ML20082C502
| ML20082C502 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 07/31/1991 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20082C500 | List: |
| References | |
| NUDOCS 9107190153 | |
| Download: ML20082C502 (169) | |
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- an n-w yenn vy EUAWEH5$y5E$$$$2$!I$..,_.m :$p .~ wdme SA M d Supplement to Applicant's Environmental Report - Post Operating j License Stage - Decommissioning l Fort St. Vrain Nuclear Generating Station O 0 Public Service Company of Colorado July 1991
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w-SUPPLEMENT TO APPLICANT'S ENVIRONMENTAL REPOFtT POST OPERATING LICENSE STAGE FOR PROPOSED DECOMMISSIONING OF THE FORT ST. VRAIN NUCLEAR GENERAflNG STATION Public Service Company of Colorado i July,1991 -
TABLE OF CONTENTS SECT!OII TITLE PAGE
1.0 INTRODUCTION
1-1 1.1 Purpose 1-1 1.2 Decommissioning Plan Description 1-2 1.3 Radioactive Waste Disposal 1-3 1.4 Regulatnry Considerations 1-3 1.5 Environmental Technical Specifications 1-4 1.6 Equipment Salvage 1-5 1.7 Need for Proposed Action 1-5 1.8 References 1-5 2.0 HISTORY AND CURRENT STATUS OF FACillTY 2-1 2.1 Nuclear Operating History 2-1 2.2 Radionuclide Inventory 2-3 2.3 Radiation Survey Results 2-8 2.4 Current Facility Status 2-13 2.5 References 2-13 3.0 PROPOSED ACTION 3-1 3.1 Introduction 3-1 3.2 Site Cleanup, Final Site Survey ard Site 3-2 Release for Unrestric'.ed Use 3.3 Criteria for Release of Equipment and 3-4 Non Radioactive Waste 3.4 Radioactive Waste 3-5 3.5 Asbestos 3-17 3.6 Manpower Levels and Schedule 3-18 3.7 References 3-18 4.0 ENVIRONMENTAL EFFECTS OF DECOMMISSIONING 4-1 ACTIVITIES 4.1 Affected Environment 4-1 4.2 RadiologicalImpact from Routine Activities 4-7 4.3 Radiation Protection Program and Occupational 4-13 Radiation Exposure 4.4 Ambient Air Quality 4-22 4.5 Effects of Chemical and Biocide Discharges 4-23 4.6 Effects of Sanitary Waste Discharge 4-24 4.7 Endangered Species 4-24 4.8 Other Effects 4-26 4.9 References 4-27 ii J
TABLE OF CONTENTS (Cont.)
SECTION TITLE PAGE 5.0 ENVIRONMENTAL IMPACTS OF ACCIDENTS 5-1 5.1 Facility Accidents involving Radioactivity 5-1 5.2 Transportation Accidents involving Radioactivity 5-20 5.3 Other impacts 5 22 5.4 References 5-22 6.0 ALTERNATIVES TO PROPOSED ACTION 6-1 6.1 Available Alternatives 6-1 6.2 No Action 6-1 6.3 Alternative Decontamination and Decommissioning 6-2 Plans 6.4 Radioactive Waste Transportation Alternatives 6-4 6.5 References 6-4 7.0 ANALYSIS 71 7.1 Proposed Action 7-1 7.2 Alternative Decommissioning Plans 7-2 7.3 Conclusions 7-3 8.0 ENVIRONMENTAL APPROVALS 8-1 8.1 Federal Requirements 81 8.2 State and Local Requirements 8-2 8.3 References 8-3 9.0
SUMMARY
AND CONCLUSIONS 9-1 iii
LIST OF TABLES TABLE TITLE PAGE 2.1-1 Fort St. Vrain Milestones and Major Events 2-15 2.2-1 Activation Analysis Results 2-16 2.2 2 PCRV Dose Rates in Air at 5 Years After Shutdown 2-18 2.2-3 Estimated Plateout Concentration of 2-19 Major Primary Circuit Components at EOC5 2.2-4 Integrated Plateout in Each Primary Circuit 2-20 Component at EOC5 2.2-5 Estimated Curie Total at Fort St. Vrain 2-21 2.3-1 Radiological Survey Summary 2-22 3.4-1 PCRV Waste Classification and Volume Reduction 3-19 3.4-2 Contaminated BOP Waste Classification and Volume 3-20 Reduction 3.4-3 PCRV Waste Volume Estimates 3-21 3.4-4 BOP Waste Volume Estimates 3 22 4.1-1 Sources of Public Water Supply 4 28 4.1-2 Fort St. Vrain Wind Speed and Duration 4-29 4.1-3 Fort St. Vrain Wind Frequency Distribution 4-30 4.3-1 Summary of Occupational Radiation Exposure 4-31 Estimates 4.3-2 Occupational Radiation Exposure Estimates 4-32 5.1-1 Summary of Postulated Decommissioning Accident 5 25 Scenarios 5.1-2 Doses Due to Postulated Decommissioning 5-26 Accidents 5.1-3 Curie Totals in Activated PCRV Concrete 5-27 5.1-4 Percentage Contribution of Activation Products 5-28 in First 6 inches of Top Head Concrete 5.1-5 Waste Volume / Activities Estimates for the PCRV 5-29 iv
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LIST OF FIGURES FIGURE TITLE PAGE 2.1-1 Fort St. Vrain Power Generation 2-24 2.2-1 PCRV and Internal Components 2-25 2.3-1 Compactor Building Radiation Survey 2-26 2.3-2 Radiochemistry Laboratory Radiation Survey 2 27 2.3-3 Turbine Building Radiation Survey - Level 5 2-28
.(Eh 4791')
2.3-4 Turbine Building Radiation Survey - Level 6 2-29 (EL. 4811')
2.3-5 Turbine Building Radiction Survey - Level 7 2-30 (EL. 4829')
2.3-6 Turbine Building Radiation Survey - Levels 8, 2-31 10 and 11 (EL. 4846', 4864', 4884')
2.3-7 Turbine Building Radiation Survey - Level 12 2-32 and 13 (EL. 4904', 4921')
2.3-8 Reactor Building Radiation Survey - Level 1 2-33 (EL. 4740')
2.3 9 Reactor Building Radiation Survey - Level 2 2-34 (EL. 4756')
2.3-10 Reactor Building Radiation Survey - Level 3 2-35 (EL. 4771')
2.3-11 Reactor Building Radiation Survey - Level 4 2-36 (EL. 4781')
2.3-12 Reactor Building Radiation Survey - Level 5 2-37 (EL. 4791')
2.3-13 Reactor Building Radiation Survey - Level 6 2-38 (EL. 4816')
2.3-14 Reactor Building Radiation Survey - Level 7 2-39 (EL. 4829')
2.3-15 Reactor Building Radiation Survey - Level 8 2-40 (EL. 4839')
2.3-16 Remtor Building Radiation Survey - Level 9 2-41 (EL. 4g49')
2.3-17 Reactor Building Radiation Survey - Level 10 2-42 (EL. 4864')
2.3-18 Reactor Building Radiation Survey - Level 11 2-43 (EL. 4881') Refueling Floor l 2.3 19 Reactor Building Radiation Survey - Level 12 2-44 l and 13 (EL. 4904' and 4921')
2.3-20 Turbine and Reactor Building Elevations 2-45 l
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j LIST OF FIGURES (Cont.)
FIGURE TITLE PAGE 3.6-1 Decommissioning Tasks Schedule 3-23 4.1-1 Area Surrounding Fort St. Vrain Site 4-35 4.1-2 Geological Structure of General Area 4-36 4.1-3 Contours of Bedrock 4-37 4.1-4 Estimated Water Table Contours Fort St. Vrain 4-38 Nuclear Generating Station 4.1-5 Plant Water Supply Systems and Effluent 4-39 Drainage Paths 4.1-6 Major Tributaries and Irrigation Ditches, 4-40 South Platte River Between Henderson and Kersey, Colorado 4.2-1 Tritium Inventory in PCRV Cavity Water 4-41 4.5-1 Liquid Waste Discharges from Fort St. Vrain 4-42 5.1-1 MAP Centered on Fort St. Vrain Site 5-31 5.1-2 PCRV Work Area - Elevation View 5-32 5.1-3 Large Scale Regionalization for Tornado 5-33 Risk Analysis vi
Supplement to Environmental Report FSV Decommissioning Section 1 !
1.0 INTRODUCTION
1.1 Purpose ;
I The Fort St. Vrain (FSV) Nuclear Generating Station is a High Temperature Gas Cooled Reactor that was operated by Public Service Company of i Colorado (PSC) from 1977 through 1989. Because of technical problems I associated with failure of the control rod drives and degradation of the l steam generator ringheaders, FSV was permanently shutdown on August 18,1989. Subsequently, a Confirmatory Order was issued by the NRC l in May 1990 confirming the permanent shutdown of FSV. PSC plans to j defuel the core and to ship all of the spent fuel to the DOE facility in Idaho :
under the provisions cf an existing contract. However, due to the uncertain schedule for shipping the spent fuel to Idaho or other DOE facilities, PSC is pursuing an alternate plan to license, construct and operate an Independent Spent Fuel Storage Installation (ISFSI) in accordance with 10CFR72. Interim storage of fuel, top reflector elements, and neutron source elements can be provided at the ISFSI. The ISFSI is located on PSC owner-controlled property and will be within the historical FSV exclusion area boundary. The ISFSI Environmental Report describes the environmental effects associated with all aspects of the construction and operation of the proposed altemative storage facility (Reference 1).
In preparation for decommissioning, a Proposed Decommissioning Plan (Reference 2) has been developed and submitted to the NRC pursuant to 10CFRRO.82. PSC is proposing the immediate -dismantlement and decommissioning of the nuclear portion of the plant, after the removal of all irradiated fuel from the Reactor Building to a DOE facility or to the ISFSI. This Environmental Report Supplement addresses all actual or poten' 'ai environmental impacts associated with the proposed decommissioning activities and is responsive to 10CFR51.53(b). The level of detail in this Environmental Report Supplement is proportional to the significance of the associated impact.
Decommissioning of Fort St. Vrain is not expected to cause any significant impact to the general public or environment. Decommissioning of Fort St.
Vrain has overall positive environmental impacts. The result of decommissioning activities will be termiriation of the 10CFR50 license.
Commitment of resources, compared to plant operational resources, is small. The major environmental impact of decommissioning is the commitment of small amounts of land for waste burial at authorized disposal sites in exchange for reuse of the facility and site for other purposes.
1-1
Supplemont to Environmental Report FSV Decommissioning Section 1 The overall environmental impact of decommissioning activities will be small when compared to continued operation of the Fort St. Vrain plant.
The following environmental benefits will result from the completion of decommissioning:
- Reduced local traffic (fewer employees, contractors and materials shipments than required to support an operating nuclear power plant).
- Elimination of the radiological sources that create the potential for radiation exposure to site workers and the general public.
- Return of the site to unrestricted use.
1.2 Decommissioning Plan Description Plans to decommission the nuclear facilities at the Fort St. Vrain plant are as follows:
- Decontamination and dismantlement of the Prestressed Concrete Reactor Vessel (PCRV).
- Decontamination and dismantlement of the contaminated balance of plant systems.
- Site cleanup and final site radiation survey.
As discussed in Section 4.2 of the proposed Decommissioning Plan, PSC has committed to comply with Reg. Guide 1.86 (Reference 3), NUREG 0586 (Reference 4) and interim NRC guidance (Reference 5) when decontaminating the Fort St. Vrain site, to allow release of the site for unrestricted use and eventual termination of the 10CFR50 license. All decommissioning activities and schedules are based on decontamination to the above limits, it is currently planned to use the remainder of the Fort St. Vrain power plant as a natural gas-fired power plant. That action will be the subject of a separate environmental evaluation. There are no plans to again use the Fort St. Vrain site for nuclear power plant operations.
1-2
Supplement to Environmental Report FSV Decommissioning Section 1 1.3 Radioactive Waste Disposal Radioactive waste disposal will follow the requirements established in 10CFR20 and 10CFR61, the disposal site criteria, and other applicable i Federal and State regulations. The waste will be classified, packaged I according to approved procedures, manifested, and shipped to a disposal facility under an approved QA/OC program.
As radioactive waste is generated, it will be processed, packaged, and :
stored onsite only until a convenient or efficient transportation limit (i.e., J weight, volume, curie content) is reached, l The Classes A, B, and C waste will be packaged, stabilized if required, and shipped in accordance with relevant State and Federal regulations and burial site criteria. The Greater than Class C (GTCC) waste, if any, will be packaged for interim storage and retained onsite until such time as it can be transported to a DOE facility, such as Hanford, or as mandated by the amended license.
The steam generator primary assemblies are planned to be shipped in special shielded shipping containers by rail to a licensed disposal facility for burial. All other radioactive waste shipments will be by truck, l 1.4 Regulatory Considerations Decommissioning of nuclear facilities is a regulated process whereby equipment, structures and portions of the facility that contain radioactive material are removed and NRC licenses are terminated. The- voluntary termination of an operating license and the subsequent dismantlement of the facility requires NRC approval as specified in 10CFR50.82. Prior to the initiation of the dismantlement of radioactive components of the facility, a Decommissioning Plan must be submitted and approved.
- Pursuant to 10CFR50.82, PSC has prepared and submitted a Proposed l Decommissioning Plan (PDP) (Reference 2) to the NRC for review and l approval. The Decommissioning Plan describes an organized means for L removing all radioactive components and all radioactivity within the reactor facility. The PDP also describes how PSC will continue to protect the health and safety of the public and environment during dismantlement
, activities.
1-3
Supplement to Environmental Report FSV Decommissioning Section 1 The current requirements with regard to occupational or public doses or effluents to the environment continue to apply throughout the decommissioning period until the part 50 license is terminated by the Commission. The decommissioning planning requirements are considered appropriate means of assuring that radiation exposures to both plant personnel and the public will be maintained as-far below the 10CFR Part 20 limits as is reasonably achievable. The transportation of decommissioning wastes will not Involve any additional technical considerations beyond those for transportation of existing radioactive material. The shipment and disposal of all radioactive wastes will be governed by the existing NRC regulations in 10CFR parts 20, 61, 71 and 73 and appropriate Department of Transportation regulations in 49CFR.
A Quality Assurance program will be applied to decommissioning activities to ensure compliance with all applicable regulations and the approved Decommissioning Plan.
All activities involving asbestos will be conducted in accordance with Occupational Safety and Health Administration regulations 29CFR Parts 1910 and 1926 and Environmental Protection Agency regulations 40CFR Part 61, Subpart M.
1.5 EnvironmentalTechnical Specifications The requirements of the Radiological Effluent Technical Specifications (RETS) and the Radiological Environmental Monitoring Program (REMP) have historically been provided in the FSV Technical Specifications. For decommissioning activities, PSC submitted proposed Decommissioning Technical Specifications (Reference 6) that will totally supersede the historical FSV . Technical Specifications. Consistent with the guidance provided in Generic Letter 89-01 (Reference 7), the FSV Decommissioning Technical Specifications will contain the programmatic controls and reporting requirements for the RETS and REMP. Associated procedural details,- such as the methodology for calculating offsite doses _ due to radioactive gaseous and liquid effluents and monitoring instrumentation operability and surveillance requirements, will be included in the Offsite Dose Calculation Manual (ODCM) and in the Process Control Program (PCP).
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Supplement to Environmental Report FSV Decommissioning Section 1 1.6 Equipment Salvage Each piece of equipment or material removed during the decommissioning process will be assumed to be contaminated until determined otherwise.
Equipment found by radiation surveys to be suitable for unconditional release will be' free released (References 8 and 9). All material contaminated with radioactivity will be disposed of in accordance with NRC regulatory requirements.
1.7 Need for Proposed Action Decommissioning of the Fort St. Vrain Nuclear Generating Station is a regulatory requirement under 10CFR 50.82, which allows a licensee to l choose one of the three available decommissioning alternatives: DECON, !
SAFSTOR or ENTOMB. Of these alternatives, PSC has selected DECON '
(immediate dismantlement) as the basis for its Proposed Decommissioning Plan. Further discussion of the decommissioning alternatives and basis for selection of the DECON alternativo is provided in Sections 6 and 7 of this report.
1.8 References
- 1. Independent Spent Fuel Storage Installation (ISFSI) Environmental Report, Rev. O, Public Service of Colorado, June 1990.
- 2. Proposed Decommissioning Plan for the Fort St. Vrain Nuclear Generating Station, Public Service of Colorado, November,1990.
- 3. USAEC Regulatory Guide 1.86, " Termination of Operating License for Nuclear Reactors," June 1974.
- 4. Final Generic Environmentallmpact Statement on Decommissioning of Nuclear Facilities, NUREG 0586, August 1988.
- 5. Residual . Radioactive Contamination from Decommissioning.
NUREG/CR-5512 Draft Report, January 1990.
- 6. PSC letter, Crawford to - Weiss, dated December 21, 1990 (P-90367),
Subject:
Decommissioning Technical Specifications.
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Supplement to Environmental Report FSV Decommissioning Section 1 7.- " Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or the Process Control Program," NRC Generic Letter 89 01, January 21,1989.
l
- 8. NRC Circular 81-07, " Control of Radioactively Contaminated Materials."
- 9. NRC Information Notice 85 92, December 1985, " Surveys of Wastes Before Disposal From Nuclear Reactor Facilities." !
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Supplement to Environmental Report FSV Decommissioning Section 2 2.0 HISTORY AND CURRENT STATUS OF FACILITY 2.1 Nuclear Operating History Fort St. Vrain is a High Temperature Gas Cooled Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). Fort St.
Vrain's location is approximately 35 miles north of Denver and three and one half miles northwest of the center of the town of Platteville in Wold County, Colorado. The site consists of 2798 acres owned by the Licensee.
Construction of Fort St. Vrain was authorized by the Atomic Energy Commission (AEC) by issuance to PSC of a provisionalconstruction permit on September 17, 1968, in AEC Docket No. 50-267. Table 2.1-1 provides a brief summary of plant milestones and major events, in 1968, the Colorado Public Utilities Commission (CPUC) issued a -
certificate of public convenience and necessity to build Fort St. Vrain.
However, in its order, the CPUC stated that the authority to build a nuclear plant rather than a fossil fueled plant was " subject to the condition that the Commission may disallow portions of investment and operating expenses which are due to the fact that the plant is a nuclear poworod plant rather than a fossil fuel powered plant, if the allowance of such portions of Investment and operating expenses would adversely affect the ratepayer."
Fort St. Vrain was initially scheduled for commercial r ;aration in 1972.
Although PSC received a full power operating license in 1973, extensive pre-operational testing mandated by the NRC-and resulting engineering i modifications delayed the commercial operation of the plant until 1979.
General Atomics (GA), the prime contractor for Fort St. Vrain, reimbursed PSC for all increases in electric operatmg expenses incurred by PSC due to this delay.
On June 27,1979, PSC and GA settled all contracts and claims between
- them relating to Fort St. Vrain by entering into a Settlement Agreement and associated agreements. Pursuant to the GA Settlement Agreement, PSC accepted Fort St. Vrain for commercial operation at a reduced capacity of 200 Mwe (at 60% capacity factor) instead of 330 Mwe (at j 80% capacity factor), as originally designed. GA paid PSC $60 million as l
2-1 l
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Supplement to Environmental Report FSV Decommissioning Section 2 an adjustment to the cost of the plant to reflect the 130 Mwe reduction in capacity. In the period 1980 through 1984, GA paid PSC approximately $97.1 million to compensate PSC for the cost of replacing the 130 Mwe reduction in capacity with other generating facilities.
Fort St. Vrain was first included in PSC's rate base in a general rate case in December 1980, in that rato decision, the CPUC allowed PSC to collect approximately $39 million of revenues (an amount subsequentlyincreased to $46 million) to cover operating expenses and provide a return on investment.
As illustrated in Figure 2.11, Fort St. Vrain's record of operation after initial criticality has been inconsistent, with a historical capacity factor of less than 15% This inconsistent record of operation was the result of technical problems and implementation of regulatory requirements, including:
- core thermal and neutron flux oscillations
- moisture ingress problems a multiple control rod drive failures to automatically scram
- implementation of an environmental qualification program
- helium circulator material failures
- major turbine building fire damage e inadequate original design analyses (which limited maximum capacity to 82% power)
- recent technical problems related to control rod drive material failures and cracking of steam generator steam outlet piping in response to Fort St. Vrain's historically reduced levels of generation, the Colorado Public Utilities Commission (CPUC) instituted penalties against PSC to reduce the revenues recovered from its customers. The Office of Consumer Counsel (OCC) filed a complaint with the CPUC against PSC alleging that in light of its operating history, Fort St. Vrain was not "used and useful" in rendering a utility service. In view of the various legal and administrative proceeding regarding Fort St. Vrain, PSC entered into a Stipulation and Settlements Agreement in September 1986 with the 2-2
'T Supplement to Environmental Report FSV Decommissioning Section 2 CPUC, and OCC and other parties. Significant provisions of the 1986 Settlement Agreement included:(1) removalof Fort St. Vrain from the rate base; (2) a provision for the sale of future energy produced at Fort St.
Vrain to PSC customers at a rate of 4.8 cents per Kwh; and (3) recovery over 5 years of $11.5 million of decommissioning costs. This effectively made Fort St. Vrain an independent power producer, with the associated risks of operation assumed solely by the PSC shareholders.
As a result of its unfavorable plant operating performance, Fort St. Vrain did not produce revenues adequate to offset expenses during 1987 -
1989. Shortfalls of approximately $24.5 million (1987), $35.6 million (1988) and $30.1 million (1989) were recorded in unrecoverable operating i and capital expenditures.
The latest annual budget while operating (1989) was $77.9 million, which included all O&M costs, as well as capitalimprovement expenses required to meet regulatory requirements and NRC commitments. Assuming a maximum power limit of 82% (based on safe shutdown reanalyses), Fort St. Vrain would have to operate at a capacity factor of 68% to break even based on the 4.8 cents per Kwh allowed by the 1986 Settlement Agreement. In order to purchase new fuel (annual expense - $26 million) for continued future operations, a capacity factor of greater than 90% is l required. To cover eventual defueling and decommissioning expenses l (decommissioning present value - $137 million), the required capacity factor exceeds 100%. Reanalyses of the safe shutdown limits to allow full power operation (330 Mwe) would still require a capacity factor in excess of 85% in order to break even. Without third party support, Fort St. Vrain is unable to generate enough electricity to support annual expenses. Even if_ the known technical problems were resolved, the benefits of plant operation do not justify this use of resources.
2.2 Radlonuclide inventory 2.2.1 Activated Components Within the PCRV l
An activation analysis was completed in October 1990, for the PCRV and associated internal components. Calculational details of the activation analysis are included in Appendix ll of Reference 1 (the activation analysis l is also identified as Reference 2), and in Reference 3. The analysis was l performed to estimate the isotopic composition, magnitude and extent of L residual radioactivity which could be present in the PCRV after the end of l operations. The actual operating history of the plant was used in the analysis by considering the total effective full power days (EFPD) l l
2-3
Supplomont to Environmental Report FSV Decommissioning Section 2 generated by the plant untilits shutdown in August 1989. The analysis consisted of three sections: neutron flux estimates in the PCRV; activation analysis of the PCRV and internal components; and calculation of gamma dose rates (in air)inside the PCRV due to non-removable (fixed) components.
2.2.1.1 Computer Codes l
The activation analysis required the use of several computer codes and i various input data libraries. The ANISN code (Reference 4) was used to determine the neutron flux throughout the reactor core and outward I through the reflectors, helium flow paths, insulation, PCRV liner and PCRV l concrete. The activation of selected components within the PCRV was then determined using the REBATE computer code (Reference 5). Finally, gamma doses (in air) within the PCRV were calculated using the REBATE, ANISN and other data manipulation codes.
2.2.1.2 Material Compositions Material comp:s;tions of components were determined from a varie sources. In most cases, material compositions were identifieu wom component drawings which referenced standard material specifiutions.
Assumptions for the number densities of trace elements, such ses europium (Eu), cobalt (Co), and niobium (Nb), were based on design maauals, previous analytical investigations and recent regulatory guidance (Reference 6). Details of the actual compositiorr and , ace element assumptions are found in Reference 2.
2.2.1.3 Activation Analysis Results The results of the activation analysis are summarized in Table 2.2-1. The nuclides of importance - as well as the *,otal estimated radionuclide inventory for activated components inside the PCRV are listed in this table.
Detailed results can be found in References 2 and 3.
The dominant nuclides for metallic components are Fe-55, Co 60, Ni 63 and Mn-54. The dominant gamma emitter in the stainless steel components was determined to be Co 60, although Nb-94 is also present.
Due to the high concentrations of Co 00 in the boronated spacers blocks, these components are the primary dose contributors inside the PCRV.
2-4
Supplement to Environmor.al Report FSV Decommissinning Section 2 The activity in graphite components is dominated by tritium (H-3) and Fe-55, which were generated due to impurities in the graphite. Due to the large volume of graphite and the curie content of H 3 and Fe 55, these components are the largest contributors to the overall radionuclide inventory. No credit was taken for the migration of H 3 out of the graphite.
The Kaowool insulation and silica blocks were determined to have fairly low activities. The carbon steel cover plate contains almost all the activity in the Kaowool/ cover plate assemblies. The silica block activity is dominated by Fe 55.
The PCRV concrete /rebar matrix contains many activation products due to the presence of trace elements. In the short term, Co 60 is the dominant gamma emitter, while Eu 152 and Eu-154 are the dominant long term gamma emitters. The nuclide contributing most to the total activity is Fe-55. Other nuclides present in lower activities were: Cs-134, Co 45, Ag-110m, H 3, C-14, Fe 59, Ni-59, Ni-63, Nb 94, Mn 54, and Ca-41.
As indicated in Table 2.21, the majority of the activity in the concrete is contained in the first 1.5 feet in all directions. Table 2.2-2 indicates the estimated required amount of concrete which must be removed to achieve the recommended release limit for unrestricted use (5 microR/hr above background), Table 2.2-2 lists dose rate estimates for each direction; therefore, the total dose inside the center of the PCRV (in air) is the sum of the dose rate for all three directions. Table 2.2-2 also indicates the estimated dose rate contribution (in air) for various stages of component removal at a location in the center of the PCRV.
2.2.2 Plateout Analysis for PCRV Internal Components 2,2,2.1 Basis of Computer Code Analysis A plateout distribution analysis of radioactive nuclides produced in the reactor -core was performed for the PCRV and internal components (Reference 7). The purpose of this analysis was to estimate the plateout concentrations and distributions in the primary coolant circuit. Analyses were conservatively performed from the Beginning Of Cycle (BOC) 1 to the End of Cycle (EOC) 5. The axial and radial core power distributions through fuel cycle 5 were calculated and used with flux distribution data
- as input to fission product release codes. Full-core fuel and graphite temperature distributions, fuel failure and release of key fission gases and metals were then calculated. Based on the full core analysis for key 2-5
Supplement to Environmental Report FSV Decommissioning Section 2 fission gases and metals, the total plateout and helium purification system inventories of radioactive nuclides were estimated.
Platcout distributions were calculated uslng the PADLOC computer code (Reference 8). _The PADLOC code performs a mass transfer calculation using mass transfer correlations and sorption isotherms to determine the partitioning of condensable radionuclides between the flowing coolant and the fixed surface in a recirculation loop. The plateout model in PADLOC is limited to one dimensional cylindrical geometry, such that all components of the primary circuit must be modeled as an equivalent series 4 of coupled sections, parallel bank cylindrical tubes. Reference sorption isotherms were used to describe the sorptive capacity of the primary circuit materials for the radionuclides of concern. )
2.2.2.2 Platcout Methodology Typically, the two dominant sources of fission products released from the core are heavy metal contamination (heavy metal outside the coated fuel particles) and fuel particles whose coatings failin service, in addition, the volatile metals (Cs and Sr) can, at sufficiently high temperatures and over long periods of time, diffuse through the silicon carbide (sic) coatinOs and be released from the intact fuel particles.
Calculations were performed for the following key nuclides: Sr 90,1-129, 1-131, Cs 137, Cs-134 and Te-127m. The source terms for fission product plateout analysis include both a direct release contribution and, where applicable, a procursor contribution, in the case of the cesium isotopes, there is a direct release of both Cs-137 and Cs-134 metal from the core. Cs-137 plateout also results from the release and subsequent decay of its precursor contributor, Xe-137. Cs-134 has no gaseous precursot _.milarly for Sr-90, there is a direct Sr-90 metal release as well as the contribution from its Kr-90 precursor. Only direct release
!- contributions are considered for 1-129,1-131 and Te-127m.
r 2.2.2.3 Analysis of Results It is anticipated that any internal PCRV component that had come in contact with primary coolant is contaminated and will be removed for disposal as radioactive waste. This includes not only the core graphite
. and structural components (which are also activated), but also the steam
- generator modules, helium circulators and Kaowool insulation. The preliminary results of the plateout analysis are shown in Tables 2.2-3 and
- 2.2-4. Table 2.2 3 lists the plateout concentration (Ci/cm 2) on primary 2-6
Supplement to Environmental Report FSV Decommissioning Section 2 - l l
l circuit components for the key nuclides, Cs 137 and Sr-90. Table 2.2-4 identifies the integrated plateout (Cl) of primary circuit components for the following nuclidos: Cs 134, Cs-137, b .31,1-129, Sr-90 and Te-127m.
Additional information on the analysis results, analytical models and comparisons with measured data is provided in Reference 7.
The accuracy of the predicted fuel performance and gas release will be assessed by comparison to measured R/B (Release to Birth Rate) data.
The accuracy of the predicted fission metal release data will be assessed by comparison to measured plateout probe data. Plateout distributions and concentrations will then be calculated for the primary circuit.
2.2.3 Contaminated Systems, Structures and Components An engineering analysis of the total curie inventory at Fort St. Vrain was completed -in June 1989 and the results of this analysis have been summarized in Table 2.2-5. This analysis is based upon past survey results, activation analysis, plateout analysis and general estimation of contamination levels occurring in the various systems. The survey results and estimation of contamination levels were then applied over the estimated surface area of the r ssociated system. This analysis accounts for all expected radioactivity at Fort St. Vrain with the exception of fuel.
Section 2.3 contains a detailed summary of the radiation survey results.
These surveys were performed to identify general radiation and contamination levels in frequently accessed areas of the facility. More detailed surveys of individual areas will be required when determining specific work plans during actual decommissioning.
2.2.4 Initial Site Characterization Plans The initial site radiological characterization will be performed to determine the radiological status of Fort St. Vrain balance of plant systems, auxiliary systems, buildings and site. Radiological measurements for direct radiation, residual contamination (fixed and removable) will be conducted and recorded. Biased and unbiased samples (i.e., sample locations are or are not influenced by previous data or historicalinformation) will be taken, analyzed and recorded. Results from current Fort St. Vrain radiological data will be used as part of the initial characterization, where appropriate.
2-7
Supplernent to Environmental Report FSV Decommissioning _
Section 2 The results of the initial site radiological characterization will assist in '
accurately planning the decommissioning activities; determining of the level of effort for decontamination of systems, structures, etc; and determining of final survey plans, extent of surveys and instrumentation to be used, it will also be_ used as a general performance indicator to assess the effectiveness of the overall site decontamination. The data will be utilized for radioactive waste management, assessing potential hazards during the decontamination and decommissioning work and for determining safety controls.
l 2.3 Radiation Survey Results 1 In August -1990, baseline radiation and contamination surveys were performed in the Reactor and Turbire Buildings. These surveys focused on identifying the major contributors to ied!ation levels above background and areas containing both fixed and loose surface contamination.
Historical radiological surveys have shown that alpha contamination (both fixed and loose surface) is not present above natural background levels at Fort St. Vrain. Surveys for elpha contamination are performed on a routine basis to confirm this.
Generally, the results of these surveys demonstrated that greater than 95% of the' plant areas have radiation levels corresponding to natural background (in the 0.004 to 0.032 mrem /hr range, as determined from historical surveys). In the results summarized in Table 2.3-1, only those areas with radiation levels above background are noted.
Additionally, fixed contamination levels are generally less than 1000 dpm/15 cm 2and loose surface contamination levels less than 1000 dpm/_100 cm2 . Most loose survey results are less than 100 dpm/100 cm 2.
In some locations, tritium may be present as fixed contamination. Due to the low energy beta activity emitted by tritium (Eavg = 0.005 MeV),
normal survey methods will not detect the tritium and therefore actua; tritium levels were not considered in this survey.
Figures 2.3-1. to 2.3-19 provide specific results of these area radiation surveys. Table 2.3-1 provides a summary of the survey results with a description of the major contributors to the radiation levels. Reactor and Turbine Building elevations are shown in Figure 2.3-20. Where results are not listed, contamination and/or radiation levels are not greater than background levels. Systems 2-8
Supplement to Environmental Report FSV Decommissioning Section 2 which are potentially contaminated are identified in Table 2.3-1 by system number for each elevation on which they are located.
2.3.1 Turbine Building Survey Results General area radiation levels throughout the Turbine Building are primarily due to natural background. Contamination levels (both fixed and loose) are less than 1000 dpm/100 cm2 in alllocations and generally less than 100 dpm/100 cm 2. Piping from the potentially internally contaminated Systems 11 (PCRV and Internal Components) and 73 (Reactor Building Ventilation) extends from Level 7 (El. 4829') to the roof of the Reactor Building.
2.3.2 Radiat!an Cources Outside the Reactor and Turbine Buildings Radioactive materials are stored on a temporary basis inside Sea-Vans, cargo trailers and designated radwaste areas. Varying amounts of radioactive materials may be stored in these trailers, but external radiation levels are typically less than 0.2 mrem /hr.
The only contaminated area outside the Reactor and Turbine Buildings is the Compactor Building directly east of the main cooling tower (see Figure 2.3-1). General area radiation levels vary from 0.2 to 0.5 mrom/hr primarily due to residual contamination inside a radioactive waste compactor. Loose surf ace contamination levels are generally less than 100 dpm/100 cm 2. The compactor contains loose surface contamination of 50,000 dpm/100 cm 2and fixed contamination levels of 50,000 dpm/15 cm2. There are two concrete bunkers in the Compactor Building which have loose surface contamination levels of 5,000 dpm/100 cm2 and fixed contamination levels in the first few centimeters of the concrete averaging approximately 20,000 dpm/100 cm2 . The presence of tritium is also suspected in the fixed contamination of the bunkers. This building is I so used for staging of radioactive wastes (including liquids) and materia' .
Piping associated with the Radioactive Liquid Waste System (System 6.'s) also runs underground from the exit point from the Reactor Building to the main cooling tower blowdown line. Sample results of oil collected in an associated oil separator have occasionally shown trace amounts of tritium, Co-60, Cs-137 and Cs-134.
2-9
Supplement to Environmental Report FSV Decommissioning Section 2 Routine surveys do not indicate any radiation or contamination levels above background in the Radiochemistry- Laboratory located in the Technical Support Building (See Figure 2.3-2), although small amounts of radioactivity are expected in drain pipin0 from this facility to the Radioactive Liquid Waste System (System 62).
2.3.3 Current Environmental Radiological Status 2.3.3.1 Beta-Gamma Radiation in Surrounding Environs The environmental radiologicalstatus of the site and surrounding areas has been monitored during the entire pre-operational, operational, and post-operational phases of the plant through the Radiological Environmental Monitoring Program (REMP). This program includes surveillances in surrounding areas to gather environmental data in the following areas:
external gamma activity levels, air sampling data, water sampling data, milk data, aquatic pathways, and food products. Sample locations are situated near the site boundaries and in outlying areas. Details of the results of these surveillances can be found in Reference 9 and in past REMP reports, which are provided annually to the NRC.
During the spring and summer of 1990, additional data were taken to further characterize the site. Soil samples were taken inside and outside the protected area, gamma radiation surveys were performed inside the protected area, and downwind air samples were taken with respect to the predominant wind direction (from the NE). Environmental radiation surveillance data from all past REMP reports and the recant -
characterization data indicate that the predominant source terms found above natural background levels are due to Chernobyl and past nuclear weapons test fallout. External radiation sources to area residents are due to naturally-occurring background radiation and atmospheric fallout.
The recent characterization data included the exposure rate from gamma-ray emitting radionuclides and were measured using thermoluminescent dosimeters (TLD). The TLD stations were constructed at 72 different locations inside and outside the controlled area boundary. Each station contained packets with two chips of CaF,(Dy), which are identical to those used in the REMP. The measurement period for the TLDs was 92 days. The mean of the two chips in each station was used to determine the mean exposure rate. The overall mean exposure rate from the TLD packages was 0.32 mrem / day. This value is not statistically different from the mean value found in the 1989 REMP report (Reference 9) of 0.38 mrem / day for the Fort St. Vrain facility area. Reference 9 indicated that 2-10
Supplomont to Environmental Report FSV Decommissioning Section 2 sinco the inception of power production by the reactor, there has boon no detectable increase in the external exposure rato duo to planned or ;
u.iplanned reactor releases.
The concer trations of Oross bota activity duo to the combination of naturally occurring radionuclidos and fission product radionucildos woro determined from air samplos at two locations downwind from tho ,
predominant wind direction. A particulatn filte r ic.' 0'r.Js beta analysis and ;
an activated charcoal cartrid0'a for 1 131 or ncble Das radior.uclido analyses woro in the samplo line. Tritium in atmospheric water vapor was collected passively by silica gol at cach of those locations. Samplir.g methodolo0y was identical to that utilized in the REMP. Fort St. Vrain operational Technical Specifications no lonDor requito measuromont of gross alpha activ!ty. Gross bota activity measured in air particulatos was principally i duo to naturally occurrinD radionuclidos or from soll resuspension. The mean wookly activity air concentrations measured at the northorn and southern monitors woro 16 femto Curios (1.0E 15)/r.0. Those concentrations are comparable to those found in the REMP pro 0 ram. Past REMP data has shown that thoto has never boon a significant difference observed betwoon facility and roierenco sitos (Referenco 9). It is concluded, therefoto, that based on the current radiolo0l caldata and past REMP data, the reactor air offluents of particulato fission products or activation products are not a soutco of doso commitment for the Fort St.
Vrain environs population.
2.3.3.2 Soll Samplos Soll samplos were taken at 124 locations insido and outside the controlled area. Samplos were taken at each location from a depth of ton contimaters and an area of 95 cm'. Two samplos were taken at each location to produce a sample sizo sufficient to fill a one quart volume.
Samples woro dried, ground to a constant density, and scaled in the quart container. After a throo wook parlod, each container was counted using Go(Li) gamma ray spectroscopy to datormino the activity concentration of important fission products, activation products and naturally occurrin0 radionuclidos. No bota analysis wns performed.
l 2-11 v,-- , - - - ,'----,.,y,,...-+-.w.-wi,-1,-.-,r.--,.-.r- , . . . _ ,...-....--..-.___-,4...,.---.c r,---.,-v..-w-- - .,__mr, - - ---
- _ _ ~- _ -_ - _ - . _ - - _ - - - - - - -.. - - , _ . .- _ _. - ..
l Supplomant to Environmental Roport FSV Decommissioning Section 2 Deep core sampics (taken at approximately 12 percent of the soil samplo locations) were taken to appro'timately ISO contimeters In depth. The coro samples woro collected in polyethylono tubes, which were frozen and sectioned to obtain samples at various depths. Tto doop coro samplos woro analyzed using the same techniques as the sotl samples.
Results of the soll samplos indicate the presence of statistically significant Cs 137 concentrations. These concentrations are due to world wido fallout remaininD from the United States, USSR and Chinci,e nuclear l weapons tohts and the Chernobylaccident. U $ supported by the fact that the Cs 137 concentrations are the som ' 3 ontire front range of )
Colorado, and other reactor generated fission products or act'vation products were not present in the samples.
2.3.3.3 Results of REMP Surveillances Tritium is the only radionucilde that was detected in concentrations above background in any offluent pathways that could be attributed to reactor operation. Since tritium is released as tritlated water, the dliution 'u/ the surroundin0 hydrosphere is significant. Elevated levels of tritlated water (Reference 2) were detected in downstream surface water samples on occasion, but the yearly mean values of downstream surface water was not statistically greater than upstream concentrations. Tritium concentrations measured in milk were all less than the lower limit of detection (LLD). However, slight increases in the downstream tritium levels, which were discussed in the 1986 Annual Radiolog! cal Environmental Monitoring Pro 0 ram (REMP) report, showed that the radiation dose commitment that can be calculated as a result of the increases was found to be negligible as compared to natural background radiation doss rates.
The REMP pro 0 ram over the years has been shown to be of adequato scopo and sensitivity to detect any accidental releases from Fort St. Vrain operation it is concluded that the doso commitments calculated for the
- closest inhabitants or other parts of the nearby ecosystems due to reactor operations are negligible. In addition to the REMP data, the most recent characterization data both inside and outside the controlled area boundary supports this conclusion.
I 1
l 2-12
Supplomont to Environmantal R:: port FSV Decommissioning Section 2 2.4 Current Facility Status The Fort St. Vrain physical facility is described in Section 2.1 of the Proposed Decommissioning Plan (Reference 1), submitted to the NRC on November 5,1990. The reactor core is in the process of being defueled.
Presently, one third of the core has been defueled to the adjacent fuel storage wells (FSWs). Further defueling is on hold pending Department of Energy (DOE) permission to begin shipment of spcat fuel to its DOE ldaho or other DOE facilities.
As noted above, the Proposed Decommissioning Plan was s. ,mitted to the NRC for approvalon November 5,1990. PSC has also submitted and received NRC approvalto reduce fire protection, emergency preparedness and physical security programs consistent with plant requirements during defueling. Other non nuclear facilities on the FSV site are maintained and operated by PSC as needed to support PSC electric power demands.
The reactor building is the only nucleai f acility to be decommissioned. An access control plan, included as part of the Proposed Decommissioning Plan, will provide suitable controls to prevent unauthorized or uncontrolled access to either the decommissioning site, or to radiologically controlled areas within the decommissioning site.
Building services, including electric power, cooling and domestic water, HVAC, lighting and other services are available and will continue to be provided throughout the decommissioning effort. These functiona!
services and systems are identified in Section 2.2 of the Proposed Decommissioning Plan, 2.5 References
- 1. Proposed Decommissioning Plan for the Fort St. Vrain Nuclear Generating Station, Public Service Company of Colorado, November 1990,
- 2. Engineering Evaluation, " Fort St. Vrain Activation Analysis,"
EE.DEC-0010, Rev. B, October 1990.
- 3. PSC Letter, Crawford to Weiss dated April 26, 1991, (P-91118); Sub}ect: PSC Response to NRC Request for Additional Information on the Fort St. Vrain Proposed Decommissioning Plan, 2 13
Supplomont to Environmental Report FSV Decommissioning Section 2
- 4. Engle, W.W., " ANISN P Multigroup One Dimensional Discroto Ordinatos Transport Code with Arisotropic Scattoring,"
Radiation Shiciding information Contor, Oak Ridge National Laboratory, Oak Ridge, Tennessoo.
S. Ebasco Servicesincorporated," REBATE A Computer Program for Calculation of Decay Gamma Source Strength For One or Two Dimensional Gamma Transport Analysis," Ebasco Services Incorporated, New York, New York.
- 6. "Long Lived Activation Products in Reactor Materials,"
NUREG/CR 3474, August 1984.
- 7. GA International Services Corporation, Report 909658, " Fort St.Vrain Platoout Analysis for Decommissioning Study," Issuo A, February 27,1989 (GP 3282).
- 8. Hudritsch, W.W., "PADLOC, A One-Dimensional Computer Program for Calculating Coolant and Platoout Fission Product Concentrations," Gontral Atomic Report G A- A14401, September 1981.
- 9. PSC lotter, Crawford to NRC, dated April 23,1990, (P 90139);
Subloct: " Fort St. Vrain Nuclear Generating Station Radiological Environmontal Monitoring Program Annual Summary Report for 1989".
2-14
%9 Supplement to Environmental Report FSV Decommissioning Section 2 TABLE 2.1 1 FORT ST VRAIN MILESTONES AND MAJOR EVENTS December 1973 Plant construction completed.
December 21 1973 Facility Operating License No. DPR 34 issued to PSC.
December 20 1973 initial fuel loading.
January 31 1974 Initial nuclear criticality.
1974 1979 Startup testing, low power operation, and requited plant modifications.
July 1979 FSV committed for commercial operation.
May 1986 NRC mandated Environrnental Qualification (EO) outage.
September 1980 Stipulatioa and Settlement Agreement removing FSV from the rate bate.
October 1986 Steam generator reanalyses performed which reduced maximum attainable power level to 82% of rated power (270 Mwe).
May 1987 Plant restart following EQ outage.
July 1987 Shutdown following helium circulator failure.
October 1987 Hydraulic fire during 9 tant restart.
December 5 1988 PSC informs NRC that FSV will be permanently shutdown not later than June 30,1990.
June 30 1989 PSC submitted the FSV Preliminary Decommissioning Plan.
July 1989 Plant record for Kw generated for one month period.
August 29 1989 PSC Doard of Directors announce decision to terminate operations at FSV effective that date.
July 2 1990 The Westinghouse Taam selected as the FSV cocommissioning contractor.
November 5 1990 PSC submitted the FSV Proposed Decommissioning Plan.
December 21 1990 PSC submitted the Decommissioning Technical Specifications.
2-15
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Supplem:nt to Environment:1 R: port FSV Decommissioning Section 2 TABLE 2.2 2 PCRV DOSE RATES IN AIR AT 5 YEARS AFTER SilVTDOWN' Gamma Dose Rate Badial Pem/Hr All components (from large side 9.7E (01) reflector to PCRV concrete)
Large side reflectors removed 2.3E (02)
(from spacers to PCRV concrete)
From core barrel to PCRV concrete 2,1E ( 02)
PCRV liner and concrete only 8.8E ( 03)
PCRV concrete only 4.5E ( 03) 22' PCRV concrete removed 0.3E ( 06) 24' PCRV concrete removed 3.4E ( 06)
Axial.Un All components (from Knowool 1.7E ( 01) insulation to PCRW concrete)
PCRV liner and concrete only 4.4E ( 01)
PCRV concrete only 1.7E ( 01) 32" PCRV concrete removed 7.6E ( 06) 34* PCRV concrete removed 4.4E ( 06) 36" PCRV concrete removed 2.6E ( 06)
/Lxial Down All Components (from core support blocks 6.1 E ( 02) to core support floor)
PCRV liner and concrete only 2.5E ( 01)
PCRV concrete only 1.8E ( 02) 20" PCRV concrete removed 5.3E ( 06) 22" PCRV concrete removed 2.7E ( 06)
' The calculated dose rates for worker access inside the PCRV after all internal components have been removed is tentatively scheduled to occur approximately 5 years after reactor shutdown.
2-18 l
,,J
Supplement to Environmsntal Riport FSV Decommissioning Section 2 TABLE 2.2 3 ESTIMATED PLATEOUT CONCENTRATION OF MAJOR PRIMARY CIRCulT COMPONENTS AT EOC5' Plateout Concentration Sr90 CS 137 dpm/ dpm/
Comoonent 1Q0 cm' C]!cm' 1Q0 cm C#1m' Lower Reflectors 1.11 E(07) 5.0E( 08) 2.22E(08) 1.0E( 06)
Steam Generator "* 2.22E(07) 1.0E( 07) 3.33E(08) 1.5E( 06)
(Reheater Section)
Circulator 4.44E(07) 2.0E( 08) 5.55E(07) 2.5E(-07)
Circulator Outlet 3.33E(05) 1.5E( 09) 9.90E(05) 4,5E( 09)
Core Barrel Annulus 2.22E(06) 1.0E( 08) 7.77E(00) 3.5E( 08)
Upper Reflectors 2.00E(07) 9.0E( 08) 4.44E(07) 2.0E( 07)
EOC5 End of Fuel Cycle #5
" Highest estimated concentration on the component Steam generator component with the highest estimated plateout concentration.
2 19
TABLE 2.2-4
- . INTEGRATIO PLATEOUT IN EACH PRIMARY CIRCUIT COMPONENT AT EOCS*
BRANCH Cs-134" CS-137" I-131 "
- I-129' " Sr-90" Te-127m*" I NAME (Curies) ICuries) (Curies) (Curies) (Curies) (Curies)
Active Core O.OOOE( + 00) 0.OOOE( + OO) 0.OOOE( + OO) 0.OOOEt + 00) 0.OOOEt + OO) 0.OOOEt + 00) ;
Low 7r Reflector 1.510E( + 01) 1.326Et + 01) 6.077E(-03) 1.037E(-08) 5.820E(-01) 3.196Et + 01) 1 Core Support Blocks 3.668E(-01) 3.224E(-01) 2.923E(-04) 4.987E!-10) 1.028E(-02) 5.664E(-01) !
Core Exit Plenum 8.166E(-02) 7.195E(-02) 2.028E(-04) 3.461 E(-10) 1.937E(-03) 1.OOOE(-010 Steam Generator inlet 1.617E(-03) 1.443E8-03) 9.125E(-03) 1.557E(-08) 1.770E(-02) 8.049E(-01) l Steam Generator Reheater 1.029E( + 01) 3.771 E(-01) 6.272E(-01) 1.066E(-06) 4.527E(41) 1.923E( + 01)
- Superheater 2.699Et + 00) 2.305El + 00) 1.478Et + 00) 2.516E(-06) 1.544E(-01) 6.909E( + 00)
Economizer 7.513Et + 001 6.339Et + 00) 2.788Et + 01) 4.341 E(-05) 2.244E(-02) 9.067E(-01)
Evaporator 7.398E( + 00) 6.879Et + 001 1.120' - < :.868E(-04) 8.131 E(-03) 1.G96E(-01)
Steam Gen. Outlet Plenum 2.277E(-02) 2.272E(-02) 1.164d .. 2.531 E(-05) 6.615E(-04) 4.924E(-04)
Circulators 1.458E(-01) 1.388E(-01) 4.432E( + 00) 7.501E(-06) 8.754E(-03) 3305E(-03)
Circulator Outlet Plenum 5385E(-03) S.882E(-03) 6.929E( + 00) 2.507E(-05) 1.274E(-03) 1.161E(-04)
Core Barre!!Lirer Annutus 1.196E(-01) 1.886E(-01) 4.172Et + 01) 8.176E(-05) 5.435E(-02) 2.579E(-03)
Core intet Plenum 1.41 OEl-02) 3302E(-02) 1.943Et + 01) 7.128E(-05) 1315E(-02) 3.041E(44)
Upper Reflectors 8.109E(-01) 2.448Et + 00) 2.702E(-01) 6322E(-07) 1.069E( + 00) 1.752E(-02)
Side Reflectors 1.053E(-03) 5.767E(-03) 1.761 E(-01) 5.441 El-07) 1.838E(-03) 2.210E(-05)
Purification System 1.106E(-03) 2.902E(-03) 2.853E( + 00) 3.542E(-04) 1.293E(-03) 2.404E(45)
TOTAL (EOCS) 3.530Et + 01) 3.290Et + 01) 1.237Ei + 03) 1.100E(-03) 2.400E( + 00) 6.OG7Et + 01)
TOTAL (3-YEAR DECAY) 1.290E( + 01) 3.071 Et + 01) -
1.100E(-03) 2.234Et + 00) 5.739E(-02)
EOCS - END OF FUEL CYCLE 5
- Based upon the source rate calculated from the xenon data using the square root of half-life dependence.
Plateout distribution based upon sorption isotherms for unoxidized aDoy steel surfaces.
2-20
- - - - . - - -- - _ __a
Supplement to Environmental Report FSV Decommissioning Section 2 TABLE 2.2 5 ESTIMAYED CURIE TOTAL AT FSV (Three Years After Shutdown)
NOTE: The systems listed below are those systems which are known to be contaminated. On-going maintenance, def ueling and component removal may transfer contamination to other systems and/or locations.
Total Curles System From from Loose No. System Activation Contamination' 11 PCRV and internal Componen.1 7.94 E( + 05) 2.54 E( + 02) 12 Controls Rods a .d Drives 1.84 E( + 04) N/A 13 Fuel Handling Equipment N/A 8.95 E( 03) 14 Fuel Storage Facility N/A 2.08 E( 02) 10 Auxiliary Equipment .
N/A 9.05 E( 03) 17 Reactor Removable Reflector 4.82 E( + 05) N/A 21 Primary Coolant N/A 6.01 E(4 01) 22 Secondary Coolant N/A 5.00 Et 4 03) 23 Helium Purification N/A 0.33 E( 01) 01 Decontamination Systems N/A. 1.00 E(-05) 02 Radioactive Liquid Waste N/A 4.00 E( 05) 63 Radioactive Gas Waste N/A 8.15 E( 05)
- Includes an estimate of loose surf ace contamination due to activated corrosion products.
N/A Not Applicable.
2 21
_ _ _ _ = _ - _ _ _ _ .
TABLE 2.3-1 RADIOLOGICAL SURVEY
SUMMARY
Reactor Internally Source of Contamination Levels
- Building Radioactive Radiation / Radiation Loose Fixed Level Elevation Systems Contamination levels (mr/hr) (OPM/100cm2) (OPM/15 cm2)
Level 13 4916*-8* 46,47 4921*-O" l
Level 12 4904*-O* 46,47 4906'-8" l Level 11 4881*-O* 11,12,13, Shine through Gen. Area - O.044 l' 14,15,16, FSW's and ESW's 21,23,46, 47,72,93 Level 10 4864*-O* 11,13,14 New Fuel Loading Gen. Area - 0.8 30,000 5,000-
- 16,21,23, Port Contact - 6.0 10,000 1 46,47,72 t
93 Hot Service (See results from level 9 below) j Facility Purge Vacuum Gen. Area - O.032 Pumps i
- Level 9 4849*-O* 11,14,16, Regeneration Gen. Area 0.15 21,23,46, System 47,61,62, 63,72,93 Hot Service Gen. Area - 0.5 100,000 10.000 -
, Facility 50,000 Level 8 4839'-O" 11,14,16, Access to Hot Gen. Area - 0.5 50,000 4846*-O* 23,46,47 Service Facility 61,62,63, 72,93 Hot Service Gen. Area - 50 100,000 10,000 -
Facility sump Contact - 200 50,000 j level 7 4829'-O* 11,14,16, Irradiated Gen. Area - 2.8 1 46,61,62. Thermocouples Contact - 4.0
, 63,72,93 5
Level 6 4811*-O" 11,46,61, Gas / Liquid Waste Gen. Area - 0.02 4816*-O* 62,63,72, System Piping Contact - 1.2 93 Turbine Building 2-22
TABLE 2_3-1 RADIOLOGICAL SURVEY
SUMMARY
(Continued)
Reactor Internally Source of Contamination Levels Bo,7 ding Radioactive Rat *iatior ! Radiation Loose Fixed Level Elevation Systems Contami,ation Levels (mr/hrt (DPM/100cm2) (DPM/15 cm'}
f1occrn-i j Level 5 4791'-O" 11,46,61, Gas / Liquid V/aste Contact - 0.25 5,000 -
62,63,72, System Piping 10,000 93 Compactor B3dg.
Level 4 4781*-O" 46,47,61, Decontamination Gen. Area - 0.5 600 500 -
62.63,72 System Contact - 3.0 1,000 93 Level 3 4771*O" 46,47,61 Decontamination Gen. Area - 0.4 1.400 100 -
62, 63, 72., Laundry Contact 2.2 500 93 Floor of Vault 500 -
Containing T-6101 1,000 Level 2 4756*-O- 46,47,61 62,63,72, 93 Level 1 4740'-6* 21,46,47, Gas Waste Compressor 100 -
61,62,63, Drains (3) 500 72,93 Liquid Waste Sump 5,600 30,000 Level 1 Below Floor 72 Reactor Building Sump 100 -
Level 500 2-23
- - - - . -__ - _ _ _ _ _ - _ _ _ _ _ _ ______ __. . .i
Supplement to Environmental Report FSV DecommissioninD bross ilectric 160,000 H g 140,000 120,000--
p 100,000- ,
80,000- y 60,000- p 40,000-
'"'"[ i . . . [I .
. . . I 19 77 1978 19 79 1980 19fil 1982 1963 198 4 1365 19ht> 1987 19b8 1989 Fort St. V n Po ver Generation 2-24
_ _ _ ___ __- _ _ - - _ _ _ . l
Supplomont to Environrnental Report FSV Decommissioning Section 2
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Suppl:mont to Environmental Roport FSV Decommissioning Sovjon 2
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Supplement to Environmentcl R: port FSV DecommissioninD Section 2 l
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Figure 2.3-2 Radiochemistry Laboratory Radiation Survey 2-27
Suppicmont to Environrnental R: port FSV DecommissioninD Section 2 6 , ammus Oe Q J j l c
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( ) denotes radiation level at 18 inches o denotes contamination survey point, < 100 dpm/100 cm' t>ose surface level unless otherwise noted; # denotes fixed contaminat.on level in dpm/15 cm' Figure 2.3 3 Turbine Building Radiation Survey -
Level 5 (El,4791')
l 2-28
Supplement to Environment:1 R: port FSV Decommissioning Section 2 1
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( ) denotes radiation level at 18 inches o denotes contamination survey point, < 100 dpm/100 cm' looso surface level i unless otherwise noted; # denotes fixed contamination levelin dpm/15 cm' I
l l Figure 2.3-4 l Turbine Building Radiation Survey -
Level 6 (El. 4811')
1 2-29 l
l
Supplement to Environmental Report FSV Decommissioning Section 2
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Figure 2,3 5 l Turbine Building Radiation Survey -
Level 7 (El. 4829')
2-30
Supplement to Environmental Report FSV Decommissioning Section 2 e.ae.
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( ) denotes radiation level at 18 inches o denotes contamination survey point, < 100 dpm/100 cm' loose surf ace level unless otherwise noted; # denotes fixed contamination level in dpm/15 cm' figure 2.3.6 Turbine Building Radiation Survey -
Levels 8,10 & 11 (El. 4846',4864',4884')
2-31
Supplomont to Environmental Report FSV Decornmissioning Section 2 l
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( ) denotes radiation level at 18 inches O denotes contamination survey point, < 100 dpm/100 cm' loose surface level unless otherwise noted; # denotes fixed contamination level in dpm/15 cm' Figure 2.3 7 Turbine Building Radiation Survey -
Levels 12 & 13 (el 4904',4921')
2-32
Supplomont to Environmental Report FSV Decommissionin0 Section 2 01 " "" b f
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denotes contact radiation reading
() denotes radiation level at 18 inches o denotes contaminatica survey point, < 100 dom /100 cm' loose surface level unless otherwise noted; # denotes fixed contamination level in dpm/15 cm' Figure 2.3 0 Reactor Building Radiation Survey -
i.evel 1 (El, 4740')
2 33
Supplomant to Environmentcl R3 port FSV Decommissioning Section 2 l
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Level 2 (El, 4756')
b 2 34
- Supplement to Environmental Report l FSV Decommissioning Section 2
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( ) denotes radiation level at 18 inches o denotes contamination survey point, < 100 dpm/100 cm' loose surf ace level unless otherwise noted; # denotes fixed contamination level in dpm/15 cm' Figure 2.310 Reactor Building Radiation Survey -
Level 3 (El. 4771')
2-35
Supplement to Environmental Report FSV Decommissioning Section 2
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Level 4 (El. 4781')
2 36
Supplomont to Environmental Report FSV Docommissioning
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( ) den'>tes radiation level at 18 inches O denotes contamination survey point, < 100 dpm/100 cm' loose surface level unless otherwise noted; # denotes fixed contamination level in dpm/15 .m' Figure 2,312 l Reactor Building Radiation Survey -
Level 5 (El 4791')
l 2 37 ,
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~ ( ) denotes radiation level at 18 inches O denotes ~ contamination survey point,- < 100 dom /100 cm' loose surface level .
unless otherwise noted; # denotes fixed contamination levelin dpm/15 cm
l- Figure 2,313 Reactor Building Rcdiation Survey -
Level 6 (El, 4016') +
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2 38 l-
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( ) denotes radiation level at 18 inches q O - denotes contamination survey point, < 100 dpm/100 cm' loose surface level '
unless otherwise noted; # denotes fixed contamination level in dpm/15 cm' Figure 2.314 Reactor Building Radiation Survey -
Level 7 (El. 4829')
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2-40
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2-41
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2-42
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Level 11 (El. 4881') Refueling Floor 2 43
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Levels 12 & 13 (El. 4904' & 4921')
2-44
Supplement to Environmental Report FSV Decommissioning Section 2
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Supplement to Environmental Report FSV Decommissioning Section 3 3.0 PROPOSED ACTION 3.1 Introduction Pursuant to 10CFR50.82, PSC has submitted a Proposed Decommissioning Plan (PDP) (Reference 1) to the NRC for review and '
approval. PSC has selected the DECON option for decommissioning Fort St. Vrain and is proposing the immediate dismantlement and decontamination of the activated and contaminated portions of the Fort St. Vrain Nuclear Generating Station. The decommissioning of nuclear facilities is a regulated process, whereby radioactive material will be removed from the plant site, the site will be decontaminated to limits established by the NRC to allow release of the site for unrestricted use and the NRC license will be terminated.
To accomplish the decommissioning of Fort St. Vrain, substantial portions of the existing plant will be decontaminated, dismantled and/or removed to release all site areas for unrestricted use. However, components and structures which are not activated or radioactively contaminated or which have been decontaminated will remain and may be utilized in the conversion of the balance of plant (BOP) systems to a natural gas-fired plant.
An overview of the decommissioning work scope is:
- 1. Removal of the Prestressed Concrete Reactor Vessel (PCRV) internal radioactive components remaining after the defueling of the reactor.
- 2. Dismantle those portions of the PCRV structure and radioactive balance-of plant systems which exceed limits for unrestricted releasa of residual radioactive materials. Decontaminate those struco res or systems where only low level surface contamination exists.
- 3. Ship all radioactive waste offsite for disposal.
- 4. Perform a final site radiation survey to confirm that all site areas can be released for unrestricted use.
- 5. Terminate the 10CFR50 operating license.
I 3-1
I'
~
Supplement to Environmental Report FSV Decommissioning Section 3 The detailed dismantling activities that -.will- be ' performed during
~ decommissioning are described in Section 2.3 of the PDP (Reference 1).
3.2 Site Cleanup, Final Site Survey and Site Release for Ur .estricted Use Upon completion of all PCRV dismantling operations and rei!, oval of the
- radioactive wastes from the site, the remaining portions of the site will be cleaned up by performing the following major activities:
Contarninated tools and equipment will be cleaned for reuse or packaged for disposal.
All Surface Contamination will be removed to the levels required for release for unrestricted use.
Radioactive liquid, -gas and-' solid waste systems will be ;
- removed and disposed. I A final radia' tion survey of the site and environs, to verify that all radioactive material has been removed, will be performed.
A: final ~ decommissioning program report and request- for termination of the Part 50 License will be prepared and submitted to the NRC.
As discussed in Section 3.3, " Criteria for-Release of Equipment-for Unrestricted Use," materials and equipment released from the site for unrestricted use will be released in accordance with regulatory requirements, if contamination of materials and/or equipment is found below the surface, the material will be disposed of as radioactive waste unless. it is determined to: be cost effective to decontaminate to unrestricted use criteria.
All temporary structures and modifications to-permanent site structures will be removed or returned to their original condition, as required. The potentially contaminated structures.inside the controlled areas such as platform scaffolding will be' decontaminated to release limits for-unrestricted use or shipped offsite as low-level waste. Temporary structures such as the laundry and respirator cleaning trailers will be decontaminated as required and returned to their licensed vendor. The 3-2 H
1
Supplement to Environmental Report FSV Decommissioning Section 3 contaminated - materials will be - packaged and shipped to' a - volume reduction facility, or to a burial site for final disposition. The trallers used ,
for personnel occupancy will. be surveyed and decontaminated, if
.necessary, for unrestricted release and shipment offsite.
After all activated . systems have been removed from the site, the remaining structures will be surveyed. All contaminated surfaces such as floors and ' sumps -wiH be decontaminated to remove residual contamination. The structures ve91 be surveyed to ensure all activated and-contaminated material has been rerneved.
The provisions of Regulatory Guide 1.86 (Reference 2) will be used for
~
residual fixed and removable contamination on the surface of materials, equipment and facilities _ left at the site, in addition, radiation levels of
~
5 R/hr above background measured at a distance of 1 meter will be used as the upper limit-for external radiation levels. The criteria for total 4
concentrations of radioactive materials above background in soll will be based upon those established in NUREG/CR-5512 (Reference 3). The use of this methodology will ensure an average total effective dose equivalent of less than 10 mrem /yr to an individualin a given population group.
Pursuant to Regulatory Gulde 1.86 (e), o comprehensive final radiation
- survey will.be conducted at the facility upon completion of all activities.
The goal of the survey is to ensure that ambient radiation levels and surface contamination levels are below the limits specified in Regulatory
- Guide 1.86 and to verify that the site can be released for unrestricted use e by meeting the acceptance criteria of 5 R/hr above background at 1 meter from the surface, f
- .The survey will cover all pertinent structures, surfaces, systems and t - components, focusing on those items identified as potentially troublesome L 'd u rin g the pre-decommissioning and during_ the I decontamination / dismantling phases. The survey will include
e PSC property,. soil, stream, and pond sediment, and water i sampling outside the fenced area for radioisotopic analysis o PSC property soll, lagoon sediment, and water sampling inside-
[: the fenced area for radioisotopic analysis f
e Radiological surveys for the PCRV and reactor building n 3-3 l
I Supplement to Environmental Report FSV Decommissioning Section 3 NUREG/CR 2082 (Reference 4), " Monitoring for Compliance with Decommissioning Termination Survey Criteria" will be used as a guide in developing the comprehensive survey program for quantifying residual radioactivity and radiation levels.
T he current Radiological Environmental Monitoring Program (REMP) will be continued in part specifically tailored to accurately monitor the environmental radiation and radioactivity levels, and to determine the effect on the radiological conditions of the environment due to decommissioning activities. The current REMP contains provisions for sampling and analyzing airborne, waterborne, ingestion, and direct radiation exposure pathways. The results of the REMP will be included in the final radiological survey report, in addition, the Offsite Dose Calculation Manual (ODCM) will provide the methodologies to assure compliance with the Fort St. Vrain Decommissioning Technical Specifications related to liquid and gaseous radioactive effluents. This program will demonstrate compliance with 10 CFR 20, 10CFR 50 Appendix A and Appendix 1, and 40 CFR 190.
At the completion of the decommissioning effort, PSC will prepare a final project summary report. The report will document the decommissioning activities that were performed and will be submitted with the application for termination of the Part 50 License and along with the final radiological survey report, will provide the basis for the approval to terminate the license.
3.3 Criteria for Release of Equipment and Non-Radioactive Waste The guidance provided in NRC Circular 81-07, " Control of Radioactively Contaminated Materials" (Reference 5) will be used to ensure appropriate survey methods are employed for the unrestricted release of decontaminated items (e.g., tools and equipment) and scrap materials.
The guidance provided in NRC IEN 85-92, " Surveys of Wastes before Disposal from Nuclear Reactor Facilities" (Reference 6) will be used to ensure appropriate survey methods are employed for the monitoring of segregated waste prior to disposalin a sanitary landfill. Procedures that have been prepared with reference to NRC Circular 81-07 and NRC IEN 85-92 will be used for surveying material to be released off-site.
3-4
Supplement to Environmental Report FSV Decommissioning Section 3 3.4 Radioactive Waste Ouring the decommissioning activities radioactive mater!als (radwaste)in liquid, solid and gaseous forms are expected to be generated.
Management of these wastes is an integral part of the decommissioning plan and includes provisions for minimizing the quantity of waste generated, waste collection treatment, volume reduction, packaging and off-site shipment for disposal.
Some solid materials may be volume reduced onsite using the existing Fort St. Vrain compactor, a mobile compactor, or other equipment as needed and as supplied by the decommissioning contractor (Westinghouse), if an onsite compactor is used, its effluent diccharge will be filtered by HEPA ventilation and airborne contamination surveys will be conducted during operation. Compactor operation will be controlled by a specific radioactive waste procedure to ensure safe operation. Compacted waste will be placed in appropriate containers such as 52- or 55-gallon drums, compacted, overpacked if required, and shipped offsite for disposal or additional volume reduction in order to reduce disposal quartities, some solid materials may also be shipped offsite for decontamination. All activities will be performed in accordance with applicable radiation l protection and radioactive waste procedures.
I Supercompaction, incineration, and melting for volume reduction wil! be l accomplished offsite at a facility licensed for these specific waste volume I
reduction techniques.
3.4.1 Classification of Radioactive Wastes l
l Classification of radioactive waste is required by 10CFR20,10CFR61, and disporal site requirements.
A Waste classification compliance program will be developed and l implemented to assure proper classification of waste for disposal. This program will ensure that a realistic representation of the distribution of radionuclides in waste is known and that waste classification is performed in a consistent manner. Any of the following basic methods, used individually or in combination, will be used to achieve this goal: materials accountability (inc!uding process knowledge and activation analysis),
classification by source, oss radioactivity measurements, and measurement of specific rad, uclides.
3-5
Supplement to Environmental Report FSV Decommissioning Section 3 Materials will be categorized as follows:
- 1. Non contaminated or minor spot decontamination required:
Materials that, on a gross basis, appear to be uncontaminated and have geometries with all surfaces easily accessible, with a small surface area-to-weight ratio will be surveyed to determine if the material can be released for unrestricted use without decontamination or with minor decor. amination effort. For example, a small surface area with coly spot and/or smearable contamination can easily be decmtaminated by such means as wiping, grinding, or removing the hot spot.
- 2. General contamination with accessible surfaces and a low area-to-weight ratio: Materials that have good geometries for purposes of surveying and decontamination, and that possess a low surface area to-weight ratio, may be shipped directly to a licensed facility for disposal or decontamination of the surfaces and subsequent release for unrestricted use or decontaminated on site to the criteria for release for unrestricted use.
- 3. General contamination / inaccessible surfaces /high surface-area-to-weight ratio: Smaller metallic-scrap or metals that have poor geometries for performing surveys (e.g., previously sheared material) will be assumed to be contaminated and be packaged for shipment for further processing at a licensed facility or shipped directly to burial.
- 4. Activated or other non-recoverable materials: activated materials and high specific activity materials will either be packaged and shipped directly to a disposal facility or to a licensed facility for further processing and volume reduction.
Radioactive materials identified above will be evaluated to determine the optimum method for release, decontamination, or shipment off-site for further processing or for burial, generally, as described in Tables 3.4-1 and 3.4-2. Items not considered for decontamination or items that, following decontamination, are considered to have too high a specific activity for off-site volume reduction, will be packaged and shipped directly for disposal at a disposal facility. Greater than Class C (GTCC) wastes will be packaged for on-site storage or for subsequent shipment to a facility designated by the DOE.
3-6
l Supplement to Environmental Report FSV Decommissioning Section 3 3.4.2 Projected Waste Generation The initial estimate of the processed and volume reduced radioactively contaminated waste for disposalis 100,072 cubic feet, with 99,219 cubic feet from the PCRV and associated operations, and 853 cubic feet from the balance of plant (BOP) Systems. The waste from the PCRV consists of activated concrete, graphite blocks, other activated components, miscellaneous equipment and piping, and concrete rubble. PCRV waste is contaminated principally with Fe-55, tritium, and Co-60. The waste from the BOP consists of tanks, pumps, HVAC filters, and miscellaneous equipment and piping.
After processing and volume reduction,it is estimated that the volume of radioactive waste will be segregated into the following categories:
_Cjaga Volume (cubic feet)
A 70,768 8 28,293 C 1,011 Due to uncertainties in the analysis, as much as 400 cubic feet of Class C wastes may be reclassified as GTCC. Waste class and volume estimates may change as onnoing planning and decommissioning operations proceed. Tables 3.4-1 and 3.4-2 identify the radioactive wastes that may be shipped for further processing. The pre-volume reduction totals and number of waste containers are delineated on Tables 3.4-3 and 3.4-4.
There will be approximately 20,000 cubic feet of uncontaminated concrete and metallic scrap. This will be transported by truck to locallandfills for disposal or for re-use as scrap materials.
3.4.3 Waste Treatment 3.4.3.1 Steam Generators Because of the anticipated 2.2 rem / hour contact dose rate associated with the steam generator primary assembly (economizer, evaporator, and superheater sections), a special shielded shipping container will be i
3-7 l
u .
l Supplement to Environmental Report FSV Decommissioning Section 3 required. The following methodology will be 3.nployed to remove and ship the steam generator primary assembly.
As the primary assemblies are lifted from the PCRV by the Reactor Building crane, the outer shroud and tube surfaces will be washed down to remove as much contamination ar.J cutting debris as possible, and will be allowed to drain as necessary over the PCRV cavity.
The steam generator primt ry assemblies will then be placed in a shipping container, located in the tru7k bay. The shipping container will consist of a metal culvert section seven foot in diameter by 27 feet long. The culvert section will be cut in half lengthwise to provide a hollow half-cylinder. Structural supports will be welded to the half section of culvert to provide structural support and allow the culvert to be upended with the primary module inside.
Support saddles will be mounted inside the culvert and serve a dual purpose. First, the saddles provide a means of attaching the steam generator primary assembly to the culvert and transmitting the load to the structural supports on the outside of the culvert. Second, the saddles will keep the steam generator primary assembly centered in the culvert with an annular space of about 8 inches between the inside diameter the culvert and the outside diameter of the steam generator primary assembly.
If required, the annular portion of the steam generator between the shroud and the cold reheat piping will be filled with grout which will encapsulate the tube bundle of the steam generator. In addition, grout may be pumped into the feedwater and steam tubes of the primary assembly, if necessary due to the high contamination levels, the 8 inch annular region between the outside of the steam generator shroud and the inside of the culvert will be filled with grout for shielding.
The combined weight of the shipping container, steam generator assembly, and groot will be approximately 195,000 pounds. If actual contamination levels in the steam generator primary assemblies are lower than expected, the shielding grout in the annular space between the steam generator shroud and the container may be omitted with a weight savings of about 56,000 pounds. A final radiological survey will be performed to ensure DOT requirements are met before the steam generators are released for shipment. The steam ganerators will be shipped by rail for disposal at a licensed burial site.
3-8
Supplement to Environmental Report FSV Decommissioning Section 3 3.4.3.2 On-site Processing of Liquid Wastes Radioactive liquid wastes which exceed 10CFR20 limits for the general public will not be released to the environment. To the maximum extent practicable, prior to discharge to the environment, the liquid waste will be processed to decrease the concentrations of radioactive material to levels below the limits established in 10CFR20. Processing will include filtering, domineralization and/or dilution. The liquid effluent will be monitored to assure that the liquid effluent released to the environment does not exceed the 10CFR20 limits. The method for processing on-site liquid wastes is described in Section 4.2. This method provides for proper dilution while also allowing a liquid effluent release rate that is greater than the 10 gpm value previously described in the Final EnvironmentalStatement (Reference 7).
Liquid wastes will be proces ted and disposed of in accordance with written approved procedures. Typicalliquid waste expected includes oils from piant systems, water from PCRV bleed and feed operations, and sludges from diamond wire cutting and sump clean-outs. Disposal of contaminated oils is expected to be accomplished by transfer of the oil to a licensed vendor for incineration. The chemical / hazardous nature of oils will be known prior to transfer for incineration. Water from the PCRV will be processed through the PCRV Waer Cleanup / Clarification System and discharged through normal plant effluent systems.
Chemicals used for decontamination will be evaluated for hazardous constituents using 40 CFR and the Material Safety Data Sheets (MSDS).
Decontamination chemical wastes expected include acids, caustics, detergents and non-hazardous solvents. The specific chemical for a particular application will depend on the material to be decontaminated.
Acids or bases will be neutralized and solidified or used for water chemistry control in the PCRV water clean-up system. Detergents and other water based solvents will generally be associated with damp rags or wipers. The wiping material will be dried prior to packaging for disposal or volume reduction.
All hazardous chemicals and materials will be subjected to a chemical control review to determine if a non-hazardous or a less toxic chemical can be substituted to prevent generation of mixed wastes, in the event that hazardous chemicals or materials must be used, procedures will ensure that all waste minimization techniques will be applied durin0 usage. Steps 3-9
4 Supplement to Environmental Report FSV Dacommissioning i Section 3 I will be taken to ensure that if hazardous materials must be used, the necessary controls will be in place so these materials will not inadvertently become radioactively contaminated. If some hazardous material does inadvertently become radioactively contaminated, it will be considered as mixed waste and subject to applicable regulations, if mixed wastes are generated, they will be managed according to Subtitle C of RCRA to the extent it is not inconsistent with NRC handling, storage, and transportation regulations.
If technology, resources and approved processes are available, PSC and the Westinghouse Team will evaluate the processes for rendering mixed waste "non-hazardous" to determine its adaptability to Fort St. Vrain decommissioning activities. PSC does not intend to petition the EPA to delist any mixed waste. However, if PSC determines it is necessary to delist any mixed waste, the procedures outlined in 40 CFR 260.20 and 260.22 will be used to exclude that waste form from regulations.
3.4.3.3 Airborne Contamination / Gaseous Radwaste Some low-level airborne . wastes may be generated during decommissioning. Disposal of airborne radioactive wastes will be accomplished by filtration (of particulates) and disposal of the filter as solid waste.
l Any gaseous effluent in the Reactor Building flows through the existing l ventilMion exhaust system. The Reactor Building ventilation exhaust filter system is designed to filter the Reactor Building atmosphere prior to release to the vent stack during both normal and most accident conditions during decommissioning.
l The system consists of three trains, one of which is normally in l
continuous operation. The design flow rate for each train is 19,000 cfm.
L Allowing 10% for degradation, the minimum flow rate is 17,100 cfm.
i One train is sufficient to maintain the Reactor Building subatmospheric and thereby rninimize unfiltered fission product release from the building. With only one exhaust fan operating, the ventilation system controls will throttle fresh air supply to the air handler in order to maintain the building pressure subatmospheric. The ventilation exhaust system is equipped with roughing filters, to capture large size particles, and with high efficiency particulate air filters (HEPA's) to provide up to 95% particulates 3-10
1 Supplement to Environmental Report FSV Decommissioning Section 3 retention. HEPA filters will be changed upon high radiation levels reading / alarm in the exhaust duct, or based on maximum pressure differential readings indicating that the filters are filled with dust.
Radiation monitoring is provided for the exhaust air to atmosphere and readings above prescribed limits are alarmed.
The Reactor Building is maintained in a subatmospheric condition to ensure that all air leakage will be inward and to minimize unfiltered fission product release from the building. The ventilation system was designed to maintain a subatmospheric condition approximately 1/4 inch water gauge negative. In actual practice, the Reactor Building pressure is normally 0.15 to 0.20 inches water gauge negative, depending on building activities and ventilation system configuration. There is an alarm at approximately 0.08 inches water gauge negative, and the outside air supply will fully close if the building pressure increases to atmospheric.
Any ventilation units serving localized confined enclosures will be mobile type having in line HEPA filters and connected with flexible ducts to the enclosure and to the exhaust duct.
3.4.3.4 Piping Contaminated piping will be packaged as LSA material in strong tight containers. The piping will be sectioned to fit into commercially available steel LSA boxes. Within the boxes, when feasible, smaller bore pipe will be nested within larger bore pipe to enhance packing efficiency, in addition, when practical the boxes will be topped with lighter dry active waste such as paper, plastic and clothing to maximize the box weight limits.
3.4.3.5 Dry Active and Solid Waste Material such as barrier plastic sheets, rags, paper and protective clothing will probably be compacted and packaged.
Sludges will be dewatered and dried, or solidified / absorbed using disposal site approved solidification /absorbentmedia. Solidification,if required, will be controlled by an approved Process Control Program (PCP).
3-11
Supplement to Environmental Report FSV Decommissioning Section 3 Solid wastes will be processed in accordance with written procedures. A general plan for solid waste processing is as follows:
-1. The waste will be initially identified at the point of generation as to the type of material and exposure rate.
- 2. The material will be segregated to allow for decontamination on-site, shipped to an off site vendor for volume reduction or packaged for disposal at an approved disposal site.
3.4.3.G Equipment and Tools This includes such items as saws, Jack hammers, monorail, forklift, shovels, pumps, tanks, ventilation system components, filters, piping, etc.
Only some, and parts of some of this equipment are expected to be discarded as radwaste. A determination of volumes of solid radwaste generated from this category will be possible only during the clean-up task when final measurement of contamination level and evaluation for decontamination will be made.
3.4.4 Radioactive Waste Disposal The radioactive waste disposal program will follow the regulations established in 10CFR20 and 10CFR61, the disposal site criteria, and other applicable Federal and State regulations. Most packaged waste will be put
-in strong tight metal containers suitable for Class A Low Level Waste (ALLW). Examples of the waste containers that may be used are drums
. (52-gallon, 55-gallon), boxes (2'x4'x6', 4'x4'x6'), liners, HIC's, sea / land containers, casks, and specialty containers. The capacity and weight limitations are governed by the activity levels and classification of the enclosed materials as specified in 49CFR and 10CFR71. The waste container will be determined by the size, weight, classification, and activity level of the different types of materials. Guidance for this activity will be contained'in radioactive waste procedures.
'GTCC waste, if any, will be stored in the adjacent ISFSI or in a structure which meets the design requirements to handle GTCC waste. The waste will be stored until such time as it can be transported to a facility licensed to accept the GTCC waste.
3-12
Supplement to Environmental Report FSV Decommissioning Section 3 in general, radioactive waste will be packaged and shipped off site for disposal on a continuous basis. Long-term storage of weste is not anticipated. Packaged waste ready for shipment will be temporarily stored onsite so that shipping loads can be maximized on a practical basis, Waste storage facilities planned for use during decommissioning activities include:
- 1. The Independent Spent Fuel Storage Installation (ISFSI) will be used for greater than Class C wastes (GTCC), if any, pending.
approvalof an appropriate disposal site. (No GTCC wastes are currently expected.)
- 2. The New Fuel Storage Building will be used as a processing and storage area for dry low level wastes.
- 3. The IACM Building (Comoactor Building) will be used as a processing and storage area for dry and dowatered low level Wastes.
- 4. The Reactor Building will be used for the storage of liquid and solid wastes.
- 5. Trailers and sea / land containers will be stored and used onsite to house dry and solidified low level waste.
0.. Selected yard areas will be used for short term storage of packaged waste staged for transport.
The activity levels of wastes stored in these areas will be controlled to levels as evaluated in the accident analysis, Section 3.4 of the PDP (Reference 1).
Safety evaluations have been performed that assess and permit storage of low level radioactive waste on the Fort St. Vrain site corisistent with the guidelines of Generic Letter 81-38 and the Standard Review Plan
-(NUREG-0800), Appendix 11.4-A. The Fort St. Vrain Possession Only-License and Technical Specifications permit possession and use of byproduct, source, and special nuclear material in quantities as required pursuant to 10 CFR 30,40 and 70.
3-13
Supplement to Environmental Report FSV Decommissioning Section 3 Due to the building seismicity and other drainage and collection requirements for the storage of wet radwaste, PSC does not intend to store wet / liquid radwaste outside the Reactor Building. The Reactor Building was designed and built with drainage systems that route spillage to collection points / sumps that are monitored for radioactivity and properly processed. Other forms of radwaste may also oc stored in the Reactor Building without significant concern, due to the building's additional features relative to fire detection and suppression, and its filtered ventilation system.
The Compactor Building is a steel building constructed on a concrete foundation, with its own " wet-pipe" fire suppression and fire detection systems. This building has two concrete basins that may be used to store barrels of dowatered wastes, consistent with the recommendation of Generic Lotter 81-38. Other dry and solidified wastes may be stored in this building in amounts consistent with limitations of the decommissioning accident analyses. A radwaste compactor, with a self-contained HEPA-filtered ventilation system, is also housed in this building.
The New Fuel Storage Building will also be used to store packaged dry and solidified low level radwaste. The New Fuel Storage Building is a single level concrete structure which is designed to withstand a 202 mph tornado wind and can withstand the design basis tornado missile. A safety evaluation has determined that no increase in an accident probability will result from radwaste storage in this location. As stated in the decommissioning accident analysis, fire detection systems will be provided before combustible radwaste will be stored in the New Fuel Storage Building.
Trailers and sea / land containers have been evaluated to house dry and solidified radwaste. Accident scenarios have been postulated and the total allowable activity levels for storage are controlled accordingly. Yard fire hydrants are available for use if necessary.
Certain large radioactive compcnents (such as helium circulators packaged for shipment) may be stored outside within the protected area while awaiting shipment offsite. Tle down systems will be considered for components stored outside, and will be installed when needed. Steps will be taken to protect containers from external corrosion as required.
3-14
l Supplemer; to Environmental Report FSV Decommissioning Section 3 All interim low level radwaste storage locations described above exist within the plant's protected area. Radiation protection procedures will be implemented that will specify requirements for processing and packaging radwaste. Procedures will also be implemented to provide suitable instructions to establish radiologically controlled areas to ensure occupational and public exposure are kept within the requirements of 10 CFR 20,10 CFR 100, and 40 CFR. Procedures will be implemented to monitor storage areas periodically and to check waste container integrity.
As a radioactive waste generator in the Rocky Mountain Compact, PSC is currently required to dispose of radioactive waste at the Beatty, Nevada disposal site. When the Beatty site is no longer operational at the time of decommissioning, the NW compact disposal site at Richland, Washington currently plans to accept waste from the Rocky Mountain Compact states.
The commercial disposal site at Barnwell, South Carolina will also be considered as an alternative for burying items that have been processed, where economically feasible and approved by the appropriate compacts.
3.4.5 Transportation Plan Before shipment from Fort St. Vrain, each package will be inspected to determine its adequacy for retaining the radioactive contents and its proper condition for shipment. Bar codes capable of being read by computerized scanners will be affixed to the container and the corresponding lid.
Waste will be packaged into waste packages that meet 49CFR and 10CFR71 requirements. Certain wastes may require use of an approved shipping cask due to radiation levels or limits for quantities of radioactivity.
Trucks will be the primary mode of transportation during this decommissioning project. It is anticipated that the shipping containers with the steam generators and grout will be shipped by rail for disposal.
Transportation surveys and documents will be prepared prior to any shipment offsite. Analyses will be performed to determine isotopic inventory and concentration for classification.
The actual routing of shipments may vary with weather and highway conditions. Additionally, local and state restrictions pertaining to radioactive material transport may affect some route selections, particularlyin congested metropolitan area.s. The carrier is responsible for 3-15
Supplement to Environmental Report >
FSV Decommissioning Section 3 selecting the appropriate route, which must conform to applicable federal, state, and local regulations. Trained persor".el will inspect and oversee shipping, in accordance with DOT anc' NRC ,egulations.
During decommissioning operations, PSC does not expect that truck shipments will exceed normal highway axle load limits, it is planned that the large concrete blocks being removed from the PCRV will be further cut and reduced in size, first so that only the radioactive portion of the concrete is disposed of as radioactive waste, and secondly so that the concrete blocks will be of a small enough size for truck haulage within normal axle load limits. Trucks meeting the highway axle load limit requirements will be used to carry heavy loads such as concrete blocks or steam generator packages.
The large steam generator packages are planned to be shipped by rail,-
although special heavy hauling equipment may also be used. If the steam generators are shipped by rail, the rail spur near the plant will be refurbished.
The onsite road within the Fort St. Vrain protected area (that leads to the Reactor Building Truck Bay) is_ a paved road and is in good physical condition. No additionalimprovements to this onsite road are necessary.
The present access roads from the Fort St. Vrain site to the main interstate P.ghway (1-25) are paved roads,in good physical condition, and are well maintained by Weld County and the State of Colorado. These access roads have been in continuous use duilng the life of the Fort St.
Vrain facility and have accommodated routine truck traf fic over the routes.
Transportation af decommissioning radioactive waste will not involve any-additional considerations beyond those for transportation of existing radioactive waste material. Existing regulations for trarisporting radioactive waste material are covered under NRC regulations in 10CFR 20, 71 and 73 as well as DOT regulations in 49CFR 170-189.
in accordance with 10CFR71.5(a) all off-site waste shipments will meet the applicable DOT requirements of 49CFR 170 - 189. This includes:
l Shipments will be loaded by the consignor (licensed shipper) and unloaded by the consignee, L
l 3-16 l
l
Supplement to Environmental Report FSV Decommissioning Section 3
- Licensco-shipper will be responsible for assuring that the shipment is properly loaded and secured.
3.5 Asbestos The Fort St. Vrain Plant was completed in the early 1970s and continuin0 maintenance and modification have been going on since that time. As a result, there is asbestos containing material in the plant. Thus, all decommissioning activities must recognize this potential and appropriate precautions taken to assure the health and safety of the workers and the public.
As part of the preparations for the initiation of actual decommissioning activities, a characterization survey was performed of those systems and areas that will be involved with the dismantlement and decontamination of the plant. That characterization found that 60 out of 155 samples taken contained asbestos. The asbestos was associated almost entirely with two plant systems, the steam generators and the helium circulators.
None of the samples were radiologically contaminated. Thus, asbestos is not expected to present a major problem for decommissioning.
During decommissionin0, only that asbestos that is associated with the decontamination, dismantlement, and removal of radioactive systems and components will be disturbed and removed.
All activities involving asbestos will be conducted in accordance with federal and state regulations (OSHA 29CFR 1910 and 1926, EPA 40CFR61 Subpart M). An Asbestos Removal Specification will be developed for the site consistent with the National Institute of Building Sciences Format, All asbestos removal work will be performed by a competent asbestos removal contractor, with appropriately trained and certified personnel. All asbestos will be packaged for shipment and disposed of at an authorized disposal site.
It is not expected that any of the asbestos will be contaminated with radioactivity. However,if anyis found that has radioactive contamination, it will be disposed of at a commercial radioactive waste disposal facility.
At the present time, the operators of the low-level radwaste disposal facilities near Beatty, Nevada, and Richland, Washington, accept asbestos for disposal.
3-17
I Supplement to Environmental Report FSV Decommissioning Section 3 3.6 Manpower Levels and Schedule Figure 3.G-1 is a schedule of the major decommissioning tasks which include PCRV dismantling and decontamination, and balance of plant BOP) and site decommissioning. This schedule is used as the top level view of the project milestones and detailed schedules. Throughout the project, dismantling the PCRV is the critical path activity, with the BOP dismantling activities scheduled to coincide with periods of reduced PCRV efforts as a means of workload leveling. The overall decommissioning project is expected to be completed over a 57 month period. The planning phase occurs over an 18 month period and the actual dismantling and decommissioning activities are planned for a 39 month period. This schedule is based on 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> work weeks. The workforce will consist of a combinMion of PSC and Westinghouse Team personnel and is estimated at 300 people. This manpower level does not include personnelinvolved in conversion activities.
3.7 References
- 1. Proposed Decommissioning Plan for the Fort St. Vrain Nuclear Generating Station, Public Service of Colorado, November 1990.
- 2. USNRC Regulatory Guide 1.86 " Termination of Operating Licenses for Nuclear Reactors," June 1974.
- 3. NUREG/CR-5512 " Residual Radioactive Contamination from Decommissioning," January 1990. Draft Report for comment,
- 4. NUREG/CR 2082, " Monitoring for Complience wit h Decommissioning Termination Survey Criteria." Draft Report, January 1990.
- 5. NRC Circular 81-07, " Control of Radioactively Contaminated Materials." May 1981.
- 6. NRC Information Notice 85-92, " Surveys of Wastes Before Disposal from Nuclear Reactor Facilities " December 1985.
- 7. Final Environmental Statement Related to Operation of the Fort St.
Vrain Nuclear Generating Station of Public Service Company of Colorado, August 1972.
3-18 1
Supplement to Environmental Report FSV Decommissioning Section 3 Table 3.41.
PCRV Waste Classification and Volume Rcduction A 8 C< *C kkki COMPsC1 ~ pct t OVERFILL Regicn constraint devices X X RCD Pins X X Metal Control Rod Reflectors X X Metal block non control rod X X Defueling elements X X X Top reflector graphite blocks X Sottcrn reflector grarmite blocks X Radial reflector gregelte blocks X Large reflector blocks X X 1/2 size reflector blks X V Linger reflector keys X M X sosegr blocks w/beron pins SipB
[bs i$hpinsremoved X X Bottcm gflector blocks "I X X Bfo!ellhcEns'ckswfoutcans X Lower reflector keys X X Core sg: port blocks X X Core suroort posts X . X X Core s4 port floor coltrns X [ X X Misc steel beneath CSF X X X Metal on large side reflector X X X Core barrel X X
Lower plenm insulatim X X Silice blocks X X Concrete - top X Concrete - CSF X Concrete - side X Concrete rubble X X Misc. Inconel parts (CSF) X X Concrete cuttire debris X X beti m purifters (PCRV head) X X Neli m diffusers X X X Helim circulators X X 3 Circ shutof f valve asserely X X Heli m bellows X X Steam generators X Lower floor /acTurtenances X X X Platfors/ tools / jib cranes X X X Crane cable /detr:/3 bucket invrtr X X kesins X
} X
! Mist.ellaneous scft waste N I X 3-19
l l
Supplement to Environmental Report FSV Decommissioning-Section 3 Table 3.4-2.
Contaminated BOP Waste Classification and Volume Reduction Alt C 4 Nb Whkhk COMPACT Miki OVERdL{ E N tiisceltaneous sof t won,te x X
' ~
Neactor isolatiori valves X X X Aefueling sleeves M ,
X X Refueling sleeve sand K use es 0/f Sard f rom f $ws X use as o/f j aTC X X !
ATC sand X use as 0/f 13W sard X tse as o/f not service facility X X t'SF sard X use stab. ~'
Core supp1rt vent f1iters X X
~
Gaseous waste surge tarts X X Gaseous waste surge tar
- sand X use as o/f Liquid drain tank X X Gas waste vacu.m tank X X Gas waste vacuin tank sacd X use as o/f Gas W.iste conpressors X X ,
I Ges 'este ce g ressor sard X j use as o/f Licuid monitor tank X X Limid waste monitor tant sand X tce as o/f timid waste demineralizers X X Licpid waste receivers X X
~ ~
Licuid waste receivers sard ' X
~~
use as o/f Licpid waste strp sard X~ use as o/f
. Llauid transfer pmps X X ,
^
Limid waste swp sups X X Liquid waste resins X use as o/f Lfcpid waste filters X X Decon solution tank X X Decon solution tank sand X use as o/f Decon *ecycle prp X X Decon chemical stoply prp X X Purified helita filters X X Hetita temoval filter X X Helita Gstter tritts X X Mpc filters X X Smali & targe bore piping X X X
^
FHM X X FHM CO W Mits X X g fMM band X urse a5 O/f j W td waste cwpactor X1 X l _
3-20
Supplomont to Environmental Report FSV Decommissioning Section 3 Table 3.4-3.
PCRV Waste Volume Estimates ITEM /STSTEM
~
CLAS$ CbCT L$A NLMSE R VOLLME (PT))** CMAINRs Wegion constraint device L pin C 70 Ivo 84 235 2
~
Metal control rod reflectors C 3000 No 37 401 3 Metal block con co.' trol rod
~
8 '
300 No 276 2025 13~
Defweiing blocki A Contam. Yes 16a2 7200 75 Top Reflector gropnite blocks A 0.5 ho 1215 1515 8 Bottent reflector grarritt e blocks A 0.5 No 1215 1414 8 Racie' reflector perm. & reble A 0.5 no 450 1903 9 Large reflector b ockt 9 <30 to 312 12600 , 50 ~
1/2 site reflecto blocks ~ 5 <30 No 3 12 4160~l 8 Upper reflector keys (carton steel)
A 7 0.1 bo 24 j 192 i 2 Side Spacer blocks w/ boron rod: 30 11 " l boren rods 8 60 No 309792 .
Blocks w/ruds recoved A <0.1 No 1152 l 2393 10 Bottom ref blocks w/nostelloy cans 3C0 27e
. Hastelloy C 10000 No 20061 Blocks with carn rernoved A 0.5 2 76 816 14 Lower reflector keys (Hastelloy) 8 1000TNo 24 180 1 tore s @ port blocks A <0.1 ~ Yes 61 1468 '~ 15 Core succort posts a <0.1 Yes 183 1 74 2 1
~
Core stoport floor colmns A <0.1 Yes 12 636 7 Misc. cteel f rca beneath CSF A <0.1 Yes 960 1C hetal on large side reflector iA <0.1 Yes 24 % t Core barret A .02 Yas 1 1400 31 Lower plenun insulat ton A <0.0G1 Yes 10 940 ~ 1 Silica blocks (25,000 tbc.) 4 0.5 Y es 503 12 Concreto top 4 <0.2 Yes 3744 9 Concrgte - CSF A 4.015 Yes 6240 15
~
Concrete side A < 005 Yes 18720 45 Concrete rubble
- jucahanmer A <.015 Y es 706 15 Misc. Inconet parts en CSF A 0.4 tio 415 ~I concrete cutting cobrts - top A <0.2 Yes 210 Concrete cuttino debris 'CSF A <.G15 Yes 200 ' l' Concrete cutting debris sidn A 4.005 Yes 323 Mellun purif ters in PCRV bead A j 0.15 Yes 10 480 Hellus dif fusers A l <0 .1 Yes 4 1752 4 ~
kettun cire, shutof f volve asmely - A <0.1 Yeo ) 4 192 2 Metius beltows A <0.1 Yes 12 1560 12 Steam generators A 2 Yes 12 20736 12 Thernuco@les and Guide tubes Lower floor / opurte m es 9
A 50 so.01 No Yes 105 1200 42 1_
Platforin.fhandLing tools / jib Cranes A <0.01 Yes 576 l 6 Crane catate/ tun /3 txxket inverters A <0.01 Yes 312 3
. Misc $ Containers A <0.01 Yes 2S8 3
. PCAV Wscer Systern A <0.01 Yes 2080 2 Resins solidify. whip, tury Aa 15 No _[ 20 2720 ~ 20 Mi:,c. Soft weste A <0.01 Yes 12000 125 i
PCRV TOTALS _ { 113,972 628
- Estittated Burial Class - Specific burial class ident!.fication may require additional analysis in accordance with 10 CFR 61
- - P re -vo ltune reduced quantitj.
3-21
1 1
Supplomont to Environmental Report FSV Decommissioning Soction 3 Table 3.4 4.
Es0P Wasto Volume Estimatos 11188/$f 5114 tLAll NhkCT tim Aikdf R Vollet (fil)* CCNil teactor isolation valves A <0.01 Tes 5 900 10 Refusting sleeves A <0.01 tes 2 192 2
~
land f r:rs fuel storage wells A ~~<0.01 Y es 750 hote 1
, lard f rm ewtgs,ent storage wells a <0.01 fes 225 hote 1 Sand f rors he t t us regerer et t on pit A <0.01 Y es 135 hote 1 Aunillery trarmfer cast send A <0.01 tes 15 hete 1 Not cett facility A <0.01 Yes los 4
, land f run hot cet t f acility, A <0.01 Yes 500 hote 1 (ore surport vent fiLte* A <0.01 Yes 15 2 Gaseous wante surge tares A <0.01 Yes 2 2646 2 tiavid drain tar
- A =0.01 Yes 1 20 1 Gaseou,s waste "actus tar
- _ A <0.01 Yes 1 980 1 Gaseous waste ccrttressors A <0.01 _ _ Y es_ 2 2058 2 L tculd waste aunt tt,r tore A <0.01 Yes 1 $/6 ~
i Liquid watts chrminerallaers A =0.01 Yes 2 192 2 Liould waste receivers A <0.01 Yes 2 1152 2 Liquid waste sum (tard) A <0.01 Yes 23 hote 1 Lic.ald transfer ptess A =0.01 f4 2 % 1 L t wid waste surp sursa A *0.01 Yes 2 5 hete 2 tiquid west 6 filters A 50.01 Y es 2 15 2 0econsoluttor[t3 A <0.01 T es 1 364 1 Deccrs recycle guy A
<0.01 Tes ___ t 2 hote 3 Decon chers surply tum a <0.01 Ves 1 2 hote 3
~
PurMied Het tus filters A <0.01 Yes 2 14 tote 3 Met tua retroval filter & <0.01 '.' es 1 i M 1 l
hetlus getter unt to A <0.01 Yes 2 __] _ 4 hote &
MvAC f 6 Iter s A <0.01 Yes 1010 1 f uel hardtirg machine A =0.01 Yes 192 2 l
Fuel hardling machine sand A <0.01 Yes 420 hote 1 Small ard large bure pipirg A <0.01 Ves 576 6 Reactor Bldg Drain System A <0.01 f ee 125 1 Instru'entation & controls A <0.01 Yes 225 2 TOTAtt 13.991 46
- - Pre volume reduced quantities.
Notes:
1- Will be used as overfill
,. 2- Will be packaged with liquid transfer pumps l 3- Vill be packaged with Decon solution tank n -
Vill be packaged with helium emoval filter 3 22
Supplement to Environmental Report FSV DecomtnissioninD Section 3 g .. .. . . y..
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i 3-23
Supplomont to Environmental Report FSV Decommissioning Section 4 4.0 ENVIRONMENTAL EFFECTS OF DECOMMISSIONING ACTIVITIES 4.1 Affected Environment 4.1.1 Overall Affected Environments The following areas, not rostricted to the immediato area of the plant, will be directly affected by the decommissioning of Fort St. Vrain:
- 1) Fort St. Vrain Sito Same a'oas as disturbed during plant construction and operation.
- 2) R9 placement power sources Samo power sources as utilized durin0 plant operation,
- 3) Radioactivo Low Lovcl Wasto (LLW) site Samo sites as huvo boon utilized during plant operation.
Samo facility as utilized during previous refuelings.
4.1.2 Demo 0raphy The population density in the rural areas surrounding Fort St. Vrain is relatively low, The nearest resident is located approximately one half milo north of the Reactor Buildin0, with the nearest town of Plattovillo located approximately 31/2 miles southeast. This is well outside Exclusion Area Boundary (EAB) and the Emergency Planning Zone (EPZL The population of Plattoville, based on preliminary 1990 consus figures, is 1515. The r.carest population contors with a population over 25,000 aro Grooley (60,399), Lon0mont (51,288), and Loveland (37,173), all based on preliminary 1990 consus figures, 4.1.3 Goo 0raphy and Land Use The site is located in Vveld County, Colorado. The area surroundin0 the sito is shown in F10 uro 4.1 1 with reference circles of 10,20 and 30 miles radil. The sito is located in the South Platte River Valley, approximately thirty five miles north of Denver, it is located in an a0riculteral area with Oontly rollin0 hills. Grado elevation at the plant is 4,790 foot, The foothills of the Rocky Mountains start to rise about twenty miles west of 4-1
Supplomont to Environmental Report FSV Decommiss!oning Soc, tion 4 the sito, and the Cont;nontal Divido is prominentiy identified by Lon0's Peak, located forty miles directly west of the site. ,
The South Platte River and St. Vrain Crook both pass through portions of '
the sito. Those two streams, which join near the northern tip of the sito, are not lar00 anough to be used for water transportation.
The general aron and land uso surrounding the sito is predominantly agricultural. The major farm products include grain, food corn, sugar boots, vogotables, boof cattle, shoco and turkeys. Thoro is also a limited nrnount of dairy forming in the area.
The industrial facilitios in the immodlato area are primarily located in the town of Plattovillo. There are 13 oillgas wells within a one milo radius of ;
the Reactor Building on Company prop 9 tty.
4.1.4 Geology and Solsmology The geologic structure of the general area in which the site is located is shown in Figure 4.1-2. The area lies on the east flank of the Colorado Front Range which is a complexly faulted anticlinal arch on which are superimposed numerous smaller folds and faults. The rocks of the core of the anticlinal arch are Procambrian crystallines, including gneiss, schists, and quartzitos which have boon intruded by granitic rocks that ,
rango in ago from Procambrian to Tortlary. On the east flank of the arch -
are Paleozoic and Mesozoic sedimentary rocks.
The regional structure of this part of Colorado is characterized by sedimentary rocks dipping eastward into the Denver Basin. Along the mountain front the regional structural pattern is interrupted by relatively small, on ochelon anticlines that plungo to the southeast. In addition to the fold axes, two groups of faults have boon recognized. The most notable occurs along the mountain front and includes a series of faults extending in a generally northwest southo6st diraction from the Procambrian into the Paleozoic Mosozoic sediments. The second group of faults has boon recognized primarily in coal minos, located generally cast of Boulder. These faults have a northeast-southwest orientation. Both groups of faults are relatively high angle faults.
The faults and the minor folds are related to the uplift of the Front Range which began in Lato Crotaccous and continued into the Tertiary. The original field examination and photo interpretation of the area surrounding the sito location failed to indicato any evidence of recent movement along 42
Supplomont to Environm2ntal Report FSV Decommissioning Section 4 any of the known faults. There is no known evidence of any recent l seismic acdylty in the immodlate area to have caused any subsequent movement.
The subsolls at the sits are St. Vrain Platte River alluvial sands and gravel overlying Plotto shalo uor! rock Generally,3 to 8 foot of loose to very loose clean sands (with occasional silty and clay lenses) are underlain by 30 to 35 foot of medium donso, fino alluvial sands. Thoso sands are underlain with 4 to 11 foot of medium danse to denso, sil0 htly clay, sandy gravel. Continuing under the gravel, hard to very hard interlayered sandstono and claystono bedrock is found at depth 45 to 51 foot. Froo water was found at a depth of about 23 foot. Estimated contours of the surface of the bedrock and the froo water lovel are shown in Figures 4.13 and 4.1-4. The shallow loose sands aro capablo of supporting only low foundation pressures, the medium densa sand will support moderato foundation pressures, and the bedrock will support high foundation pressures.
4.1.5 Hydrologv The sito location is betwoon the South Platte River and St. Vrain Crook about two miles south of the confluence of those two streams. Surface water rights are owned in four ditches which traverso portions of the site area, in addition, ninotcon shallow we!Is are located on the site area.
Flow of ground water on the site is toward the alluvial deposits of both the South Platte River and St. Vrain Crook. The contours of tha water table indicato that the flow of ground water is predominatoly toward the South Platte River Valley (Figure 4.15). Much of the ground water comes from the South Platte River and St. Vrain Crook, such that the water table changes with the flow rato (elevation) in the two streams. Total precipitation, mostly in the form of rain,in the South Platto Valloy is small and contributos relatively little to the ground water.
4.1.5.1 Plant Water Supply When the plant was operating, cooling water for the plant was supplied by the main cooling tower and the service water tower. Mako up water for the main cooling tower was obtained from water diverted from the South Platte River and St. Vrain Crook, and supplomonted by water from a system of six shallow wells. Make up water for the service water tower is suppfled by the domestic water system, with back up from the shallow well system. Potable water and water for closed systems in the plant, 4-3
Supplomont to Environmentel Report FSV Decommissioning Section 4 such as the secondary coolant system, is supplied by the domestic water district. The local water district is the Central Wold County Water District, whose source of supply is Colorado Big Thompson Project water from Carter Lake, which is iocated about twenty miles west of the site. The arrangement of the various water supply systems is shown in Figure 4.1 6.
4.1.5.2 Plant Effluent Liquid effluent from the plant is discharged primarily from alther the plant building drains or the cooling tower blowdown line.
Miscellaneous turbine plant drains such as floor drains, the Turbino Building sump, and yard drains, are normally directed to the South Platte River via the continuation of the Goosequill ditch to the farm pond. A diversion box is provided in the Turbine Building drain line, where of fluents are normally directed into the Goosequilt ditch. Under abnormal conditions which prevent dischargo via ths Goosoquill ditch, offluent is alternatively directed to the St. Vrain Creek via a slough. Similarly, the reactor plant drains flow to a diversion box from which the flow is normally directed to the South Platte River via the continuation of the Goosequill ditch, or alternatively to the St. Vrain Crook via a slough.
Further downstream from the plant, the Goosequill ditch flows into the Jay Thomas ditch and the combined stream flows into a 25 acto farm pond.
The overflow from the farm pont' flows into the South Platte River close to its confluence with the St. Vr 'in Creek. The drainage path via the Goosequill ditch and the pond is normally used.
Three lined evaporation ponds (total surface area of 3.6 acres) are present and are utilized to receive chemically treated effluent (primarily produced by periodic regeneration of plant domineralizers). Two ponds are located a few hundred feet northeast of the plant building. The other pond is located south of the switchyard.
Use of surface water downstream from the site is limited almost entirely to irrigation. A diagram of the major tributaries and irrk alon ditches on the South Platte River between the gaging stations Henderson and Korsey is shown on Figure 4.1-7. The plant site is located just upstream of the junction with the St. Vrain Creek, adjacent to the Jay Thomas Ditch.
4-4
Supplemsnt to Environmental Report !
FSV Decommissioning l Section 4 1 Analyses for the reactor site were conducted on the amount of diversion and stream flows of the nearby waterways. From these originalanalyses, j lt was concluded that effluent from the plant would be carried primarily by ,
the South Platte River except during the irrigation season with allowance for reservoir storage. Effluent in irrigation water would enter ground water i in the alluvium and would eventually be transported back into the strata bed of the South Platte River. There have been no significant changes in ;
the waterway flows or diversions to require new analyses.
The sources of public water supplies within thirty' miles of the site are given in Table 4.1 1. There are two towns downstream within this radius
- that presently obtain part or all of their water from wells in the alluvium of the South Platte Ifver: Gilcrest and LaSalle. It h:;s been common practice for farmers to obtain domestic water from shallow wells in the alluvium. Many of those who formerly used shallow wells as their source of domestic water now obtain water from the Central Weld County Water ,
District. This same district is the source of domestic water for the plant. t 4.1.6 Meteorology l
4.1.6.1 General Climate The general climate around the Fort St. Vrain reactor site is typical of ths Colorado eastern slope plains region. In this seml-arld. region the precipitation averages 10 to 15 inches a year, mostly from thunderstorms in late spring and summer. The annual free water surface evaporation rate is about 45 inches per year (Reference 1).
The wind records show no dominant direction, although winds out of the north by northeast segment do occur with the greatest frequency. The .
Winds are generally light (10 mph), with higher velocities occurring during
- various atmospheric disturbances.
The weather is generally mild. Most seasons are characterized by low humidity and sunny days, with occasional, short lived storms bringing precipitation into the area. Relative humidity averages about 40 percent during the day and 65 percent at night. Thermal radiation losses resulting from lack of cloud cover provide considerable variation in temperature from night to day. Although. snowfall may be significant, the snow cover ,
is usually melted in a few days.
45
--mU-----_-_-i_---_-.e.J.-..mr_ ..A;-..e--m .s%- .--. # -, ,-.--,----.#--,yw-.-,.3,.~wm-,3 ..c%.,%[py am,,,.,v-q--c,---,,-,.,,m-,,,-,.,v...p,-a-7,-y.+, ,
- - . . ..- . _ - - . _ _ ~
Supplomant to Environmental R: port FSV Docommissioning Section 4 4.1.6.2 Severo Weather Tabulated below are temperature and precipitation records for three cities within 20 miles of Fort St. Vrain (soo Figure 4.1-1). The recording periods woro 19731988 (Brighton), 19311988 (Longmont), and 1967 1988 (Greeley).
Briahton Lonomont Greelev Max. Temp. (dogrees F) 101 100 103 Min. Temp. (dogrees F) 23 36 25 Max. Precip. Day (In.) 2.73 4.04 3.20 Max. Snowfall - Month (in.) 22.1 32.1 37.3 Based on information extracted from archived weather data collected from Fort St. Vrain's 60 motor meteorological tower for the period 1986 through 1989, the following weather extremos were observed:
Maximum Temperature = 104 degrees F Minimum Temperaturo = 26 degrees F ,
Maximum Wind Velocity = 48 mph at wind direction G.5 degrees (NNE)
Seasonally, winds tend to be strongost in the late winter and spring, the season with high chinook frequency, and again in the summer, when thunderstorms occur frequently.
Stron0 winds, especially under chinook conditions, have been observed on various occasions in eastern Colorado. The chinook winds are strongest immediately to the east of the mountain ridge and diminish rapidly over the plains with increasing distanco from the mountains.
The measutomont records at the site from July 1986 to December 1989 reveal a prevalence of northerly and southerly winds caused by the shallow depression of the St. Vrain Creek and the South Platte River and by the proximity of the Rocky Mountains. The meteorologicaldata for this period for the wind speed and duration and frequency of distribution is contained in Tables 4.12 and 4.13, respectively.
Northeastern Colorado has moderate thunderstorm activity. The region near Fort St. Vrain averages 50 days / year in which tht.nder and lightning occur. The majority of those thunderstorms are present from late spring through the summer.
4-6 i N --.
1
Suppl:mont to Environm:ntcl Roport FSV Decommissioning Section 4 The Fort St. Vrain sito is located in a region that typically exporloncos 5 tornadoes por year por 10,000 square mitos. The peak tornado activity occurs in the month of June. According to the National Weather Service, 117 tornadoes occurred in Wold Coun'.y during the porlod 1950-1987. ;
4.2 Radiologicalimpact from Routine Activities The current Radiological Environmental Monitoring Program (REMP) will be continued in part specifically tailored to accurately monitor the environmental radiation and radioactivity levels, and to dotormlne the offect on the radiological conditions of tho environment due to decommissioning activities. The results of the REMP will be included in the final radiological survey report.
4.2.1 Onsite Processing of Liquid Wastos During the Fort St. Vrain decommissioning proJoct, contaminated water will be generated through several processos (such as flooding of the PCRV, rinsing of contaminated components removed from the PCRV) and through decontamination operations. Flooding the PCRV will result in the release of radionuclidos that exist in the PCRV as a result of activation and platoout into the water. Of primary concern is tritium, sinco a fraction of the tritium inventory is expected to teach out of the graphlto reflector blocks into the water. It is anticipated that some gamma omitting isotopes, Co 60 and Cs 137, will also enter the water. Based on data analysis and research of tritium release rates from graphite, the tritium levelin the water is expected to reach maximum concentration within 10 days of the initial flooding of the reactor. The PCRV will be flooded prior to removal of the top head of the PCRV. The radioactivo particulatos, Co-60 and Cs 137, will be removed under controlled conditions through filtration and domineralization processos and the tritium concentration reduced by a " feed and blood" process using the PCRV liner as a reservoir.
The tritiated water will be released after dilution via the normal plant liquid offluent pathway. During unforosoon instancos or circumstances that preclude use of the normal discharge path, the water may be directed to interim short term storage until a controlled discharge can be performed.
The water used to flood the PCRV wi!I be processed through the PCRV water -cleanup / clarification system. Water used to decontaminate components and structures may be processed through the plant liquid radioactive waste processing system (System 62).
4-7 e . -- . . - _ - - -. - - . - -. .. .- - . , _ _ - _
Supplom:nt to Environmental Report FSV Decommissionin0 Section 4 4.2.1.1 PCRV Water Cleanup / Clarification System The PCRV Water Clounup/ Clarification System will be installed to maintain water clarity and control radioactivo material concentrations in the PCRV.
The system will consist of two parallel trains. Each train will consist of a coarse strainer deslD ned to removo Gross debris and to protect downstream equipment. A standard dimension process pump will provido the driving head for the purification flowrato through two banks of filters located downstream of the pump. A profilter will remove larger suspended solids, and a final filter will provido the degroo of filtration necessary to ensure acceptable water clarity. Sultable valving and cross connection betwoon trains will enhance system flexibility and availability. Each train of the PCRV water cleanup / clarification system will have a recirculation
(" full flow filtration") capacity of 500 gpm. In addition to the capability for full flow filtration of the PCRV water inventory, the system design will also include partial (sido strearn) domineralization for controlling dissolved solids. " Food and blood" connections for adding clean makeup water and for removing contaminated water will also be provided to control the tritium concentration. Chemical addition tanks are included for chemistry and Ph control, and to suppress biological Growth in the water system.
The system design will also includo instrumentation, controls, and sampling points. These will enable proper operation in monitoring the system and offectiveness of its components. The purified wate. will return to the top of the PCRV cavity by means of a distribution header designed to minimizo local velocities and turbulence to maintain underwater visibility.
The equipment will be appropriato for the radioactive nature of the process fluid. Equipment that can generato a high radiation field, such as filters and domineralizers, will be shielded and provided with remote handling capability. Equipment fluid drains and leakoffs will be collected, treated and disposed of as discussed.
The PCRV water cleanup / clarification system will dischargo to the outlet piping of the Reactor Building sump pumps, during food and bleed operations. Ef fluents from the system will be diluted by the dilution flow from the circulatlog water makeup pumps to the main cooling tower blowdown line prior to release to the surrounding surface waters. The rate of offluent release from the PCRV, along with the dilution flow, will be controlled to assure that the radionuclido concentrations following dilution do not exceed the limits specified in 10 CFR 20. The same methods for controlling release of radioactive liquids have boon, and will continue to be, used for releases from the Reactor Buildin0 sump and the l
I 4-8
Suppl:m:nt to Environmental R port FSV Decommissioning Section 4 radioactivo liquid waste syntom.
The effluent release flow rate from 'ho PCRV is adjustable, with a controlled maximum flow rate of approximately 100 gpm. Likewise, the dilution flow rato from the circulating water makeup pumps to the main cooling tower blowdown lino is adjustable, with a typical dilution flow rato of approximately 1500 to 2000 gpm. Those flow rates are greator than those discussed in the historical FSV Final Environmental Statomont, but flow will be controlled to ensure that discharge concentrations do not excood 10CFR20 limits.
Since the PCRV water cleanup / clarification system dischargo outlet will connect to the outlet piping of the Reactor Building sump pumps, the automatic monitoring and protection features which currently govern releases from the Reactor Building sump and radioactivo liquid wasto system will also govern releases from the PCRV (Figuro 4.51). These features include two redundant activity monitors in the radioactivo liquid wasto dischargo lino, which monitor gross gamma activity in the lino, although those monitors will not be useful for detecting tritium concentrations. The signals from those activity monitors are arranged in one-out of two logic, so that if olther monitor detects high concentrations of gross gamma activity, the monitor will alarm and automatically close the block valvos in the radioactivo liquid wasto dischargo lino. The block valves are also interlocked with main cooling tower blowdown flow and will automatically close when the flow drops below a prodotormined setpoint.
The PCRV water cleanup / clarification system discharge flow rato, low dilution water flow switch setting, and the sotting of the radioactivity monitors will be established in accordance with the Offsite Doso Calculation Manual (ODCM), prior to radioactive liquid release to assure the limits of 10CFR20 are not excooded.
Prior to release and periodically during food and blood releasos, representativo samplos obtained from the PCRV cleanup / clarification system will be analyzed for gross alpha and beta activity, principal gamma emitters (including dissolved and entrained gases), tritium and other radioisotopos of concern to determine the dilution factor required to assure that the concentration in the cooling tower blowdown flow will not exceed the values specified in 10CFR20. Once the liquid wasto release rato, blowdown flow, and activity monitors have boon set, the liquid waste offluent will be released at a controlled and monitored flow rate to the cooling tower blowdown line for dilution and release to the environment.
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Supplomont to Environmental Report FSV Decommissionino Section 4 4.2.2 Tritium Roloasos The amount of tritium anticipated to be released to the environment during decommissionin0i s approximately 535 Cl. This liquid release will occur over a period of several months. Approximately 150 million Daltons of water are expected to be used to provide tritium dilution prior to dischar00. The sourco for this water is unticated water from the circulating water makeup storago ponds, pumped into the main cooling tower blowdown line by one or more circulating water makeup pump.
This is shown schematically in FIOure 4.51. The total amount of tritium calculated to be present in the PCRV as a result of reactor operations (using upper limit concentrations for impurities in the Draphlto) is approximately 1 0 0 ,0 0 0 C l. However, it is expected that the actual amount of tritium will bo sl 0nificantly loss than this. Based on data on measured tritium releaso ratos from Oraphlto and assuming the conservativo estimate of 100,000 Cl, lovels of tritium in the water system were estimated as a function of time for various pur00 ratos from the system. The tritium will be removed by a food and blood operation, using the PCRV shived water filtrat on and purification system, and released to 8
the surrounding surface waters following dilution to ensuro concentrations are below 10CFR20 limits. The maximum concentration in the PCRV water system is expected to be about 0.3 microcuries/cc, or a total of 300 Cl. After pur0 nD i for 40 days, the levelis expected to drop to below 0.1 microcurlo/cc and the lovelis expected to continue to fall thoroafter. The integrated total tritium released from the graphite blocks is predicted to be approximately 535 Cl. A plot of total tritium predicted to be in the PCRV water cleanup / clarification system vs. timo is shown in Figure 4.21.
4.2.3 Complianco with 10CFR20 in order to show compliance with 10CFR20 (Referenco 2), the concentrations of radionuclidos in liquid offluents were dotormined and compared with the of fluent concentration limits defined in 10CFR20. The guidelines provided in the NRC approved Fort St. Vrain Offsito Doso Calculation Manual (ODCM) woro followed. It is planned to release tritium from the dischar00 line at a rato such that followin0 dilution, tritium concentrations will be below the 10CFR20 limits.
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Supplomont to Environmental Report FSV DecommissioninD Section 4 The PCRV water cleanup / clarification system will be used to remove other radionuclidos, specifically Co 60 and Cs 137 associated with the PCRV cavity water No s1 0nificant activity from those radionuclidos is exp . sd.
Furthermore the domineralizers in the PCRV water cleanup /clariticatlun systern will reduco Co 60 and Cs 137 concentrations to non-dotectable lo w 's All liquid wasto from Fort St. Vrain decommissioning will be routed throu0 han existinD radioactivo liquid wasto dischargo lino. Simultaneous releases from the Liquid Westo System (System 62) and the Roactor Building sump (System 72) are not possible sinco a throo way ball valvo is installed at the connection botwoon the Reactor Buildin0 sump dischar00 lino and the radioactivo liquid wasto dischar00 line. All roloasos are monitored by the Liquid Waste Radioactivity monitors (RT 6212 and RT-6213).
4.2.4 Compliance with 10CFR50 The dosos to an individual at the liquid dischar00 point have boon calculated for tritium. No potable water pathway exists at Fort St. Vrain.
Gascous radionuclido releases from normal decommissioning operations have boon datormined to bo insignificant (close to background or non-detectable lovel.) However the reactor building ventilation system with its HEPA filtration will be used durin0 dismantlin0 of the PCRV. For those reasons no 10CFR50 Appendix l doso calculations for Oaseous tritium, particulates, and lodino will be performed.
4.2.4.1 Methodology The methods used to calculate the dosos from a liquid tritium discharge are based on Regulatory Guido 1.109 (Reference 3). These methods datormino dosos due to in00stion of fish located downstream of the offluent release point, as documented in the Fort St. Vrain Offsite Doso Calculation Manual (ODCM) (Reference 4).
i 1
4-11
Supplement to Environmental Report FSV Decommissioning Section 4 4.2.4.2 Dose Results The offsite doses to a maximally exposed adult from projected liquid tritium discharges resulting from decommissioning, as displayed in Figure ,
4.21, are: i l
Dose (mrom)
,,_Qo s e 1st Otr. 2nd. 3rd 4th Otr (each) Total Totrl Body 4.19 E(-3) 9.87E ( 4) 7.16E ( 3)
Organ
- Liver 4.19 E( 3) 9.87 E( 4) 7.16 E( 3)
- Thyroid 4.19 E(-3) 9.87 E(-4) 7.16 E( 3)
- Kidney 4.19 E(-3) 9.87 E(-4) 7.16 E( 3)
- Lung 4.19 E( 3) 9.87 E(-4) 7.16 E( 3) )
- GI LLI 4.19 E(-3) 9.87 E( 4) 7.16 E( 3)
The organ doses are all the same because none of the organs has a disproportionate affinity for tritium (water), All doses are well within the 10CFR20 limits of less than 3 mrom total booy and less than 10 mrom for any organ (calendar year) and less than 1.5 mrom total body and 5 mrom for any organ (quarter).
Note these doses represent the total dosc projections for the feed and bleed liquid tritium discharges. The feed and blood operations for tritium control are planned over one calendar year. This exposure is a one time dose and will not cause an impact during the entire decommissioning project life. The graphite blocks within the PCRV are the major tritium sources. After they have been removed and the feed and blood operation has been completed, it is expected that any additional tritium releases will be negligible.
l The above dose calculations are based upon the assumption that 535 Ci of tritium leaches out of the graphite blocks that were exposed to a neutron flux over the life of the core, which is considered to be i conservative. The feed and bleed operation can effectively process larger I
inventories of tritium in the PCRV shield water, should more than 535 Ci of tritium leach out of graphite blocks, if it were assumed that tritium was
!. released in the effluent at one-half the Maximum Permissible l Concentration of 10 CFR 20 (1/2 X 3E-3 Cl/ml = 1.5 E-3 Ci/ml) for six months, a total body dose of 1,48 mrem would result, using the dose 4-12
Supplement to Environmental Hoport FSV Decommissioning Section 4 ;
calculation methods described above. This dose, and the organ dosos, are within the above specified limits. Assuming a continuous effluent flow rate of 2,000 gpm, which is easily achievable based on the dilution water source,2,984 Ci of tritium would be released. This is more than a factor of fivo greater than the conservatively predicted inventory of tritium in the water. It is thus concluded that dosos resulting from liquid tritium releases will be below the applicablo limits, even if much more tritium loaches out of the graphlto than is projected based on available data.
In summary, the dose calculations indicate that the impact of projected liquid trit ium discharges f/om decommissioning activities is very small and in compliance with appicablo NRC discharge limitations.
4.3 Radiation Protection Program and Occupational Radiation Exposure 4.3.1 Introduction The primary objective of the Fort St. Vrain radiation protection program is to protect workers, visitors and the general public from unwarranted exposurc to radiation or to radioactive matorials during the decommissioning of Fort St. Vrain, in order to moet this objectivo, efforts will be made to reduce and maintain personnelexposures and radioactive offluents to levels that are as low as is rossonably achievable ( ALARA) and to ensure that all occupational radiation exposures are within the limits of 10CFR20.
Project management in conjunction with a radiation protection organization will be responsible for developing, maintaining and implementing a comprehensive program to ensure minimization of radiation exposure to personnel, the public and the environment. The guldance and recommendations contained in Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupation Radiation Exposures at Nuclear Power Stations will be As Low as is Reasonably Achievable", Revision 3, June 1978, have been considered in formulating the ALARA Program for the Fort St. Vrain Decommissioning Project. This program is outlined in more detail in Section 3.2 of the Propoced Decommissioning Plan.
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Supplement to Environmental Report FSV Decommissioning Section 4 The radiation protection program will be conducted in full compliance with all applicable NRC regulations, and will assure that all occupational radiatiot exposures are maintained within the limit., specified in 10CFR20 and are reduced to a level as low as reasonably achievable. A dedicated radiation protection rtaff will perform the following functions:
- 1. Assure all decommissioning personnel receive appropriate radiation protection training in exposure control and emergency procedures.
- 2. Assure that radiologletI survey maps of the work areas are developed and assure that surveys are performed to meet the requirements of 10CFR20.
- 3. Designate and segregate available work areas as olther contaminated or clean.
- 4. Assure that access controls are implomonted to prevent contamination of clean areas and to prevent unnecessary access to controlled areas.
- 5. Ensure that work areas are monitored and posted based on the evaluation of survey results and the requirements of 10CFR20,
- 6. Ensure that personnel exposure records are maintained.
To minimize potential worker exposure to tritium, portable ventilation will be utilized on the PCRV work platform to exhaust the local area above the flooded PCRV. This system will move the tritiated water vapor away from the work area. The need for other protective measures, such as " wet suits", will be evaluated based on the radiologicalconditions on the work platform.
4.3.2 Occupational Exposure Estimates Based on the tasks outlined in Section 2.3 of the Proposed Decommissioning Plan, estimates were made to determine the duration of each task in the radiation environment. Estimates were also made to determine the average radiation levels in the areas that each of these tasks will be performed. Estimates of the radiation levels were performed based on calculated activities for each activated component and on estimated plateout activities for contaminated components. These calculated radiation levels were then reduced in accordance with standard ALARA considerations that reflect the benefits of maintaining maximum distance 4 14
Supplement to Environrnental Report FSV Decommissioning Section 4 from the source and utilizing the shielding introduced by the water system or other local shielding.
Actual measurements of radiation levels at Individual work sites will be performed prior to commencing each individual task. These measurements may necessitate changes in work procedures. The design of the water shielding system for the PCRV dismantlement and decontamination activities provides some flexibility in radiation protection, through adjustment of the water level.
The projected occupational exposure for each major activity involving radiation exposure is given in Table 4.31. The detailed breakdown of each major activity to the WBS level and the associated projected exposures are provided in Table 4.3 2. The total cumulative occupational '
exposure for the entire decommissioning project is estimated to be 433 person-rem, due almost entirely to PCRV dismantlement and associated waste handling activities, as shown in Tables 4.31 and 4.3 2.
The general area background dose rate associated with the Balance-of-Plant (BOP) systems are expected to be less than 1.0 mr/hr. However,it is anticipated that Individual components such as filter housings, valves and piping that process fluids from the PCRV may have radiation levels that exceed this dose rate. Therefore, the exposures estimated for the individual systems are based on the removal of some components that exceed the 1.0 mr/hr and are expected to result in the exposures estimated for the Individual systems as listed.
The radiation environments for packaging and shipping radioactive waste are expected to vary over the course of decommissioning operations. The use of shielded transfer containers, the Hot Service facility, long handled tools and tag lines for handling radioactive materials and componoms will be among methods used to minimize worker . exposures. In addition, shielded shipping containers will be used for packaging radioactive waste that exceeds dose limits for shipping. It is expected that worker exposures recorded for the packaging and shipping of radioactive waste will result from working extended periods in low radiation backgrounds.
Consequently, personnel exposure estimates for BOP and radioactive waste processing are based on experience from similar nuclear industry projects rather than direct does calculations.
The methods for calculating the occupational radiation exposures consisted of, (1) utilizing the calculated dose rates for reactor internal components, (2) utilizing projected scheduled time for completing each task where the potential for radiation exposures exists and (3) utilizing 4-15
.. ._.m. _ _ _ _ _ _ . _ _ _ _ - __m _ -- . . _ ,
Supplement to Environmental Report FSV Decommissioning Section 4 engineering experience gained from similar projects in operating nuclear plants. The assumptions that were used to calculate the values shown in Tables 4.31 and 4.3 2 include:
- 1. For PCRV operntions only, the time workers spend in the work station radiation environment was assumed to be 50% of the scheduled time to complete the task. The total scheduled time was used in the Balance of Plant (BOP) and radwaste handling areas since radiation levels are expected to be low in comparison to the PCRV internals.
- 2. The " crew averaged" radiation levels were determined by assuming that the total exposures estimated for completing a task would be uniformly distributed among the crew.
- 3. The majority of graphlto reflector blocks will be removed without the use of shleided transport casks.
- 4. The highly activated components, such as boronated pins and large sido reflector blocks, will be loaded into shielded containers under water and transferred to the Hot Service Facility for processing.
- 5. The water levelin the PCRV will be maintained such that the general area dose rate on the work platform will typically be less than 2.0 mr/hr.
- 6. Personnel exposures for Health Physics (HP) and Quality Assurance (O A) coverage for PCRV decommissioning opt rations were assumed to be 10 percent (HP) and one percent (QA) of craft personnel exposures for a total of 11 percent. An estimate of 10 percent for HP coverage was assumed for BOP and radioactive waste packaging operations. Radiation
(. exposure for QA coverage of BOP and radioactive waste
- operations was assumed to be minimal.
l
- 7. The " crew averaged" dose rate for BOP systems decommissioning is expected to be less than 1.0 mr/hr resulting in minimal exposures. The exposures listed for each system are estimates based on the potential for worker exposure at work stations in proximity of the PCRV shield / pipe penetrations and contaminated components.
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Supplomont to Environmental R: port FSV Decommissioning Section 4
- 8. The estimate of exposures for redwaste handling were based on experienco galnod from packa0i ng and shippin0 radwasto in other operating nucloor plants. The dose ratos are expected to vary. Sources will be shlolded to acceptable lovels to moet shipp;ng requiromonts.
- 9. Tho workers will be trained in ALARA principals and good work practicos to minimize occupational exposures.
Other techniques such as mockup training for workors on potentially high exposure tasks, use of video and remoto monitoring equipment and onDi ncored controls will also be applied as appropriate to reduco personnel exposure.
Furthormoro,in order to comploto each individual task and koop personnel exposures to a minimum, the following work practicos will bo implomonted:
- 1. Pro-job briefings will bo hold with craf t and radiation protection personnelto assure that ALARA practicos have boon adequately factored into the work packa00s for completing tasks.
- 2. Personnel exposures will be monitored on a regular basis for potentially high exposure tasks to identify any irregularities that may indicato excessivo personnel exposures. In the event that an irre0ularity is found it will be investigated immediately and correctivo actions implomonted.
- 3. Tag lines will be attached and used when ri O0l nD and lifting hl0h exposuro rato components (e.g., steam generators) from the PCRV to koop workers as far from the source as possible.
- 4. Only essential personnel will be allowed insido the Reactor Building when bl0h exposure rato components are being removed. Casual observers will not be permitted.
- 5. Bag 0l ng techniques will be designed for quick installation.
G. Lung handled wipo tools wili be used when appropriate to wipo
-down wot components removed from the PCRV.
- 7. Shadow sh olds (lead blanket curtains or equivalent) will be used, where appropriato, to reduce radiation fields at the work stations.
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Supplomont to Environmentol Report l FSV Decommissioning Section 4 l
t
}
4.3.2.1 Estimated Radiation Exposure to Conversion Workers t The radiation exposure to converblon workors due to decommissioning ,
activities will be very small. Those workers, while still within the 1 rostricted area, will be outsido the area controlled for occupational radiation protection purposes. Docause of this, no single worker will recolvo more than 100 mrom por year, the maximum exposure permitted under the now revisions to 10 CFR 20. This doso represents a worker standing at the controlled area fence for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> in an averago dose into of 0.050 mrom por hour, in addition, radiation dose rates and concentrations at tho boundary established for controlof decommissioning activities will be such that the average exposure to all conversions workers will not exceed 10 mrom por year from all sources.
The FSV conversion work is expected to involvo a total of about 500,000 person hours over a period of 30 months. Considoring that the ambiont background radiation in the vicinity of the Fort St. Vrain sito is about 0.016 mrom por hour, the conversion personnel will recolvo an approximate total doso from background radiation of olght person rom for the period. They may recolve an additional projected 2.5 person rom from incidental radiation due to decommissioning activities, based on the averago 10 mrom por year.
Because thoro are no known radiation sourcos in the Turbino Building (other than controlled calibration sources used by Radiation Protection 1 personnel, which will be removed prior to convorsion activities), dose rates in the Turbino Building will romalr. essentially the somo as at present.
Measured dose ratos are about 0.008 to 0.010 mrom por hour, based on data provided in the Proposed Decommissioning Plan, Figuros 3.13 through 3.17. Individual workers in those areas will recolve from 15 to 25 mrom por year while working in those areas, somewhat loss than the average ambient background exposure.
If 50 workers are involved in conversion work in the turblno building for .
eight hours por day for two years, then the total doso for conversion work will be about two person-rom for the two year period. This doso is slightly loss than the throo person rom doso those workers would otherwise receive from background radiation for the same period.
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Supplomaat to Er vironnental Report FSV Decommissionlog Section 4 l 4.3.3 Contamination Control Contamination controlis required to provent the apread of contamination and in order to minimizo the number and volume of contaminated areas, tools, and components within the project, and to minimizo the potential for creating mixed wastes. Major elements of this program include area survoillances, containment, identification and control of hot particles, decontamination of areas, tools and equipment, rcuse of contaminated tools within radiologically controlled arons, leak detection and repair, and the uso of performanco indicators in trending data to measure program offectiveness.
Examplos of contamination contml motheds that have boon incorporated into the decommissioning plan includo:
- 1. The PCRV will be filled with water to control radioactive particulates that would normally be released when handled in air.
- 2. Containment or enclosures of appropriato sito, equipped with HEPA ventilation, will be used as necessary to provent the spread of contamination wliilo contaminated graphlto blocks and other components are being removed from the PCRV or otherwiso handled.
- 3. A work platform will be installed on the PCRV af ter the PCRV head has boon removed. The platform will be equipped with a HEPA-filtered ventilation system that will exhaust air from bonaath the work platform. This altflow will minimizo the spread of contamination.
- 4. A dobris collection system will be used in concreto cutting operations to minimizo the spread of contaminatlon.
- 5. Strippablo paint or other suitable onc'osures will be applied to some radiologically clean components or areas to provent cross-contamination.
t l
Additional cetamination control methods will be considered during job l
planning and work packa00 review. Isolation containments may be used if the surrounding work area is uncontaminated or is much cicancr than the work area to minimize the spread of contamination.
I 1
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Supplomont to Environmental Report FSV Decommissioning .
Section 4 Radiological surveys will routinely be conducted to identify and measuro contamination lovels. Those data will be used to identify if additional controls are warranted.
4.3.4 Respiratory Protection Program 4.3.4.1 Program Description An NRC approved Respiratory Protection Program will be used during the decommissioning project.
The PSC Radiation Protection Manager is designated to be responsible for directing all aspects of the respiratory protection program, it is the Intent of project management to minimiro personnel exposure to air contaminated with radioactivo dusts, mists, fumos, gases, or vapors.
The primary means to achlove this goal will bo to prevent or mitigato the hazardous condition at the source. Every reasonablo offort will be modo to achlove this objective by the use of engincorod controls, including process modification, containment and ventilation techniques. Rospirator protection will be specified in writing for worker protection against potential airborno contamination at the work stations, i ,
i 4.3.4.2 Radioactive Materials Control Radioactivo materials will be controlled through the uso of inventory and accountability procedures. The radiation protection organization will be responsible for surveillance, posting and access control to all radioactivo materials areas, as well as inventory and accountability of radioactive calibration and reference standards.
4.3.4.3 Radiation Protection Facilities and Equipment Project management will provide facilities, equipment and instrumentation adequato to permit the staff to function officiently. The facilities and I
decommissioning equipment to be provided will be adequato to moet the noods of the FSV project.
i 4-20
_ _ _ _ _ _ __ , _ . , _______ _ .~ .- _, . _ , ~ -
1 Supplomont to Environmental Report FSV Decommissioning Section 4 4.3.4.3.1 Faciiities The following facilities will be required by the radiation protection organization:
- 1. Sample counting facilities.
- 2. Whole body count room.
- 3. Dosimetry issue and reading area.
- 4. Facilities to clean, repair and decontaminato protectivo equipment, monitoring instruments, tools and other materials.
- 5. Change areas with provisions for segregation of contaminated from non-contaminated clothing.
- 6. Control station to control wasto movements and movement of equipment and instruments.
- 7. Communication equipment to facilitato communication with personnel in controlled work areas.
- 8. Calibration facilities for surveillance and monitoring instruments.
4.3.4.3.2 Assessment and Reporting 4.3.4.3.3 Identification and Correction of Problems A system will be established for identifying and reporting radiation protection problems and for instituting correctivo actions. The system will include provisions for: identification, reporting, investigation, ovaluation, Identification of root cause, and implomonting correctivo actions.
Problems and resolutions will be documented. Supervisory involvement in resolution and correctivo actions will be required as appropriato.
Procedures will be developed which implement these provisions.
4.3.4.3.4 Program Reviews A comprehensive system of planned and porlodic reviews will be implemented to verify compliance with all aspects of the radiation protection program.
At least once por calendar year, a technical review of the radiation protection program will be conducted by technically qualified personnel who are independent of the project radiation protection program.
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Suppl::mont to Environmentel Report FSV Decommissioning Sectinn 4 4.3.4.3.5 Porformanco Indicators Based upon nuclear industry practico and the recommendations in NRC and INPO guidanco documents, sovetal paramotors have boon selected to old in monitoring radiation protection performanco goals. Goals for performanco indicators will be recommended by the project staff and approved by the ALARA committoo. Key overallIndicators willincludo:
- 1. Annual collectivo doso.
- 2. Percent of project radiolo0 1cally controlled area entries in respirators.
- 3. Rato of personnel clothing and skin contamination events.
- 4. Percent of jobs for which the measured collectivo dose varios from the pro Job estimato more than 25%.
- 5. Number of radioloolcalincident reports.
4.4 Amblent Air Quality 4.4.1 Fugitive Dust During various demolition and dismantling operations, some fugitivo dust will be generated it is planned that the nonradioactivo portions of the PCRV tophead concreto will be cut usin0 diamond wiro techniques. This technique usos water as a lubricant which minimizes the generation of fu0l tivo dust. Any dust that is gonorated from this operation or similar oporations involvin0 concrete will be filtered by the existinD reactor building ventilation system. This system consists of throo trains, one of which is normally in continuous operation. This system contains exhaust filters composed of banks of moisture separators, HEPA filters and charcoal absorbors. Each bank contains 16 Individual HEPA clomonts, in addition, portable air handling units with HEPA filters will be used to ventilato localized work areas, such as when cuttin0 or sectionin0 operations are conducted in isolated areas. Finally, Donoral dust control will be maintained by the use of HEPA-filtered vacuum cloanors and the use of scabblers equipped with shrouds connected to HEPA filtered vacuums as nooded. Thorofore, it is not anticipated that fugitivo dust generated from decommissionin0 activities will have an adverso af f act on ambient air quality.
4-22
Supptoment to Environmental Report FSV Decommissioning Section 4 4.4.2 Exhaust Emissions ,
Exhaust omissions from diesel powered equipment and vehicles may have a slight Impact on air quality. It is estimated that approximately 350 truck shipments of radioactive wastes will be made. In addition, several hundred shipments of nonradioactive wastes will be made to locallandfills.
However, the total number of vehicles at the site has been significantly reduced from the levels required to suppoit continued plant operation for the duration of operating license.
The primary source of plant omissions during decommissioning activities will be the auxiliary boiler. While the plant was In operation, the auxiliary boilers, rated at 160,000 lbm/hr (total), (Reference 5) were rolled upon during shutdown conditions and start up operations to provide the steam motive power for the main feedpumps and hellum circulators. The auxillary boilers were also rolled upon to provide steam heating for plant buildings and the PCRV. However, during decommissioning, one auxillary boiler has been modified and dorated to maximum capacity of 15,000 lbm/hr and will only be used for building heating. Consequently, the exhaust emissions during decommissioning activities will be significantly less than those emitted during plant operation. Therefore, with respect to exhaust emissions, decommissioning will have a positive environmental impact.
4.4.3 Asbestos In the areas where it has been determined that asbestos must be removed or disturbed for decommissioning activities, removal of asbestos will be porformed within control ventilated areas equipped with exhaust filtration to minimize the release of asbestos. The public will be prohibited from entering work areas where asbestos removal and packing operations are being performed. The concentrations of asbestos fiber will be maintained within allowable levels. It is expected that no public exposure to asbestos will result from asbestos dismantling and packaging operations. All asbestos that is removed from nonradiologically controlled areas will be trucked to an approved landfill. All requirements of 29CFR Parts 1910 and 1926 concerning asbestos removal, handling and packaging will be met.
4.5 Effects of Chemical and Blocide Discharges The decommissioning process at Fort St. Vrain does not use chemical decontamination to any significant extent. The only other effluent of a chemical or biocido nature expected is water containing the normal 4-23 i
1 _. . - - , - - . , _ _ _ - ._ _ , . -, ._ -
Supplement to Environmontcl R:rport FSV Decommissioning Section 4 amount of cleaning fluids (dotorgent) that inicht be usert to decontaminato walls and floors in the facility, During plant operation, waste water was produced by domineralizer regonoration, blowdown from the main cooling tower and blowdown from the service water cooling tower. However, during decommissioning activitios the primary source of dischargo, the main cooling tower, will bo out of service. Domineralizer regonerat!ons as well as wasto water discharged from the service water tower will be greatly reduced. Itis therefore concluded that there will be greatly reduced consumption of chemical / blot,ldos and resulting offluents during decommissioning. Waste water produced during decommissioning activities wil continuo to bo discharged in accordance with the sito NPDES permit. Figure 4.51 schematically shows the origins and pathways for various liquid effluents.
4.6 Effects of Sanitary Wasto Dbchargo Sanitary and sink drains are combined and trooted in an activLted sludge process which consists of a pro-aeration chamber, two aeration lagoons and a polishing pond. The final stage of treatment involves chlosination by the addition of calcium hypochlorito Ca (OCL) 2. The offluent from this pond (3 gpm, average) will be combined with the blowdown from the service water cooling tower and discharged from the Station (Figure 4.51) via the Goosequill Ditch to the South Platto River as it was dono during plant operation.
Sanitary wastes from approximately 320 PSC workers and contractors on a daily basis have been previously treated by the site sewago aoration system as described above. The number of people required on sito during .
decommissioning is expected to be a maximum of approximately 300 por '
day. With this decrease in personnel from that of the normal plant operations staff, the actual sanitary wastes to be treated and discharged will also be reduced. Thorofore there will be less environmental impact from sanitary waste discharges.
4.7 Endangorod Species Based on References 5 and 6, no significant impact on the flora and fauna of the site or its surrounding area has been experienced as a result of ,
construction or operation of FSV Accordingly, the impact on flora and fauna due to the decommissioning of FSV is expected to be negligible.
No natural habitat exists on land areas directly involved with decommissioning activities which harbors sensitive species. Noise and l
4-24
Supplement to Environmental Report FSV Decommissioning Section 4 other activity associated with decommissioning activities is not expected to have significant impects on wildlife species in the surrounding non-developed areas.
PSC has obtained two separate reports from the Colorado Department of '
Wildlife, Northeast Region (Reference 7). The area reported covers 3500 square miles and is bounded (approximately) on the north by the Colorado- -
Wyoming border, on the south by the city o' Brighton, on the east by the town of Goodrich and on the west by Colorado Interstate Highway 25.
Actual boundaries are south - Latitude A0 degrees; north Latitude 41 degroos; east - Longitude 104 degrees; and west Longitude 105 degrees.
For this entire area, the following is the listing of threatened or endangered wildlife including birds, mammals, reptiles, and amphibiana:
Bald Eagle Endongorod (Federal and Stato) i Peregrine Falcon Endangered (Federal and Stato)
Whooping Crane Endarigorod (Federal and State) in the aond listing provided to PSC, the Department of Wildlife prepared a seperate list which included all speclos which occur in habitat types similar to those in the Fort St. Vrain area. The list is as follows:
Bald Eagle Endangered (Federal and Stato)
Peregrine Falcon Endangorod (Federal and Stato)
The Department of Interior, Fish and Wildlife Service prepared a list of Faderal threatened and endangered speclos which may occur within the area of influence of Fort St. Vrain and the site to be decommissioned, The list is as follows: '
Bdd Eagle Bl,'ck footed Ferret References 5 ar.d 6 were prepared for and approved by the Atomic Eneigy Commission in 1972. Both the bald eagio and the peregrine falcon were identified in those documents as ondangerod. The Atomic Energy Commission granted PSC a licence to construct and operate FSV based on their findings. The whooping crano does not occur in habitat types similar to thost. In the FSV area. Regarding the black-footed forret, the Fish and Wildlife Service states "the standard that is used by the Service for determining possible project effects to black-footed forrets it; the disturbance of cerrontly occupied prairio dog habitat." Thore is no such habitat on the Fort St. Vrain site.
l 4-25
_-..,.__m. . . . . - , - . , . - , _ . _ . - , , . . . _ , . .
l Supplement to Environmentel Hoport FSV Decommissioning Section 4 4.8 Other Effects 4.8.1 Nolso There will be very litslo noise impacts as a result of the proposed dismantling project. In general, demolition activities will normally be conducted by one shift of up to 300 workers during the day. The exception to one shift operation is the diamond wiro cutting of PCRV concrote which will be on a throo shift basis. This cutting method is quiet and is not expected to be heard offsito. Thore will be no significant noise impacts during the evening hours.
The areas in the vicinity of the plant sito are sparsely populated, ".,nd thu nearest population contor, Plattevillo, is epproximately 3 1/2 miles Southeast of the sito.
4.8.2 Effects of Runoff from Decommissioning Durir a the decommissioning activities there are no additional buildings constructed and no extra parking nooded for workers or equipment, Decontamination work is planned to take placo Indoors so -as to limit the amount of dust roloosed to the environment (see Section 4.4.1). The precipitation runoff resulting from decommitstoning is not expected to be greater than the runoff from Fort St. Vrain during normal plant operations. (
4.8.3 Socioeconomic Impacts Decommissioning is expected to be completed over a 57 month period and to employ a peak work force of approximetely 300 (this includes PSC management, support staff and contractor personnel, but does not includo any personnel involved with potential conversion activities). The temporary nature of the project and the limited work force support the conclusion that no significant demographic shif ts or significant effects on the regional economy will result from the decommissioning.
The decommissioning work force will include administrative and clerical employees, technicians and engineers, security personnoland construction workers, including laborers, craf tsmen, equipment operators, etc.
PSC plans to obtain all of its clerical, technicai, and engineering staff from 4 the existing work force. Most of the contractor technical and support
-personnel wil' be relocated to the Fort St. Vrain area. It is expected that most of the dec,ommissioning construction workers will be drawn from the Denver regionallabor pools.
4-26 )
1
-Supplement to Environmental Report FSV Decommissioning Section 4 Decommissioning is not expected to have a significant impact on regional or local employment and unemployment rates, whether the work force is drawn from the Denver and Front Range population or relocated to the area. Due to the specialized nature of the task, and the smalllabor force required, no significant impacts c , local or regional labor markets or demand for services are anticipatoJ The expenditure of greater than $137 million will contribute to the economy of the project. Assuming that 20 percent of the project cost will be spent locally, and using a total impact multiplier of 1.3, the project's total impact on the local economy will be in excess of $35 million.
4.9 References l
- 1. " Evaporation Atlas for the 48 Contiguous United States," NOAA 1 lechnical Report NWS-33, Department of Commerce,1992.
- 2. Appendix B to 10 CFR Part 20, Sections 20.1001 to 20.2401, Table 2 (as revised in May,1991)
- 3. Regulatory Guide 1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Colaplience with 10CFR50, Appendix 1, October 1977.
- 4. Fort St. Vrain Offsite Dose Calculation Manual (ODCM) Public Service Company of Colorado, August 1990.
- 5. c!nv 6 *onmentcl Statement related to operation of the Fort St.
V' ' ' - N - lest Generating Station of Public Service Company of C JorMc, Docket No. 50-267, August 1972.
- 6. Public Service Company of Colorado, " Fort St. Vrain Nuclear Generating Station, Applicant's Environmentel Report Operating License Stags," dated December 1970.
- 7. United States Department of the Interior, Fish and Wildlife Service,
" Endangered Species List - Fort St. Vrain Power Station," dated November 21,1990.
4-27
Supplement to Environmental Report FSV Decommissioning Section 4-TABLE 4.1 1 SOURCES OF PUBLIC WATER SUPPLY pownstream in the Platte River Valley Distance Pop;tation Typ of Miles M ricipal(ty Served M Snu+ce of Sply 5 10 Gilcrest 382 Wetts tuwitsn of South Platte River 10 15 LaSalle 1,300 Wells Alluvi m of South Platte River 10 20 Greeley 39,000 Surface Cache la Poudre River; Colorado Big ,
Thorpson Project; NJY) and Deadron Creeks Other Municipalities 34 Plattevitte 950 Wetts Attuvium 5 10 Mead 900 Surface Big Thonoson River and St. Vrain Creek 10 15 Johnstown 1,200 surface Big Thorrpson kivsc
, 10 15 Fort LLoton 4,000 Wetts Alluvium of South Platte River 10 15 Frederick 1,500 Surface Boulder Creek
'[ 10 15 Longsmt 50,000 Surface horth and South St. Vrain Creek ib ' 10 20 Loveland 30,000 Surface 81g Thonpson River
.-k' 10 15 Berthout 3,200 Surface Big 'honpson River f- 15 20 Hudson 540 Wetts Attuvi m f: = 15 20 Brighton 13,000 Wetts At twim af Suth Platte River f 15 20 Erie 1,3 75 Surface Sexath Boulcer Creek ,
15 20 Windsor 1,500 Surface Greeley 20 25 Eaton 1,500 Wetts Altwium 20 25 Keenesburg 475 Wetts Laramie and Fox Hitts formations 20 25 Broomfield 20,000 Wetts Fox Hills sandstone -
20-25 Lafayette 10,000 Surface South Boulder Creek, Waneka Reservoir
?" 45 Louisville 6,000 Surface Saath soulder Creek 40 25 Lycs* 1,340 Surface horth St. Vrain Creek 20 25 Timath 150 Surface creeley m 20 30 Fort Col 1 ins 80,000 Surface Cache ta Poudre R tver 25 3^ South Adams Water 25,000 Wells Altuvi s of South Platte River and Sanitary Olst.
(Camerce City) 25 30 Lochbuie 1,000 Wells Attuvi m of South Platte River 25-30 horth Hi;ron Weter 80 Wells F ox Hi t t a sandstorw Dist- (near Broonrf ield) 25 30 Northwest utilities 15,000 Wells Arapahoe and Fon Hills formations; Conpany attuvi m of South Platte River 25 30 Federal Heights 8,000 Wetts Arapahoe and Fox Hills formations 25 30 Westminster 60,000 Surface Clear Creek Wells Arapahoe ard Fcx Hills formations 25 30 Bautder 96,000 Surface horth Boulder Creek 25 30 Jarrest own 230 Ground Alluvi m 4-28 l
l
Supplomont to Environmental Report FSV Decommissionin0 Section 4 TV\BLE 4.12 FORT ST. VRAIN WIND SPEED AND DURATION PERIOD OF RECORD: 1986 1989 STABILITY CLASS: All Classes Wird Speed (mph) at 10m level Wind Direction 1-3 47 8 - 12 13 - 18 19 - 24 >24 TOTAL N 545.13 579.12 355,24 272.14 76.43 21.50 1849.56 NNE 738.41 729,08 420.54 217.72 72.83 25.22 2202.80 i NE 803.04 964.27 353.48 101.27 19.11 3.53 2244.70 ENE 820.38 1051.09 303.93 38.20 4.26 1.26 2219.12 i
E 597.95 845.41 227.74 27.21 2.77 0.76 1701.84 ESE 570.52 748.32 256.38 41.60 4.52 1.51 1622.85 l SE 526.77 584.33 231.54 61.27 6.04 2.77 1412.72 ,
SSE 637.06 666.42 265.01 68.02 23.41 9.85 1669.77 S 872.38 805.30 228.23 56.06 19.41 7.31 1988.69 SSW 1072.95 937.43 120.03 23.18 2.92 2.36 2158.87 SW 1204.10 1537.78 157.65 24.11 5.54 2.27 2931.45 i WSW 867.01 1113.02 166.28 62.69 11.57 6.03 2226.60 W 369.11 263.26 75.50 50.46 26.98 10.84 796,15 WNW 205.06 169.15 84,86 90.29 29.78 20.91 600.05 NW 278.20 299.73 160.76 87.95 29.96 8.83 865.43 NNW 388.84 380.16 221 87 129.06 36.58 4.28 1160.79 VARIABLE 0.00 0.00 0.00 0.00 0.00 0.00 0.00 '
Total 10496.91 11673.87 3629.04 1351.23 371.11 129.23 27651.39 Periods of calm (hours): 1241.77 c Hours of missing data: 1728.56 l
I 4-29 l
Supplomont to Environmental Report FSV Decommissioning Section 4 T/LBLE 4.1-3 FORT ST. VRAIN WIND FREQUENCY DISTRIBUTION PERIOD OF RECORD: 1986 1989 STABILITY CLASS: All Classes Wind Speed (mph) at 10m level Wind Direction 1 3 4-7 8 - 12 13 - 18 19 - 24 >24 TOTAL N 0.020 0.021 0.013 0.010 0.003 0.001 0.067 NNE 0.027 0.026 0.015 0.008 0.003 0.001 0.080 NE 0.029 0.035 0.013 0.004 0.001 0.000 0.081 ENE 0.030 0.038 0.011 0.001 0.000 0.000 0.080 E 0.022 0.031 0.008 0.001 0.000 0.000 0.062 ESE 0.021 0.027 0.009 0.002 0.000 0.000 0.059 SE 0.019 0.021 0.008 0.002 0.000 0.000 0.051 SSE 0.023 0.024 0.010 0.002 0.001 0.000 0.060 S 0.032 0.029 0.008 0.002 0.001 0.000 0.072 SSW 0.039 0.034 0.004 0.001 0.000 0.000 0.078 SW 0.044 0.056 0.006 0.001 0.000 0.000 0.106 WSW 0.031 0.040 0.006 0.002 0.000 0.000 0.081 W 0.013 0.010 0.003 0.002 0.001 0.000 0.029 WNW 0.007 0.006 0.003 0.003 0.003 0.001 0.022 NW 0.010 0.011 0.006 0.003 0.001 0.000 0.031 NNW 0.014 0.014 0.008 0.005 0.001 0.000 0.042 VARIABLE 0.000 0.000 0.000 . 0.000 0.000 0.000 0.002 Total 0.380 0.422 0.131 0.049 0.013 0.005 1.000 Periods of calm fraction: 0.045 Fraction of missing data: 0.063 4 30
l Supplement to Environmental Report j FSV Decommissionin0 Section 4 TABLE 4.3-1
SUMMARY
OF OCCUPATIONAL RADIATION EXPOSURE ESTIMATES ,
wuRM2 #TMrA&R (f)
RA rt Att1EG L%PtMURE kVORKACTib77Y crus nas> <ron -aux >
2.3 PCRV DISMANTLING AND DECONTAMINATION 2.3.1 PCRV INITIAL PREPARATION / DIS ASSEMBLY 23,733 7.37 2.3.2 SillELDED ACCESS TO PCRV 20,578 70.38 -
2.3.3 DISMANTLE PCRV CORE 49,368 157.34 2.3.4 CORE SUPPORT FLOOR, !!ARREl. AND INSULATION D/D 9,213 103.35 2.3.5 PCRV LOWER PLENUM D/D 16,103 59.88 2.3.6 FIN AL PCRV DISM ANTLINO, DECONTAMIN ATION & CLEANUP 15,047 17.67 SUHTOTAL (PCRV) . 134,042 306.00 2.4 DOP CONTAMIN ATED SYSTEMS DISMANTLINO/ WASTE PACKAGINO 1,2,1,1 R ADIOLOGICAL CII ARACTERIZ ATION 7,279 0 25 2.4.2 DISMANTLING OPERATION 58,684 1,43 SURTOTAL (BOP) 65,963 1.68 2.6 RAD WASTE PROCESSINO A SillPPING 33,055 65.37 GRAND TOTAL PCVR,BOr', PACK AGINO A S!!!PPINO 233,060 433.06 KEY:
(1) Exposure work time (worker efficicney) is estimated to bc 50% of achcdaled work time for PCRV tasks where the potential for radiation exposure exists.
1 4-31
Supplement to Environmental Report FSV Decommissioning Section 4 TABLE 4.3 2 Occupational Radiation Exposure Estimates KEY:
(1) Exposure work time (worker efficiency) is estimated to be 50% of scheduled work time for PCRV tasks uhere the potential fc,< radiation exposure exists. ner oa n v.wK WBS WORKBREAIDOWN STRUCTURE saapuun won e cwvaa wc m .sto NUMBER woux nus arowns knurum uPasens nue runs (913 HR5) (Pt2 ilk.t) (MR11R) (Pt.R RM) 2.3 PCRV DISMANTLING AND DECONTAMINATION l
2.3.1 PCRV INITIAL PREPARATION / DISASSEMBLY 2.3.1.5 Modify Main Crane 682 341 0.1 0.03 l
2.3.1.7 Ddension PCRV Tendons 25.590 12.795 0.1 1.28 2.3.1.8 Remove Core Elements with Fuel llandling Machine 2.3.1.8.1 PCRV Region Constraint Devicca 2.208 1,104 1.7 1.88 2.3.1.8.2 Remove Metal Clad & Control Rod Reflector Bhwks 12.267 6,134 0.4 2.45 2.3.1.9 lie Purincation Cornponent Wells 2.013 1.007 1.0 1.00 i 2.3.1 SUBTOTAL 21.381 6.64 l
2.3.1 HP & OA COVETlAGE(11 %) 2.352 0.73 2.3.1 TOTAL 23.733 7.37 2.3.2 SillELDED ACCESS TO PCRV l 2.3.2.3 Seal PCRV C<xding Tubes & Tendon Conduita 4.110 2.055 1.0 2.06 l 2.3.2.4 Center Access Penetration 1.190 595 1.1
' 0.65 2.3.2.5 PCRV Shicided Water System 4.980 2.490 1.0 2.49 2.3.2.6 Airboruc Contamination Control 3.633 1.816 0.3 0.54 2.3.2.7 Cut Core Top 11 cad 21.660 10.830 1.1 11.91 2.3.2.8 Flood PCRV 180 90 0.6 0.05 2.3.2.9 PCRV Cavity.Shicided Work Platform 1.325 663 1.0 0.66 2.3.2 SUBTOTAL 18.539 18.36 2.3.2 HP & OA COVERAGE (11 %) 2.039 2.02 2.3.2 -- TOTAL 20.578 20.38 2.3.3 DISMANTLE PCRV CORE 2.3.3.3 Defueling Eleme,ts 16.683 8.342 1.7 14.18 2.3.3.4 Replacement and Perm flex Reflector Blocks 16.683 8.341 3.7 30.86 2.3.3.5 Large Side Reflector Blocks 27.782 13.891 3.6 50.00 2.3.3.6 Baronated Spacer Elements 16.683 8.342 1.9 15.85 2.3.3.7 Ilastalloy Can lica Reflector Bhwks 8,341 4.170 6.8 28.36 2.3.3.8 Core Support Bhwks and Ponta 2.780 1.390 1.8 2.50 2.3.3 SUBTOTAL 44.476 141.75 2.3.3 HP& OA COVERAGE (11 %) 4.892 15.59 2.3.3 - TOTAL - 49.3G8 157.34 4-32
{
SupplemOnt to Environm0ntal R0pcrt FSV D0 commissioning Section 4 TABLE 4.3 2 Occupational Radiation Exposure Estimates KEY:
(1) Exposure work time (worker efficiency) is estimated to be 50% of scheduled work time for PCRV tasks where the potential for radiation exposure crists.
WBS WORKBREAKDOWN S7RUO7URE sau.1nuv wonxu cnwm woam to)
NUMBER woex rius exmsvas arzww sxourvas r1Ms F1121k1 (90 itks) (PG IIRS) (AIMIR) (P0 kDt) 2.3.4 CORE SUPPORT FLOOR, BARREL AND INSULATION D'D 2.3.4.3 Core Barrel and Keys 2.3.4.3.1 Core Barrel and 24 Outer Keys 4.913 2,456 9.2 22.59 2.3.4.3.2 24 Core Barrel to Graphite Lower Key Rcrnoval 952 476 13.3 6.33 2.3.4.3.3 24 Core Barrel to Graphite Upper Key Removal 1.624 812 10.0 8.12 2.3.4.4 Core support Floor 7.762 3.881 12.5 48.51 2.3.4.5 Top CSF Insulation 1,350 675 11.2 7.56 2.3.4 SUBTOTAL 8,300 93.11 2.3.4 HP& OA COVERAGE (11 %) 913 10.24 2.3.4 TOTAL 9.213 103.35 2.3.5 PCRV LOWER PLENUM D/D 2.3.5.3 Twelve Stearn Generator Modules 2.3.5.3.1 secondary . Steam oenerator Modules 14,688 7,344 0.8 5.88 2.3.5.3.2 Primary - Steam Generator Modules 7,594 3,797 4.7 17.85 2.3.5.4 IIelium Circulators 2.3.5.4.1 Remove liclium Circulators Primary Modules 1,056 528 11.1 5.86 2.3.5.4.2 Remove licliurn Circulators Secondary Modules 1,144 572 3.5 2.00 2.3.5.5 Core Supp. Floor Columns, Lower Floor & Flex Col. 1,144 572 22.2 12.70 2.3.5.6 PCRV luside Top, Rot.& Side Insul.a Cover Plates 3.388 1,694 5.7 9.66 2.3.5 SUBTOTAL 14,507 53.95 2.3.5 HP& OA COVERAGE (11 %) 1,596 5.93 2.3.5 TOTAL 16,103 59.88 2.3.6 FIN AL PCRV DISMANTLING, DECONTAMINATION & dLEANUP 2.3.6.1 Remove Beltline Activated Concrete 17,284 8.G42 1.7 14.52 2.3.6.2 Decon Lower PCRV Lines 984 492 0.8 0.39 2.3.6.3 PCRV Wall and Liner Penetrations; PCRV 7,404 3,702 0.2 0.79 Safety Valve Instrur:.cntation & Piping 2.3.6.4 Dernobilize and Clean Up Area 1,440 72b 0.3 0.22 2.3.6.5 Decon PCRV for Final Relcase Survey 1,440 0 0.0 0.00 2.3.6 SUBTOTAL 13.556 15.92 2 2.3.6 HP& OA COVERAGE (11 %) 1,491 1.75 2.3.6 TOTAL 15,047 17.67 i GRAND TOTAL PCRV 134.042 366.00 )
4-33 !
s
Supplement to Environmental Report FSV Decommissioning Section 4 TABLE 4,3-2 Occupational Radiation Exposure Estimates WBS. WORKBREAKDOWN STRUCTURE saunta wonra cuwan wonwt NUMBER wca nus a u vas asmsrw sa m vas nus nsus _
(PLR -NRS) (F62 NAS) (MRSIR) (N2 -afM) 2.4 BOP CONTAMINATED SYSTEMS DISMANTLING / WASTE PACKAGING 1.2.1.1 RADIOLOGICAL Cil ARACTERIZ ATION 7.279 7.279 <1 0.25 2.4.2 DISMANTLING OPERATION 2.4.2.2 Sydern 13 Fuel llandling System 4.648 4.648 <1 0.22 2.4.2.3 Systern 14 Fuct Storsge Wells 3,742 3.742 <1 0.10 2.4.2.4 Systern 16 Ilot Service Facility Autiliary Transfer 4,477 4,477 <1 0.10 Cask & Equipment Storage Wells 2.4.2.5 Sygern 23 licliurn Purincation Systern 5.448 5.448 <1 0.23 2.4.2.6 System 46 Reactor Plant Cooling Water 1.500 1.500 <1 0.06 2.4.2.7 System 47 Purincation Couling water 250 250 <1 0.02 2.4.2.8 System 6! Decontamination Systern 3.493 3.493 <1 0.10 2.4.2.9 Systern 62 Liquid Waste Systern 10.610 10.610 <1 0.25 2.4.2.10 System 63 Gas Waste system 9,948 9.948 <1 0.13 2.4.2.11 Systern 72 Reactor Building Drain 4.577 4.577 <1 0.02 2.4.2.12 Systern 73 Reactor Plant Ventilation 1,694 1.694 <1 0.02 2.4.2.13 System 93 Controls & Instrumentation Piping, 1,370 1,370 <1 0.02 last*umentation, and Structure External to PCRV 2.4.2.14 Contaminated Laundry Facility & RW Comp. 930 930 <1 0.01 SUBTOTAL BOP ,
59,966 1.53 ;
BOP HP COVERAGE (10 %) 5,997 0.15 GRAND TOTAL BOP 65.963 1.68 2.6.2 RAD WASTE PROCESSING AND SHIPPING 30.050 30.050 VARIES 59.43 2.6.2 HP COVERAGE (10 %) 3.005 5.94 2.6.2 GRAND TOTAL PACKAGING & SHIPPING - 33.055 65.37 GRAND TOTAL PCVR, BOP, PACKAGING &
SHIPPING ' 233.060 433.06 4-34
i I
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Supplement to Environmental Report FSV Decommissioning Section 4' R.67 W.
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Supplement to Environmental Report FSV Decommissioning Section 4
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L
' Supplamant to Environmental Report FSV Decommissioning Section 5 m^
5.0' ENVIRONMENTAL IMPACTS OF ACCIDENTS
{ 5.1 Facility Accidents involving Radioactivity 5.1.1 Introduction The radiologicalimp t on the general public withiri 30 miles of Fort St.
- Vrain for six accident scenarios (Reference 1), were analyzed using the AIRDOSE EPA computer code (Reference 2).
The- AIRDOSE EPA computer' code is a methodology that estimates radionuclide concentrations in air; rates of deposition on ground surfaces; ground surface-concentrations; intake rates via inhalation of air and ingestion of meat, milk, and fresh vegetables; and radiation doses to man
-from airborne releases of radionuclides. These doses are presented in terms of one year committed effective dose equivalents.
-A modified Gaussian plume equation is used to estimate both horizontal and vertical dispersion of as many as 36 radionuclides released from one
. to six stacks or area sources.. Radionuclide concentrations in meat, milk
[ 1 and fresh produce consumed by man are estimated by coupling the output-
, - of the atmospheric transport models with the U.S. NRC, Regulatory Guide
- 1 1.109- (Reference 3) terrestrial food chain models. Dose conversion
- factors are input to the code, and dose to man at each distance and o di' 7 tion specified are estimated for red bone marrow, lungs, endosteal ,
- . - bone tissuei celts, breast, thyroid,' and gonads through the following exposure modes
j e immersion in air containing radionuclides, _
e exposure _ to ground surfaces contaminated by deposited
- radionuclides, - '
e' immersion in' contaminated wator, e inhalation of radionuclides in air, and
- e ingestion of feed produced in the area. }
Meteorology data used in the analysis was'obtained from the Fort St.
Vrain meteorological station. Population data was based on the 1980 census. Estimates of significant water areas were obtained from map data supplied by the United States Geological Survey (USGS).
51 1
j
f-Supplement to Environmental Roport FSV Decommissioning Section 5 -
The accident scenarios used are taken from Section 3.4 of the Proposed Decommissioning Plan.- The tritium, Eu 154, Co 60 and Fe 55 source terms are the same. However, the dose analysis to a maximally exposed individual has been performed with the AIRDOSE-EPA code.
The area shown in Figure 5.1-1 is a circular area of 7319.2 square kilometers (2826 square miles) with the Fort St. Vrain plant at the center.
The population within this square was estimated at 1,531,600. The sectors making up this area are 4828 m. (5280 yards) on a side. The ,
dose output from the AIRDOSE-EPA computer code is keyed to a radial distance from the plant as well as a compass direction for the radionuclides mentioned.
The: risk of ~ accidents resulting in a radiological release drring decommhsloning activities is considerably less than during plant operation, due to th.. removal of irradiated fuel from the Reactor Building. Since the reacto" e be defueled prior to the commencement of decommissioning
- operations and all fuel will be removed from the Reactor Building, only nor.-resctor accident scenarios are evaluated in this section. The focus of these decommissioning accident analyses will be on public health and safety.
. The . following postulated accident - scenarios have been analyzed,
-considering activation levels and isotopic composition of components to
- be processed, and the anticipated dismantling activities
.1. Dropping of contaminated concrete rubble 2.- Conversion construction near PCRV dismantlement 3.- Heavy load drop
- 4. Fire
- 5. Loss of PCRV shielding water
- 6. Loss of Power
- 7. Natural disasters The components with the highest activation levels were used in the accident analyses. Therefore, accidents that were analyzed bound the radiological consequences from other postulated accident scenarios, in l evaluating the postulated accidents, conservative assumptions were made when data or knowledge to support more realistic analyses were lacking.
- Conservatism in this context is defined to mean that the radiological
[ consequences from the postulated accidents will be overestimated rather than underestimated, 5-2
Supplomant to Environmantal Raport FSV Decommissioning Section 5 A capsule summary of the accident scenarios is given in Table 5.1-1. The
- ort term (0 2 hr) doses from these accidents are discussed in Chapter 3 u ;he Proposed Decommissioning Plan. The doses to a maximally exposed off site individual from the postulated accidents, are presented in Table 5.1-2. From this table, the limiting accident is a tornado induced missile resulting in a red bone marrow dose of 7.2 E( 5) mrem and a dose of 2.0 E(-4) mrem to the lungs. These doses are miniscule and well within the 25 rem whole body dose and 300 rem to any specific organ guidelines to a maximum individual established by 10CFR100.
These doses are also a small fraction of the 1 rom whole body dose and 5 rem to any specific organ dose guidelines cited in EPA Protective Ac"on Guidelines (Reference 4).
The lifetime fatal cancer risk to the rnaximum Individual from the postulated tornado-induced missile accident is 1.0 E(-4). This is insignificant when compared to the naturally occurring cancer rate in Colorado.
The following natural disasters were considered in the accident analyses:
Radiological External Event Mitigatino Feature Conseouences Earthquake Low Probability Not postulated; See of Occurrence Section 5.1.9 High Winds, Bounded by Tornado See analysis in Hall Section 5.1.9 Rainfall, Flood Site Location No release Range Fire Plant buffer No release The acVvity concentrations of the various components used in the following accident analyses were derived from the detailed neutron activation analysis (Reference 5). Where chemical impurities were involved.in neutron activation reactions, the maximum impurity levels permitted by the pertinent specifications were conservatively assumed to exist. With the exception of tritium concentrations, the radioisotope concentrations of interest used in the accident analyses have been taken directly from the activation analysis. Tritium concentrations predicted by 5-3
Suppl 3msnt to Environmsntal Report FSV Decommissioning
-Section 5 the activation analysis were considered extremely unrealistic for the following reasons:
- 1. In the activation analysis, the dominant source of tritium was from activation of lithium impurities. The activation analysis assumed that no tritium formed by lithium activation migrated out of the graphite into the primary coolant. The lithium concentration assumed to be present prior to irradiation in the graphite blocks was based on the maximum concentratlun permitted - by the specifications, in actuality, lithium is relatively volatile and tends to migrate out of the graphhe during the high temperature graphitization process. Therefore, it is considered probable that the lithium impurity concentrations in the graphite used to form the large side reflectors and side spacer blocks were an order of magnitude lower than the maximum specification limit.
- 2. Thc large graphite side reflectors ai.. ,
w blocks were exposed to relatively low temperatures (300 600 degrees C) during reactor operations. These low temperatures preclude a significant amount of tritium from being chemically absorbed in the graphite and retained. Since tritium has a small atomic radius, it is likely that tritium formed by activation of lithium (Li-6 and Li-7) will migrate out of the graphite. Due to this temperature dependence of chemical absorption, it is considered that tritium concentrations are two or three orders of magnitude below those predicted by the activation analysis.
3, in the presence of moisture, hydroger atoms from water nl.oleculos compete with and replace tritium atoms at active carbon sites in the graphite matrix, releasing tritium from the graphite. Before the graphite blocks are remcved from the PCRV, they will be submerged under water when the PCRV is flooded, which is expected to result in the release of a substantial fraction of tritium.
Based on the effects noted above,it is considered that a value of 10 uCilg of tritium represents a conservative estimate of tritium concentration in the_large side reflector and side spacer blocks (Reference 6). While this l concentration is a factor of approximately 40 below that projected in the activation analysis for these blocks, it provides a more realistic representation of the tritium concentration of the graphite blocks after l
they are removed from the PCRV. Therefore, a tritium concentration of 5-4 f
l l
y Suppism3nt to Environmental Raport FSV Decommissioning Section 5 10 Ucl/g in t_he large side reflector and side spacer blocks is assumed for the postulated decommissioning accident scenarios.
5.1.2 Assumptions l
The following are the major assumptions used in the analysis of postulated accidents which may occur during the dismantling activities:
- 1. The reactor is defueled and all irradiated fuel is removed from the Reactor Building.
i
- 2. Since al' fuelis removed from the reactor, there will be no need for shutdown /cooldown systems such as decay heat removal.
- 3. The Reactor Building ventilation system will remain operable, providing filtration of effluents to the environment, while the potential exists for drop of a large activated graphite block.
5.1.3 Dropping of Contaminated Concrete Rubble Accident 5.1.3.1 Identification of Cause After the majority of the PCRV top head concrete is removed in large pieces by diamond wire cutting, the last six inches (just above the PCRV top head liner) will be removed by utilizing a mechanical breaker to break up the concrete around the parameter of the PCRV top head liner, enabling the removal of the remaining concrete wafer in sections. This accident ;
scenario assumes that radioactivity is released from the drop of a rubble transport container due to a faulty crane or operator error.
5.1.3.2 Accident Description An activation analysis performed for Fort St. Vrain (Referenct 6) shows that the highest concentration of radioactivity in the PCRV concrete is in the six inch increment of the PCRV top head immediately above the top head liner as shown in Table 5.1-3. The values in Table 5.1-3 are based on three years decay, the approximate time frame in which the dismantling work is expected to take place. The percentage contribution of activation 55
Supplement to Environmental Report
' FSV Decommissioning Section 5 products within this concrete is given in Table 5.1-4. As shown, nearly 100% of the total activity is accounted for by the nuclides listed. it is assumed that Fe-55 and Co-60 are created from activation of contamination in the concrete rubble. No activation products in the embedded robar are assumed released. Of the entire PCRV 5.8% by weight (2% of volume)is rebar. It is conservatively assumed that 10%
of the totel activity resides in the concrete, it is also assumed that, of the total releasable activity in the concrete rubble,40% is due to tritium and 60% is due to the remaining nuclides.
The analysis conservatively assumes that 10% of the concrete in the six-inch thick concrete segment is involved in the accident. Of this amount, only 1% of the Fe-55, Co-60, tritium and Eu 154 are assumed to be released.
The airborne activity was calculateo to be 32.8 millicuries of Fe-55,1,43 millicuries of Co-60,3.93 millicuries of tritium and 5.90 millicuries of Eu-154. No credit was taken for particulate filtration by the Reactor Building ventilation system.
5.1.3.3 Analysis of Effects and Consequences This scenario was modeled as an elevated stack release using the AIRDOSE-EPA computer code. The maximum individual dose is 7.1 E(-7) mrem red bone marrow and 1.8E(-6) mrom to the endosteal bone tissue.
5.1.4 Conversion Construction Accident Near PCRV Dismantlement 5.1.4.1 Identification of Causes
- 1. Crane Failure An evaluation was performed on the potential impact of a construction crane toppling which would impact the Reactor Building. Due to the proximity of the planned new boiler building to the Reactor Building, it will be possible for a crane boom to strike the Reactor Building above the refueling floor level.
A crane boom is relatively light and fragile. An impact with the Reactor Building is not expected to cause structural damage to the building. At worst, the crane boom could drape over the Reactor Building siding. No radiological impact is expected from such an accident. LSA containers located outside the Reactor Building will I
l 5-6
Supplem:nt to Environmsntal R: port FSV Decommissioning Section 5 be protected if they are stored within the fall radius of the construction cranes. This accident is bounded by the heavy load drop (Section 5.1.5) and tornado (Section 5.1.9).
- 2. Explosion / Fire Due to Natural Gas Line Leak:
Fort St. Vrain will be repowered by a natural gas-fired boller. The most severe accidents that can be postulated during decommissioning activities involve a natural gas line leak resulting in an unconfined vapor explosion or fire, or an explosion of the gas-fired boiler itself. The decommissioning and repowering schedules have been reviewed. There is over a year between completion of the removal of highly radioactive components (graphite blocks) from the PCRV and introduction of natural gas on site. In the event of a slippage in the dismantling schedule, administrative controls will be implemented to prevent charging the gas-fired boiler natural gas line on site concurrent with handling of the activated graphite blocks from the PCRV. Therefore, given the actual schedule and administrative controls, an explosion or fire due to a natural gas line leak is not credible during the decommissioning process.
Accidental release of activity caused by a postulated explosion of a container of flammable gas, such as those used to support decommissioning (e.g., propane or acetylene tank or bottle), was taken into consideration. Flammable liquids and gases will be administratively controlled during decommissioning and conversion to prevent use or storage of substantial quantitles of flammable liquids or gas near areas containing highly activated wastes.
However, even if it were postulated that an explosion did occur near rac'ioactive waste containers, this event would not produce consequences exceeding those analyzed in this section for a heavy load drop, tornado or fire. This conclusion is based on the relatively small size of the missiles resulting from such a postulated explosion, and the relatively large amounts of activity postulated to be released in the above mentioned accidents.
5.1.5 Heavy Load Drop Accident The dismantling of the PCRV will be accomplished with the aid of three types of hoist systems. These systems include the main Reactor Building bridge crane, the auxiliary 17-1/2 ton hoiu on the bridge crane, and three 1-1/2 ton jib cranes on the refueling floor level. The Reector Building l
crane will be re reeved to allow the 170 ton main hook to tri <el from the 5-7
i Supplamant to Environmsntal Roport FSV Decommissioning Section 5 l
refueling floor to ground level. An elevation view of the PCRV work area is shown in Figure 5.12. There will be many heavy loads removed during the dismantling process. These loads include:
- 1. Large side reflector blocks
- 2. Large concrete sections
- 4. Helium diffusers
- 5. Concrete Core Support Floor or CSF sections The accident scenarios developed for heavy load drops in nucles' power plants consider the dropping of a heavy lo9d (e.g., fuel shipping cask) on a very large radionuclide inventory such as fuel or spent fuel (Relcreace 7), in the case of Fort St. Vrain, all fuel will have been removed from the Reactor Building prior -to commencement of dismantling operations.
Therefore, the full spectrum of heavy load drop accidents is much less severe than in an operational nuclear power plant.
The most severe heavy load drop accident is postulated to consist of dropping the component containing the largest inventory of dispersible radioactive material. Table 5.1-5 has been compiled to show the various components and their respective radioactive inventories. Sampling will be performed prior to waste movement to determine and verify the radionuclide composition and total Curie content. Review of this table indicates that the large side reflector blocks contain the largest radioactive inventory. The use of an entire large side reflector for this accident analysis is conservative since the predicted activity inventory exceeds the LSA Curie limit specified in 10CFR71, Table A-1, for Type A-2 waste containers.
The drop. of a heavy load onto a highly radioactive component was evaluated and determined not to represent the worst case scenario. For instance, the dropping of one of the 240 large side reflector blocks back into the PCRV might crush portions of adjacent reflector blocks. However, since all highly radioactive components are kept under water unless they are being removed, the debris and its attendant activity would remain in the water. This activity would be cleaned up in the PCRV water cleanup and clarification system. Any " slosh" created by the block drop would drain back to the PCRV cavity or drain down inside the Reactor Building, eventually to the Reactor Building sump and keyway, which have a capacity of approximately 350,000 gallons. These accident scenarios are bounded by the Loss of PCRV Shielding Water accident described in Section 5.1.7.
5-8
Supplsmsnt to Environm3ntal Rsport FSV Decommissioning Section 5 Alternatives for removal of the 270 ton concreto CSF from the PCRV include sectioning it into pieces within the PCRV, and removing the pieces by means of the Reactor Building crane, or raising the entire CSF above the PCRV with specially installed high capacity strand jacks. Since the activated graphite blocks would have been removed from the PCRV prior to removal of the CSF, and since the CSF concrete is predicted to contain only 6 Curies of activity, a heavy load drop during this operation does not have the potential for release of significant quantitles of radioactivity, if the entire CSF is raised by high capacity Jacks, drop of the CSF is not considered credible since such an accident would require inultiple jack failures.
5.1.5.1 Identification of Cause A heavy load drop accident is a relatively low probability event. A failuro of the hoisting cable could cause a drop of the load, in accordance with Reference 7, the probability of this event is on the order of 1.0E(-5) to 1.0E(-6) per demand (holst lift). The loss of the crane brakes could be due to mechanical failure, operator error, or an incorrect maintenance operation. Since the Fort St. Vrr.;n Reactor Building bridge crane does not qualify as a Single-Fallure-Proof crane in accordance with NUREG-0554 (Reference 8) guidelines, the loss of crane brakes is postulated as a credible failure mode.
5.1.5.2 Accident Description For this accident it is postulated (hat the Reactor Building bridge crane is hoisting one of the 240 large side reflector blocks. It is currently planned to section these reflector blocks into smaller pieces for packaging in LSA shipping containers. Moreover, it is conservative to ast,ume that a single reflector block may be transported intact in its own shipping container.
After aporopriate radiation surveys and removal of surface contamination, the container with the single unsectioned side reflector block is lowered down the enlarged equipment hatch. Failure of the crane is postulated at this point. This results in the side reflector block container falling approximately 100 fest to the level of the truck loading bay. The shipping container ruptures, spilling its contents on the truck loading bay floor.
Administrative controls will be in place that will prevent the tractor of the tractor tiailer from being in the loading bay during lowering of the container, and will ensure that all the truck loading bay doors are closed.
It is conservatively assumed that one percent of the activity of a single i
{
l 5-9
1 Suppl 3 ment to Environmsntal Report
- FSV Decommissioning Section 5 -
large reflector block is dispersed from the drop. The dust is postulated to remain airborne and will escape the immediate area through the Reactor Building ventilation exhaust. Credit is taken for cleanup afforded by the Reactor Building ventilation system.
J The Fort St, Vrain activation analysis (Reference 5) indicates that the major contributors to the activity in these large side reflector blocks are Fe-55, tritium, and Co-60. The total activity in each of the large side ,
reflector blocks has been calculated to be 1477 Curles. A one percent i release for this scenario results in M.)i Curies becoming airborne in the ,
Reactor Building. Of this amount,14.6 Curies are Fe-55,0.06-Curles are l tritium and 0.11 Curies are Co-60. These activities are based on a three year decay period. Credit is taken for a 95 percent filter efficiency for Fe-55 and Co 60. Tritium is released unfiltered.
5.1.5.3 Analysis of Effects and Consequences This scenario was modeled as an elevated stack release using the AIRDOSE-EPA computer code. The maximum individual dose is 7.3E(-7) mrem red bone marrow dose and 1.8E(-6) mrem to the lungs.
5.1.6 Fire 5.1.6.1 Identification of Cause During decommissioning and repowering activities, there are many possible fire initiators that could result in a release of radioactive materials.
These possible fire initiators include:
- 1. Fires started from cutting torches.
- 2. Contamination control tent fire.
-3. Fires associated with component processing activities on the refueling level.
- 4. Electrical fires.
The most likely initiator has been determined to be a cable tray fire started from a spark during PCRV random cutting operations. The fire would be quickly extinguished by the fire. watch on duty for the cutting operations.
The radiologicalconsequence of this accident would be negligible since the cable trays contain virtually no radioactive contamination.
5-10
-. - - - - . . .. . - - - - - - _ . . .- . - ~ =
)
Supplemsnt to Envhonmantal Report i FSV Decommissioning Section 5 The postulated fire accident involves a fire enveloping LSA waste containers. The greatest exposure for a fire accident to occur is during ?
the approximate six month period when the highly radioactive large side reflector blocks and side spacer blocks are being removed from the .PCRV.
Controls will be implernented prior to the storage of th'e LSA contalnurs, LSA containers will be limited.to groupings with no more than the equivalent Curie content of 230 side spacer blocks. This represents the largest inventory of activated graphite blocks vulnerable to fire. Suf ficient spatial separation will be imposed to- preclude fire propagation to an adjacent group of LSA containers. The packaging of these boxes and/or drums is planned to be completed inside the Reactor Building. Temporary
-storage or staging of these containers prior to shipment is also expected.
it is assumed that interim radioactive material storage will be available for up to 15 LSA boxes and 200 drums in the former new Fuel Storage Building.
Fire detection capability _ will be Installed in the LSA container storage area
- prior to the storage of the LSA containers. There will be no uncontrolled combustible materials in this building. The controls defined above will be
. implemented prior to the storage of th;., containers to limit the grouping of-LSA packages containing combustible- materials. These controls will
' ensure sufficient spatial separation is available to preclude fire propagation to an adjacent group of LSA containers and precludes the possibility of a fire with consequences greater than that which is analyzed.
5.1. '.2 Accident Description
~
For.the fire accident it is postulated that a tractor trailer begins- to transport packaged waste from the Reactor Building truck loading bay to an - off-site burial _ ground / processing facility. The shipment is conservatively postulated to consist of 230 side spacer blocks wn., their boron pins removed. There are 1152 side spacer blocks to be removed-during the decommissioning process.
This 3706 Ci source term is- the largest that will be contained in a transport truck destined for a burial ground.
it is postulated that an engine fire develops on the transport tractor and the fire spreads to the tractur's diesel fuel tanks. Based on work at the
' Waste Isolation Pilot Plant, the frequency of an unsuppressed truck fire is in the range of 1.0 E(-4) to 1.0 E( 5) per year (References 9 and 10). The tractor diesel fuel tanks may contain a combined capacity of up to 300 5-11
Supplemsnt to Environmental Report '
FSV Decommissioning Section 5 gallons of fuel.-The fuel tanks are postulated to rupture from the heat and engulf the entire tractor tialler and the LSA containers in a diesel fuel pool
- fire. It is conservatively assumed that graphite side spacer blocks are '
enveloped by the diesel fuel fire.
A fire involving 300 gallons of diesel fuel spilled onto a relatively flat surface will burn out within thirty minutes. The resultant fire temperature will be. bounded by the ASTM E119 (Reference 11) standard fire curve.
Most of the graphite will be exposed to temperatures well below the fire temperatures due to insulation provided by adjacant graphite blocks and some protection afforded by the shipping containers. ;
Under.these conditions, it is conservative to assume that 50 percent of the graphite inventory on a shipping trailer is oxidized during the 30-minute fire, it is assumed that all of the tritium in the oxidized fraction (50 percent of the total tritium inventory), is released, in addition to tritium release, it . is- assumed that 0.015 percent = of the balance .of the
. radionuclide inventory is released in the form of particulates (Reference 12). The accident is assumed to occur- at ground level immediately outside oflthe Reactor Building truck loading bay. The radioactive inventor / for the 230 graphite side spacer blocks is calculated to be 3,706 total Curies. This total inventory consists of 3556 Curles of Fe 55,122 Curles of tritium and 28.Curles of Co 60. Fifty percent of the Tritium is assumed to be released (approximately 61 Curies). The additional release of the remaining radionuclides will be 0.534 Curies of Fe-55 and 0.0042 Curies of Co-60.
5.1.6.3 Analysis of Effects and Consequences This scenario was modeled as a surface area release 14.6 m. In diameter-using the AIRDOSE EPA computer code. The maximum individual dose is
- 2.1E(-5) mrom red bone marrow dose and 3.1E(-5) mrom to the lungs.
l 5.1.7. Loss of PCRV Shielding Water Accident l
- 5.1.7.1 Identificat;on of Causes During a portion of the Fort St. Vrain decommissioning, the PCRV cavity will be flooded with water. This water will be circulated and purified by the PCRV water cleanup and clarification system to gradually decrease the
-radioactivity in the water. This system is expected to be in operation during the period when the PCRV internals are being removed.
5-12 L__ _ _ _ .___z_._ _ _ _ _ _ _ _ _ _ _ _ _ . . _ ._ __ . _ .__- -
Suppl:m:nt to Environment:1 R: port FSV Decommissionin0 Section 5 This rccident scenario assumes that there is a loak or rupture of the PCRV water cleanup and clarification piping resulting in a liquid rolooso duo to a mechanicalimpact or a mechanical falluto of a wold or flango.
5.1,7.2 Accident Description This accident scenario assumos that a mechanical failure of the PCRV water cleanup and clarification system pipin0 to the PCRV cavity occurs, resultin0 ni a pipo rupture. Other looks/b; oaks ca 1 be postulated (l.o., seal f ailures). However, the results are the same. Tritlated water with dissolved coslum, iron and cobalt would be spilled into the Roactor Building sump and keyway. Assuming the worst caso (comp',oto emptying of the PCRV), calculations indicato that 423,500 gallons could fill the Roactor Buildin0 sump / keyway, and flood the basement floor to a holght of two foot. This water woisld bo 49 foot below gradt- .nd would be contained bv the Ronctor Buildin0 sump /koyway and wr# . No credit is taken for the Roactor Building ventilation system foi this accident scenarlo.
Since the non Dascous activitLs will be retained in the spilled water, tritium (roloased throu0h ovaporation) is the only significant activity available. This will be ovaporated from the surface area of the spilled water in the Reactor buildin0 basemont.
The PCRV liquid release will not soop throu0h the sump concreto seams as tho water tablo 15 well a^oove the 40 foot below grado lovel. To dato, no known in loaka00 of ground water has boon observed into the Reactor Buildin0 sump.
The Reactor Buildin0i s approximately 120 foot long and 70 foot wide wnich consoivatively provides (no0locting equipment) a surface area for the spilled water of 9120 square foot (848 square motors), From Westin0h ouse Report WC AP 11002 (Reference 13). the best fit ovaporation rato at 70 percent relative humidity and an air circulation spood of 1 m/ soc is 0.046 g/m' soc or 0.046 cc/m' soc (assuminD 1 gram
= 1cc of watorl, it is prcJicted that tritium levels in the PCRV water will be loss than approximately 535 Curios. However, for this analysis, it is conservatively assumod that the theoretical maximum amount of tritium is transferred to iho PCRV shloiding water from the Ornphlto blocks, which is approximately 1 E(5) Curies, Thoroforo, the tritium concentration in the spilled water is calculated to be 62.4 uCl/cc.
5-13 i
l 1
Suppl::msnt to Environmental R:: port FSV Decommissionin0 Section 5 With an ovaporation rat 9 of 0.046 cc/m'-soc and a tritium concentration of 62.4 uCl/cc, the tritium release rato is about 2.5 mCl/ soc over the 848 square motors of surface arts. Over a two hour period,18 Cur:os of l tritium would be released to the etmosphero.
5.1.7.3 Analysis of Effects and Consequencus This scenario was modeled as a surface area releaso 29,1 m. In diamotor using the AIRDOSE EPA computer code. The maximum Individual doso is 4.6E( 6) mrom red bono marrow doso and 4.6E( 6) mrom to the lunt,s.
5.1.8 Loss of Power During the plant decommissioning, power will be normally supplied by of f.
sito sources. No backup power is assumed availablo durin0 a loss of power. The primary machinery using power during the decommissioning will be:
Pumps: Domolition Tools:
Dolonized Water System Plasma Arc Torch Firo Water Pumps Diamond Wire Cutter Servico Water Pumps Water Jet Cutter Water Treatment Drills PCRV Cloanup Water Pump Mobile Laundry Cranos PCRV Work Platform Lighting: HVAC:
Underwater Li0htinD Ventilation Fans Building HEPA Filters / Fans Plant Area HEPA Vacuums /
Portable Cleanors 5.1.8.1 Identification of Causes This accident postulatos the loss of off sito power due to weather related events. Such events could includo downed power lines due to strong winds or heavy icinD conditions. The likollhood of this cccurrence is remoto since off sito power can be suppilad to the sito through six separato lines.
5 14
_ __ ,- - -- -.~ .. , _ . . _ _ _-
Suppl:m:nt to Environmental Report FSV Decommissioning Section 5 5.1.8.2 Accident Description Loss of power would result in the loss of plant ventilation (HVAC) systems, lighting, plant water systems, and demolition power.
Decommissioning activitics would conso until power is rostored.
Loss of power to the PCRV water cleanup and clarification pumps will not result in a radioactivo release siitco the flow of blood water to the ovaporation ponds will be stopped. While loss of ventilatinn will force personnel from radiological control areas, no off site consequences are anticipated.
The postulated accident scenario is the loss of power to the HVAC while a largo sido reflector block has boon removed from the PCRV for cuttinD.
Those graphite blocks will be grappled and holsted by a lib crono to a refueling floor work station. At a work station a block will be cut into sections in preparation for packaging into LSA containers. The loss of power is assumed to occur after the cutting / cleaving onoration, it is assumed that those processing operations (korfing debris) roloaso 1.5 po' cont of the total activity of a single largo sido reflector block. Itis cc.morvatively judged that the combination of radiologicalcontrols in place at the work station (e.g., confinement through tonting) and the confinement function provided by the Roactor Building itself will result in rotontion of 99 porcent of the Fo 55 and Co 60 korfing dobris in tho ;
Roactor Building. It is assumed that one percent of the Fo 55 and Co 60 in the korfing debris and 100 percent of the tritium in the korfing debris are released at ground lovel from the Reactor Building. No credit is taken for the Reactor Building ventilation system.
The total activity in each of the largo sido reflector blocks has boon calculated to be 1477 Curios. A release of 1.5 porcent of the radioactivo material is assumed from the korfina debris in each block. Of that
- amount, one percent of the Fe 55 and Co 60, and 100 poucnt of the tritium is released, resulting in a total of 0,31 Curlos released to the environment. This total release consists of 0.213 Curios of Fe 55,0.091 Curios of tritium and 0.0017 Curies of Co GO. Those activities are based on a throo year decay period.
l 5-15
l Supplement to Environmental Report FSV Decommissioning Section 5 5.i.8.3 Analysis of Effects and Consequences This scenario was modeled as a surface area release with en offluent velocity of zero using the AIRDOSE EPA computer code. The maximum individualdose is 2.4E( 7) mrom red bone marrow dose and 5.8E( 7) rnrem to the lungs.
5.1.9 Natural Disasters For the effects of natural disasters, the following externallnitiating events were considered:
- 1. Earthquake The Reactor Building is designed to withstand the Design Basis Earthquake of 0.10 g horizontal ground acceleration at the site without unsafe damage or fa: lure to function. For decommissioning, it is required that the Reactor Building continue to perform its confinement function following a seismic event. The seismic qualification of the Reactor Building will be maintained during decommissioning. No other new or existing systems or equipment are required to function during or following an earthquake.
The most severe event which could result from a large earthquake is considered to be a drop of a radioactive waste container holding a highly activated graphite block (see heavy load drop accident).
However, the simultaneous occurrence of an earthquake and the hoisting of a heavy load is not considered credible. The consequences of this simultaneous earthquake and heavy load drop scenario were not analyzed due to the low probability of such an event.
- 2. Tornado and Wind Effects From Reference 14, Section 14.1,2, the basic dusign wind velocity for the plant is 90 mph. The equipment and structures exposed to wind load are designed to support design wind load combined with functionalloads within the specified allowable stresses.
The tornado danger at the plant site is extremely remoto. However, the Reactor Building was designed to withstand wind loadings developed by a tornado of 202 mph (total horizontal wind velocity) 5 16
Supplomont to linvironmental Report FliV Decommissioning '
Section 5 without exceeding yield stresses in the basic building structure.
The Reactor Building was also designed to withstand a maximum ;
tornado of 300 mph (total horizontal wind velocity) acting on tho l full area of all structures and a drop In atmosphoric pressure of 3 psl within a period of 3 seconds, without excooding ultimato stress levels in the main structural members. At the 300 mph wind spood, the siding on the Turbino and Reactor Buildings above the turbino dock and refueling floor lovels may be carried away, but the basic building structure will not collapse.
- 3. Floods j l
From Reference 14, Section 14.1.3, the plant sito is protected from i excessive runoff and flood by design of the yard dralnage system.
Grado level is approximately 17 foot above the highest observed flood level, and from 10 to 13 foot above the rnaximum probable flood lovel. The walls of the structures extending below grado lovel are watertight, and buoyancy offects were taken into account in their construction. Thorofore, there will be no further consideration of accidents due to flooding during decommissioning activities.
- 4. Ranto Fire The Fort St. Vrain sito is located in an area of Wold county devoted to agriculture. The sito itself is surrounded by pasture land and irtlqated fields. Within the plant exclusion area is a fire buffer area concisting of maintained grass and ornamentallandscaping. A 20 foot wide concreto pad rings the sito. Thorofore, a brush or range fire is not a credible accident during decommissioning activities.
5.1.9.1 Idor+1fication of Causes The risks from a tornado at Fort St Vrain duri.lg decommissioning are quito low for two reasons. First, the obability that a tornado will strike the sito is diminishingly small. Second, the plant specific vulnerability to a tornado and its consequences are also small. Unlike an operating nuclear power plant with active safety systoms to contain largo quantities of radioactive materials- at high energy levels, all spent fuel will be removed from Fort St. Vrain and the PCRV will essentially be a passive container of radioactive material. Possible loss of power caused by a tornado is specifically analyzed in Section 5.1.8.
5 17
Supplomont to Environmental R: port FSV Decommissioning Section 5 The Roactor Dullding roof and siding above the refueling floor are designed to withstand a tornado with a wind spood up to 202 mph. The probability of exporloncing a tornado with wind spoods abovo 202 mph during decommissioning is extremely low based upon Information end methodology provided in the draft Individual Pl ant Examination of External Events (!PEEE), NUREG 1407 (Roforence 15).
Based on the work of Abbey and Fujita (Reference 16), the continental United States was broken down into 20 distinct tornado hazard regions.
Thoso region', wc n genere rod into 4 broad areas shown in Figure 5,13, ranging from t highou M in region A to the lowest risk in Region D. The Fort St. Vrain Ute is iassiflod into Roglon C.
Referenco 17 is used to establish the occurrenco rate for different classifications of tornadoes. The National Sovero Storms Forocast Contor (NSSFC) national database for the years 1950 1978 was used as the basis for the occurrenco rato analysis. The NSSFC data are categorized by Fujita intensity scales (F scalos). To prodlet the probability that a tornado with maxirnum windspood will strike a nuclear power plant requires adjusting the F scslos for: tornado reporting trends, F scalo classification errorr, path longth intensity variata: t and occurrence ratos and windspood relationships adjusted for intensity vark: tion. The adjusted, or updated, tornado scales are denoted by "F", Tornado wind velocities for the F- and F' scales are compared as follows:
Maximum Windspood Maximum Windspood fEEcala inictyM mnN E'-Santo intonaLimahl FO 40 - 72 F'O 40 73 F1 73 - 112 F'1 73 103 F2 113 -157 F'2 103 135 F3 158 206 F'3 135 - 168 F4 207 - 260 F'4 168 209 F5 261 - 316 F'S 209 277 The following evaluation demonstrates the low probability of occurrence of a tornado with wind velocity exceeding 202 mph at Fort St. Vrain, by comparing the frequency of occurrence of tornadcas in Wold County with the NSSFC data The occurrence rato of a F4 tornado is 3,4 E( G) mi'/yr (Reference 17). According to the National Weather Bureau's historical data for Wold County from 1950 through 1987, thoto was only one tornado in the F3 rango, That single F3 tornado is the only tornado in the vicinity of Fort St. Vrain of the 256 tornadoes recorded by NSSFC for all 5-18
Supplomont to Environmental Report FSV Decommissionin0 Section 5 of rod on i C that had estimated windspoods greator than 158 mph. Based on this samplo from the population, it can be inferred that the probability of a tornado at Fort St. Vrain in the F3 rango is much loss than 3.4 E( 6) mi'/yr.
The occurrenco rato for a F5 tornado in Ro0l on C is 3.5 E( 7) mi'/yr (Referenco 17). The Natloaal Weather Bureau's Wold County data show no tornado occurrence whh intensity of F4 cr greator. Thus, the 56 F4 and nino F5 tornadoes recorded by NSSFC all occurred outsido the Fort St.
Vrain area.
From this data, it con be concluded that the probability of occurrence of an F4 or greator tornado is less than 3.5 E( 7) mir/yr. According to the draf t IPl!EE (Reference 15) " Plants Designed A0alnst NRC Current Critoria", those events pose no sl0nificant threat of a sovero occident because the current deslan critoria for wind are dominated by tornadoos having a frequency of excoodance of about 1 E( 7). The following section contains a specific accident analysis for a postulated tornado with winds less than 202 mph.
5.1.9.2 Accident Description Temporary storago or staging of radioactive wasto containers prior to shipment is expected, it is assumed that interim radioactivo material storage will be available for 15 LSA boxes and 200 drums in the Fort St.
Vrain Fuel Stora00 Building. Calculations demonstrato that neither forces generated by 202 mph wind loadina, nor the impact from the tornado-driven design basis missilo, will result in breach of the walls or roof of this buildin0 In this scenario,it is assumed that a 202 mph tornado strikes the Fort St.
Vrain site. At this lower wind level, the walls of the Reactor Building onclosing the PCRV will remain intact.
The tornado driven dosion basis missilo is a 12 foot x 12 inch x 4 inch thick fir plank, weighing 105 pounds, which impacts and ponotrates tho Reactor Building above the refuelin0 floor lovol. It is assumed that this missile strikes and ruptures a contalr.or with 46 graphite sido spacer blocks. It is conservatively assumed that one percent of the act'/ity n the container is dispersed and released to the environment. E ' f% tion credit is assumed.
l 5-19
- - - . , - , - - . - -y . - - . , , y---- ,.9.gf , ,..mn ,y --e,,;,-,ew .2-.,- nw,+ ,+w,..,0w9,-. -,we + y
Supplomont to Environmental Report FSV Decommissioning ,
Section 5 The total radioactivity inventory for the 40 sido spacer blocks is approximately 741 Curles. This totalinventoryis comprised of 711 Curios of Fo 55,24.4 Curios of tritium and about 5.5 Curies of Co 60. Assuming a one porcent release, this results in 7.41 Curlos released to the environment. Those activities are based on a throo year decay period.
The major exposure path was assumed to be air inhalation to an adult standing at the Exclusion Area Boundary (EAB).
5.1.9.3 Analysis of Effects and Consequences This scenario was modoled as a surface area release of 0.18m in diamotor using the AIRDOSE-EPA computer code. The maximum individual dose is 7.2E( 5) mrom red bone marrow doso and 2.0E(-4) mrom to the lungs.
5.1.10 Summary The results of the proceding accident scenarios, postulated for Fort St.
Vrain decommissioning activities, indicate that the radiation exposures to the general public will be minisculo. Those ovaluations have determined that, in all casos, the radiological consequences are well within the 10CFR100 guidelinen of 25 rom whole body doso and 300 rom to the thyroid from lodino exposure, of a maximally exposed Individual. These dosos are also a small fraction of the 1 rom whole body doso and 5 rom to any specific organ dose guidelines cited in the EPA Protective Action Guidelines (Reference 4).
These scenarios are considered to have a low probability of occurrence and their radiological consequences bound other loss severo accidents scenarios. The sfore, it is concluded that the Fort St. Vrain decommissioning activities do not pose any undue risk to the health and safety of the general public.
5.2 Transportation Accidents involving Radloactivity All shipments of waste from Fort St. Vrain are expected to be transported by truck except the upper portions of the steam generators.
The putential exists for truck accidents which could lead to radiation exposure to- transportation personnel and the general public. Truck accidents also could result in non radlological injuries. Typical truck transportation of waste generated by Fort St. Vrain decommissioning activities is not expected to easult in radiation dosos above background exposure to transportation personnel or to the general public.
5 20
Supplomont to Environmental Report FSV Decommissioning Section 5 A worst case on site truck accident was analyzed in Section 3.4.6 of the Proposed Decommissioning Plan. (This was for a tractor trailer loaded with 230 graphite side spacer blocks.) The results of this conservative analysis for a maximally exposed offsite individual (at the EAB) was a whole body doso (0 2 hr.) of 121 mrom and a lung dose of 215 mrom.
The population dose from this accident scenario has also been calculated using the AIRDOSE EPA code. The results are documented in Section 5.1.6.3. ,
The on site truck fire can also be postulated to occur off site. Application of a conservative air dispersion factor of 3.0E( 2) s/rn8 (Reference 19) results in a projected dose to an onlooker at 100 motors of 103 mrem to the whole body and 183 mrom to the lung. Potential doses to a transportation worker or a firefighter are estimated to be a factor of 3 higher.
These doses are based on the very conservative assumption that the diesel truck fire is not extinguished but burns itself out after consuming all the truck's diesel fuel. It is unlikely that such a fire would not be contained and extinguished within 30 minutes.
While these consequences are not trivial, they are low and very unlikely.
The number of fire accidents involving transportation of the most highly radioactive material (graphite side spacer blocks) from Fort St. Vrain is estimated to be 4.2 E( 5). This value is based on accident probabilities (Reference 20) for 5 shipments with a one way distance of 913 miles, it is estimated that to dispose of all radioactive waste from Fort St. Vrain decommissioning will require 160 one way and 225 round trip truck trips.
The one way distance to the disposal site in Beatty, Nevada, is 913 miles, although other disposal sites may be used. The number of traffic accidents was estimated to be seven with two nonradiologicat injuries and no fatalities. These values were also based on Reference 19.
It is currently planned to ship the upper primary (contaminated) portions of the twelve steam generators by rail to Richland, Washington, for burial.
A shipping container will consist of a metal culvert section seven feet in diameter by 27 feet long. The culvert section will be cut in half lengthwise to provide a hollow half cylinder. Structural supports will be welded to the half section of culvert.
5 21
Suppismsnt to Environmsntal Rsport FSV Decommissioning Section 5 The region between the outside of the steam generator shroud and the inside of the culvert can be filled with grout for shielding. The combined >
weight of the shipping container, steam generator assembly and grout could be as much as 195,000 pounds, it is possible that a railroad accident could result in the steam generator shipping container falling into a river enroute to Richland, Washington.
We have not analyzed this accident because we feel that the risk from this accident scenario is very low. First, the probability of a railroad accident occurring near a river or large stream is very low. The estimated probability of immersion due to accidents occurring on railroad bridges over deep rivers is 2E( 11) per car mile (Reference 20). Additionally the consequence from this accident would be minuscule, i.e., less than 0.001 mrom whole body. The low consequence is due to the fact that the steam generators are shipped in strong containers and loose crud and surface activation associated with the steam generators will have already been removed during the period that the steam generators are purposefully submerged in the PCRV water.
5.2.4 Summary Results of transportation accident scenarlos postulated for Fort St. Vrain decommissioning activities demonstrate that radiation exposure to the general public will be very low. Thus, activities related to the Fort St.
Vrain decommissioning project will not pose any undue risk to the health and safety of the public.
5.3 Other impacts We have evaluated the impacts of radiologicalaccidents on the health and safety of the public in the two previous sections. There are no large supplies of hazardous chemicals for decommissioning that could result in a catastrophic release. Approximately 20,000. cubic feet of non radioactive concrete and metallic scrap will be disposed of in local land fills, We conclude that there are no other significant activities resulting from decommissioning operutions that could result in a substantialimpact on the environment.
5.4 References
- 1. Proposed Decommissioning Plan for the Fort St. Vrain Nuclear Generating Station, Public Service Company of Colorado, November 1990.
5-22
Supplomont to Environmontal Report FSV Decommissioning Section 5 -
4
- 2. AIRDOSE EPA: A Computerized Methodology for Estimating '
Environmental Concentrations and Doso to Man from Airborno Releases of Radionuclidos, U.S. Environmental Protection A0 0ncy, December 1979. ,
- 3. Regulatory Guido 1.109 " Calculation of Dosos to Man From Routino Releases of Reactor Effluents."
4 " Manual of Protective Action Guidos and Protective Actions for Nuclear incidents," EPA 520/1-75 00,-A, U.S. Environmental Protection Agency, January 1990.
- 5. EngIncoring Evaluation, " Fort St. Vrain Activation Analysis,"
EE DEC 0010, Rev. 8, Public Service Company of Colorado, October 1990.
- 6. General Atomics Internal Correspondence, M.B. Richards to F.C.
Dahms, " Tritium Sourco Terms In Fort St. Vrain Permanent Sido Reflector and Spacor Block Graphlto," dated October 10, 1990; attachment to General Atomics Lotter GP 3487, dated October 11, 1990.
- 7. " Control of Heavy Loads at Nuclear Power Plants," NUREG-0612, USNRC, July 1980
- 8. "Singlo-Failuro-Proof Cranes for Nuclear Power Plants," NUREG-0554, USNRC, May 1979.
- 9. "Wasto isolation Pilot Plant Integrated Risk Assessment," DOE 89-10, June 1990,
- 10. "Special Analysis: Heavy Truck Firos, 1982-1986," National Fire Protection Association Firo Analysis and Research Division.
- 11. " Standard Test Methods for Fire Tests of Building Construction and Materials," ASTM E119-88,
- 12. Mishma, J. and Schwendiman " Fractional Airborne Roloaso of-Uranium During the Burning of Contaminated Wasto," BNWL-1730, April 1973.
5-23
i Supplement to Environmental Report FSV Decommissioning Section 5
- 13. WestinghouseWCAP-11002,"Evaluationof SteamGeneratorOverfill Due to a Steam Generator Tube Rupturo Accident," February 1986 (proprietary).
- 14. Fort St. Vrain Updated Final Safety Analysis Report, Revision 8, Public Servico Company of Colorado.
- 15. Draft; individual Plant Examination of External Events (IPEEE),
NUREG 1407, USNRC,1990.
- 16. Abbey, R.F., and Fujita, T.T.; "Regionalization of the Tornado Hazard, Tenth Conference on Severo Local Storms," American Mateorological Socloty, October 1977,
- 17. " Tornado Missile Simulation and Design Methodology, Volumo 2:
Model Verification and Data Baso Updatos," EPRI NP 2005, Volemo 2, ProJoct 616 2, Final Report, August 1981.
- 18. NUREG CR-0672, Technology, Safety and Costs of Decommissioning a Reference Bolling Water Reactor Power Station, Pacific Northwest Laboratory,1984.
- 19. WASH 1238, EnvironmentalSafety of Transportation of Radioactive Materials to and from Nuclear Power Plants, U.S. AEC, Directorate of Regulatory Standards,1972,
- 20. Donnis, A. W. ot al. "Soverities of Transportation Accidents involving Large Packages", Sandla Laboratorics SAND 77 000, May 1978.
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Supplement to Environmental Report FSV Decommissioning Section 5 TABLE 5.1 1
SUMMARY
OF POSTULATED DECOMMISSIONING ACCIDENT SCENARIOS Epstulated Accident Descriotion Dropping of Contaminated Rubble from PCRV top Concrete Rubble head concrete is dropped during processing.
Conversion Construction Natural gas explosion /
Near PCRV Dismantlement crane falling.
Heavy Load Drop Container drop to loading bay.
Fire Truck diesel fuel pool fire.
Loss of PCRV Shielding Watei Pipe rupture in the PCRV water cleanup / clarification system.
Loss of Power Release of graphite cutting debris from refueling floor work station.
Natural Disasters Tornado generated missile striking LSA waste container.
5 25
i Supplement to Environmental Report FSV Decommissioning Section 5 TABLE 5.12 DOSES DUE TO POSTULATED DECOMMISSIONING ACCIDENTSd8 Maximum '
Individual Red Bone Maximum Marrow Individual Postulated Dose Organ Dose Accident (mreml'2' (mrem)
Deopping of Concrete Rubble 7.1 E( 7) 1.8E( 6)
(endosteal)
Heavy Load Drop 7.3E( 7) 1.8E( 6) (lung)
Fire 2.1 E( 5) 3.1 E( 5) (lung)
Loss of PCRV Shielding Water 4.6E( 6) 4 6E( 6) (lung)
Loss of Power 2.4E( 7) 5.8E( 7) (lung)
Natural Disaster (Tornado) 7.2E( 5) 2.0E( 4) (lung)
- 1. All maximum doses occur in the first sector 4828m WSW of the plant. Refer to Figure 5.1 1 for location.
- 2. The most recent release of AIRDOSE-EPA does not calculate a whole body dose.
Red bone marrow doses are presented in its place.
5-26 L_ _ . . . _ _ _ . _ __.,. _ _
Supplement to Environmental Report FSV Decommissioning Section 5 TABLE 5.13 CURIE TOTALS IN ACTIVATED PCRV CONCRETE (3 YEARS DECAY)
LEJil!0Il Danill Cutica Top Head Axial Up 1st 6 inches 9.83 E(1) 2nd 6 inches 2.56 E(1) 3rd 6 inches 2.52 E(0) 4th 6 inches 2.70 E( 1) 5th 6 inches 3.68 E( 2) 6th 6 inches 6.35 E( 3) 7th 6 inches 131 E( 3) 8th 6 inches 2.85 E( 4)
Radial 1st 6 inches 8.89 E(0) 2nd 6 inches 3.13 E(0) 3rd 6 inches 3.66 El 1) 4th 6 inches 4.10 El 2) 5th 6 inches 5.94 E( 3) 6th 6 inches 1.08 E( 3) 7th 6 inct as 2.31 E( 4) 8th 6 inches 5.22 E( 5)
Core Support Floor 1st 6 inches 5.69 E(0)
Axial Down 2nd 6 inches 3.80 E( 1) 3rd 6 inches 3.33 E( 2) 4th 6 inches 3.60 E( 3) 5th 6 inches 4,66 E( 4) 6th 6 inches 7.67 E(-5) 7th 6 inches 1,42 El 5) 8th 6 inches 3.08 E( 6) 9th 6 inches 6.69 E( 7) 10th 6 inches 1.25 E( 7) 5 27
Supplement to Environmental Report FSV Decommissioning Section 5 TABLE 5.14 PERCENTAGE CONTRIBUTION OF ACTIVATION PRODUCTS IN FIRST 6 INCHES OF TOP HEAD CONCRETE (3 YEARS DECAY data from Reference 5)
Sionificant Nuclidga fgreent of Total (H H3 2.89 Ca41 0.05 Cs45 0.18 Fe 55 89.29 Co 60 3.43 Cs-134 0.24 Eu 152 3.51 Eu 154 _UD 99.95 Note 1: 98.3 curies totalin 1.44 E(7) cc of top head concrete.
I l
l 5 28 L -_ . _. . . _ . . . _. . __ _ . . . _ _ . . _ _ _ _ _ _ . . _ _ _ _ _ _ . . . . .
Supplement to Environmental Report FSV Decommissioning Section 5 TABLE 5.15 WASTE VOLUME / ACTIVITIES ESTIMATES FOR THE PCRV (3. YEARS DECAY)
Estimated Total Curies /
lle.rndyllcm Quantity Curies ,jgg, Region constraint device & pins 84 122 1.4 Metal clad reflector block CR 37 23100 624 Metal clad reflector block NCR 270 173000 640 Defueling blocks 1482 <0.01 Top reflector graphite blocks $89 2700 4.58
.Dottom reflector graphite blocks 902 4000 4.43 RaGal reflector hex graphite blocks removable & permanent 396 3300 8.3 Large side reflector blocks 240 354500 1477 Half site reflector blocks 96 83700 872 Upper reflector keys (carbon steel) 24 0.0144 0.0000 Side spacer blocks (no rods) 1152 18550 16,1 Boron rods 309792 36800 0.12 Lower reflector keys (Hastelloy) 24 470 19.6 Core support blocks 61 120 2.0 Core support posts 183 36.5 0.2 Core support floor columns 12 1 0.08 Misc. steel from beneath CSF 2 Metal on large side reflectors 24 0.014 0.001, ~
Core barrel 1 8.4 8.4 Lower plenum insulation <0.01 Silica blocks (25,000 lbs.) 250 Concrete top 130 l Concrt.te CSF 6 Concrete side 12
! Misc. Inconel parts on CSF 15 l
Hastelloy Cans 20061 3800 0.19 Concrete cutting debris top 16 Concrete cutting debris CSF 0.45 Concrete cutting debris side 0.44 l
l Helium purifiers in PCRV head 10 0.9 0.09 Helium ditfusers 4 20 5 Helium cire, shutoff valve assy. 4 2 0.5 5 29
l l
Supplement to Environmental Report FSV Decommissioning Section 5 l TABLE 5.15 (Continued)
Estimated Total Curies /
Item /Sts.Lem Quantity Curies _ltgm, j Heliurn bellows 12 20 1.06 Thermocouples & guide tubes 0.8 Steam generators 12 5676 473 Lower floor appurtenances 2 Platform / handling tools / jib cranes <0.01 Crane cable / drum /3 bucket inverters <0.01 Helium circulators 5 438 87.6 Orifice valves 37 415 11.2 Control rod drive assembf / 44 233 5.3
- Control rod absorber assembly 88 2.8 0.03 CSF Kaowool & Cover Plates 90 90 CSF Liner 142 142 Radial PCRV Liner 10 10 Top Cover Plates 5.7 5.7 Top Knowool <0.01 <0.01 Ton Head Liner 105 105 5-30 2._ .. ,- _ . . _ . . - ~ . . - _ . . - . . . _ _ _ , . _ _ _ . - . . _ _ _ . . . . . . . _ . _ _ . _ _ _ . - _ _ _ . .- _. _
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, Map Centered on Fort St. Vrain Site i
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5-31 L -
Supplement to Environmental Report FSV Decommissioning Section 5 (3) 1-1/2 Ton -,
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Fi0 ure 5.12 PCRV Work Area - Elevation View l 5-32 l
Supplomont to Environmental Report FSV Decommissionin0 Section 5
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I Supplement to Environmental Report l FSV Decommissioning l Section 6 l
6.0 ALTERNATIVES TO PRDPOSED ACTION 6.1 Available Alternatives The proposed decommissioninD alternative for Fort St. Vrain is the DECON, or immediate dismantlement ailernative. Upon approval of the Proposed Decommissioning Plan and removal of allirradiated fuel from the Reactor Building, this alternative will be implemented to immediately decontaminato and dismantle as necessary all plant systerns and areas to allow release of the facility for unrestricted use.
The followin0 were also considered as possible alternatives to the DECON decommissioning alternative, and will be evaluated in the following paragraphs:
- No action
- The SAFSTOR Decomrnissioning Alternative
- The ENTOMB Decommissioning Alternative The Decommissioning final rule allows the licensco to select any of the approved decommissionin0 alternatives. As each of these alternatives have been evaluated and determined to be acceptable (with limited exceptions; sco- ENTOMB below), licensees rnay choose any approved decommissionin0 alternative. Current regulations in 10CFR50.82 do not require that the licensee provide justification for selection of one approved decommissionin0 alternative over another.
6.2 No Action Decommissioning of the Fort St. Vrain Nuclear GeneratinD Station is a regulatory requirement under 10CFR50.82, which requires that a licensee must apply for authority to surrender its license and to decommission the facility within two years following permanent cessation of operations, and in no case later than one year prior to expiration of the operating license.
Therefore, takin0 no action is an unacceptable alternative.
6-1
Supplomont to Environmental Report FSV Decommissioning -
Section 0 0.3 Alternative Decontamination and Decommissioning Plans 6.3.1 Delayed Dismantlement (SAFSTOR Option)
An alternative to the DECON (Immodlato dismantlemont) decommissioning alternativo is SAFSTOR, in which the plant is placed and maintained in a condition that allows the nuclear facility to be safely stored and subsequently decontaminated to lovels that permit release for unrestricted uso, SAFSTOR can be utilized for a porlod of up 00 years, based on the expected amount of radioactivo decay during on approximato 50 year storogo period. SAFSTOR beyond 60 years will only be allowed when necessary to protect public health and safety.
PSC has considered SAFSTOR, which was the decommissioning >
alternative originally selected and submitted to the NRC in the Preliminary Decommisaloning Plan (P 89228, dated June 30,1989). The preliminary plan was based on a SAFSTOR parlod of 55 years, followed by a 3 year dismantlement and decommissioning period.
The following list identifies soveral of the factors that woro evaluated in datormining that the SAFSTOR alternative is less desirable than the DECON attornativo:
- 1. Selection of DECON reduces the risk associated with impacts of the regulatory process, a process still very much in a stato of development. This process could have a significant impact on future cost of decommissioning or duration of approval process.
2 Financial Risks:
- a. Low Lovel Radioactivo Wasto (LLRW): much uncertainty exists and is associated with future costs of low level radioactive waste (LLRW) disposal. Based on annual escalation ratos of disposalcosts of up to 15%, this may cause additional financial uncertainties in the final cost of decommissioning the longer the plant is placed in SAFSTOR.
- b. Decommissioning Cost Estimates: uncertainties and contingenclos can be much more accurately predicted in the immodlate future, than for 50 - 60 years into the future.
l l
62 l
t
Supplomant to Environmental R: port FSV Decommissioning Section S
- 3. Engineering evaluations performed by PSC and Indopondent contractors, confirmed during the competitive bid process for selecting the decommissioning contractor, have clearly demonstrated that the technology is available today to dismantle, decontaminate, and decommission Fort St. Vrain in a cost offective manner. PSC does not nood to rely on any major technological breakthrough or advances to perform the completo decommissioning and sito release of Fort St. Vrain for unrestricted uSe.
Selection of the DECON decommissioning alternativo is contingont upon approval by the Colorado Public Utilities Commission of PSC's app!! cation to repower Fort St. Vrain as a conventional fossil fueled generating unit.
The approval is necessary to ensure that sufficient revenue recovery conditions will be present to of fset the additional expenditures necessary to implernent the DECON alternativo, as well as successfully repower Fort St. Vrain.
6.3.2 Encasement for Radioactive Decay (ENTOMB Option)
A second decommissioning alternative to DECON is the ENTOMB decommissioning alternative. ENTOMB is defined as the alternative in which radioactivo contaminants are encased in a structurally long lived material, such as concreto. The entombed structure is appropriately maintained and continued surveillance is carried out until the radioactivity decays to a level permitting unrestricted release of the propcrty.
Although the Decommissioning Rule does not specifically preclude selection of the ENTOMB alternative for power reactors, ENTOMB was provided to allow the NRC flexibility when dealing with smaller reactor facilities, reactors which do not run to the end of their lifetimes, or other situations where long lived isotopos do not build up to significant levels or where there are other site specific factors affecting the safe decommissioning of the facility (e.g., presence of other nuclear facilities at the site for extended periods). Additionally, the NRC has indicated that the ENTOMB attemative is not a viable chc!co for power reactors.
6-3
Supplomont to Environmental R: port FSV Decommissioning Section G ;
PSC has completed preliminary activation analysos (Reference 1) of the PCRV and reactor core region, and determined that even after 00-100 years of decay, sufficient radioactive contaminants will be present to precludo sito release for unrestricted use. Thorofore, ENTOMB is not considorod to be a viablo decommissioning alternativo due to the presence of long-lived nuclidos on the Fort St. Vrain site, and the reluctance of the regulatory agency to allow oncasement of tlio large radioactivo quantitles on sito for an extended period beyond 100 years.
Thorofore, the ENTOMB decommissioning alternative is not considered to be an acceptablo alternative for decommissioning Fort St. Vrain, and was not considorod.
G.4 - Radioactive Waste Transportation Alternatives The primary mothod of transportation of the radioactive wasto from the decontamination and decommissioning of Fort St. Vrain will be truck.
Because of size and weight constraints, the special shlolded shipping containers with the steam gonorator modulos will be transported by rail for disposal.
The location of Fort St. Vrain precludes the consideration of convoyance by water as an alternative mode of transportation.
G.5 References
- 1. Engineering Evaluation- Fort St. Vrain Activation Analysis, EE DEC-0010 Rev. B, Public Service of Colorado, October 1990.
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q Supplement to Environmental Report !
FSV Decommissioning l Section 7 t 7.0 ANALYSIS r 7.1 Proposed Action ,
The proposed decommissioning plan to decontaminate and dismantle
$1ctivated and/or contaminated portions of the facility will result in several ,
positive impacts. All radioactive contamination as well as asbestos in the l
areas to be decommissioned will be removed from the site and transferred !
to an approved disposal facility, thereby allowing the FSV site to be released for unrestricted use. Because of the existing low radiation levels +
at the plant and considering the escalating costs and uncertainties associated with -disposal, there is no advantage in delaying decommissioning activities.
Approved disposal sites and space are currently available. Regulatory requirements are constantly changing, and the availability of a disposaisite !
in the future is uncertain._ Over the past several years, the costs of :
disposal havo been increasing approximately 15% per year, and at this rate, they will almost double avery five years. Disposal costs are a :
significant portion of decommissioning ' costs, so delays in ;
decommissioning will most certalnly result in higher costs for completing ;
this project, ;
No radiologicalIrnpacts are expected to the public in the vicinity of the ,
FSV site as the result of decommissioning activities. !
There will be some worker exposure _to radiation during various decommissioning activities, it is estimated that plant workers will be exposed to an estimated total dose of approximately 433 person rem from :
various dismantling and cleanup activities. Procedures will be implemented to ensure that worker exposure will be maintained as' low as '!
reasonable achievable (ALARA).
There may be some radiation exposure to transportation personnel and the general public from the transportation of radioactive wastes to the
. disposal site. An estimated 385 truck shipments (160 one way and 225 >
-- round trip) will be made over a_one way distance of 913 miles. These r: shipments will .be packaged. in accordance with . Department of Transportation requirements for allowable radiation levels, and the ,
radiation dose impact to workers and the public will be minimal.
71
)
l
'.--.. - . ,~a_~_._. . - - . - _ . _ _ _ . _ . . . _ _ _ _ _ . _ , . - . . _ _ . _ - , . ~ . , _ _ _ . _ . _
Suppisment to Environmental Report FSV Decommissioning Section 7 Asbestos removal will be carried out in a safe manner according to the requirements of 29 CFR 1910.1010, and no exposure to the general i public is expected. Work will be carried out in ventilated areas, and the exhaust will be filtered to remove asbestos particles.
The exhaust from vehicles and equipment will have minor impact on air-quality during dismantling operations. These impacts will be minor and limited to the duration of the dismantling activities.
Some fugitive dust will be generated by the dismantling activities.
Impacts on air quality will be minimal and temporary in duration.
No known archaeological or historic resources will be Impacted by the project.
There are no known threatened or endangered plant or animal species on ;
or near the FSV site that will be impacted.
~
The expenditure of greater than $137 million will contribute to the economy of the project. Assuming that 20 percent of the project cost will be spent locally, and using a total Impact multiplier of 1.3, the project's total impact on the local economy will be in excess of $35 million.
7.2 Alternative Decommissioning Plans
- Alternative decommissioning plans evaluated were found not to be as acceptable technically, economically, or environmentally as the proposed plan. A brief discussion of each alternative considered is provided below. ,
i
-7.2.1 ENTOMB As discussed in Section 6.3.2, this alternative involves encasing the-radioactive contaminants in a structurally long lived material, such as the concrete..PCRV, until the radioactivity decays to . a level permitting unrestricted release of the property.
- Since the NRC-Indicated that this alternative is not a viable choice for power reactors like Fort St. Vrain, this alternative is not evaluated further.
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_:.-~..,._-_ - - . . -. . - - - _ _ . -- _ -
Supplcmont to Environmental Report FSV Decommissioning Section 7
/.2.2 SAFSTOR As discussed in Section 6.3.1, this alternativo involvos maintaining the plant in a condition that allows the nuclear facility to be safety stored and subsequently decontaminated to levels that permit release for unrestricted uso. Based on the 60 year elowance, PSC ovaluated placing FSV into a SAFSTOR condition until 2043, and then decommissioning the facility from 2043 until 2046.
Althou0 hradiation lovels would decrease by 2043, they would still require remoto handlin0 In addition, as previously discussed, future low lovel radioactive wasto disposal costs are not kr.own and, based on recent 15%
annual escalation rates, can be expected to increase dramatically. Prico.
Anderson Act liability would continuo and other liabilities and regulatory requirements are uncertain. Further, the continuing site security and maintenanco costs contribute to the unattractiveness of this attornativo.
7.2.3 Of.! CON As alscussed in Section 3.1, this alternative has been chosen for the proposed decommissioning alternativo. It involves the decontamination and dismant!amont as necessary of plant systems and areas to allow release of the facility for unrestricted uso, This alternativo results in worker radiation exposures that are greater than the S AFSTOR alternativo, but the exposures are still substantially loss than the exposure estimates for light water reactor dismantlement projects.
7.3 Conclusions The DECON alternative is the most acceptable of the various feasible alternatives ovaluated. The proposed plan will result in the-decontamination or removal of contaminated equipment and materials from the FSV site, to an approved vwsto disposal facility with minimal adverse environmental impacts. The overall positivo environmental impacts resultin0 from the proposed plan will far outwel 0h adverse impacts. The FSV facility will under00 r horou0h cleanup, and unrestricted access will be permitted. The expenditure of greater than
$137 million will also have a positive environmentalimpact.
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A... . . . . . . . . . . . ..
. ..r L
> Supplement to Environmental Report FSV Decommissioning Section 8 0.0 ENVIRONMENTAL APPROVALS Decommissioning of I ort St. Vrain wi!; require the authorization of several Federal, State and locel agencas. Some activities, including tho decommissioning itself, will require specific authorization. Others may involve permits and approvals already in effect for operation of the facility.
Federal, State and local requirements are identified, and the status for each is leviewed below.
8.1 Federal Requirements Decommissioning octivities that are subject to Federal regulations, permits, licenses, nctifications or approvals include:
- Initiation of decommissioning e Handling, packaging and shipment of radioactive waste e Radio communications e Worker health and sbfoty e Worker radiation protection e Handling and removal of asbestos
- Hazardous waste generation The majority of these activities fall under the purview of Nuclear Regulatory Commission (NRC) regul .)ns: Title 10 of the Code of Federal Regulations (CFR). Applicable Titic 10 regulations are:
- Part 50 -
for decommissioning
- Part 20 - for protection against radiation
- Part 51 - for environmental protection
. Part 61 - for disposal of radioactive wasto e Part 71 - (and 49 CFR Parts 171 through 174) for packaging and transportation of radioactive waste.
The Decommissioning Plan requires review and approval by the NRC.
Once the Decommissioning Plan is approved and all irradiateo fuel has been removed from the Reactor Building, decommissioning will proceed under the conditions established by the Plan. The Proposed Decommissioning Plan for Fort St. Vrain was submitted to the NRC in November 1990. Decommissioning will be performed under the existing 10 CFR 50 regulations.
8-1
I-Supplement to Environmental Report FSV Decommissionin0 Section 8 Worker health and safety protection durin0 der w issionin0 alls f under Occupational Safety and Health Administration (d. Q regulations. These are 29 CFR Parts 1910 and i926 regulations appuable to construction activities. These regulations include requirements for respiratory protection (non-radiolo0l cal), hearing protection, illumination, scaffold safety, crano and rigging safety.
Asbestos handling and removal falls under OSHA regulations 29 CFR 1910 and 1926, and Environmental Protection A 0cncy (EPA) regulations 40 CFR Parts 61, Subpart M. In the State of Colorado, the State Health Department administcts the EPA regulations dealin0 with asbestos handling and remcJal.
Federal Communications Commission (FCC) licenses are required for radio communications equipment used at the Fo.1 St. Vrain site. This would include any radio communications equipment used in the reactor dismantlement and radwaste processing areas.
8.2 State and Local Requircments Permits and approvals from or notifications to several State and local -
aDencies are required for safety and environmental protections purposes.
Some of these are for specific decommissioning activitics, and others are for existin0 FSV site facilities and ongoing ectivities that are necessary to support decommissioning. Decommissioning activities and related site operations that fall under State and local jurisdiction include:
. Asbestos removal e Asbestos disposal e Fuel oil stora00
- Air emissions e Plant service water wella e Site liquid effluents (non-radiological)
. Building permits
. Hazardous waste generation At the State level, Colorado Department of Health's (CDOH) Air Quality Contro! Division regulates the installation, removal and encapsulation of friable asbestos containing materials.
Diesel fuel used during decommissioning is expected to be drawn from existing underground on site storage tanks. These are regulated by the CDOH's Hazardous Material and Waste Management Division.
1 i
82
l Supplsment to Environmental Report FSV DecommissioninD Section 8 Air emissions from the burnin0 of diesel fuel is regulated by the CDOh's Air Quality Control Division.
The site make-up water wells are operated under permits from the Colorado Depcrtment of Natural Resources.
The site sowa00 and non-radioactive liquid effluents are regulated by the CDOH's Water Quality Control Division.
At the locallevel, building permits will be required from deld C(unty for any temporary field of fico f acilities constructed on the pisnt site to su;; port decommissioning activities. Wold County uses the Uni.or.n Dull 6n0 Code for evaluatin0 permit applications.
Hazardous waste generation is re0ulated by the Colorcoo Department of Health's Hazardous Materials and Waste IWana]en,cx: Division.
Notification of the generator status and annual repermg nro conducted in accordance with Colorado state regulations.
8.3 - References 1, 10CFR Part 20, Standards for Protection Against Radiation
- 2. 10CFR Part 50, Domestic Licensing of Production and Utilization Facilities.
- 3. 10CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions.
- 4. 10CFR Part 61, Licensing Requirements for Land Disposal of Radioactive Waste.
- 5. 10CFR Part 71, Packaging and Transportation of Radioactivo Material.
- 6. 49CFR Parts 171 through 174, Departrnent of Transportation, Hazardous Materials Regulations. ,
- 7. 29CFR Part 1910, Occupational Safety and Health Standards.
8, 29CFR Part 1926, Safety and Health Regulations for Construction.
- 9. 40CFR Part 61, National Emission Standards for Hazardous Air Pollutants.
C-3
L Supplement to Environmental Report FSV Decommissionin0 Section 9 9.0
SUMMARY
AND CONCLUSIONS Public Service Company of Colorado proposes to decommission (DECON) nuclear facilities at the Fort St. Vrain Plant. The Fort St. Vrain High Temperature Gas Reactor operated from 1977 through 1989. Followin0 this decommissioning the 10CFR50 operating license will be terminated and the reactor site restored to unrestricted use.
The major dismantlement and decontamination activities to be performed during decommissioning are described in detail in Chapter 3. The decommicsioning project is divided into the followin0 major work areas:
- 1. Decontamination and dismantlement of the PCRV.
- 2. Decontamination and dismantlement of the contaminated balance of plant (BOP) systems.
- 3. Site cleanup and final site radiation survey.
- 4. Terminate the 10CFR50 operating license.
Site cleanup is described in Section 3.15 and the final site radiation survey is described in Section 2.3.
The twelve upper, primary portions of the steam generators will be shipped to a dirci, sal site by rail. Otner waste will be transported by truck to waste disposal sites.
Baseline radiation and contamination surveys were performed on the Fort St. Vrain Reactor and Turbine buildings in August 1990. The results showed that fixed contamination levels are generally less than 1000 dpm/15 cm 2and loose surface contamination levels are generally less than 1000 dpm/100 cm2 in the Reactor Building. In the Turbine Building, contamination levels (both fixed and loose) are less than 1000 dpm/100 cm2 in all locations and are generally less than 100 dpm/100 cm 2.
No future nuclear power operations at Fort St. Vrain are planned.
Radiological impacts to the public are expected to be insi0nificant.
Routine dismantling and packaging operations at the Fort St. Vrain site are not expected to produce releases above a small fraction of 10CFR20 and 10CFR50 Appendix I limn:
The planned decommissioning aperations that potentially could produce radioactive raleases will be conducted only when plant and local ventilation systems are in service.
9-1
9 Supplement to Environmental Report FSV Decommission;ng Section 9 Workers carrying out various dismantling and demolition activities wit! be i exposed to some radiation. However, radiation exposures of occupationally exposed personnel will be maintained as low as reasonably achievable (ALARA) and in compliance with 10CFR20. The total estimated dose for all decommissioning activities at the site has been estimated at 433 person rom. Decommissioning personnel will be protected against altborne radioactivity by the health physics controls and local environmental controls such as portable ventilation exhaust systems with HEPA filters. An NRC approved Respiratory Protection Program will be used during the decommissioning project.
Accident scenarios were developed for onsitu decommissioning accidents.
These included: dropping of contaminated concrete rubble, conversion construction near PCRV dismantlement, heavy load drop, fire, loss of PCRV shielding water, loss of power, and natural disasters such as tornadoes. The components with the highest radioactive inventories were used in the accident analyses.
No postulated accident has potential onsite or offsite radiologicc:
consequencesin excess of a small fraction of the EnvironmentalProtection Agency's Protective Action Guide or above 10CFR Paa 20 I;mits for routino occupational exposure.
The maximum potential onsite accident dose is due to 3 truck transportation accident and ensu3ng fire involving graphite side spacc-t blocks. Based on worst case site meteorology the postulated dose is 121 mrom whole body and 215 mrem lung dose at the Exclusion Area Boundary. Based on generic worst case meteorology, such a transport accident occurring offsite could lead to exposure of an onlooker situated 100 meters from the accident location of 103 mrem whole body and 183 mrem lung dose.
It is not expected that decommissioning activities will have an adverse impact on air quality. The reduced staff and derating of the auxiliary boiler will actually improve air quality relative to past power operation. The impact of the planned natural gas-fired power plant on air quality will be the subject of a separate environmental evaluation.
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Supplement to Environm:ntal Raport FSV Decommissioning Section 9 Asbestos removal operations will only involve that asbestos which must be disturbed to decommission the plant, and are expected to result in no public _ exposure to asbestos. Work areas will be ventilated and the exhaust filtered. All requirements of 29CFR1910.1010 concerning asbestos removal will be met.
Noise impacts from decommissioning will be minimal. Th , bulk of PCRV ,
concrete will be removed by diamond wire cutting rather than by jackhammer. Demolition will be conducted during the day and the nearest population conter, Platteville is located 31/2 miles southeast of the site.
There should be no impacts on wildlife or plants. No threatened or endangered species will be impacted.
Radioactive discharge of tritium is planned. 535 Cl of tritium are expected to be released during processing of the PCRV cavity water over a one year period. This process will utilize the existing liquid waste discharge system. Water from the plant blowdown will be used to sufficiently dilute the tritiated water to the 10CFR20 maximum permissible concentration limit.
The proposed decommissioning plan will have a positive socioeconomic impact. A torce of approximately 300 PSC and contract workers will be required for decommissioning activities. Over $137 million will be spent on the decommissioning project and as much as 20 percent of that amount is expected to accrue to the local economy.
s The proposed decommissioning plan for Fort St. Vrain is environmentally 4 sound and is expected to have no significant impact on the environment.
The decommissioning plan will resuit in the removal of radioactively contaminated equipment, materials, and waste from the site and permit unrestricted use of the decommissioned facilities.
9-3 1 1
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