ML20054H893
| ML20054H893 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/16/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054H885 | List: |
| References | |
| NUDOCS 8206250146 | |
| Download: ML20054H893 (10) | |
Text
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ENVIRONMENTAL IMPACT APPRAISAL BY THE OFFICE OF NUCLEAR REACTOR REGULATION MAINE YANKEE ATOMIC POWER COMPANY MAINE YANKEE ATOMIC POWER STATION DOCKET NO. 50-309 1.0 Introduction The spent fuel storage capacity of the Maine Yankee Atomic Power Plant was 318 fuel assemblies (1 and 1/3 cores) when the plant was licensed in 1972. This licensed capacity was increased in 1975 to 953 fuel assemblies (4 and 1/3 cores) by reracking the spent fuel pool. This limited increase in storage capacity was in keeping with the expectation generally held in the industry that commercial fuel reprocessing would not provide near-term relief from diminishing available storage locations.
Comercial reprocessing of spent fuel has not developed as had been originally anticipated.
In 1975 the Nuclear Regulatory Comission directed the staff to prepare a Generic Environmental Impact State-ment (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 l
of Spent Light Water Power Reactor Fuel (NUREG-0575), Volumes 1-3 (the l
FGEIS) was issued by the NRC in August..1979.
In the FGEIS, consistent i
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 spent fuel pools.
In the discussion of this alternative it is noted that onsite storage capacity could be increased by fuel pin com-paction. Applications for approximately 76 spent fuel pool capacity increases have been reviewed and approved. 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.
C206250146 820616 PDR ADOCK 05000309 P
In addition to the alternative of increasing the storage capacity of the existing spent fuel pool, the FGEIS discusses in detail other spent fuel storage alternatives. The finding of the FGEIS is that the environ-mental impact costs of interim storage are essentially negligible, re-gardless of where such spent fuel is stored. A comparison of the impact-costs of various alternatives reflect the advantage of continued genera-tion of nuclear power versus its replacement by coal fired power generation.
In the bounding case considered in the FGEIS, that of shutting down the reactor when the existing spent fuel storage capacity is filled, the cost of replacing nuclear stations before the end of their normal life-time makes this alternative uneconomical.
This Environmental Impact Appraisal (EIA) addresses only the specific environmental concerns related to the proposed expansion of the Maine Yankee spent fuel storage capacity. This EIA consists of three major parts, plus a sumary and conclusion. The three parts are:
(1)descrip-tive material, (2) an appraisal of the environmental impact of the proposed action, and (3) an appraisal of the environmental impact of postulated accidents. Additional discussion of the alternatives to increasing the storage capacity of existing spent fuel pools is contained in the FGEIS.
1.1 Description of the Proposed Action The currently proposed reracking program would increase the number of storage cells from 953 to 1476. The licensee proposed in references 1 through 9 to further increase the storage capacity by disassembling the spent fuel assemblies and reassembling the fuel pins into bundles having a more consolidated fuel pin array. Each consolidated fuel bundle will contain up to 285 fuel pins as compared to the 176 fuel pins in a standard fuel assembly.. Both the standard and consolidated fuel assemblies are sized such that they can be stored in the proposed spent fuel racks. By consolidating fuel pins in this manner the storage capa-city will be increased from 1476 assemblies to the equivalent of 2038 standard assemblies plus the capacity for a full core discharge of 217 assemblies. This is also equivalent to 2390 standard assemblies without full core discharge capability.
The licensee also proposes to use an " emergency" rack located in the spent fuel cask laydown area for temporary storage of up to 121 spent fuel assemblies in their "as discharged" configuration. This rack will only be used for the temporary storage of fuel during a full core discharge should it be needed.
This would allow an additional 121 storage cells in the permanent racks to be used for standard / consolidated fuel assembly storage while retaining the capacity for a full core discharge.
The environmental impacts associated with the operation of Maine Yankee, as designed, were considered in the NRC's Final E,nvironmental Statement (FES) issued July 1972. The licensee was later authorized to increase the storage capacity from 318 to 953 assemblies. The environmental impact of that action was considered in an Environmental Impact Appraisal issued with Amendment 11 to the operating license dated October 31, 1975.
. In this EIA we have evaluated only additional environmental impacts which are attributable to the currently proposed increase in the spent fuel storage capacity of the plant.
1.2 Need for Increased Storage Capacity The plant now has a licensed fuel storage capacity of 953 spaces. Cur-rently 577 spent fuel assemblies are stored in the pool. The licensee's estimated spent fuel pool capacity requirements through the end of Maine Yankee's operating license are presented in Table 6.1 of reference 6.
Based on these estimates a full core reserve capacity at Maine Yankee will be lost in 1984 and normal reload discharge capacity will be lost in 1987.
The estimated storage capacity needed for plant operation through the end of the current operating license (year 2008) is 2450 standard assemblies.
The proposed expansion program, including new racks, fuel consolidation and the use of either the " emergency" rack or the reactor vessel.will provide this capacity for storage.
1.3 Fuel Reprocessing History Currently, spent fuel is not being reprocessed on a commercial basis in the United' States. The Nuclear Fuel Services (NFS) plant at West Valley, New York, was shut down in 1972 for alterations and expansion; in Septem-ber, 1976, NFS informed the Commission that it was withdrawing from the nuclear fuel reprocessing business.
The Allied General Nuclear Services (AGNS) proposed plant in Barnwell, South Carolina,'is not licensed to operate.
l The General Electric Canpany's (GE) Morris Operation (MO) in Morris, Illinois is in a decommissioned condition. Although no plants are licensed for reprocessing fuel, the storage pool at Morris, Illinois and the storage pool at West Valley, New York are licensed to store spent fuel. The storage pool at West Valley is not full, but NFS is l
presently not accepting any additional spent fuel for storage, even from those power generating facilities that had contractual arrange-ments with NFS.
On May 4, 1982, the license held by GE for spent fuel storage activities at its Morris operation was renewed for another 20 years; however, GE is also not accepting any additional spent fuel for storage at this facility.
2.0 The Facility The principal features of spent fuel storage at Maine Yankee, as they relate to this action, are described here as an aid in following the evaluations in subsequent sections of this environmental impact appraisal.
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. 2.1 The Spent Fuel Pool (SFP)
Spent fuel assemblies are intensely radioactive due to their fresh fission product content when initially removed from the core; also, they have a high thermal output. The SFP is designed for storage of these assemblies to allow for radioactive and thermal decay prior to shipping them to a reprocessing facility. Space permiting, the assemblies may be stored for longer periods, allowing continued fission product decay and thermal cooling. The Maine Yankee spent fuel pool is approximately 37 ft. wide X 41 ft. long X 38 ft, deep and is a stainless steel lined pool with 6 feet thick concrete walls and floor located in the Fuel Building.
2.1.2 Radioactive Waste Treatment Systems The plant contains waste treatment systems designed to collect and pro-cess the gaseous, liquid, and solid waste that might contain radioactive material. The waste treatment systems have been evaluated in the Final Environmental Statement (FES) dated July 1972. There will be no change in the waste treatment systems, as decribed in Section III-D of the FES, because of the proposed modification.
2.3 Spent Fuel Pool Cleanup System The spent fuel pool cleanup system consists of a filter demineralizer (precoat filter material and powdered anion and cation resin), filters, and associated piping, valves, and fittings. The system is designed to.
remove corrosion products, fission products, and impuritie's from the pool water.
Pool water purity is monitored by monthly chemical and
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radiochemical analysis. Demineralizer resin will be replaced when pool water samples show reduced decontamination effectiveness. The licensee has indicated that no change or equipment addition to the spent fuel pool cleanup system is necessary to maintain pool water quality for high den.
sity fuel storage.
The licensee has indicated that portable water filtra-tion equipment could be used to augment the installed cleanup system if deemed necessary to reduce general area dose rates during pin compaction.
2.4 Spent Fuel Pool Cooling and Makeup System There have been no modifications proposed to the existing MYAPC spent fuel pool cooling system (SFPCS) as a result of this proposed expansion program.
It is a single loop cooling systgm consisting of two parallel pumps, each rated at 750 gpm (3.75 x 100 po0nds per hour) and one heat exchange'r.
The design is such that the rupture of any one line connected to the pool will not result in draining the pool by siphoning. A flow of 850 gpm of primary component cooling water cools the heat exchanger. The rating of the SFPCS heat exchanger depends upon the pool water temperature.
0 Assuming the component cooling water inlet temperature is.85 F and a pool water temperature of 100oF, the heat exchanger rating is 4.72 x 106 BTU /HR. When the pool water is at its limit 1540Ftheheatexchangerisratedat22x10gngtemperatureconditionof BTU /HR.
. Three sources of makeup water have been identified: 1) the normal makeup system is the chenical volume control system (CVCS) which can provide 150 gpm in less than 15 minutes; 2) backup makeup sources of water are three primary grade water hose connections, each capable of supplying 20 gpm in less than 15 minutes; and 3) an emergency makeup source is the fire main system, which is capable of providing 150 gpm in less than 20 minutes.
3.0 Nonradiological Environmental Impacts of the Proposed Actions The nonradiological impacts of Maine Yankee operation, as designed, were considered in the FES issued July, 1972.
Increasing the amount of spent fuel stored in the existing spent fuel pool will not cause any new nonradiological environmental impacts which were not previously considered.
The amount of waste heat emitted by the plant as a result of the proposed action will increase slightly but will result in no ceasurable increase in impacts upon the environment.
4.0 Radiological Environmental Impacts of the Proposed Actions 4.1. Introduction The potential radiological environmental impacts associated with the expansion of the spent fuel storage capacity were evaluated and determined to be environmentally insignificant as addressed below.
The spent fuel pins which will be transferred and stored in the con-solidated fuel elements will have decayed a minimum of 120 days prior to consolidation (compaction will be initially performed with fuel which has been in the pool for a period greater than three years).
During the transfer of pins under water, both volatile and nonvolatile radioactive nuclides may be releasad to the water from either the surface of the pins transferred from the reactor to the SFP or from defects in the fuel cladding. Most of the material released from the surface of the assemblies will consist of activated corrosion products such as Co-58, Co-60, Fe-59, and Mn-54 which are nonvolatile. The radionuclides that might be released to the 'ater through defects in the cladding, such as w
Cs-134, Cs-137, Sr-89 and Sr-90 are also predominantly nonvolatile.
The primary impact of such nonvolatile radioactive nuclides is their contribution to radiation levels to which workers in and near the SFP would be exposed. The volatile fission product nucl. ides of most concern that might be released through defects in the fuel cladding are the noble gases (xenon an,d krypton), tritium and the iodine isotopes.
f Experience indicates however that there is little radionuclide leakage (volatile or nonvolatile) from spent fuel stored in pools after the fuel has cooled for several months. The predominance of radionuclides in the spent fuel pool water appears to be radionuclides that were present in the reactor coolant system prior to refueling (which become mixed with water in the spent fuel pool.during refueling operations) or crud dislodged from the surface of the spent fuel during transfer from the
reactor core to the SFP. During and after refueling, the spent fuel pool cleanup system reduces the radioactivity concentrations sufficiently.
It is theorized that most failed fuel contains small, pinhole-like perforations in the fuel cladding at the reactor operating condition of approximately 8000F. A few weeks after refueling, the spent fuel cools in the spent fuel pool so that fuel clad temperature is relatively cool, approximately 1800F. This substantial temperature reduction should reduce the rate of release of fission products from the fuel pellets and decrease the gas pressure in the gap between pellets and f
clad, thereby tending to retain the fission products within the gap.
In addition, most of the gaseous fission products have short half-lives and decay to insignificant levels within a few months. Based on the operational reports submitted by the licensee and discussions with the operators, there has not been any significant leakage of fission products from spent light water reactor fuel stored in the Morris Operation (MO)
(formerly Midwest Recovery Plant) at Morris, Illinois, or at the Nuclear Fuel Services' (NFS) storage pool at West Valley, New York. Spent fuel has been stored in these two pools which, when it was in an operating reactor, was determined to have significant leakage and was therefore removed from the core. After storage in the onsite spent fuel pool, this fuel was later shipped to either M0 or NFS for extended storage.
Although the fuel exhibited significant leakage at reactor operating conditions, there was no significant leakage from this fuel in the off-site storage facility.
4.2 Radioactive Material Released to the Atmosphere With respect to gaseous releases, the only significant noble gas isotopes attributable to storing consolidated assemblies for a longer period of time would be Krypton-85. As discussed previously, experience has demonstrated that after spent fuel has decayed 4 to 6 months, there is no significant release of fission products from defective fuel. However, we have conservatively estimated that up to an additional 16 curies per year of Krypton-85 may be released from the SFP when the modified pool is completely filled. This increase would result in an additional total body do of less than 0.10 mrem / year to an individual at the site boundary. This dose is insignificant when compared to the approximately 100 mrem / year that an individual receives from natural background radia-tion. The additional total body dose to the estimated population within a 50-mile radius of the plant is less than 0.001 man'-rem / year. This is small compared to the fluctuations in the annual dose this population would receive from natural background radiation. This dose represents an increase of less than 0.6% of the dose from the plant evaluated in the FES in Table 13. Thus, we conclude that the proposed modification will not have any significant impact on doses offsite.
Assuming that the spent fuel will be stored onsite for several years, Iodine-131 releases from spent fuel assemblies to the SFP water will not be significantly increased since the Iodine-131 inventory in the fuel will decay to negligible levels between refuelings.
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. The SFP cooling system is capable of maintaining the SFP water at a temperature of 1540F. During some full core discharges the rate of fuel movement from the reactor to the pool is limited by procedure.
Since the administrative limit of 1540F is not being changed as part of this action, we do not expect any significant change in the annual release of tritium or iodine, as a result of the proposed modification, from that previously evaluated in the FES.
Most airborne releases from the plant result from leakage of reactor coolant which contains tritium and iodine in higher concentrations than the spent fuel pool. Therefore, even if there were a slightly higher evaporation rate from the spent fuel pool, the increase in tritium and iodine released from the plant as a result of the increase in stored spent fuel would be small compared to the amount normally released from the plant.
If levels of radiciodine become too high, the release can be diverted to charcoal filters for the removal of radiciodine before its release to the environment. The plant radiological effluent Technical Specifications, which are not being cha.nged by this action, restrict the total gaseous releases of radioactivity from the plant including the SFP.
4.3 Solid Radioactive Wastes The concentration of radionuclides in the pool water is controlled by the filter and ion exchanger and by decay of short-lived nuclides. The activity is higher during refueling operations while reactor coolant water is intro-duced into the pool and decreases as the pool water is processed through the filter and ion exchanger.
The increase of radioactivity in the pool during the con,solidation of fuel pins from a low density fuel element to a high density one should be insignificant because the spent fuel to be consoli-dated is relatively cool thermally, and radionuclides in the fuel pins will have decayed significantly.
Should the pin compaction process produce excess amounts of crud or radio-activity from damage to a pin during transfer, the filter-demineralizer could be augmented with portable water filtration equipment positioned in close proximity to the particulate source. At present, the build-up of crud along the sides of the pool has not affected exposure rates along the sides of the pool. However, if crud build-up should become a problea, the licensee could remove this source by underwater vacuum.
cleaning. We therefore expect onlysa small in~ crease in radioactivity released to the pool water as a result of the proposed modification, as..
discussed in Section 4.1, because consolidation will be performed primarily with pins in zircalloy clad fuel elements which retain significant ductility.
Since potentially brittle fuel elements would not be consolidated, the SFP radioactivity levels should not be affected by the consolidation program.
We therefore conclude that the SFP purification system will keep concen-trations of radioactivity in the pool water levels reasonably close to that which exist prior to the modification.
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. While we believe that there should not be an increase in solid radwaste due to the modification, as a conservative estimate, we have assumed that the amount of solid radwaste is not expected to be increased by more than ghe current value of solid radioactive waste generated by the SFP of 20 ft per year. The annual average amount of solid waste shipped from a single pres-surized water reactor between 1973 and 1977 was about 6900 cubic feet per year.
If the storage of additional spent fuel does increase the amount of solid waste from the SFP purification system by about 20 cubic feet per year, the increase in total waste volume shipped would be less than 0.3 percent and would not have any significant environmental impact.
Disposal of the empty cut-up fuel cageg after the compaction process would add an annual volume of about 200 ft of solid radioactive waste to the total volume shipped. This would increase the total radioactive waste volume shipped about 3%. However, this volume should not.be considered as an increase in solid radwaste nor an impact on the environmental assessment of themodification since the fuel elements and cages would need to be disposed of eventually.
The major impact on the solid radioactive waste volume will be from disposal of the low density racks and associated boral poison cans.
Cutting and packaging the racks and cans after the modification is complete will add an additional 1400 ft3 to the solid radwaste volume for the year the modification is effected, assuming it is completed within a year. However, this is a non-repetitive burden and only re-presents about a 20% increase of solid radwaste volume for that year.
4.4 Radioactivity Released to Receiving Waters There should not be significant increase in the liquid release of radio-nuclides from the plant as a result of the proposed modification. The l
amount of radioactivity on the SFP filter and demineralizer might in-crease due to the additional spent fuel in the pool, but this increase of radioactivity should not be released in liquid effluents from the The plant radiological effluent Technical Specifications, which plant.
are not being changed by this action, restrict the total liquid releases of radioactivity from the plant.
I Visual observations can be made to determine if there are leaks in the SFP liner. To date, no water leakage from the SFP has been observed.
4.5 Occupational Radiation' Exposures We have reviewed the licensee's plan for modification of the spent fuel pool (SFP) by reracking with high density racks coupled with con-
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The occupational collective dose equiva-solidation of fuel assemblies.
l lent for the modification is estimated by the licensee to be about 12
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man-rems for removal and disposal of the existing racks and installation i
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9 of high density racks.
In addition, an annual dose equivalent of 12 to 40 man-rems, depending on the work area to be occupied, for pin compac-tion of 200 assemblies is estimated. We consider these to be reasonable estimates because they are based on realistic dose rates and occupancy factors for individuals performing a specific job during the reracking and consolidation operations. These operations are expected to provide a small increment to the plant's annual collective occupational man-rems.
2 Based on relevant experience, we find that the increment in onsite occupa-ti6nal dose, resulting from the proposed increase in stored fuel assemblies, l
has been reasonably estimated by the licensee based on dose rates in the spent feel pool area from radionuclide concentrations in the SFP water.
and from the spent fuel assemblies.. The spent fuel assemblies themselves will contribute a negligible amount to dose rates in the pool operations area because of the depth of water shielding the fuel. Consequently, the occupational radiation exposure resulting from the additional spent fuel in the pool is negligible. Based on present and projected operations in the spent fuel area, we estimate that the proposed modification should add less than about 13 percent to the total collective annual occupational radiation dose at this facility. Thus, we conclude that storing. additional-fuel-in the SFP should result in a small increase in doses received by occupational workers.
4.6 Radiological Environmental Impact of Postulated Accidents Although the new high-density racks and the use of consolidated fuel assemblies will accommodate a larger inventory of spent fuel, we have determined that the installation and use of the racks and the consolidation of fuel will not change I
the radiological consequences ~of a postulated fuel handling accident in the SFP l
area from those values reported in the Maine Yankee FES dated July 1972.
Additionally, the NRC staff has under way a generic review of load handling operations in the vicinity of spent fuel pools to determine the likelihood of a heavy load impacting fuel in the pool.and if necessary, the radiological consequences of such an event. Becausethelicenseeis prohibited from lifting P. spent fuel shipping. cask over the pool, until e approved by the NRC, and racks may not pass over spent fuel in the SFP, we have concluded that the likelihood of any heavy load handling acci-der.t is sufficiently small, that the proposed modification is acceptable, and no additional restrictions on load handling operations in the vicinity of the SFP are necessary while our review is under way.
5.0 Summary As discussed in Section 4.0, expansion of the storage capacity of the SFP will not create any significant additional radiological effects. The additional total body dose that might be received by an individual at the site boundary or the estimated population within a 50-mile radius I
is less than 0.10 mrem /yr and 0.001 man-rem /yr, respectively. These 1
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. doses are small compared to the fluctuations in the annual dose this population receives from background radiation. The population dose re-presents an increase 6f less than 0.6 percent of the dose from the plant evaluated in the FES. The occupational radiation dose of workers during the modification of the present storage racks is estimated by the licensee to be about 12 man-rem. This is a small fraction of the total men-rem burden from occupational exposure at the plant. Based on'present and projected operations in the spent fuel pool area, we estimated that the proposedmodification should add less than about 13 percent to the total collective annual occupational radiation dose at this facility. Thus, we conclude that storing additional fuel in the SFP should result in a small increase in doses received by occupational workers.
6.0 Basis and Conclusion for Not Preparing an Environmental Impact Statement We have reviewed this proposed facility modification relative to the requirements set forth in 10 CFR Part 51 and the Council on Environmental Quality's Guidelines, 40 CFR 1500.6. We have determined, based on this assessment, that the proposed license amendment will not significantly affect the quality of the human environment. Therefore, the Comission has determined that an environmental impact statement need not be pre-pared and that, pursuant to 10 CFR 51.5(c), the issuance of a negative daclaration to this effect is appropriate.
7.0 References 1.
Maine Yankee Atomic Power Company (MYAPCo.) letter to U.S. Nuclear Regulatory Commission (USNRC), dated September 18,' 1979, : Proposed Change 70.
2.
MYAPCo. letter.to USNRC, dated October 18, 1979.
3.
MYAPCo letter to USNRC, dated September 29, 1980, Proposed Change 70 Supplement No. 1.
4.
MYAPCo. letter to USNRC, dated July 28, 1981.
5.
Licensee's Status Report dated September 29, 1981.
6.
MYAPCo. letter to USNRC, dated October 5, 1981..
7.
MYAPCo. letter to USNRC, dated October 26, 1981, Proposed Change 70, Supplement No. 2.
8.
MYAPCo. letter to USNRC, dated February 10, 1982.
9.
MYAPCo. letter to USNRC, dated May 7, 1982.
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