ML18052B199
| ML18052B199 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 07/14/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18052B197 | List: |
| References | |
| NUDOCS 8707230712 | |
| Download: ML18052B199 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENVIRONMENTAL ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO EXPANSION OF THE SPENT FUEL STORAGE CAPACITY CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255 1.0 IDENTIFICATION OF PROPOSED ACTION The proposed amendment would permit the increase in the licensed storage capacity from 798 spent fuel assemblies to 892 spent fuel assemblies for the Palisades Plant spent fuel pool (SFP).
This would extend the full core discharge capability from the year 1988 to the year 1991.
The application for license amendment was submitted by letter dated February 20, 1986.
2.0 THE NEED FOR THE PROPOSED ACTION Commercial reprocessing of spent fuel has not developed as had been originally anticipated.
In 1975, the Nuclear Regulatory Commission directed the staff to prepare a Generic Environmental Impact Statement (GEIS, the Statement) on spent fuel storage.
The Commission directed the staff to analyze alternatives for the handling and storage of spent light water power reactor fuel with particular emphasis on developing long range policy.
The Statement was to consider alternative methods of spent fuel storage as well as the possible restriction or termination of the generation of spent fuel through nuclear power plant shutdown.
A Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel (NUREG-0575), Volumes 1-3 (the FGEIS) was issued by the NRC in August 1979.
In the FGEIS, consistent with long range policy, the storage of spent fuel is considered to be interim storage to be used until the issue of permanent disposal is resolved and implemented.
One spent fuel storage alternative considered in detail in the FGEIS is the expansion of onsite fuel storage capacity by modification of the existing SFPs.
Applications for approximately 118 SFP capacity increases have been received and 116 have been approved.
The remaining ones are still under review.
The finding in each case has been that the environmental impact of such increased storage capacity is negligible.
However, since there are variations in storage designs and limitations caused by the spent fuel already stored in some of the pools, the FGEIS recommends that licensing reviews be done on a case-by-case basis to resolve plant-specific concerns.
In the case of the Palisades Plant, the proposed expansion of fuel storage capacity will allow for two more cycles of operation before losing full core off-load capability. There are no existing, independent, licensed sp~nt fuel storage facilities available to store r *a7o12307f2.. 870714T'1 :. -
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The replacement power costs for those two cycles would be approximately $110 million at today's rates of purchase which would have to be paid by the Michigan rate payers in the Consumers Power Company service area.
3.0 ENVIRONMENTAL IMPACTS OF THE PROPOSED ACTION 3.1 RADIOLOGICAL IMPACTS 3.1.1 3.1. 2 Radioactive Wastes The plant contains radioactive waste treatment systems designed to collect and process the gaseous, liquid, and solid wastes that might contain radioactive material.
The radioactive waste treatment systems have been previously evaluated and found acceptable, and are discussed in the Final Addendum to the Final Environmental Statement (FES) dated February 1978 (U.S. NRC, NUREG-0343).
There will be no change in the radioactive waste treatment systems as a result of the additional storage racks.
The conclusions of the previous evaluation of the radioactive waste treatment systems are unchanged by the installation of new spent fuel storage racks.
Radioactive Material Released to the Atmosphere With respect to releases of gaseous materials to the atmosphere, the only radioactive gas of significance which could be attributable to storing additional spent fuel assemblies for a longer period of time would be the noble gas radionuclide Krypton-85 (Kr-85).
Experience has demonstrated that after spent fuel has decayed 4 to 6 months, there is no longer a significant release of fission products, including Kr-85, from stored spent fuel containing cladding defects.
To determine the average annual release of Kr-85, we assume that all of the Kr-85 released from any defective fuel discharged to the SFP will be released prior to the next refueling.
The enlarged capacity of the pool, therefore, has no effect on the calculated average annual quantities of Kr-85 released to the atmosphere each year.
Iodine-131 releases from spent fuel assemblies to the SFP water will not be significantly increased because of the expansion of the fuel storage capacity since the Iodine-131 inventory in the fuel will decay to negligible levels between refuelings.
Most of the tritium in the SFP water results from activation of boron and lithium in the primary coolant and this will not be affected by the proposed changes.
A relatively small amount of tritium is contributed during reactor operation by fissioning of reactor fuel and subsequent diffusion of tritium through the fuel and the iircaloy cladding.
Tritium release from the fuel essentially occurs while the fuel is hot, that is, during operations and, to a limited extent, shortly after shutdown.
Thus, expanding the SFP capacity will not significantly increase the tritium activity in the SFP.
3.1.3 3
Another potential source of airborne activity due to storing additional spent fuel assemblies in the SFP is through evaporation.
However, this is not expected to be a significant source of radioactivity for the following reasons:
(1) Storing additional spent fuel assemblies in the SFP is not expected to increase the bulk water temperature above those found during normal refuelings as used in the design analysis.
Therefore, the expected evaporation rate is about the same as before, and thus, there will not be any significant change in the annual release of tritium or iodine from the SFP.
(2)
On an annual basis, most airborne releases from the plant result from leakage of reactor coolant which contains tritium and iodine in higher concentrations than the SFP.
Therefore, even if there we~e a higher evaporation rate from the SFP, the potential increase in the release of tritium and iodine would be small compared to the amount normally released from the plant and that which was previously evaluated in the FES.
(3) Regardless of the sources, the plant is limited in its total releases of gaseous activity by the radiological effluent Technical Specifications.
Accordingly, the staff has assumed, for dose calculation purposes, that there will be no significant increase in the release of tritium or radioiodine due to evaporation from the SFP.
The concentration of radionuclides in the pool water is continuously controlled by the SFP cleanup demineralizer and by decay of short-lived isotopes.
The activity is highest during refueling operations when reactor coolant water is introduced into the pool, and decreases as the pool water is processed through the demineralizer.
Thereafter, the activity concentration has been and should continue to be dependent on the demineralizer resin cycle, with no long-term buildup.
The increase of radioactivity, if any, due to the proposed modification, should be minor because of the capability of the cleanup system to continuously remove radioactivity in the SFP water to acceptable levels.
Solid Radioactive Wastes We do not expect any significant increase in the amount of solid waste generated from the SFP cleanup systems due to the proposed modification.
About 50 cubic feet of 'dewatered resin has been produced approximately annually since 1976 at Palisades as a result of changing the demineralizer resin.
The trend is expected to continue.because of the chemistry controls applied to the primary coolant and the high integrity of the fuel cladding.
If the amount of solid waste is assumed to increase by one additional resin bed (50 cubic feet) per year due to the increased operation of the SFP cleanup system, the storage of additional spent fuel would increase the amount of solid waste by an average of about 50 cubic feet per year.
The annual average volume of solid wastes shipped offsite for burial from Palisades has been approximately 20,000 cubic feet.
- Thus,
3.1.4 3.1.5 4
even this postulated increase in annual waste volume shipped from Palisades would be less than 1% of the total annual waste volume.
This is a negligible increase and would not have any significant additional environmental impact.
The staff has also examined other potential sources of solid waste resulting directly from the reracking operation.
In the event the present spent fuel racks to be removed from the SFP because of the proposed modification are contaminated, they may be disposed of as low level solid waste (approximately 4,000 cubic feet).
We assumed that approximately 300 filters from a portable vacuum cleaner (approximately 2,000 cubic feet of waste) used to decontaminate the SFP floor will need to be disposed of because of the reracking.
Therefore, we have estimated that approximately 6,000 cubic feet of solid radwaste will be generated directly because of the proposed modification.
Averaged over the lifetime of the plant, this would increase the total waste volume shipped from the site by less than 1.%.
This is a negligible increase and will not have any significant additional environmental impact.
Radioactive Material Released to Receiving Waters There should not be a significant increase in the liquid release of radionuclides from the plant as a result of the proposed modification.
Since the SFP cooling and cleanup systems operate as a closed system, only water originating from cleanup of SFP floors and resin sluice water need be considered as potential sources of radioactivity.
It is expected that neither the quantity nor activity of the floor cleanup water will change as a result of this modification.
The SFP demineralizer resin removes soluble radioactive materials from the SFP water.
These resins are periodically sluiced with water to the spent resin storage tank.
The amount of radioactivity in the SFP demineralizer resin may increase slightly due to the additional spent fuel in the pool, but the soluble radioactive material should be retained on the resins.
If any radioactive material is transferred from the spent resin to the sluice water, it will be removed by the liquid radwaste system.
After processing in the liquid radwaste system, the amount of radioactivity released to the environment as a result of the proposed modification would be negligible.
Radiological Impact Assessment/Public Radiation Exposure This section contains the staff's estimates of the impacts on the public from the proposed SFP modification.
Major sources of radio-activity and principal environmental pathways were considered in preparing this section.
The principal source of radiation doses to individual members of the general public from releases from the SFP is exposure to Kr-85 releases from the SFP during subsequent storage.
The licensee expects no additional Kr-85 releases due to the SFP storage capacity modification as discussed in Section 3.1.2.
1 2
5 Regarding the environmental impacts of accidents, this expansion will allow the retention of 94 spent fuel assemblies that were discharged from the core thirteen to fifteen years ago.
The evaluation of the environmental impact of accidents in the spent fuel storage pool in the FES for Palisades assumed that the damaged fuel had been discharged from the core for only 30 days.
The fission product release from the thirteen to fifteen year old spent fuel would be insignificant in comparison to that previously evaluated.
The effects of a fuel handling accident during discharge is unchanged.
Accident sequences initiated by natural phenomena such as tornadoes, floods or seismic events are usually not included in the preparation of Final Environmental Statements since the radiological risk associated with such events (the probability of such events times the consequences of such events) would not be different in kind from those accident scenarios which have been analyzed and would be within the range of risk associated with dominant risk scenarios discussed in such statements. 1 Further, it is the staff's judgement, based upon design requirements relating to effects of natural phenomena, that there is a large uncertainty in the probability of the events occurring.
However, the conclusions contained in the draft report prepared by Brookhaven National Laboratory (BNL),
"Beyond Design-Basis Accidents in Spent Fuel Pools" (Generic Issue 82),
dated January 1987, necessitate additional staff review and discussion.
The draft BNL report (the report) concluded that the estimated risk results from beyond design basis accidents in spent fuel pools are comparable to the estimated risk posed by severe core damage accidents and that the uncertainty in the estimate is large (greater than a factor of 10). 2 The report estimated the probability of a spent fuel_gool fire resulting in a severe release to be approximately 10 per reactor year (Ry) when a_~eismic event was the initiating event and approximately 10 /Ry when initiated by a cask drop event.
The NRC staff has reviewed the draft report and concluded that the risk estimates for seismic initiated events are substantially lower than those indicated by the report, maybe by as much as two orders of magnitude.
The factors which were reviewed and are the bases for this conclusion are considerations of (1) the structural aspects of the spent fuel pool in conjunction with seismic hazards, and (2) the transport behavior of fission products potentially See for example the discussion of risk of accidents contained in the Final Environmental Statement for San Onofre Nuclear Generating Station, Units 2 and 3.
Final Environmental Statements issued prior to June 1980 did not contain a discussion of radiological risk from beyond design basis events.
The NRC staff has reviewed and commented upon the draft BNL report.
A final report is anticipated to be issued in July 1987.
The report examined accident scenarios which were postulated to cause failure of the spent fuel pool resulting in a fire of the fuel cladding and subsequent release of radioactivity.
6 released during a fire.
The first factor involved seismic initiated events. The report concluded that the frequency of a spent fuel pool fire resulting i§ a severe release from seismic induced events was approximately 10 /Ry for both BWRs and PWRs.
The report noted that available historical data is insufficient for obtaining meaningful site specific estimates of severe seismic events.
A methodology to derive an estimate of seismic hazards was developed by BNL to be used in their analysis.
The staff has reviewed the seismic assessment in the BNL study and concluded that these considerations are conservative.
The report noted that fragility curves for the massive reinforced concrete structure of LWR spent fuel storage pools have never been developed and, therefore, fragility assessment for other structures were used.
The BNL report assumed that the fragility of the spent fuel pool walls was comparable to that of the Zion auxiliary building shear walls.
The staff previously concluded in NUREG/CR-4334, 11An Approach to the Qualification of Seismic Margins in Nuclear Power Plants, 11 that the fragility of shear walls (for both BWRs and PWRs) is deemed to be conservative because 1) the limit on inelastic energy absorption capability (ductility limit) was considered to correspond to the onset of significant structural damage; and 2) the capacity for an individual wall corresponds to the lowest capacity wall among a number of walls used to resist seismic forces.
Thus, no credit was taken for energy absorption beyond the nominal ductility limit and the actual redistribution of the forces in the configuration of the walls.
Based upon the NRC staff's review of the assumptions, data and conclusions concerning.the postulated initiating seismic ey§nt and spent fuel pool fragility, the staff concludes that the 10 /Ry estimate contained in the report for the probability of a spent fuel pool fire resulting from a seismic event is likely overestimated by more than one order of magnitude.
The second factor considered by the staff in its review of the overall risk of spent fuel pool fires involved the retention of fission products on structures.
The method utilized in the draft BNL report to determine radionuclides released assumed that 100%
of noble gases, halogens and alkali metal~ would be released.
These radionuclides are the principal contributors to exposure and, therefore, to risk.
The staff has reviewed the assumptions contained in the report and concluded that the release values of radionuclides may be too conservative and the study failed to consider that a portion of these radionuclides could 11plate-out 11 on various structures and, therefore, reduce the consequences of the event.
The staff further concluded that the consideration of these factors could result in a reduction of the source term by approximately an order of magnitude.
Based upon the above, the staff believes that the BNL estimate of the risk of a spent fuel pool fire caused by a severe seismic event resulting in a severe release may be approximately two orders of magnitude too high.
7 Additionally, the staff reviewed the report and the assumptions for fires initiated by a cask drop event.
The report concluded that the probability of a spent fuel pool fire resblsting in a severe release for these events was approximately 10 /Ry.
Further, as noted in the report, the likelihood of cask drop accidents can be reduced by improving procedures, administrative controls and installing more reliable equipment already implemented in operating pl ants as discussed below.,
The staff considers the estimate in the BNL report to be at least an order of magnitude too high on the basis of plant specific improvements in the handling of heavy loads at nuclear power plants that have been implemented by compliance with the guidelines of NUREG-0612, 11Control of Heavy Loads at Nuclear Power Plants", as noted in Generic Letter 85-11 dated June 28, 1985.
These improvements include upgrades to the cranes and lifting rigs, operator training and identification of safe load paths.
These features were not included in the analysis of the risk resulting from a cask drop event.
As identified by BNL in their draft report, BNL arbitrarily us~~ a conditional probability of structural failure of the.pool of 10 given a cask drop on the edge of the pool.
Based on licensee's meeting its commitments to properly 1) maintain and test equipment,
- 2) train personnel, and 3) follow procedures, especially those procedures that do not allow transit over the pool edge unless the height of the cask is less than six inches above the floor, there should be at least one to two orders of magnitude reduction in the BNL estimated probability of pool failure caused by cask drop.
The previously discussed considerations regarding the conservative estimates of fission product transport under the beyond design basis seismic event scenerio from the BNL draft report also apply to the cask drop event.
In addition, mitigation is available from make-up water systems unaffected by the cask drop.
Therefore, the risk of a spent fuel pool fire caused by a cask drop event resulting in a severe release may be approximately two orders of magnitude too high.
Thus, the staff finds that the draft BNL report overestimates both the probability and the consequences of a spent fuel pool fire.
Specifically, it overestimates 1) the probability of a spent fuel pool fire resulting in a severe release from seismic considerations by approximately two orders of magnitude; and (2) the probability of a spent fuel pool fire resulting in a severe release from a beyond design basis cask drop by approximately two orders of magnitude.
Based on the above, the NRC staff concludes that continued re-racking of spent fuel pools until Generic Issue 82 is resolved does not represent a significant contribution to the overall plant risk to the health and safety of the public.
The draft BNL report indicates that the estimates of overall risk due to beyond design basis spent fuel pool accidents are comparable to present estimates for dominant core melt accidents.
Based on the discussion above, pending resolution of the generic issue, the NRC staff concludes that the risk associated with the potential for severe releases resulting from a spent fuel pool fire due to a loss
3.1. 6 3.1. 7 8
of pool water may be approximately two orders of magnitude smaller than the risks associated with dominant risk sequences for pressurized water and boiling water reactors.
Thus, the staff concludes that the risk associated with the potential for such fires is not a significant increment in radiological risk resulting from nuclear reactors, and does not result in a significant impact on the quality of the human environment.
Radiological Impact Assessment/Occupational Exposure This section contains the staff's evaluation of the estimates of the additional radiological impacts on the plant workers from the proposed operation of the modified SFP.
The occupational exposure for the proposed operation of the modified SFP is estimated by the licensee to be 3.3 person-rems based on the licensee's detailed breakdown of occupational dose for each phase of the operation.
This dose is a small fraction of the average annual occupational dose of 500 person-rems for all plant operations.
The small increase in radiation dose should not affect the licensee's ability to maintain individual occupational doses within the limits of 10 CFR 20, and is as low as reasonably achievable.
Normal radiation control procedures in accordance with the guidelines of Regulatory Guide 8.8 should preclude any significant occupational radiation exposures.
Based on present and projected operations in the SFP area, we estimate that the proposed operation of the modified SFP should add only a small fraction to the total annual occupational radiation dose at this facility.
Thus, we conclude that the proposed storage of spent fuel in the modified SFP will not result in any significant increase in doses received by worker.s.
Summary Based on its review of the radiological impacts of the proposed expansion of the SFP at Palisades, the staff concludes that:
- 1.
There will be a negligible increase in gaseous, solid and liquid radioactive material as a result of the SFP expansion itself or the continued storage of the additional fuel assemblies.
- 2.
There is no impact on the public since there is no increase in the calculated average annual quantities of Kr-85 released to the atmosphere.
- 3.
Total occupational exposure for the SFP modification is only a very small fraction of the average total annual occupational dose, and the licensee has taken appropriate steps to ensure that occupational doses will be maintained as low as reasonably achievable and within the limits of 10 CFR Part 20.
Therefore, the staff finds that there will be no significant additional environmental radiological impact attributable to the proposed reracking and modification to increase the spent fuel storage capacity.
9 3.2 NON-RADIOLOGICAL IMPACT The spent fuel storage racks that will be removed from the pool will be decontaminated and will be disposed of either as low level radioactive waste or as non-radioactive waste, depending on the effectiveness of decontamination.
Because of the small quantity (approximately 4,000 cubic feet), this should pose no significant environmental problem.
The new assemblies will be fabricated at a Westinghouse facility at Pensacola, Florida, and moved directly to the fuel pool area for installation.
Installation is not expected to impact terrestrial resources not previously disturbed during original plant construction.
The only non-radiological discharge altered by the fuel pool modification is the waste heat.
The contribution of the thirteen year old and older fuel assemblies to the total plant heat discharge will be negligible.
Heat is removed from the fuel pool by the SFP cooling system.
This is a completely closed system which uses a heat exchanger to transfer the removed heat to the Component Cooling Water System.
This system transfers the heat to the Service Water System which takes suction from and discharges to Lake Michigan.
The total heat load will be increased by less than one percent.
Because there is no significant environmental impact attributable to the discharge of waste heat from the plant as indicated in the FES dated June 1972 and the very small increase which will occur as a result of the fuel pool expansion, the staff finds the impact of the additional heat load to be negligible.
The licensee has not proposed any change in the discharge of chemicals nor changes to the National Pollutant Discharge Elimination System permit in conjunction with the fuel pool modification.
No increase in service water usage is proposed.
Therefore, we conclude that the Palisades Plant spent fuel storage expansion will not result in non-radiological environmental effects significantly different from those already reviewed and analyzed in the FES.
4.0 ALTERNATIVE USE OF RESOURCES This action does not involve the use of resources not previously considered in connection with the Nuclear Regulatory Commission's Final Environmental Statement dated June 1972 related to this facility.
5.0 AGENCIES AND PERSONS CONSULTED The NRC staff reviewed the licensee's request and did not consult other agencies or persons.
6.0 BASIS AND CONCLUSIONS FOR NOT PREPARING AN ENVIRONMENTAL IMPACT STATEMENT The staff has reviewed the*proposed modification to the Palisades Plant relative to the requirements set forth in 10 CFR Part 51.
Based upon the environmental assessment, the staff concludes that there are no
10 significant radiological or non-radiological impacts associated with the proposed action and that the proposed license amendment will not have a significant effect on the quality of the human environment.
Therefore, the Commission has determined, pursuant to 10 CFR 51.31, not to prepare an environmental impact statement for the proposed amendment.
Dated July 14, 1987 Principal Contributors:
T. Wambach, Project Manager C. Nichols, Radiation Protection Branch