ML19318C791

Safety Evaluation Re Postulated Fuel Handling Accident Inside Containment

ML19318C791
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	06/17/1980
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19318C789	List: ML19318C788 ML19318C791 ML19318C795
References
NUDOCS 8007020290
Download: ML19318C791 (6)

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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING THE FUEL HANDLING ACCIDENT INSIDE CONTAINMENT TROJAN NUCLEAR PLANT PORTLAND GENERAL ELECTRIC COMPANY DOCKET NO. 50-344

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Introduction By letter dated January 14, 1977, the staff requested Portland General Electric Company (the licensee) to evaluate the previously unevaluated poten-tial consequences of a postulated Fuel Handling Accident Inside Containment (FHAIC) at Trojan. The licensee submitted the evaluation of the FHAIC by letter dated March 7, 1977.

The licensee stated that the potential conse-quences of this accident are 29.8 Rem thyroid and.13 Rem whole body at the Exclusion Area Boundary (EAB). The licensee concluded that these doses are within the guidelines of 10 CFR Part 100.

The staff reviewed the evaluation and by letter dated September 21, 1977, requested additional information concerning containment isolation in the event of an accident.

By letter dated October 19, 1977, the licansee responded to the September 21, 1977 request.

The licensee stated that using radioactivity release assump-tions of Regulatory Guide 1.25, together with what was judged to be a reasonable model for limited mixing of radioactive materials in the air above the refueling cavity, it was shown that the offsite doses are well within the guidelines of 10 CFR Part 100. The staff reviewed the licensee's response and found it to be inadequate, and by letter dated November 15, 1977, the licensee was requested to justify his mixing model over the refueling cavity.

The licensee submitted by letter dated July 7, 1978, smoke tests of air-flow patterns at a similar refueling cavity.

Subsequently, on October 13, 1978, we requested by telephone that additional justification be provided of the mixing model used by the licensee.

Preceding our telephone request, measurements were made of thermal convection velocities above the Trojan refueling cavity.

By letter dated November 2, 1978, these results were applied to the evaluation of a FHAIC and were found by the licensec to substantiate the mixing model within the containment.

Evaluation We have completed our review of the licensee's March 7 and October 19, 1977, and July 7 and November 2, 1978, submittals which address the potential conse-quences of a spent Fuel Handling Accident Inside Containment (FHAIC). We have reviewed the July 7 and November 2,1978, letters in which the licensee 1

presented data to justify a dose reduction factor, due to mixing in the con-tainment, of 2.4 in the potential consequences of the FHAIC. We conclude that the data presented by the licensee does not provide adequate justification for the dose reduction factor of 2.4.

We agree with the licensee that (1) thermal convection upward from the pool and (2) the reduction in the flow rate into the ventilation exhausters around the refueling cavity with distance from the exhauster will allow some of the radioactivity to rise above the refueling cavity and become mixed inside containment. We do not agree that measurements 1

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made by the licensee at Trojan and the smoke tests made by another licensee inside a containment support a conclusion that more than half of the radio-activity released into the containment during the FHAIC would be kept inside containment by action of ventilation effluent monitors outside the containment ventilation isolation valves. We do not agree that the walls around and above i

the refueling cavity will cause thermal updrafts of about 300 fpm from the refueling cavity because such a flow rate was not measured at the test run at Trojan. We do not agree that the flow rate into the ventilation exhausters around the refueling cavity will fall off as rapidly as predicted by the licensee because of the existence of the pool water surface near the exhaus-ters and the proximity of the exhausters to each other.

These matters have not been addressed by the licensee in his submittals.

At present the sample line to the radiation monitors which automatically isolate the containment at Trojan is located downstream of the containment isolation valves. The licensee stated in his March 7,-1977 submittal that the time interval between t'a puff release reaching the isolation valves and the isolation valves shutting is 23.1 seconds.

Therefore, we concluded that the present automatic isolation system at Trojan could allow all the radioactivity from the damaged fuel assembly to be released before the containment could be isolated.

We have considered the placement of radiation monitor (s) above the refueling pool inside containment to cause containment isolation of the purge exhaust system before one half of the radioactivity released from a damaged fuel assembly is released to the environment.

By letter dated January 10, 1977, the licensee stated that the minimum transport time for radioactivity from the damaged fuel assembly to the inboard isolation valve is 2.8 seconds and the maximum allowable closure time for the inboard isolation valve is 5 seconds (Technical Specification 3/4.6.2).

The time for the monitor (s) to respond to the radioactivity and act to shut the valves should be about one second.

The outboard isolation valve which is required to close in 3 seconds is considered to fail leaving only the inboard isolation valve closing in the required 5 seconds.

Based on this, we conclude that this method to reduce the poten-tial consequences of the FHAIC is not acceptable because the estimated minimum time to isolate the containment purge exhaust system is greater than the minimum ti:ae for the radioactivity to reach the inboard isolation valve.

Therefoie, all the radioactivity could be releaFed from Containment before isolation could occur.

We have performed an independent analysis of the FHAIC.

Our assumptions and the resulting potential consequences at the EAB are given in Tables 1 and 2.

Table 1 is for use of a modified sauty grade charcoal filter for at least l

285 hours0.0033 days <br />0.0792 hours <br />4.712302e-4 weeks <br />1.084425e-4 months <br /> after shutdown and Table 2 is fuel handling operations without the charcoal filter after 215 hours0.00249 days <br />0.0597 hours <br />3.554894e-4 weeks <br />8.18075e-5 months <br /> after shutdown.

Inese charcoal filters would be in the containment purge exhaust system to filter the radioiadine released l

from containment during the FHAIC.

These filters must meet the design and testing requirements of Regulatory Guide 1.52 (Revision 2).

However, the filters need not meet certain require-ments that are associated with safety grade equipment.

The new filters need only meet the same seismic requirements of the existing ducts a d fans of the purge exhaust system because loss of the purge exhaust system in went of an 2

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The earthquake would mean no radioactivity could be released from this system.

filters and fans need not be_ redundant if the licensee incorporates a Limiting Conditions of Operation into the. Technical Specifications requiring operation of the filters during fuel handling operations for at least 285 hours0.0033 days <br />0.0792 hours <br />4.712302e-4 weeks <br />1.084425e-4 months <br /> after shut-The filters and fans also need not be on the emergency diesel because lors down.

of offsite power will shut down the purge exhaust system and prevent any release of radioactivity.

The charcoal filters for the purge exhaust system must be installed as discussed above, and suitable Technical Specifications must be implemented to govern their operation and surveillance. Appropriate Technical Specifications for the operation and surveillance of the. Containment Building Purge Filtration System are given in Enclosure 2.

Once installed and Technical Specifications implemented, the calcu. lated potential consequences of the postulated FHAIC will be appropriately within the guide-lines of 10 CFR Part 100 and will, therefore, be acceptable.

Appropriately 1

within the guidelines of 10 CFR Part 100 has been defined as less than 100 Rem i

to the thyroid. This is based on the probability of this event relative to other events which are evaluated against 10 CFR Part 100 guidelines. Whole bcdy doses were also examined, but they are not controlling due to decay of tha short-lived radioisotopes prior to fuel handling.

The potential conse-quences of this postulated accident at the low Population Zone Boundary are l

1ess than those given for the EAB in Table 1.

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1 A recent study 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 essumed for evaluating the Fuel Handling Accident In' side 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 raatation in the core.

Tha probability of the postulated fuel handling accident inside containment is small.

Not only have there been'several hundred reactor years of plant opera-ting experience with only a few accidents involving spent fuel being dropped 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 based on the assumption that a spent fuel assembly falls about 14 feet directly onto another assembly in the core; an impact which results in the greatest energy available 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 i

I vertical fall path.

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 assem-blies is sufficiently small that refueling inside containment is not a safety concern which requires remedial action beyond that addressed here.

Wa have, however, conservatively calculated the potential radiological con-f sequences of a fuel assembly drop onto the reactor core with the rupture of IJ. N. Singh, " Fuel Assembly Handling Accident Analysis," EG&G Idaho Technical Report RE-A-78-227, October 1978.

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L all the fuel pins in two fuel assemblies. We have also assumed for this pos-tulated accident that the source term for both spent fuel assemblies is that given in Regulatory Guide 1.25.

.This.is conservative becauss (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 recommended in Regulatory l

Guide 1.25.

The calculated potential radiological consequences at the exclu-sion area boundary and low population zone for the complete rupture of fuel 1

pins in two assemblies are twice the values given in Table 1.

Because these potential consequences are within the guidelines of 10 CFR Part 100 using the conservative ~ assumptions of Regulatory Guide 1.25, we have concluded that the potential consequences of this postulated accident are acceptable ar.d no additional restrictions on fuel handling operations and plant operating procedures are needed beyond those discussed above for limiting the dose from a single dropped fuel assembly.

Environmental Considerations The environmental impacts of an accident involving the handling of spent fuel inside containment have ber.n addressed in Section 6.1 of the Final Environ-mental Statement dated August 1973 for the operation of Trojan.

Conclusion As discussed above, the staff has evaluated the licensee's analysis of the postulated FHAIC. After performing an' independent analysis of the radiolog-ical consequences of a FHAIC to any individual located at the nearest exclusion boundary, 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.

For our conclusion to be valid, the licensee must either install charcoal filters on the containment purge exhaust system t

and implement suitable Technical Specifications like those in Enclosure 2 concerning their operation and operability for at least 285 hours0.0033 days <br />0.0792 hours <br />4.712302e-4 weeks <br />1.084425e-4 months <br />'during refueling operations, or implement revised refueling procedures to limit the consequences of FHAIC. These procedures would be to establish a minimum fuel decay time between shutdown and irradiated fuel movement of 285 hours0.0033 days <br />0.0792 hours <br />4.712302e-4 weeks <br />1.084425e-4 months <br />, or to conduct spent fuel movement during this period of time with the containment parge/ exhaust fans not in operation.

Dated:

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Table 1 ASSUPPTIONS FOR AND POTENTIAL CONSE00ENCES OF THE POSTULATED FUEL HANDLING ACCIDENTS AT THE EXCLUSION AREA BOUNDARY FOR TROJAN NUCLEAR PLANT Assumptions:

Guidance in Regulatory Guide 1.25 Power Level 3558 Mwt Fuel Exposure Time 3 years Power Peaking Factor 1.65 Equivalent Number of Assemblies Damaged 1

Number of Assemblies in core 193 Charcoal Filter Efficiency Organic and Elemental combined 95%

Decay time before moving fuel 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> 0-2 hours X/0 Value, Exclusion Area Boundary (ground level release) 6.9 x 10~4 sec/m 3

Doses, R<rm Thyroid Whole Body Exclusion Area Boundary (EAB)

Consequences from Accidents Inside Containment 9.8 0.7

• Assigned Activated carbon decontamination efficiency for an air filtration system designed to operate outside primary containment per Table 2, Regulatory Guide 1.52, Revision 2, March 1978.

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Tab'le 2 ASSUMPTIONS FOR AND POTENTIAL CONSEOUENCES OF THE POSTULATED FUEL HANDLING ACCIDENTS AT THE EXCLUSION AREA BOUNDARY FOR TROJAN NUCLEAR PLANT Assumptions:

Guidance in Regulatory Guide 1.25 Power Level 3558 Mwt Fuel Exposure Time 3 years Power Peaking Factor 1.65

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Equivalent Number of Assemblies Damaged 1

Number of Assemblies in Core 193 Charcoal Filter Efficiency Elemental and Organic Combined None Decay Time Before Moving Fuel 285 hours0.0033 days <br />0.0792 hours <br />4.712302e-4 weeks <br />1.084425e-4 months <br /> 0-2 hours X/0 Value, Exclusion Area Boundary (Ground Level Release) 6.9 x 10"4 sec/m3 Doses, Ren Thyroid Whole Body Exclusion Area Boundary (EAB)

Consequences from Accidents Inside Containment 100 0.25

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