever irradiated fuel is being handled." The licensee has agreed to this modification.
A recent studyII as indicated that dropping a spent fuel assembly h
into the core during refueling operations may potentially cause damage to more fuel pins than has been assumed for evaluating the Fuel Handling Accident Inside Containment. This study has indicated that up to all of the fuel pins in two spent fuel assemblies, the one dropped and the one hit, may be damaged because of the embrittlement of fuel cladding material from burnup in the core. This may affect spent fuel assemblies with burnups of less than 5000 MWD /MT.
The probability of the postulated fuel handling accident inside containment is small.
In the several hundred reactor-years of plant operating experience there have been only a few cases of a spent fuel assembly being dropped into the core.
None of these dropped assemblies has resulted in measurable releases of activity.
The potential damage to spent fuel estimated by the recent study was based on an assumption that a spent fuel assembly falls about 14 feet directly onto one other assembly in the core and that the impact results in crushing the fuel pins in both assemblies. This type of impact is unlikely because the falling assembly would be subjected to drag forces in the water which should cause the assembly to skew out of a vertical fall path.
E. N. Singh, " Fuel Assembly Handling Accident Analysis," EG&G Idaho d
Technical Report RE-A-78-227, October 1978.
. Based on.the above, we have concluded that the likelihood of a spent fuel assembly falling into the core and damaging all the fuel pins in two assemblies is sufficiently small that refueling inside containment is not a safety concern which requires imediate remedial action.
We have, however, conservatively calculated the potential radiological consequences of a fuel assembly drop onto the reactor core with the rupture of all the 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 Guide 1.25.
This is conservative because, (1) these two assemblies should not have the power peaking factor and clad gap activity recommended in Regulatory Guide 1.25 and (2) the pool decontamination factor for inorganic iodine should be greater than that recomended in Regulatory Guide 1.25.
The calculated potential radiological consequences at the exclusion area boundary for the complete rupture of fuel pins in two assemblies are twice the values given in Table 2.
These conservatively calculated potential consequences are within the guidelines of 10 CFR Part 100; consequently, we have concluded that the potential consequences of this postulated accident are acceptable.
Based on the above considerations, we conclude that the proposed change to keep the containment equipment hatch open during refueling does not involve a significant hazard to the public and, therefore, is acceptable.
Environmental Considerations We have determined that the amendments do not authorize a change in effluent types or an increase in total amounts of effluents nor an increase in power leve' and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 651.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do 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 Comission's regulations and the issuance of these amendments will not be inimical to the comon defense and security or to the health and safety of the public.
Dated: March 2,1979
Table 1 Assumptions Used in Calculating the Radiological Consequences of a Fuel Handling Accident inside Containment
- Power Level (Mwt) 3250 Fuel Exposure Time (yrs) 3 Equivalent Number of assemblies damaged 1
Number of assemolies in core 193 Decay Time before Fuel Movement (hrs) 100 Power Peaking Factor 1.65 4
Volume of air activity mixed in (ft )
9.54 x 10 4
Purge exhaust rate (cfm) 4 x 10 Time Elapsed to Detection of Activity (sec) 32 Purge Valve Closure Time (sec) 60 Fraction of activity released from con-tainment 0.64 Fraction of activity released from spent fuel area building 0.36 Auxiliary building filter efficiency (%)
(from October 6, 1972 SER) 90 3
-4 Exclusion Area Boundary y /Q (s/m )
5.1 x 10 Other Assumptions per Regulatory Guide 1.25 Table 2 Radiological Consequences of Postulated Fuel Handling Accident Inside Containment Doses (Rem)
Thyroid Whole Body Equipment hatch closed 84 0.3 Equiprient hatch open 90 0.5
- Containment /Auxiliarj~ Building equipment hatch open.