ML20246F242
| ML20246F242 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/30/1989 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20246F224 | List: |
| References | |
| NUDOCS 8907130170 | |
| Download: ML20246F242 (102) | |
Text
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PRELIMINARY DECOMMISSIONING PLAN FOR FORT ST. VRAIN NUCLEAR GENERATING STATION TABLE OF CONTENTS 1
l 1. INTRODUCTION 1
- l. 1.1 PSC Decision to Decommission Fort St. Vrain ....... 1-1 l 1.2 NRC Criteria ...................................... 1-1 1 1.3 Preliminary Decommissioning Plan Contents . . . .. . . . . 1-2 1.4 Summary of Decommissioning Activities ............. 1-2
- 2. GENERAL DESCRIPTION OF FORT ST. VRAIN 2.1 General Description ............................... 2-1 2.2 Comparison of HTGR and LWR's ...................... 2-10
- 3. MAJOR TECHNICAL ACTIONS 3.1 Introduction ...................................... 3-1 3.2 Overview of Defueling and Spent Fuel Storage ...... 3-3 3.3 Fuel Issues ....................................... 3-7 3.4 Preliminary Plateout and Activation Studies ....... 3-8 3.5 Disposal Plan for Reactor Components .............. 3-14 3.6 SAFSTOR Issues .................................... 3-17 3.7 Decontamination and Dismantlement Plans . . . . . . . . . . . 3-19 3.8 Decorumi ssioning Cost Estimate . . . . . . . . . . . . . . . . . . . . . 3-35
- 4. RADI0 ACTIVE WASTE ISSUES 4.1 High Level Radioactive Waste ...................... 4-1 4.2 Low Level Radioactive Waste ....................... 4-2 4.3 Other Radioactive Wastes .......................... 4-3 4.4 Transfer and Shipment of Unirradiated Fort St.
Vrain Fuel ........................................ 4-4
- 5. RESIDUAL RADI0 ACTIVITY CRITERIA 5.1 Introduction ...................................... 5-1 5.2 Basis for Limit ................................... 5-1 5.3 Method to Establish a Residual Release Limit ...... 5-1 5.4 Site Release Survey ............................... 5-2
- 6. FINANCIAL FUNDING PLAN 6.1 N RC C r i t e r i a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 Corporate Financial History with Respect to Fort St. Vrain .................................... 6-1 6.3 Decommissioning Funding ........................... 6-1 6.4 Defuel i ng Pl an Fundi ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 8907130270 890630 F i PDR ADOCK 05000267 P PNU 7
. 4 PRELIMINARY l
DECOMMISSIONING PLAN TABLE OF CONTENTS i
1 i
i !
APPENDICES A External Trust Fund Contract LIST OF FIGURES ,
p l 2-1 Fort St . Vrai n Pl ot Pl an . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 l 2-2 Reactor and Turbine Buildings - Plan View ............ 2-6 l 2-3 Reactor and Turbine Buildings - Elevation View ....... 2-7 2-4 Prestressed Concrete Reactor Vessel .................. 2-8 2-5 PCRV General Prestressing Arrangement . . . . . . . . . . . . . . . . 2-9 2-6 Reactor Core and Reflector Arrangement . . . . . . . . . . . . . . . 2-11 2-7 Reactor Core - Pl an Vi ew . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-12 2-8 Reactor Support Structures ........................... 2-13 3-1 Fuel Handling Machine ................................ .3-4 3-2 Refueling Floor Arrangement (Elev. 4881') . . . . . . . . . . . . 3-6 3-3 Plateout Analysis Primary Circui t Model . . . . . . . . . . . . . . 3-9 3-4 Core Drilling Operations ............................. 3-23 3-5 Removal of Concrete Cutout ........................... 3-25 3-6 Removal of Dummy Fuel Bl ocks . . . . . . . . . . . . . . . . . . . . . . . . . 3-28 3-7 Removal of Core Support Blocks and Posts ............. 3-30 -
3-8 Removal of the Core Barrel . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-32 LIST OF TABLES 1-1 Fort St. Vrain Approximate Decommissioning Timeline .. 1-4 2-1 Major Fort St. Vrain Milestones and Events ........... 2-2 2-2 Total Radiation Exposure for Years 1986 - 1988 ....... 2-15 3-1 Estimated Plateout Concentration on Major Primary Circuit Components ........................... 3-11 3-2 Activation Dose Rate Estimates ....................... 3-13 3-3 Decommissioning Cost Estimate ........................ 3-36 6-1 Site Specific Cost Estimate (Escalated Dollars) ...... 6-2 6-2 Decommissioning Trust Fund Estimate .................. 6-5 ii
l PRELIMINARY DECOMMISSIONING PLAN TABLE OF CONTENTS LIST OF ACRONYMS AEC Atomic Energy Commission ALARA As Low As Reasonably Achievable ATC Auxiliary Transfer Cask 1 BWR Boiling Water Reactor CCTV Closed Circuit Television ]
i CD0H Colorado Department of Health 1 CFR Code of Federal Regulations CPIU Consumer Price Index for All Urban Consumers CPUC Colorado Public Utilities Commission CRD Control Rod Drive CRD0A Control Rod Drive and Orifice Assembly CSF Core Support Floor DOE Department of Energy dpm Disintegrations per Minute DRI Data Resources, Inc.
EPA Environmental Protection Agency EQ Environmental Qualification ESW Equipment Storage Wells FERC Federal Energy Regulatory Commission FHM Fuel Handling Machine -
FSV Fort St. Vrain FSW Fuel Storage Wells GA General Atomics GTCC Greater Than Class 'C' (Radioactive) Waste HEPA High Efficiency Particulate Absorber HLWR High Level Waste Repository HSF Hot Service Facility HTGR High Temperature Gas-Cooled Reactor HVAC Heating, Ventilation and Air Conditioning ICPP Idaho Chemical Processing Plant (DOE)
INEL Idaho National Engineering Laboratories ISFSI Independent Spent Fuel Storage Installation KW Kilowatt KWH Kilowatt-hour LWR Light Water Reactor MREM Millirem (IE-3 REM)
MCB Metal Clad (Reflector) Block MPF Manipulator Positioning Fixture MWe Megawatt (electric)
MWt Megawatt (thermal)
NFS Nuclear Fuel Services NRC Nuclear Regulatory Commission OCC (Colorado) Office of Consumer Counsel PCRV Prestressed Concrete Reactor Vessel PSC Public Service Company of Colorado PWR Pressurized Water Reactor QA Quality Assurance RCD Region Constraint Device REBAR Steel Reinforcement Bars REM Roentgen Equivalent Man (Radiation Measure) iii
l PRELIMINARY-
- DECOMMISSIONING PLAN. TABLE OF-CONTENTS RERP Radiological Emergency Response Plan SAFSTOR Delayed Decontamination / Dismantlement Decommissioning Option SFSC- Spent-Fuel Shipping Cask VPA Valley Pines Associates f
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SECTION 1 INTRODUCTION
.1.1 PSC DECISION TO DEC0ptilSSION FORT ST. VRAIN By letter to the Nuclear, Regulatory Commission . (NRC) dated December 5, 1988 (Reference 1.1), Public Service. Company of Colorado (PSC) notified the NRC ;that " based on economic considerations associated with the ongoing operating costs of Fort St. Vrain,- PSC has . determined that it will be -necessary to terminate Fort St. Vrain operations early. While a final' date for terminating Fort St. Vrain operations has not been . selected at this time, Fort St. .Vrain operations will be discontinued on' or before June-30, 1990."
Based on an end of operations date of not later .than June 30, 1990, PSC is within the 5 year period in which the Preliminary-
~
Decommissioning Plan is to be submitted. Therefore, this plan is being submitted at 2he earliest opportunity to be responsive to the requirements of .10 CFR 50.75(f), which requires submittal of -- a Preliminary Decommissioning Plan. Additionally, 10 CFR 50.33 requires that any holder of an operating license. for a utilization -
facility shall submit information indicating how reasonable assurance will be provided that funds will . be available to decommission the facility. This Preliminary Decommissioning Plan, inclusive of Section 6 " Financial Plan", is being. submitted to be responsive to' the requirements of 10 CFR 50.33(k)(2) and 10 CFR 50.75(f).
1.2 NRC CRITERIA This Preliminary Decommissioning Plan establishes = PSC's plans to decommission the Fort St. Vrain Nuclear Generating Station. Contents of this Preliminary Plan are being -submitted in accordance with the criteria contained in 10 CFR 50.75(f), which are identified as follows:
"E'ach licensee shall at or about 5 years prior to the projected end of operation submit a preliminary decommissioning plan containing a cost estimate for decommissioning and an up-to-date assessment of the major technical factors that could affect planning for decommissioning. Factors to be considered in submitting this information include -
(1) The decommissioning alternative anticipated to be used. The requirements of 50.82(b)(1) must be considered at this time; (2) Major technical actions necessary to carry out decommissioning safely; (3) The current situation with regani to disposal of high level and low level radioactive waste; (4) Residual radioactivity criteria; 1-1
PRELIMINARY DECOMMISSIONING PLAN SECTION 1 (5) Other site specific factors which could affect decommissioning planning and cost; If necessary, this submittal shall also include plans for adjusting levels of funds assured for decommissioning to demonstrate that a reasonable level of assurance will be provided that funds will be available wher needed to cover the costs of decommissioning. "
1.3 PRELIMINARY DECOMMISSIONING PLAN CONTENTS Section 1 of this plan contains a description of PSC's current situation with respect to Fort St. Vrain and a schedule of future Fort St Vrain decommissioning activities.
Section 2 contains a general description of Fort St. Vrain and a comparison of a typical light water reactor (LWR) with the Fort St. Vrain High Temperature Gas-Cooled Reactor (HTGR).
l Section 3 identifies the major technical actions which will oe required to decontaminate the site and environs for eventual -
decommissioning of the site and termination of the 10 CFR bu license.
Section 4 identifies the current major high level and low level radioactive waste issues confronting PSC during decontamination and decommissioning of Fort St. Vrain.
Section 5 discusses the residual radioactivity criteria that PSC will use as its basis for decontamination of Fort St. Vrain components and facilities.
Section 6 contains the PSC Financial Plan which identifies the funds necessary to accomplish decommissioning and PSC's plan for accumulating these funds.
1.4 SUMARY OF DECOMMISSIONING ACTIVITIES PSC has selected the SAFSTOR option for decommissioning Fort St. Vrain based on occupational safety, economic and financial considerations. While this decision is based primarily on occupational exposure considerations and financial decisions, it is influenced by the unavailability of a Federal High Level Waste Repository (HLWR) to receive one segment of Fort St. Vrain spent fuel. The following is an overview of the activities which must be completed to accomplish decommissioning. A timeline for accomplishment of these activities is provided in Table 1-1.
- a. Defuel the Prestressed Concrete Reactor Vessel (PCRV):
Ship five (of six) segments of spent fuel assemblies to a Department of Energy (DOE) facility in Idaho per prior 1-2
PRELIMINARY DECOMMISSIONING PLAN SECTION 1 agreements (Reference 1.2) with the DOE. Store one segment of spent fuel in the Fuel Storage Wells -(FSWs). PSC is also conducting feasibility studies for a 10 CFR Part 72 Independent Spent Fuel Storage Installation (ISFSI).to be used at Fort St. Vrain until a Federal HLWR is available.
- b. Remove Components and Prepare the PCRV for SAFSTOR Period:
Perform system decontamination and dismantlement; remove
. helium circulators and control rod drive and orifice assemblies (CRDOAs); modify plant systems (e.g., fire protection and security) to support anticipated requirements during the SAFSTOR period.
- c. Ship Retained Fuel Segment to HLWR:
Ship the remaining fuel segment to a Federal HLWR (anticipated date: 2020).
- d. SAFSTOR Period:
Maintain the radioactive components in SAFSTOR for a period of approximately 55 years.
- e. PCRV Dismantlement and Site Release For Unrestricted Use:
Following conclusion of the SAFSTOR period, decontaminate and dismantle radioactive components to levels which are acceptable for unrestricted release.
1-3
PRELIMINARY DECOMMISSIONING PLAN SECTION 1 TABLE 1-1 FORT ST. VRAIN APPROXIMATE DEC0WIISSIONING TINELINE The dates identified in the following schedule are based on the latest possible end of operations date. PSC plans to be fully prepared to begin defueling by November 1, 1989. Other dates will be adjusted accordingly it final shutdown occurs before June 30, 1990.
The schedule for decommissioning of Fort St. Vrain is as follows:
June 30 1990 Final shutdown of reactor (latest date).
September 1990 End of 100 . day fuel decay period; comence defueling of the reactor (5 spent fuel segments to DOE Idaho, I spent fuel segment stored in FSWs).
October 1992 . Complete defueling to DOE- Idaho; comence defueling to FSWs.
April 1993 Complete defueling to FSWs; Comence component removal and preparation of the site for SAFSTOR. -
October 1993 Complete component removal; Place FSV in SAFSTOR with onsite spent fuel storage in either the FSWs or an onsite ISFSI.
1993 - 2020 SAFSTOR with one segment of spent fuel in FSWs or in an onsite ISFSI.
2020 Ship one spent fuel segment to Federal HLWR.
2020 - 2043 SAFSTOR Period (no remaining onsite spent fuel storage).
2043 End SAFSTOR; Decontaminate and dismantle the PCRV-and remaining contaminated systems.
2046 Terminate 10 CFR Part 50 license; Release the site for unrestricted use or controlled use by owner.
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l 1
PRELIMINARY
. DECOMMISSIONING PLAN SECTION 1 REFERENCES FOR SECTION 1-1.1 PSC Letter, R.0. Williams (PSC) to J. Calvo (NRC), dated December 5, 1988;
Subject:
"Early Termination of Fort St.
Vrain Operations", (P-88422).
1.2 The As-Modified Three Party Agreement, Document : Number 34426, Dated July 1,1965.
i I
1-5
c SECTION 2 GENERAL DESCRIPTION OF FORT ST. VRAIN 2.1 GENERAL DESCRIPTION Fort St. Vrain is a High Temperature Gas-Cooled Reactor (HTGR) owned and operated by PSC. Fort St. -Vrain's location is approximately 35 miles north of Denver and three and one-half miles northwest of the town of Platteville in Weld County, Colorado.
The site consists of 2798 acres owned by PSC. Approximately one mile square.within the site area is designated as _the exclusion area, and the' licensee maintains complete control over this . area.
The closest - distance from the . reactor building to the nearest-exclusion area boundary is about 1935 feet, but the reactor building is about 3500 feet from the nearest site boundary.
Construction of Fort St. Vrain was authorized by the Atomic Energy Comission (AEC) by issuance of a provisional construction permit to PSC on September 17, 1968, in.AEC Docket No. 50-267. Table -
2-1 contains a brief sumary of major 'lant milestones / events occurring during the history of the plant. -
In 1968, the Colorado Public Utilities Comission (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-fuel 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 powered 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 operation in 1972. Although PSC received a full power operating license in 1973, extensive pre-operational testing mandated by the NRC and resulting engineering modifications delayed the comercial operation of the plant until 1979. GA, the prime contractor for Fort St. Vrain, reimbursed PSC for all increases in electric operating 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 comercial operation at a reduced capacity of 200 MWe (at 60% capacity factor) instead of 330 MWe (at 80% capacity factor), as originally designed. GA paid PSC
$60 million as 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.
2-1
l PRELIMINARY-
' DECOMMISSIONING PLAN SECTION 2 TABLE 2-1 MAJOR FORT ST. VRAIN MILESTONES AND EVENTS December 1973 Plant construction completed.
December 21 1973 Facility Operating License No. DPR-34 issued to PSC.
December 26 1973 Initial fuel loading.
January 31 1974 Initial nuclear criticality.
1974 - 1979 Startup testing, low power operation, and required plant modifications.
July 1 1979 Fort St. Vrain begins commercial operation.
Novemb9r 6 1981 Achieved 100 percent full . power operation.
June 1984 Shutdown following six control rod drives (CRDs) failing to automatically scram.
February 14 1986 Plant restart following CRD Refurbishment Outage.
May 1986 Environmental Qualification (EQ) outage.
September ~ 1986 PSC removed FSV from the rate base in an agreement with the CPUC; Initial decommissioning writedown.
October 1986 Safe Shutdown Cooling reanalysis performed which reduced maximum power level to 82% of rated power (270 of 330 Mwe).
April 17 1987 Plant restart following EQ outage.
July 1987 Shutdown following helium circulator bolt failure; Replaced "D" Circulator.
October 1987 Hydraulic fire during plant restart.
June 1988 Plant record for . MWe generated for one month period.
July 1988 Shutdown to refurbish helium circulators.
September 27 1988 Reevaluation of decommissioning costs results.in revised write-down.
December 5 1988 PSC Board of Directors approves decision to shutdown and decommission Fort St.
Vrain; operations will cease on or before June 30, 1990.
2-2
PRELIMINARY DECOMMISSIONING PLAN SECTION 2 Fort St. Vrain was first included in PSC's rate base in a general rate case in December 1980. In that rate decision, the CPUC allowed PSC to collect approximately $39 million of revenues (an amount- subsequently increased to $46 million) to cover operating expenses and provide a return on investment.
In response to Fort St. Vrain's historically reduced levels of i generation, the CPUC had issued decisions which instituted penalties '
against PSC to reduce the revenues recovered from its customers.
PSC, the CPUC and interveners had taken these penalty issues through the legal appeals process to the Colorado Supreme Court.
Additionally, the Colorado Office of Consumer Counsel (OCC) had filed a complaint with the CPUC against PSC alleging, among other things, that in light of its operating history, Fort St. Vrain was not "used and useful" in rendering a utility service. The OCC sought to remove Fort St. Vrain from PSC's rate base so that PSC would no longer be able to collect its previously established annual revenue requirement of approximately $46 million with respect to Fort St. Vrain. l In view of the various legal and administrative proceedings regarding Fort St. Vrain, PSC entered into a Stipulation and ~
Settlement Agreement dated September 24, 1986, with the CPUC, the OCC and other parties. Significant provisions of the Stipulation and Settlement Agreement on PSC's part include: a reduction in its annual retail electric revenues collected through rates of $29 million effective October 1,1986; removal from rate base of PSC's investment in Fort St. Vrain assets; recovery over five years of $22 million of plant costs and $11.5 million of decommissioning costs; refunding to customers of $73 million in penalties previously incurred by PSC; a provision for the sale of energy produced at Fort St. Vrain in the future to customers at 4.8 cents per Kwh with possible limited increases; contribution of $2 million to a foundation to provide for the energy needs of low income customers; and a stipulation that PSC would not file for any rate increase to be effective before July 1, 1988. As a result of these actions agreed to by PSC, all legal and administrative proceedings relative to the ratemaking treatment of Fort St. Vrain were terminated.
2.1.1 Buildinos and Structures The completed facility is shown in Figure 2-1. The basic installation consists of a reactor building, a turbine building, cooling towers, and an electrical switchyard. The turbine building, cooling towers and switchyard do not contain any radioactive components which require evaluation as part of this Preliminary Decommissioning Plan. This plan and the associated cost estimate assume that all main buildings (reactor building, turbine building and Building 10) as well as all outlying auxiliary buildings, will not be dismantled during the SAFSTOR period.
2-3
.5 PRELIMINARY DECOMMISSIONING PLAN SECTION 2
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PRELIMINARY DECOMMISSIONING PLAN SECTION 2 2.1.2 Reactor Buildina The reactor building (Figures 2-2 and 2-3) houses the prestressed concrete reactor vessel (PCRV), fuel handling area, fuel storage wells, fuel shipment facilities, decontamination and radioactive liquid and gas waste processing equipment, and all reactor plant process and service systems.
The building is able to withstand wind loadings developed by a 100 mph wind or a tornado of 202 mph total horizontal wind velocity without exceeding yield stresses. The facility is designed to withstand a maximum tornado of 300 mph total horizontal wind velocity without damage to the PCRV structure, although the upper part of the reactor building siding may be damaged due to the tornado.
The PCRV and nuclear steam supply system are located in the west portion of the reactor building. The east portion of the reactor building houses auxiliary and support systems and facilities such as the fuel storage wells (FSWs), the hot service facility (HSF), the equipment storage wells (ESWs), storage and laydown areas for various' pieces of equipment, and the loading port for the spent fuel shipping cask (SFSC). -
2.1.3 Turbine Buildina The turbine building (Figures 2-2 and 2-3) houses the turbine generator with condensing, feedwater, and other auxiliary systems, and an auxiliary bay area housing the reactor plant ventilation equipment, the controlled personnel access to the reactor building, an area housing the control room and miscellaneous electrical services, and a service and office area which provides space for miscellaneous . shops, auxiliary steam system components, and administrative offices.
2.1.4 Prestressed Concrete Reactor Vessel (PCRV)
The PCRV (Figures 2-4 and 2-5), which contains the nuclear steam supply system, is a reinforced concrete structure prestressed with steel tendons. Following defueling, the PCRV will contain the ,
majority of the remaining radioactive materials in the reactor building. The PCRV concrete and REBAR are expected to remain activated due to direct irradiation from the reactor core. Highly ;
radioactive components will remain inside the PCRV until removed j during PCRV decontamination and dismantlement. !
Physically, the 151/2-foot thick heads and the 9-foot thick concrete walls are constructed around a 3/4-inch thick low-carbon steel liner which forms the internal cavity. The liner is anchored to the concrete at frequent intervals and provides a gas-tight membrane seal. A core support floor (CSF) is provided within the PCRV in the form of a reinforced, water-cooled, concrete wafer with a 2-5 i
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PRELIMINARY DECOMMISSIONING PLAN SECTION 2 HELIUM STORAGE PRESTRESSED o CONCRETE P S' ' REACTOR BUILNNG REACTOR VESSEL- % l l
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. DECOMMISSIONING PLAN SECTION 2 NNNNbNA4444d FI.IIIId.1111M
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PRELIMINARY DECOMMISSIONING PLAN SECTION 2
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i PRELIMINARY DECOMMISSIONING PLAN SECTION 2 TYPICAL MruCLsN4 PENETRAT100s ACCESS PENETRATIost
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PRELIMINARY DECOMMISSIONING PLAN SECTION 2 3/4-inch carbon steel outer liner, supported by 12 steel core support floor columns from the bottom of the PCRV cavity.
-The PCRV internals (Figure 2-6) consist of the following major components. The active core region consists of fuel elements, control rods, removable and permanent graphite reflector blocks and boronated graphite spacer blocks, all contained within the core barrel.. Boronated metal clad reflector blocks (MCBs) and the control rod drive and orifice assemblies (CRDOAs) are located on top of the active core region. Region constraint devices (RCDs) are located at the top of the core fuel and reflector columns to prevent relative movement between the columns in order to minimize fluctuations and temperature redistributions.
The active core is divided into 37 regions consisting of 1482 total fuel elements (Figure 2-7); these 37 regions are further divided into 6 " segments" for purposes of refueling. Removable graphite reflectors are located on the top, sides and bottom of the active fuel region. " Permanent" graphite reflectors are located on the sides of the core outocard of the removable reflectors in the shape of a right circular cylinder stacked the entire height of the fuel region. Between the permanent reflector and the core barrel is -
a right circular cylinder of graphite spacers containing boronated stainless steel pins, used primarily for shielding.
Immediately outboard of the core barrel is a helium interspace area. Outboard of this interspace area is an outer metal insulation cover plate, Kaowool (thermal) insulation, an inner metal insulation cover, another layer of Kaowool, and then the PCRV carbon steel liner, which serves as a gas-tight barrier for the primary coolant gas.
Below the active core region, . the CSF bears the weight of the active core and reflectors through the core support posts and the core support blocks (Figure 2-8). The CSF also is the bottom termination point of the core barrel and has 12 penetrations for the 12 steam generator modules. The CSF is supported from the bottom head of the PCRV with 12 core support floor columns.
2.2 COMPARISON OF HTGR AND LWR's Fort St. Vrain is the only HTGR in the United States. The 330 MWe (net) generating plant features a helium-cooled reactor with a uranium-thorium fuel cycle.
The PCRV contains the total primary coolant system including the reactor, steam generators and helium circulators (Figure 2-4).
The reactor is graphite moderated and reflected with an active core consisting of 1482 hexagonal fuel elements loaded with triso-coated uranium and thorium particles cast into cylindrical rods. Helium, at 700 psia (nominal), is discharged from four helium circulators and passes down through the core absorbing heat from the fission process.
2-10
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PRELIMINARY DECOMMISSIONING PLAN SECTION 2
' NOTES:
- 1. FUEL ZONE BOUNDARIES
- 2. FUEL REGION BOUNDARIES b R ADIAL FUEL ZO*JE 11
- 3. CONTROL ROD COLUMN RADIAL FUEL 20NE lit SHADED REFLECTOR ELEMENTS N RADIAL FUEL ZONE IV ARE NORMALLY REPLACED WITH ADJACENT FUEL REGION j f RADIAL FUEL ZONE V
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BOUNDARY l NUMBER I l SIDE REFLECTOR STEE L CORE SPACE R BARREL CORE MAP Figure 2-7. Reactor Core - Plan View 2-12
PRELIMINARY DECOMMISSIONING PLAN SECTION 2
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PRELIMINARY DECOMMISSIONING PLAN SECTION 2 The helium then flows through twelve steam generator modules, transferring heat to the secondary coolant to produce steam, and returns to the suction side of the helium circulators.
The Fort St. Vrain PCRV is designed to perform the same functions as both the reactor vessel and containment structure normally seen at light water reactors (LWRs); therefore, no need exists for a separate containment structure to surround the Fort St.
Vrain PCRV. Additionally, there are significant design and constructi3n differences between the fort St. Vrain PCRV and the steel reactor vessels normally utilized in LWRs. Although the PCRV and LWR reactor vessels are both design d to contain fission products during all credible accidents and operating conditions, the use of prestressed concrete with significant amounts of reinforcing steel has created a unique dismantlemer' problem. The concrete and its steel reinforcing bar have become ectivated in the vicinity adjacent to the active core. An additional constraint to dismantlement is
. created by the use of a PCRV with limited access to the PCRV internals. The top of the PCRV has 37 refueling penetrations (19-inch I.D.) and an access penetration (46-inch I.D.). The PCRV does not have a removable top head which is typically seen in LWR ,
design. Therefore, the PCRV top head must bv dismantled to allow - '
access into the PCRV for dismantlement.
With the exception of the unique once-through steam generators, the secondary side of the plant is similar to conventional fossil-fueled superheat units. Separation of the primary coolant from the secondary-side steam cycle is accomplished in a manner similar to that normally seen in pressurized water reactors (PWRs). However, the Fort St. Vrain steam generators are located inside the lower cavity of the PCRV and are not easily accessible. As noted above, this also presents a signi fi", ant restriction on removal of the steam generators through the limited access openings in the bottom head of the PCRV.
The radiological consequences of operating Fort St. Vrain have .
been exceptionally low, due in part to the use of helium gas as the !
primary coolant and the ability of the fuel particle coating to retain fission products. Use of helium has significantly reduced the amounts of activated corrosion products transported through the core when compared to typical LWRs. Therefore, the auxiliary support systems and the reactor building areas are much less contaminated than similar LWR systems. Personnel exposure rates and discharges to ;
the environment have consistently been orders of magnitude below i those of the average LWR power plant in the US. The major reason for j this performance is inherent in the design of the core and primary '
coolant system. Release of contamination into the reactor building environment is minimized by: (1) the unique graphite core structure, !
with ceramic-coated fuel particles, (2) the use of inert primary coolant gas (helium), and (3) the PCRV with its double sealed helium-pressurized penetrations.
2-14
f PRELIMINARY-DECOMMISSIONING PLAN SECTION 2 The equilibrium radioactivity circulating in the primary coolant was conservatively calculated to be 30,900 C1 (design level) !
with an expected value of 2630 Ci. Actual measurements. measured 439 Ci during the third fuel cycle core (Ref. 2.1).
In the area of occupational radiation exposure, total radiological exposures for each of the years 1980-1983 was less than 1 person-REM for all personne; badged at Fort St. Vrain. Total exposure levels increased to 35 persen-REM in 1935 due almost entirely to the control rod refurbishment program. Occupational exposure levels during the period 1986-1988 are identified in Table 2-2.
TABLE 2-2 TOTAL RADIf. TION EXPOSURE FOR YEARS 1986-1988 NO. OF MONITORED TOTAL EXPOSURE 1E.A_R EMPLOYEES (PERSON-REM) ,
1986 291 1.82 1987 209 1.24 1988 232 0.72 In the area of low level radioactive waste, a total of 200 cubic meters of low-level solid waste has Ren generated at Fort St.
Vrain since criticality was achieved. This includes waste generated as a result of (1) initial fuel loading,- (2) three refuelings, each I of which replaced one segment (one-sixth) of the core, and (3) numerous plant modifications and maintenance operations, including major work performed on components within the primary coolant system.
With the exception of tritium, nobic gas airborne and radioactive liquid effluent releases from Fort St. Vrain have been i more than an order of magnitude below the average of the US nuclear power industry. Higher than anticipated amounts of tritium have been ;
experienced at Fort St. Vrain primarily because of upsets in the '
l helium circulator auxiliary system, which have allowed varying
! amounts of bearing water to enter and circulate in the primary coolant system. The tritiated moisture is removed by the purification system and is discharged from the plant as liquid waste (Ref. 2.1) .
2-15 w_________--_____--_ _ - _ _ _ _ . .. . _ - - - -
( 1 :5 *
.PRELIMINA'RY DECOMMISSIONING PLAN SECTION 2 y
REFERENCES FOR SECTION 2 2.1 Brey, H.L. 'and Graul, W.A.; " Operation of the Fort St. Vrain High-Temperature Gas-Cooled Reactor Plant," Proceedings of the American Power Conference, Chicago, Illinois, April 26-28, 1982.
4 9
1 2-16 1
O-----_-___- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
SECTION 3 MAJOR TECHNICAL ACTIONS
3.1 INTRODUCTION
3.1.1 Decommissioning Alternative Selected In accordance with the decommissioning options provided in the NRC's Decommissioning Final Ruin, PSC has chosen the SAFSTOR option for the basis of the Preliminary Decommissioning Plan for Fort St.
Vrain. It is anticipated that Fort St. Vrdn will be placed in this SAFSTOR condition for a period of up to 55 years, followed by decontamination and dismantlement. Following decontamination, the site will be eligible for release for unrestricted use. PSC currently plans to maintain owner-controlled jurisdiction over the site and remaining structures following dismantlement and decontamination to levels which would allow site release for unrestricted use.
SAFSTOR, as defined by the NRC, occurs when a nuclear facility is placed and maintained in a condition that allows the nuclear facility to be safety stored and subsequently decontaminated -
(deferred decontamination) to levels that perrit release for unrestricted use. Due to the high radiation levels of activated components contained in the core and the occupational exposure involved with their removal, 55 years has been chosen to allow sufficient radioactive decay to radiation levels which will minimize occupational exposure.
3.1.2 Overview of Decommissioning Activities for the decommissioning process, the following phases and corresponding uajor technical actions are foreseen:
- a. Final Shutdown (not later than June 30,1990):
- Shutdown during the burnup of Fuel Cycle 4.
- Three of eight allowable segments of spent fuel have previously been shipped to DOE during defuelings prior to 1989.
- Six fuel segments (1482 fuel blocks) remaining in the Core.
- b. Defueling Period (September 1990 through April 1993):
- Defuel six segments from the reactor core.
- Ship five spent fuel segments to Idaho.
- Transfer the remaining spent fuel segment to the FSWs, pending eventual transfer to a licensed ISFSI.
- c. Component Removal Period (April 1993 through October 1993):
- Remove and ship CRDs and helium circulators off-site as radioactive waste.
3-1
F PRELIMINARY DECOMMISSIONING PLAN SECTION 3
- Dismantle and/or decontaminate contaminated systems outside of the PCRV.
- Complete necessary modifications .to ready the balance of plant for SAFSTOR (e.g , fire protection) and to support spent fuel storage in the FSWs.
- d. SAFSTOR and Fuel Storage Period (1993 - 2020):
- Maintain the PCRV and necessary support systems in a SAFSTOR condition.
- Maintain the remaining spent fuel segment in the FSWs (or ISFSI) for ultimate transfer to DOE in approximately 2020.
- e. SAFSTOR Period (2020 - 2043):
- Continue to m:intain the PCRV and necessary support systems in a SAFSTOR condition.
~ f. PCRV Decontamination / Dismantlement Period (1043 - 2046):
- Remove the remainir.g core internal components.
- Decontaminate and dismantle that portion of the PCRV structure which exceeds limits for unrestricted release of residual radioactive materials. -
- Ship all remair.ing radioactive waste offsite.
- Terminate the 10 CFR Part 50 license.
The dates identified above are based on a June 30, 1990, shutdown and will be adjusted accordingly if an earlier shutdown occurs. Commencement of defueling is subject to change, depending on finalized details of the defueling actions, equipment reliability, and other variables. The date for final shipment of the remaining
- fuel segment is based on opening the HLWR in 2003 and the existence of a 10 year queue of spent fuel awaiting shipment to the HLWR. The dates for final dismantlement may also be revised due to several factors, including a determination of the actual radiological conditions within the PCRV, availability of high-level and low-lev,el radioactive waste disposal and financial conditions and decisfon-within PSC.
3.1.3 Other Site-Soecific Factors There are several other site-specific factors not covered in the Decommissioning Final Rule which could impact the planning and funding for decommissioning of Fort St. Vrain. These are listed below and include some factors not directly associated with j decommissioning as defined by the NRC in the Decommissioning Final Rule. However, these factors are closely interrelated to overall PSC planning and funding, and are identified in this Preliminary
- Decommissioning Plan for information purposes only,
- a. Defueling -
Although not directly related to decommissioning, defueling has a large impact on overall PSC costs and schedule due to 3-2
1
. PRELIMINARY DECOMMISSIONING PLAN SECTION 3 the length of time (currently estimated to be 2.6 years) rcquired to complete defueling activities. Therefore, defueling is discussed-in general terms in Section 3.2.
- b. Fuel Issues -
Although not directly related to decommissioning, fuel related issues and their effects on decommissioning are discussed in Section 3.3.
- c. Conversion -
A final decision to convert Fort St. Vrain to a fossil-fired facility has not yet been made. If converted, i this would significantly impact the balance of plant equipment, staffing levels and security require:nents during SAFSTOR and decommissioning. Since this decision is not final , PSC will keep the NRC apprised of any future decisions related to Fort St. Vrain conversion. Any onsite construction activities to. convert Fort St. Vrain to fossil fuels will not commence until the PCRV has been completely defueled,
~
- d. Fort St. Vrain Removed From the PSC Rate Base -
Inability to recover full decommissioning costs through the PSC rate base and other special financial considerations specific to Fort St. Vrain are discussed more thoroughly in Section 6.
- e. DOE Research Projects -
DOE and PSC are currently evaluating Fort St. Vrain to determine if there are treas of interest (materials, programs, and/or analyses) that DOE may be interested in recovering from PSC during Defueling and Decommissioning.
3.2 OVERVIEW OF DEFUELING AND SPENT FUEL STORAGE 3.2.] h 8tplino Process Although not within the scope of this Preliminary Decommissioning Plan, defueling of the six spent fuel segments (1482 fuel assemblies) contained in the core is twended to be accomplished in the manner identified in the following paragraphs. The objective of defueling'is to place the spent fuel in its final disposition as rapidly as possible and then proceed with the shipment of other activated ahd contaminated plant components and wastes. The spent fuel will be removed region by region using the fuel handling machir,e (FHM) (Figure 3-1) and will be replaced with dumy fuel blocks to ensure structural stability of the core during defueling. To maintain subcriticality during the spent fuel removal, lumped (boron) poison pins will be loaded into the dummy fuel blocks.
Five cf the six spent fuel segments (1242 fuel assemblies) will be shipped to the DOE storage facility in Idaho. The remaining 3-3
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PRELIMINARY-DECOMMISSIONING PLAN SECTION 3 spent fuel segment (240 fuel assemblies) will be stored onsite in the FSWs, which are licensed under the original 10 CFR 50 operating license for Fort St. Vrain. If necessary, the FSWs will be utilized until the Federal HLWR is available and Fort St. Vrain spent fuel can be shipped to the repository for disposal (currently estimated for 2020).
The following briefly describes the materials intended to be used. The dummy fuel blocks will be fabricated from graphite and will be loaded with boron pins for reactivity control. Configuration of the dummy blocks will be very similar to the present fuel blocks in that they will have a center pickup hole to facilitate manipulation by the FHM and will have coolant holes to maintain adequate flow distribution though the core.
For planning purposes, the defueling and subsequent component removal periods. would begin no later than September 1990. The defueling period, accomplished first, is anticipated to last approximately 2.6 years and the component removal period is anticipated to be accomplished in the next 0.5 years.
Several factors contribute to the duration of the defueling period, including the required sequence of movements of equipment on the fuel deck (Figure 3-2), the lack of storage space for the spent fuel removed from the PCRV (the FSWs can only store two of the six segments of spent fuel from the core as part of the original 10 CFR 50 licensing basis) and the slow process for shipping spent fuel offsite. A total of 1482 fuel blocks will be removed from the core in a region-by-region sequence. After removal from the core, 1242 spent fuel blocks will be transferred by the FHM to the Fort St.
Vrain SFSCs for shipment to Idaho. A total of three SFSCs are available with a capacity of six spent fuel blocks per cask. The remaining 240 spent fuel blocks will be transferred by the FHM to the FSWs for storage. All the removed fuel blocks will be replaced by dummy fuel blocks on a region by region basis.
3.2.2 Component Removal Period In order to safely contain radioactive contaminants located at the site, certain actions are anticipated to be necessary to prepare plant systems and buildings. Radioactively contaminated systems outside the PCRV will be either: (1) abandoned (for systems with levels of contamination below the limit for release for unrestricted use); or (2) mechanically and electrically isolated, flushed, and either decontaminated or dismantled and disposed as radioactive waste. A determination of suitable security boundaries will be made.
Necessary modifications and reviews will be made to ensure that adequate fire protection will remain available both inside and outside these boundaries, and that actions are taken to contain remaining contaminated systems within these boundaries.
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3-5
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 L
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l l 3-6
PRELIMINARY DECOMMISSIONING PLAN ~ SECTION 3 The Component Removal Period will commence immediately following the completion of defueling. During this period, the CRD0As, helium circulators and possibly the RCDs will be disposed of as radioactive waste. The auxiliary transfer cask (ATC) will be used to remove the CRD0As for disposal following completion of defueling.
Helium circulators will then be removed utilizing existing equipment and well established practices. All- remaining core components will be left in place for removal during the PCRV decontamination and dismantlement period. These actions are further described in Section 3.5.
3.3 FUEL ISSUES 3.3.1 Soent Fuel Aareements The original three party agreement (Rsf. 3.1) between PSC, General Atomic (GA) and the AEC provided storage for eight segments of FSV fuel at the Idaho National Engineering Lab (INEL) facility.
To date, PSC has shipped three segments of spent fuel to INEL as a result of three previous refuelings. In 1988 the agreement was modified (Ref. 3.2) to clarify the intent of the agreement and to assure the storage of five spent fuel segments at DOE Idaho. ~
3.3.2 Disposition of Soent Fuel (Seament 9)
The remaining segment of spent fuel (Segment 9) will be stored in the FSWs for an interim period (if required), pending licensing of an ISFSI facility. The FSWs are adequate for indefinite storage of up to 2 segments of spent fuel. Use of the FSWs was evaluated as part of the original licensing basis for the 10 CFR 50 license.
Fegment 9 will ultimately be transferred to DOE for permanent storage in the HLWR in approximately 2020. The year 2020 was selected based on an HLWR inservice date of 2003 and an existing 10-year queue of spent fuel awaiting shipment to the HLWR.
Additional conservatism were incorporated to allow for uncertainties in the HLWR inservice date, possible enlargement of the spent fuel queue, and potential difficulties which may be encountered when transferring the spent fuel from PSC to D0E. It should be noted that PSC is actively pursuing other possibilities for disposal of Segment 9 which may negate the need for any onsite spent fuel storage in either the FSWs or in an ISFSI.
3.3.3 Unirradiated Fuel Issues In addition to the irradiated fuel in the core, PSC possessed one segment of unirradiated fuel onsite (Segment 10) and unirradiated material (work-in-process) at the GA Fuel Fabrication Facility. This material consisted of up to 97% highly enriched . uranium at various stages of the fuel fabrication process. All unirradiated fuel and materials have been sold to Nuclear Fuel Services (NFS) of E. win, Tennessee. All unirradiated material has been removed from the GA 3-7
l PRELIMINARY DECOMMISSIONING PLAN SECTION 3
' Fuel Fabrication Facility; however, PSC will continue to store Segment 10 at Fort St. Vrain until NFS is ready to receive this segment for storage (currently estimated for 1992).
3.3.4 PSC liabilities Associated with the GA Fuel Fabrication Facility PSC has notified'GA that PSC has no Intention of fabricating additional FSV fuel. PSC is contractually ' obligated to' provide funding if GA decides to decommission this- facility. This notification initiated action on the part of GA and Valley Pines Associates to decide the future of the Fuel Fabrication Facility.
3.4 PRELIMINARY PLATEOUT AND ACTIVATION STUDIES 3.4.1 Plateout Analysis 3.4.1.1 Purpose A preliminary plateout distribution analysis of fission product isotopes produced in the reU, tor core was performed for the _
PCRV and internal components (Ref. 3.3). The purpose of the analysis
-was to estimate the plateout concentrations and distributions in the primary circuit to better 1.lan decontamination activities and to estimate occupational exposure during decontamination' activities.
3.4.1.2 Assumptions and Methodology The analysis was conservatively performed using the end of Fuel Cycle 5 (Segment 10) as the end of plant life, as opposed to the planned end of life during Fuel Cycle 4 (Segment 9). The axial and radial core power distributions through Fuel Cycle 5 were calculated and, with flux distribution data, were used 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 analyses for the key fission gases and metal, the total plateout and helium purification system inventories of radionuclides were estimated.
Finally, the plateout distributions were calculated for the PCRV internal components. A schematic of the primary circuit model used in the plateout analysis is shown in Figure 3-3.
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 coating fail in srnice. In addition, the volatile metals (Cs and Sr) can, at sufficiently high temperatures and over long periods of time, diffuse thi9 ugh the silicon carbide (Sic) coatings and be released from intact %el particles.
Calculations were performed for the following key nuclides:
Sr-90, I-129, I-131, Cs-137, Cs-134 and Te-127m. The source terms 3-8
PRELIMINARY DECOMMISSIONING PLAN' SECTION 3 CORE EXIT LOWER -
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3-9
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 for fission product plateout analysis included both a direct reisase l contribution and, in some cases, a precursor contribution. In the I 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 while Cs-134 has no gaseous precursor. precursor Similarly,contributor for Sr-90,Xe-137, there is a direct Sr-90 metal release as well as the contribution from its Kr-90 precursor. Only direct release contributions need to be considered for I-129, I-131 and Te-127m. 1 3.4.1.3 Analysis Results l
The results of the analysis indicate that contamination I levels in the PCRV at the time of shutdown will exceed the current NRC guidelines for release for unrestricted use (Ref. 3.4) . Some nuclides with relatively short half lives, such as I-131 and Te-127m, will decay to acceptable limits during the SAFSTOR period. Other nuclides, such as Sr-90 and Cs-137, with longer half lives will not decay to acceptable limits during the SAFSTOR period. Therefore, it will be necessary to remove or decontaminate components or areas to acceptable limits before the PCRV would meet the unrestricted release limits for surface contamination. It is anticipated that any -
component that has come in contact with primary coolant will have to be decontaminated or removed for disposal as radioactive waste. This includes not only core graphite and structural components, but also the steam generators, helium circulators and Kaowool insulation.
Table 3-1 lists the preliminary determinations of fission product plateout concentrations (Cs-137 and Sr-90) on major primary circuit components.
The final analysis will contain estimates of the total curie distribution of key nuclides as well as estimates for concentrations (Ci/sq. cm.) of plateout nuclides on major PCRV components. These finalized estimates will be included in the Proposed Decommissioning Plan.
3.4.2 Activation Analysis 3.4.2.1 Purpose j
A preliminary activation analysis was performed for the i PCRV and associated internal components to determine the amount and l extent of activation of components and structures in the immediate j vicinity of the core (Ref. 3.5). This evaluation was necessary to estimate the radiation levels and amount of radioactive materials which will require eventual removal and disposal during the l dismantlement period. The analysis was performed to determine the '
isotopic composition, magnitude and extent of residual radioactivity which would be present in the PCRV at the end of Fuel Cycle 4. I 3-10 1
~;
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 TABLE 3-1 ESTIMATED PLATE 00T CONCENTRATION ON MAJOR PRIMARY CIRCUIT COMPONENTS Plateout Concentration Sr-90 Cs-137
-(dpm/ (dpm/
-Comoonent 100 cm2 ) (Ci/cm2 ) 100 cm2 ) (C1/cm23 Lower Reflectors 1.11E07 5.0E-08 2.22E08 1.0E-06 Steam Generator .. 2.22E07 1.0E-07 3.33E08 1.5E-06 (ReheaterSection)
Circulator 4.44E07 2.0E-08 5.55E07 2.5E-07 Circulator Outlet 3.33E05 1.5E-09 9.99E05 4.5E-09 Core Barrel Annulus 2.22E06 1.0E-08 7.77E06 3.5E-08 Upper Reflectors 2.00E07 9.0E-08 4.44E07 2.0E-07 Highest estimated concentration on the component.
Steam generator component with the highest estimated plateout concentration.
i i.
3-11
['
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 3.4.2.2 Assumptions and Methodology The analysis was divided into three parts: neutron flux estimates in the PCRV, activation analysis of the PCRV and internal components and the gamma dose rates inside the PCRV due to non-removable (fixed) components. Component material compositions were obtained from a variety of sources. Actual material '
specifications and certification records were used' where possible.
Industry standard compositions were used where original material data was not available. . Trace element abundances were estimated using current regulatory guides and/or any available sample data.
The activation analysis required tha use of several computer codes and various input data libraries, the ANISN transport theory code was used to determine the neutron 1 %x throughout the reactor core and outward to the PCRV concrete. The neutron activation of selected components within the PCRV was then determined using the REBATE computer code. The activation analysis results were checked with the ORIGIN computer code. Finally, the gamma ray doses' within the PCRV were calculated using the photon transport option of the ANISN code. The analysis was performed using three one-dimensional models: (1) radial (core center line outward through -
the PCRV side); (2) axial up (core center line upward through the PCRV top head); and (3) axial down (core center line downward though the core support floor) . A schematic of the PCRV and internal components is shown in Figure 2-6.
3.4.2.3 Analysis Results The results of the preliminary activation analyses show that a SAFSTOR period of up to 55 years is justifiable to reduce the occupational risk to workers. The dominant nuclide for gamma radiation dose estimates for metallic components is Co-60. The components found to have the highest dose rate were the boronated side reflector spacers (see Figure 2-6). The dose rate from these spacer blocks in the center of the core area is reduced by shielding from the large permanent side reflectors. However, if the large reflectors are removed, the dose rate in the center of the core area at five years after shutdown is estimated to be approximately 330 Rem /hr. Mter a SAFSTOR period of 55 years, this dose rate is estimated to drop to less than 0.3 Rem /hr. Therefore, after the SAFSTOR period, there could be limited access inside the PCRV to perform dismantlement activities using remote and manual techniques.
Table 3-2 shows the estimated dose rates inside the PCRV before and after the SAFSTOR period with various components removed.
l Portions of the PCRV liner and concrete will probably require removal. The amount of concrete requiring removal will depend upon the limit established for release for unrestricted use and the time after shutdown that dismantlement will occur (to allow for radionuclides decay). The dose rate from the PCRV concrete itself is dominated by Eu-152 in the long term. Co-60 is the primary dose 3-12
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 TABLE 3-2 ACTIVATION DOSE RATE ESTIMATES Dose Rates Dose Rates Removal at the Core Mid-Plane at the Core Mid-Plane Phases 5 Years After Shutdown After 55 Year SAFSTOR 1 2.l 0.0015 2 330.0 0.2300 3 1.0 0.0007 Phase 1:
Radiation levels in the center of the core area as a result of all internal PCRV components remaining in place, including: large permanent side reflectors, boronated spacer blocks, core barrel, Kaowool insulation and cover j plates, PCRV liner, PCRV concrete, core support blocks and posts, silica insulation, and CSF liner and concrete.
Ehnf._21 Radiation levels in the center of the core area following removal of the permanent large side reflectors; all other internal PCRV components identified in Phase 1 remain in pl ace.
Phase 3:
Radiation levels in the center of the core area following removal of the permanent large side reflectors, all components inside and inclusive of the core barrel (boronated spacer blocks, core support blocks and posts),
Kaowool insulation and cover plates, and silica insulation.
Remaining components include CSF liner and concrete, and the PCRV liner and concrete.
I
- h Dose Rates are in Rem /hr and are the sum of radial, axial up and axial down dose rates. No components below the CSF were included in the analytical model. i Core mid-plane dose rates for Phase 1 are less than those for Phase 2 due to the shielding provided by the large reflector blocks inside the boronated spacer blocks.
Removal of these large permanent reflectors causes a '
significant increase in core mid-plane dose rates.
I 3-13
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 contributor in the short term due to the activated REBAR in the J concrete. The volume of PCRV concrete requiring removal would be greatly decreased after a 55 year SAFSTOR period. Although no formal limits for residual radioactivity have been established, it is anticipated that delaying dismantlement for up to 55 years could reduce the amount of concrete requiring removal by a factor of one-half to one-third, than if the PCRV were dismantled soon after final reactor shutdown.
The finalized study will contain estimates of the total radioactive material inventory in the reactor vessel and internal components due to activation. If possible, samples of irradiated reactor components will be analyzed prior to component removal to aid in verification of analysis results. The Proposed Decommissioning Plan will provide estimates of the curie content of major PCRV components, as well as dose rate estimates inside the PCRV.
3.5 DISPOSAL PLAN FOR REACTOR COMPONENTS Following defueling, the component removal period will focus on removing all radioactive liquids, gases, and contaminated systems located outside the PCRV. The following discussion represents the -
most time and cost-effective scenario presently known to defuel and remove core components. Future plans submitted in support of the Fort St. Vrain defueling and decommissioning may differ from that outlined in this plan due to results of further planning to expedite defueling, component removal and dismantlement, or as a result of advances in technology allowing performance of tasks prior to presently scheduled dates.
Remote handling and packaging systems, the Hot Service Facility (HSF), and personnel shielding will be utilized to the greatest extent possible during evolutions involving highly activated or contaminated components to minimize personnel radiation exposure in accordance with good health physics principles and ALARA (As Low As Reasonably Achievable) practices.
3.5.1 Reoion Constraint Devices (RCDs)
During the defueling period, the RCDs will be removed from the reactor as defueling proceeds to each fuel region. A total of 84 RCDs will be handled during the defueling process. Final decisions on interim handling and storage of RCDs will be made after completion of detailed defueling time motion studies. It is anticipated that the RCDs will be handled in one or a combination of the following methods:
a) return to the PCRV for storage until decontamination and dismantlement following SAFSTOR.
b) retain in temporary onsite storage (not in the PCRV).
c) disposal as low level radioactive waste.
3-14
l I
l PRELIMINARY I DECOMMISSIONING PLAN SECTION 3 I l
RCDs will be classified for waste disposal according to f 10CFR61. RCDs will be packaged and (depending on waste I classification according to 10CFR61) shipped for disposal or stored until disposal capacity for Greater Than Class C (GTCC) waste is developed. Preliminary analysis indicates that the RCDs are not anticipated to be GTCC waste.
3.5.2 Control Rod Drive Orifice Assemblies (CRDOAs) and Absorber Strinas '
During the defueling period, the CRD0As will be reinstalled in the PCRV for temporary storage prior to di:,posal. After defueling is completed, the CRD0As will be transferred from the PCRV to the HSF using the ATC.
l Absorber strings will be removed from the CRD0A and will be cut to size and packaged in disposal liners inside the HSF. Present analysis indicates the absorber strings are Class C waste in accordance with 10 CFR 61 classification guidelines. If further analysis indicates the absorber strings are GTCC waste, they will be stored onsite until disposal capacity is developed. The disposal liner design allows disposal of two pair of absorber strings per -
liner. Shipment of the absorber string liners for disposal as low-level radioactive waste is anticipated to be performed using (available) licensed shipping casks and well established procedures.
Following removal of the absorber strings, each CRD0A will be packaged in an appropriate container, and shipped for disposal as low-level radioactive waste. Disposal of the CRD0A and absorber strings together is not considered feasible because licensed casks with sufficient shielding and capacity to accommodate the size of the entire CRD assembly are not economically available to offset the methods identified above.
PSC has considerable previous experience in handling CRD0As, sectioning absorber strings and packaging them for disposal, and disposing of orifice valve assemblies.
3.5.3 Helium Circulators After the shipment of the CRD0As is completed during the Component Removal Period, the helium circulators will be removed for disposal and shipment.
Circulators will be removed during this period due to the availability of experienced people and necessary equipment (including the circulator removal turntable), and proven procedures will also be available. The choice has also been influenced by the fact that removal is relatively simple and straightforward, and the disposal costs are known.
3-15
1 l
l PRELIMINARY !
DECOMMISSIONING PLAN SECTION 3 Suitable disposal techniques will be used to prepare the contaminated - portions of the circulators for shipment, including dismantlement of non-radioactive portions of the mechanism and decontamination of radioactive portions of the circulator.
3.5.4 Remainino Plant Primary and Secondary Systems 3.5.4.1 Overview At some- point after final reactor shutdown, all plant systems external to the PCRV which are not required to support spent fuel storage in the FSWs will be radiologically surveyed to identify those systems which are contaminated with radioactive materials.
Systems identified to be contaminated will be evaluated to determine the best available option ,(consistent with regulatory guidance and ALARA principles) for either: (1) decontamination; (2) abandonment; or (3) dismantlement and disposal as radioactive waste. Portions of the purification system, decontamination system, liquid waste system, gas waste systen, and other radioactively contaminated systems identified in the future will be treated in this manner. Each contaminated system to be dismantled will be evaluated to determine if decontamination can be accomplished. Those materials and -
components identified for low level waste disposal will be packaged, classified in accordance with 10 CFR 61 criteria, and shipped for disposal to an approved disposal facility. Temporary decontamination facilities will be utilized wherever practical . Uncontaminated systems will be isolated (mechanically and electrically) and abandoned after they are no longer required for shutdown operations.
Due to the high radiation and contamination levels present inside the PCRV following shutdown, the core support structure (blocks and posts), steam generators, RCDs, metal clad blocks, removable reflectors, transition blocks, dummy blocks, large side reflectors, boronated spacer reflectors, core barrel, and other internal PCRV components are not included in the initial component removal, but will be stored for decay inside the PCRV during the SAFSTOR period.
3.5.4.2 Systems Required During Defueling or SAFSTOR During the Defueling Period, decay heat removal from the core will be maintained by providing primary and secondary coolant flow as needed. Therefore, many plant systems must remain operational. Once the last spent fuel segment has been defueled to the FSWs, those systems determined to be necessary will remain operational to support spent fuel cooling in the FSWs.
3.5.4.3 Systems Not Required During Defueling or Component Removal As systems are identified which are no longer required for defueling or component removal activities, options for the disposition of each system will be evaluated. Dependent on 3-16 m_ __._ _ __m . _ _ _ . - _ . . - - - - - - - - - - - - - - - - - - - - - - - - - " "
\
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 establishment of a limit for residual radioactivity, contaminated systems outside of the PCRV will be abandoned, dismantled, or decontaminated. Contamination surveys performed in the future will determine to a large degree the final option selected for each system. The helium purification system, decontamination system, liquid and gas waste systems, fuel handling systems and components, FSWs, ESWs, HSF, and ATC will each require evaluation to determine final disposition.
It may be determined that selected plant systems may be required during the dismantlement period following SAFSTOR. Those systems identified will be flushed, isolated (mechanically or electrically) and pl aced in a lay-up condition to be available following the SAFSTOR period. The Proposed Decommissioning Plan will contain greater detail on the disposition of plant systems than is available at this time.
3.5.4.4 Temporary Systems In some cases, temporary systems may be required to replace an existing permanent system as that system is removed from service.
One example could be a temporary system designed to process -
radioactive fluids after the plant liquid waste system is removed from service. Each temporary system will be established as required, and only as needed. Emphasis will be placed on vendor-obtained skid-mounted systems, or self-contained units to be used during the interim period they are required on site, then removed to be returned to the vendor. This method would minimize radioactive waste generation and shorten the time required for decommissioning activities.
3.6 SAFSTOR ISSUES PSC intends to maintain the PCRV in a S'AFSTOR condition for a period of up to 55 years to allow sufficient radioactive decay of activated components. The primary factor in selecting a 55 year SAFSTOR period is to allow radioactive decay of highly activated components inside the PCRV. Allowing these activated components to decay for 55 years will reduce the amount of activated material that remains above limits for unrestricted use, such that it will reduce the amounts of activated PCRV concrete that will require removal and will reduce the exposure of occupational workers.
The following plant conditions subject to control under the 10 CFR 50 license are expected to be present during the SAFSTOR period.
Further detail on each of these items is provided in the following paragraphs.
- a. Segment 9 Spent Fuel:
- As previously noted, this spen.t fuel will be stored in the FSWs or in an ISFSI for an interim period during the SAFSTOR period. If the spent fuel is stored in the FSWs, 3-17
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 the required support systems will remain in effect to support spent fuel storage, including decay heat removal, fire protection, security, emergency response, and spent I
fuel handling capability.
- b. PCRV:
- Since PCRV internals will remain radioactive during the entire SAFSTOR period, they will be isolated from any possible access; penetrations into the PCRV will be seal welded shut to provide a barrier between the radioactive components and the environment.
- The PCRV internal cavity will be in an air atmosphere, at or near atmospheric pressure during SAFSTOR. With these conditions, there will be no need to maintain the top head penetration hold-down and cover plate system.
- Steam generators, and other internal PCRV components (with the exception of the helium circulators and the CRD0As) will remain in place during the entire SAFSTOR period.
- Suitable security, emergency response and fire protection capabilities will be maintained consistent with 10 CFR 50 requirements. l
- A modified environmental monitoring program (primarily -
restricted to on-site activities) will remain in effect throughout the SAFSTOR period.
- Routine radiological surveys will be performed inside the reactor building and its surrounding area onsite to detect any potential release of radioactive materials.
- Necessary monitoring and surveillance activities will be performed during the SAFSTOR period to protect the health and safety of the public.
- The 10 CFR 50 license will remain in effect (as a
" possession only" license) throughout the SAFSTOR period.
- c. Systems and Site Areas External to the PCRV: 4
- It is currently anticipated that all other radioactive material and systems external to the PCRV will be decontaminated or dismantled and removed from the site during the Component Removal Period.
- Systems which are required to remain operational will be appropriately prepared for weather conditions and appropriate surveillance tests and maintenance activities will be established to ensure operability.
- Building support and service systems will be reduced to those minimally required to support SAFSTOR conditions.
In order to prepare Fort St. Vrain for SAFSTOR, several activities must be completed. Primary and secondary systems which are not radioactively contaminated and are not required to support fuel storage must be isolated and placed in a layup condition. The reactor building then must be secured as defined by the SAFSTOR Security Plan, which will be submitted in conjunction with the Proposed Decommissioning Plan.
3-18
' PRELIMINARY DECOMMISSIONING PLAN SECTION 3' I i
A minimum of two barriers will remain in effect during the SAFSTOR period to prevent the inadvertent release of radioactive I materials from the PCRV to the environment. The primary (inner) barrier will consist of the liner of the PCRV, together with a sealed barrier at all penetrations (either primary or secondary closures, or both in some cases, such as the steam generator penetrations). A secondary barrier will be provided by the thick prestressed concrete surrounding the liner which has the ability to adequately limit the rate of leakage from a possible failure of the liner. In addition to the PCRV concrete, secondary barriers for penetrations through the PCRV concrete and liner will be provided by seal welding (or equivalent) to provide a semi-permanent barrier. Therefore, during the SAFSTOR period, at least two barriers will be available to isolate radioactive materials inside of the PCRV from the environment.
Following defueling and component removal, a fuel storage period will exist, during which many systems are required to support storage of the spent fuel in the FSWs. These systems are needed to remove decay heat, unload the fuel from the FSWs and provide HVAC to the reactor buildings. Many of these systems are sized and designed to support operation of the FSWs. PSC is currently evaluating -
possible modiNcations to increase system efficiency. Details of proposed modit: cations will be included in the Proposed Decommissioning Pla.7.
Plant systems foreseen to be required during SAFSTOR (as a minimum) are for fire protection and electric power. These systems are currently designed and sized to support normal and accident conditions at an operating nuclear power plant and, as a result, are more complex than required to support the SAFSTOR conditions. It is predicted that these systems will require modification. The extent of these modifications is unknown at this time and is intended to be provided in future submittals, such as a revision to the Fire Protection Program Plan. These plant modifications will be designed in the future and will be controlled by a suitable design control process. All applicable Technical Specifications will also be j revised at a future date, as required.
l 1
The 10 CFR 50 license requirements during the SAFSTOR period will be addressed in the Proposed Decommissioning Plan. These will include revisions to the Fort St. Vrain Quality Assurance (QA) plan, Radiological Emergency Resporise Plan (RERP) Plan, Fire Protection Program Plan, Environmental Monitoring Program, Security Plan and Technical Specifications. The staffing levels required during the SAFSTOR period will be defined in the Proposed Decommissioning Plan.
3.7 DECONTAMINATION AND DISMANTLEMENT PLANS SAFSTOR will end approximately 2043. At that time, all remaining radioactive material at the Fort St. Vrain site will be removed and the contaminated portion of the PCRV will be dismantled 3-19
_ _ . _ _ _ _ _ _ - - _ - - _ _ - - - _ - - - - - - - - + - - -
I PRELIMINARY DECOMMISSIONING PLAN SECTION 3 I or decontaminated. Although a residual radiation release criteria d has not been selected, preliminary plateout and activation analyses predict that the following PCRV components may require removal and disposal as radioactive waste during the eventual dismantlement period:
3
- dummy blocks
- graphite reflector I
- metal clad reflector
- boronated spacer blocks core barrel
- insulation and cover plates <
- core support blocks
- posts and silica blocks
- core support floor
- steam generators PCRV liner and
- RCDs adjacent concrete l
All other remaining components are predicted to be at sufficiently low radiation and/or contamination levels to allow release for unrestricted use, or require only minor decontamination prior to release for unrestricted use.
The Dismantlement Period is predicted to take approximately 3 years. By the year 2046, the FSV plant is anticipated to be at a level which will allow release of the site for unrestricted use and -
termination of the license. Although the methods for removal of the above components are not finalized at this time, preliminary dismantlement analyses performed by PSC indicate that the technology exists to allow remote dismantlement at that time. <
l PSC recognizes that the site cannot be released for unrestricted use until all radioactive materials are removed to the predetermined limit for residual radioactivity. A detailed Decontamination Plan has not been prepared for submittal with this Preliminary ' Decommissioning Pl an. Since decontamination and 1 dismantlement of the PCRV are not anticipated to occur until l approximately 2043, it would be premature to detail the decontamination and dismantlement methods or the specific organization to be used. However, PSC has performed dismantlement studies in sufficient detail (utilizing existing technology) to define a methodology for dismantlement and decontamination of the l PCRV. A detailed PCRV Dismantlement Plan and Fort St. Vrain site {
Decontamination Plan will be developed and submitted prior to '
commencing dismantlement activities associated with termination of the 10 CFR 50 license.
A conceptual plan for dismantling the PCRV and internal !
components has been prepared and is presented below. This conceptual {
plan is an element in the basis for the cost estimate for this Preliminary Decommissioning Plan. Since decommissioning ]
and 1 decontamination technology is expected to change dramatically over {
the next few years, details of any final plan will be submitted to l the NRC prior to start of the PCRV decontamination and dismantlement. j Experience indicates that further development work is desirable to '
improve the equipment and techniques, develop better remotely 3-20 {
I
-- _ - - . - - - - - - - - _ - 1
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 operated equipment, reduce costs and exposures and gain necessary experience. Experience has shown that, when considering the application of a technique, especially a complex technique, to the dismantling of nuclear facilities, expert advice on the selection and application of the right equipment or technique is essential. In addition, the equipment should undergo lengthy testing and operators should be well trained in a non-radioactive environment before the equipment is used in radioactive applications.
-The following scenario represents a generalized, but feasible. plan which could be used to decontaminate and dismantle the PCRV a, the remaining core components in a safe manner. This discussion is provided for informational purposes only. The following are the major components of the conceptual plan which will be discussed in detail in the following paragraphs:
- 1. Remove core drillings for radioactivity mapping
- 2. .- Je loose contamination from PCRV internal surfaces
- 3. hemove PCRV tendons
- 4. Remove refueling penetrations
- 5. Segment and remove PCRV top head
- 6. Chemically decontaminate lower steam generator internals -
- 7. Install Manipulator Positioning Fixture (MPF)
- 8. Remove graphite reflector blocks and dummy fuel blocks
- 9. Package graphite reflector blocks for disposal
- 10. Dispose of dummy fuel blocks
- 11. Remove side reflector block restraints
- 12. Remove side reflector blocks
- 13. Remove boronated spacer blocks
- 14. Remove core support blocks
- 15. Remove core support posts
- 16. Segment and remove core barrel
- 17. Remove silica blocks
- 18. Remove liner insulation plates
- 19. Segment and remove core support floor
- 20. Gouge grooves in liner
- 21. Section and remove PCRV liner and portions of PCRV concrete walls
- 22. Install rigging and handling equipment for steam generator removal
- 23. Dispose of steam generators
- 24. Perform final radiation survey of PCRV 3.7.1 Remove Core Drillinos for Radioactivity Manoina Based on current estimates, disposing of the volume of activated concrete will be a major portion of the PCRV dismantlement cost. To establish the actual amount required for removal, proper characterization of the PCRV's concrete walls will be performed.
Analysis of concrete core drill samples comprising a Sngitudinal profile of the PCRV wall will provide sufficient 3-21
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 radiological information to locate the lire of demarcation between activated and non-activated concrete. Core drilling operations associated with the removal of PCRV upper head and cere support floor will provide core samples for similar radiological evaluations to estimate concrete volumes.
3.7.2 Remove loose Contamination from PCRV Internal Surfaces A portion of the work required to remove PCRV internals will require worker entry into the PCRV volume. Background radiation levels inside the upper cavity of the vessel (due to internal core components) are initially too high to permit manned entry to remove loose surface contamination from the PCRV internal surfaces.
Consequently, extensive decontamination operations, first by remote means and later by manual effort, will be necessary.
The majority of loose contamination in the upper cavity will be removed with high-pressure hot water, using remotely manipulated spray heads. Follow-up decontamination operations will be manually performed with ultra-hign-pressure cleaning equipment operated from scaffolds within the PCRV.
After the upper cavity liner and activated concrete walls are removed, a final ultra-high-pressure water cleaning of the lower cavity surfaces will reduce loose surface contamination to releasable levels.
3.7.3 Remove PCRV Tendons Some tendon removal is required to facilitate the cutting of the PCRV concrete. The tendons are embedded in the PCRV, which shields them from irradiation by the reactor active core. Therefore, the tendons are not anticipated to be radioactive.
Tendon removal is accomplished as follows: (1) tendons are mechanically detensioned; (2) tendons are pulled from the tendon tube embedded in the concrete; (3) the tendon is sectioned into manageable lengths as it is being removed from the PCRV concrete using plasma arc cutting equipment; and (4) final transfer of the tendon sections to a designated storage area for disposal as clean scrap.
3.7.4 Remove Refuelino Penetrations The PCRV head contains 37 refueling penetrations (22-inch 0.0.)and a top access penetration (61-inch 0.0.). These penetrations are lined with alloyed steel. The penetration liners and adjacent concrete are likely to be radioactive from neutron activation and contaminated from primary coolant plateout.
The penetration liners will be remcVed after a series of preparatory operations described below are completed (Figure 3-4):
3-22
i PRELIMINARY DECOMMISSIONING PLAN SECTION 3 DVERHEAD H015T i
i
(, , CORE DRILL TO HYORAULICS
% TO LIQUID 1-,t a ,-
REmVAL
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i marmos
- rocinaras LIQUID CATCH DEVICE LIQUID REMOVAL
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-i B TIIll 11 II II Il ll II Il Il 11 11 Il 11 11 Il 11 11 l asc mecart rum Figure 3-4 Core Drilling Operations 3-23
1 PRELIMINARY j DECOMMISSIONING PLAN . SECTION 3 1
L .1. The upper reactor cover plate is a 31-foot diameter, 0.75-inch l thick steel plate, weighing approximately 11 tons. It is bolted to the upper extensions of the top head penetration liners. It will be removed by detaching the plate-to-liner fasteners and transferring it from the top plate area to the refueling floor. The plate will be radiologically evaluated, decontaminated if necessary, and plasma arc cut into segments {
sized to permit transport by truck.
l 2. Neutron shielding in the form of I-inch thick boronated l polyethylene sheets surrounds the upper extensions of the refueling penetrations. The shielding must be removed manually to provide access to the penetrations.
- 3. A concrete curb attached to the PCRV head extends vertically above the head surface. It constitutes a 1-foot thick concrete annulus that is 2-feet high and 31 feet in diameter; its total weight is approximately 14 tons. The curb will be removed to improve access for removal of the upper penetration l siners. It will be cut free from the head with a diamond wire cutter, segmented into quadrants for handling purposes, and transferred to the appropriate storage area for disposal. -
- 4. Removing the penetration liners will direct demolition debris downward into the PCRV. This debris can be trapped with a catch device, such as a net suspended in the PCRV below the head. Remote tooling will be used to insert and install the catch device through the refueling penetrations.
- 5. Each refueling penetration liner has an upper section that extends the liner from the top head to the upper reactor cover pl ate. These sections must be removed with plasma arc cutting equipment before the penetration liners can be removed from the top head.
Once the above steps are taken, the (39) penetration liners can be removed by core drilling, using a 24-inch diameter, water-cooled diamond core drill . The drilling stroke will be terminated when the steel liner on the top head's inner surface is penetrated. The resulting core will be removed and transferred to the radioactive waste processing area.
3.7.5 Seament and Remove PCRV Too Head i To provide sufficient access to the interior of the PCRV, portions of the head will be segmented using diamond wire cutting equipment and removed in stages (Figure 3-5).
The thermal barrier, consisting of alternate layers of insulating material and steel plate, is fastened to the head liner.
It will be removed with a remotely operated pole-mounted manipulator inserted through the refueling penetrations.
3-24 l
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 t
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-[ M 11 11 Il 1111 Il 11 Il Il 1111 il 111111 Il Il 11 Il ces samt noon Figure 3-5. Removal of Concrete Cutout 3-25
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 PCRV head segmentation will be initiated by cutting planes between refueling penetrations. By lacing the diamond cutting wire down through selected penetrations and upward through other penetrations, wire loops are formed around concrete. The abrasive action of the moving wire makes a cut between the penetrations. The process is repeated until a segment is freed from the head. The segment will be rigged and transferred to a staging area on the refueling floor. This process is repeated until the radioactive portions of the head are removed. For some cuts, refueling penetration holes do not exist; therefore, vertical holes will be core drilled for wire placement. Once the head segments are transferred to the refueling floor staging area, the lower activated portion of the concrete will be separated from the non-irradiated concrete with a diamond wire cutter. l A
3.7.6 Chemically Decontaminate Lower Steam Generator Intarnals The internals of the steam generators on the primary side are to be chemically decontaminated prior to removing the steam generators. A special closure device will wrap around the steam generator and cover the primary exit flow holes, which are located below the lower seal' bellows and above the primary closure. This -
device will provide a water-tight seal against the outer steam generator surface and will connect to the closure device via hoses.
Chemicals used for decontamination will be selected based on their cleaning effectiveness and the waste form they generate. Mixed waste forms are undesirable because of disposal problems and can be avoided by proper selection of chemicals. Reuse of solution for each subsequent steam generator will help to minimize the liquid waste volume.
3.7.7 Install Manipulator Positioning Fixture (MPF)
II, the defueling operation, the 37 regions of the reactor will be defueled. In regions 2 through 37, the fuel blocks will be replaced with boronated dummy blocks. A manipulator will be developed to remove the center region (Region 1) of dummy blocks.
The remote tooling assembly is the manipulator positioning fixture (MPF), comprising a mast with anchoring foot plates and a rotating cantilevered beam with movable carriage. The MPF and its ancillary equipment will be located on the reactor building refueling floor. The inactive refueling control room will be modified to accommodate the MPF control equipment, including control consoles and cabinets, computer, CCTV monitors, and their associated controls.
The hydraulic power supply units will be positioned on the refueling floor and interfaced with the control room equipment.
To place the remote operatSn equipment within the PCRV, the MPF will be lowered through the top head tent enclosure. Its foot will be fastened to the graphite floor blocks, with its placement and attachment closely inspected with remote CCTV cameras. After 3-26
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 attachment of the foot, the upper end of the MPF will be anchored.
Every installation step will first be performed at a mockup facility to facilitate the actual installation into the PCRV.
3.7.8 Remove Graphite Reflector Blocks and Dummy Fuel Blocks The manipulator, attached to the MPF, will be used to position a latching device to allow a hoist to remove each block from its position in the reactor core (Figure 3-6). Each block'will be placed in a drum for removal. This procedure will be repeated until all reflector and dummy fuel blocks have been removed.
3.7.9 Packaoe Graohite Reflector Blocks for Discosal PSC has considerable experience in safely removing used reflector blocks, packaging them in steel drums, and disposing of them as radioactive waste. Void spaces within the drums will be filled with solidified radioactive waste generated from liquids and solids. This operation will be conducted within the tented work enclosure located on the refueling floor of the reactor building.
Local . shielding will be utilized to attenuate work area dose rates and to achieve acceptable contact radiation levels for shipping -
containers.
3.7.10 Discose of Dummy Fuel Blocks l
The dummy fuel blocks will be placed in the reactor core to maintain core geometry during reactor defueling. Following removal from the core, provisions will be made for disposal of the radioactively contaminated portions of the dummy blocks. Appropriate volume reduction techniques will be used in an attempt to reduce the volume of radioactive waste generated.
3.7.11 Remove Side Reflector Block Restrain 11 .
The uppermost layer of permanent graphite blocks is held in place with restraints fastened to the core barrel upper ring with bolts threaded into the ring. These restraints must be removed to permit removal of the graphite blocks. The bolts holding the restraints in place are locked by metal tabs that must be bent to permit removal of the bolts. This task will be performed using the manipulator arm and appropriate txling.
The bolts will be loosened aid disengaged from the upper ring of the core barrel, freeing the restraints for removal. A waste disposal liner will be placed on the floor of the PCRV. A latch fixture designed to pick up and release the reflector restraints will ,
be suspended from the MPF hoist. As the restraints are removed, they will be loaded into the liner, one at a time. The liner will be decontaminated while exiting the top head and will then be placed in temporary storage.
3-27
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PRELIMINAP' DECOMMISSIONIhL PLAN SECTION 3 3.7.12 Egmove Permanent Side Reflector Blockt There are a total of 288 permanent side reflector blocks in the PCRV, each weighing approximately 1500 pounds. They comprise four interlocking shapes, and they will be removed, one at a time, and placed in a liner, which will accommodate six blocks. The lifting er.gagement tool will attach to a pre-selected reficctor block and the hoist will transfer the block to the prestaged liner. After the liners are loaded, they will be removed from the PCRV with the reactor building crane, and transferred to a waste processing tent.
The liners will be decontaminated as they pass through the top head.
3.7.13 Remove Boronated Stacer Blocks A total of 1,152 boronated spacer blocks are positioned between the permanent side reflector blocks and the core barrel.
Each weighs approximately 100 pounds. Multiple 0.875-inch diameter holer penetrate the upper surface of each block; each hole contains stainless steel dowels which have a boron concentration of 2 to 2.5 percent.
The spacer blocks will be removed from the PCRV in much the ~
same way as the permanent side reflector blocks. Proper sequencing will require removal of four spacer blocks for each permanent side reflector block.
3.7.14 Remove Core Support Blocks The core and reflector region is supported by a series of graphite core support blocks and core support posts laterally restrained by the core barrel (Figure 2-8). Each block is interlocked with adjacent blocks and keyed to surfaces above and below by the positioning posts.
The MPF will be rosting on a number of these support blocks and must be relocated to permit their removal (Figure 3-7). The blocks will be transferred, one at a time, into a liner, utilizing the reactor building crane, the electric hoist, and manipulator arm, as needed.
3.7.15 Remove Core Sucoort Posts As the care support blocks are removed, the graphite core support posts used to position tb blocks will also be removed. As a block is removed, its three posts become available for extraction. A special lifting device will be slung from the hoist, and with the assistance of the mafiipulatu arm, the three posts will be lifted one at a time and transferred into the liner that was utilized for receiving the core support blocks.
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PRELIMINARY DECOMMISSIONING PLAN SECTION 3 3.7.16 Seoment and Remove Core Barrel i
The core barrel is a steel cylinder 29 feet high and 27 feet in diameter; its wall thickness varies from 2.25 inches to 2.75 inches. :The total weight of the barrel is approximately 126 tons. 1 After removal of the reflector and spacer blocks and lower r. ore support blocks, ~ removal of the core barrel will begin. l The core barrel comprises three vertical regions: the upper, middle, and lower rings. These three regions will be segmented using j a remotely operated plasma arc torch. The core barrel will be cut into 48 vertical segment;;, and then cut horizontally at the ring joints (Figure 3-8). Each segment will be rigged as it is detached and then stored temporarily inside ti,e PCRV. Later, these segments j will be nested in bundles,, banded, and packed for disposal.
3.7.17 Remove Silica Blocks i
The core support floor is covend with a thermal barrier which consists of three layers of silica blocks and two layers of Kaowool.
The inner surface on the lower part of the core barrel is also '
thermally protected with a layer of Type B insulation. After -
removing the core support blocks and support posts, the manipulator will remove the thermal barrier from the core barrel. After the core barrel is removed from the cavity, the silica blocks and Xaowool insulation will be manually removed from the core support floor.
3.7.18 Remove Liner Insulation plates
> The removal of the core barrel will expose the thermal barrier, which is attached to the PCRV liner. The barrier comprises two layers of partially compressed insulating material, one layer of thin steel sheetirig, and an outer layer of 0.25-13ch thick steel plate, all fixed to the PCRV liner with studs. ,
The radiation levels within the PCRV at this stage will be low enough to permit manual removal of the thermal barrier. Work platforms will be installed to provide worker access to the inner PCRV surfaces. Decontamination of the exposed surface of the insulating region will be accomplished with ultra-high pressure washing techniques. The wash water will be collected in the lower cavity of the PCRV for processing.
The stud fasteners that hold the barrier in place will be removed with power tools. The two steel layers will be removed by hand and placed in containers for removal. After removal of the steel, a wetting agent will be applied to the exposed insulating material to control dust generation. The insulation layers will then be removed observing suitable occupational precautions. The bags and steel coverings will then be transferred out of the PCRV for waste I
processing. l 3-31 l
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' PRELIMINARY-DECOMMISSIONING PLAN' SECTION 3 The work generally will progress from top to bottom, utilizing h!PA-filtered vacuum cleaners to . control the spread of both radioactive contamination and insulation fines. As the fastming studs are uncovered, they will be removed with a plasma arc cutter.
l Work in-the lower cavity will proceed.in parallel with that in the upper cavity. Some insulation adjacent to the core support floor !
will be removed after the core support floor is removed from the O PCRV.
3.7.19 Seament and Remove Core Suocort Floor The core support floor is an approximately 5-foot thick, steel-lined concrete disc with 12 vertical holes. 'The floor will be segmented with a diamond wire cutter and transferred to the refueling floor. The radioactive portions will then be separated from the non radioactive materials with a diamond wire cutter.
3.7.20 Gouoe Grooves in Liner The PCRV liner . is 0.75-inch thick steel. As described -
previously, the steel liner will be removed with the activated concrete. In preparation for concrete cutting, the steel liner will be gouged to provide unencumbered concrete cutting paths for the diamond wire cutting process.
One to 2-inch wide gaps will be gouged in the liner by utilizing air-arc cutting technology. Circumferential1y complete horizontal cuts will be made near the core support floor and near the PCRV horizontal centerline. A series of vertical cuts, extending from the lower horizontal cut to the top of the liner, will complete the cutting pattern in the liner. Because the liner is a source of radioactivity, all cutting operations will require . continuous air processing with HEPA filtration. y 3.7.21 Section and Remove PCRV Liner and Concrete Walli The steel liner and adjacent concrete in the PCRV walls are activated. The depth of activated concrete in the PCRV walls varies over elevation and with time after shutdown. The radioactive portions of the PCRV walls will be removed as slabs cut away from the non-radioactive concrete. A diamond wire cutter will be used to cut a ring around the activated concrete in the PCRV wall, and will make vertical and horizontal cuts on the ring to form the individual slabs. Two horizontal cuts will be made: one at the core support floor elevation, and a second one at the mid-height of the reactor cavity. Vertical cuts will be made between tendon tubes. Before the diamond wire cutter is used, gaps will be cut through the liner by ,
air gouging. These gaps provide guide paths for cutting concrete with the saw.
3-33
1 l
1 PRELIMINARY DECOMMISSIONING PLAN SECTION 3 1 i
Vertical tendon tubes are located in a ring approximately 2-1/2 feet outboard of the inner surface of the PCRV wall. These {
tubes will be used to radially position the diamond cutting wire. i Horizontal gaps will be formed for wire manipulation by circumferential cutting operations, first with air gouging at the steel liner and then with wall saws at the core support floor-PCRV wall interface and also at the approximate mid-height of the reactor {
cavity. These two cuts will be made deep enough into the concrete wall to intercept the inner ring of tendon tubes mentioned above, and will be used to lace the cutting wire from tube to tube, permitting completion of the wire cutting loop. The cut will then be made between two adjacent tendon tubes. A total of 42 vertical tendon tubes will be utilized to cut the ring.
The resulting concrete slabs will be transferred to the refueling floor, where additional diamond wire cutting will separate radioactive from non-radioactive concrete. The resulting segments, rubble, and cutting residue will be segregated, processed, packaged,and transported to appropriate holding areas.
3.7.22 Install Ricaina and Handlino Eouioment for Steam Generator -
Removal The steam generators will be disconnected from the core support floor and maintained at their location for removal from the PCRV. The rigging required will depend on the capabilities of the existing rigging systems. Once rigged and braced in place, each steam generator will be det ched frc9 the core support floor by plasma-arc cutting.
3.7.23 Dispose of Steam Generators Following removal of the core support floor, each steam generator will be lifted from the PCRV using the reactor building bridge crane. In the radioactive waste processing tent, the steam generators will be decontaminated by ultra-high pressure water or chemical techniques to achieve unconditional release. If not !
achieved, the steam generators will be disposed of as radioactive I waste.
Access to internal surfaces will be attained by segmenting with plasma-arc c. utter. The resulting decontaminated pieces will be .
placed in appropriate containers and/or rigged from the refueling l floor and transferred to the ferrous metal scrap pile. Any activated portions will be cut away and discarded as radioactive waste.
3.7.24 Perform Final Radiation Survey of PCRV i l
Following completion of all decontamination / dismantlement;. 1 tasks included in the scope of the project, a final radiation survey l of the site will be conducted to verify that residual radioactivity 3-34
PRELIMINARY DECOMMISSIONING PLAN SECTION 3 levels will permit its unrestricted ' release. In the event the sampling program indicates that radioactivity levels are not within ;
acceptable limits, appropriate corrective measures will be taken, i 3.8 DECOPHISSIONING COST ESTIMATE As described previously in this section, there are three major groupings of activities in the decommissioning schedule:
- 1. Defueling and component Removal Periods to prepare the site for.the SAFSTOR Period;
- 2. SAFSTOR Period, including SAFSTOR with Fuel Storage, Fuel Shipment, and SAFSTOR without Fuel Storage; q
- 3. The final Decontamination and Dismantlement Period. j 1
The cost estimate for these three periods in 1989 dollars is !
provided in Table 3-3. This cost estimate reflects the use of information from several sources, including actual experience, in-house studies, vendor cost estimates, and past consultant studies.
A contingency (normally 25%) has been added to all estimates to cover unexpected costs.
Although not' . directly related to the decommissioning activities defined in the Decommissioning Final Rule, PSC has committed to submit information concerning PSC's financial plans to support the defueling of Fort St. Vrain. (Ref. 3.6). In addition to those decommissioning costs identified in the preceding paragraph, the cost estimate for defueling reflects the cost of removing the fuel i from the reactor, storing Segment 9 in an onsite ISFSI through 2019, {
and shipping Segment 9 to a Federal HLWR in 2020. These costs are j presented in the right-hand column of Table 3-3. I l
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DECOMMISSIONING PLAN' SECTION 3 TABLE 3-3 DECOP90SSIONING COST ESTIMATE (Costs in $000, Expressed in 1989 Dollars)
DECOMMISSIONING DEFUELING/ FUEL RELATED COSTS COSTS I.. DEFUELING AND COMPONENT REMOVAL PERIODS (1989 - 1993):
Labor $12,799 Labor $51,812 Overhead 5,513 Overhead 21,122 Burial Costs 1,036 ISFSI 17,849 System GA Facility '9,625 Modifications 213 Seament 10 5.000 TOTAL COSTS: $19,561 $105,408 II. SAFSTOR PERIOD (1993 - 2043):
- a. SAFSTOR With Fuel Storace (1993-2020). Tynical Annual Costs:
Labor $384 Labor $305 Overhead 222 Overhead 95 ANNUAL TOTALS $606 $400
- b. Fuel Shinoino in Year 2020:
Labor $384 Labor $600 Overhead 222 Overhead 95 TOTAL IN 2020 $606 $695
- c. SAFSTOR Without Fuel Storaae (2021-2043). Tvoical Annual Costs:
Labor $187 Overhead 239 TOTAL PER YEAR $426
- d. Total SAFSTOR Costs:
DECOMMISSIONING FUEL RELATED SAFSTOR COST: $26,766 SAFSTOR COST: $11,495 III. DECOMMISSIONING / DISMANTLEMENT PERIOD (2043 - 2046):
Labor $15,867 Overhead 13,319 Burial Costs 5.384 TOTAL COST $34,570 IV. TOTAL COSTS: 1
- l. -' DECOMMISSIONING DEFUELING/ FUEL RELATED TOTAL COSTS: $80,897 COSTS $116,903 f 3-36
L ll:
PRELI.MINARY I DECOMMISSIONING PLAN SECTION 3 REFERENCES FOR SECTION 3 3.1 The As Modified 3-Party Agreement, Document Number 34426, dated July 1,-1965. j 3.2 DOE letter, K.R. Hastings (DOE) to R. Husted (PSC), dated May t 11, 1988 (G-88163);-
Subject:
" Modification to Contract AT(04-3)-633: DOE Commitment to Receive FSV Spent Fuel". ]
J 3.3 General Atomic Report No. GA909658 (Initial Issue) "FSV Platecut Analysis for Decommissioning Study", dated April 15, 1988.
3.4 USAEC Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors," June 1974.
3.5 PSC Interoffice Memorandum (NDG-89-0189), V. Walker to M.
Fisher, "FSV Activation Analysis Results", dated February 21, 1989.
3.6 PSC Letter, R.O. Williams (PSC) to J. Calvo (NRC), dated December 5, 1988;
Subject:
"Early Termination of FFort St. -
Vrain Operations", (P-88422).
3-37
) SECTION 4 RADI0 ACTIVE WASTE ISSUES 4.1 HIGH-LEVEL RADI0 ACTIVE WASTE 4.1.1 Status of the Permanent Hiah-level Waste Repository No disposal capacity for high-level radioactive waste currently exists in the United States. DOE has selected the Yucca Mountain site in Nevada for characterization and possible licensing as a HLWR. Operation of a HLWR for disposal of spent reactor fuel is currently scheduled to begin in 2003, according to DOE's revised mission plan. A ten year queue of spent fuel currently exists for the HLWR which has priority over any Fort St. Vrain spent fuel.
Spent fuel from Fort St. Vrain will not be eligible for inclusion in this queue until the reactor is shutdown and the last segment It is conservatively estimatedof spent fuel is defueled to the FSWs.
that PSC will be required to store the last segment of spent fuel until 2020.
4.1.2 Hiah-level Radioactive Waste Manaaement Plan As noted in Sections 3.2 and 3.3, five segments of spent reactor fuel removed from the reactor will be shipped from Fort St.
Vrain to a storage facility in Idaho operated by 00E during the defueling period. The remaining segment will be stored in the FSWs licensed under 10 CFR 50 or in an ISFSI.
A comprehensive High-Level Radioactive Waste Management Plan will be implemented which will include the following, as a minimum:
- a. A personnel training program to include the following:
- Spent Fuel Handling Precautions and Requirements
- Radioactive Waste Shipping Requirements
- Requirements for Operation of an Independent Spent Fuel Storage Installation (if built)
- Operation of the FSW's (if no ISFSI)
- b. Procedures for handling, packaging, inventory, and storage of spent fuel,
- c. Procedures for the transportation of spent fuel.
It is noted that PSC has considerable experience in this area.
All of the administrative controls are in place and have been demor*trated to be usable.
4-1
l- !
PRELIMINARY DECOMMISSIONING PLAN SECTION 4 4.2 LOW-LEVEL RADI0 ACTIVE WASTE 4.2.1 Status of the Rocky Mountain Low-level Radioactive Waste Comoact j
The State of Colorado's responsibilities in connection with obligations under the Rocky Mountain Low-level Radioactive Waste Compact and the Low-Level Radioactive Waste Policy Amendment Act of 1985 are chronicled below:
- a. The Rocky Mountain Low-Level Radioactive Waste Compact was formed in 1982 to oversee management of radioactive waste disposal in the region.
- b. Colorado has been designated as the next host state for the Rocky Mountain Low-Level Radioactive Waste Compact. It is Colorado's responsibility to develop a disposal site replacing the Beatty facility by January 1,1993.
- c. The Beatty, NV, disposal facility is currently licensed and operational, with sufficient capacity to serve the Rocky Mountain Compact's needs until at least December 31, 1992. -
Due to the high level of uncertainty surrounding this entire issue, many scenarios for resolving this matter may be possible. PSC is committed to cooperate with the State of Colorado in its effort to find a solution to radioactive waste disposal capacity. As of January 1989, Colorado has made limited progress toward developing and licensing a low-level radioactive waste disposal site required to meet the January 1, 1993 deadline.
PSC will continue to identify options and develop alternate plans to maximize waste disposal flexibility durino decommissioning, should a licensed low level waste disposal facility not be available.
Additionally, radioactive waste generated during decommissioning will be appropriately managed such that the health and safety of the general public and occupationally exposed personnel is not jeopardized and the environment is not endangered.
4.2.2 Low-level Radioactive Waste Manaaement Plan A comprehensive Low Level Radioactive Weste Management Plan will be implemented which will include the following, as a minimum:
- a. A personnel training program to include the following: 4
- Waste Volume Minimization !
- Radioactive Waste Classification
- Radioactive Waste Shipping Requirements
- b. Procedure for collection, handling, packaging, inventory, and storing radioactive wastes.
4-2 q
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PRELIMINARY DECOMMISSIONING PLAN SECTION 4 L
- c. A program for compliance with requirements for radioactive waste classification and for verification of waste form compliance in the case of stable wastes.
- d. Procedures for shipping radioactive wastes for disposal.
l .
As noted for high-level radioactive waste, PSC. also has considerable experience in this area. All necessary administrative controls are in place and have been demonstrated to be usable.
1 4.3 OTHER RADI0 ACTIVE WASTE 4.3.1 Mixed Waste !
4.3.1.1 Definition Mixed waste is a general category of waste materials which
, are not only - radioactive, but contain chemical constituents that would by themselves be classified as hazardous under Environmental l Protection Agency (EPA) regulations. I
~ 4.3.1.2 Hixed Waste Treatment -
Currently, there are no facilities in the United States authorized to dispose of mixed wastes. There are, however, several vendors licensed to process certain types of mixed waste; these processes render the waste non-hazardous by definition of the EPA.
Due to the limited number of options for treatment of mixed wastes, emphasis will be placed on avoiding introduction of hazardous material into radioactive contaminated areas, or causing radioactive materials from coming in contact with hazardous substances. Suitable administrative controls will be implemented to achieve control over
- chemicals, solvents, and other hazardous materials to avoid introduction into areas contaminated with radioactive material.
Personnel training and periodic inspections will be performed to assure compliance to prevent creation of mixed wastes during decommissioning. If required, a program to sample radioactive wastes for hazardous materials will be implemented if administrative controls are found to be insufficient or ineffective.
l 4.3.2 Greater Than Class 'C' Radioactive Waste '
4.1.2.1 Definition Greater Than Class 'C' (GTCC) radioactive waste is low-level radioactive waste which exceeds classification limits for near surface disposal as defined in 10 CFR 61.
4-3
h PRELIMINARY DECOMMISSIONING PLAN. SECTION 4 4.3.2.2 GTCC Waste Treatment As noted 'in 10 CFR 61.55, GTCC wastes are generally not acceptable for near surface disposal. On May 18, 1988, the NRC published a proposed rule amending 10 CFR 61 to require disposal of GTCC waste in a deep geologic repository, unless disposal elsewhere has been approved by the NRC. Since no facility currently exists, PSC's preliminary plans will call for all GTCC waste to be stored onsite until disposal capacity is developed. Therefore, disposal of GTCC waste is currently anticipated to occur in conjunction with transfer of the last segment of spent fuel to the Federal HLWR.
4.4 TRANSFER AND SHIPMENT OF UNIRRADIATED FORT ST. VRAIN FUEL Although not within the scope of the NRC Decommissioning Final Rule, the following status of fuel issues affecting PSC and Fort St.
Vrain are provided for informational purposes only, since they significantly affect PSC's long term planning and overall financial exposure.
As noted in Section 3.3, in addition to the irradiated fuel in the core, PSC possessed one segment of unirradiated fuel onsite -
(Segment 10) and unirradiated material (work-in-process) at the GA Fuel Fabrication Facility. This material consisted of up to 97%
highly enriched uranium at various stages of the fuel fabrication process.
All unirradiated fuel and materials have been sold to Nuclear Fuel Services (NFS) of Erwin, Tennessee. All unirradiated material has been removed from the GA Fuel Fabrication Facility; however, PSC will continue to store Segment 10 at Fort St. Vrain until NFS is ready to receive this segment for storage (currently estimated for 1992).
4-4
SECTION 5 RESIDUAL RADI0 ACTIVITY CRITERIA
5.1 INTRODUCTION
A limit for residual i sdioactivity will be established below which radioactive materials weuld not pose a significant threat to the general public or the environment after termination of the facility operating license. Decontamination and disposal activities will remove all components contaminated with radioactive materials above this limit, but will allow materials activated or contaminated below this limit to remain at Fort St. Vrain.
5.2 BASIS FOR LIMIT Guidance obtained from Regulatory Guide 1.86 (Ref. 5.1) is generally sufficient for loose and fixed surface contamination.
However, it does not address materials made radioactive due to neutron irradiation. Therefore, it is anticipated that a limit for the unrestricted release of residual radioactive material will be established and will be based on the annual dose to an individual member of the general public. This basis would be inclusive of all -
conditions identified during decommissioning. When establishing a limit, evaluations will consider radionuclides distribution and abundance, radioactive decay, the method of disposal, interim PSC control over the disposal location, and the pathway through the environment before exposure to the general public.
5.3 METHOD TO ESTABLISH A RESIDUAL RELEASE LIMIT Requirements for establishing this limit will be fully researched and a limit developed prior to submitting the Proposed Decommissioning Plan. The following methodology will be utilized to develop PSC's proposed limit for residual radioactivity.
A limit will be proposed by PSC for the maximum yearly dose to an individual of the general public at any time after the Fort St.
Vrain site is released for unrestricted use. Agreement will then be sought by all agencies regulating the decommissioning of Fort St.
Vrain.
An analysis will be performed to determine the expected annual dose to an individual mernber of the general public based on exposure pathways for various waste streams, isotopic abundance and varying concentration levels of radionuclides, and options for methods of disposal. This information will then be used to determine a functional level for the unrestricted release of radioactive materials. As a result of this evaluation, administrative limits will be established and procedures written to implement the residual 5-1
4; PRELIMINARY-DECOMMISSIONING-PLAN SECTION 5 release limits such :that the annual dose to an individual wi'I' not' exceed the governing ~ limit agreed to by various regulatory authorities.
5.4 SITE RELEASE SURVEYI The plan. for performing final surveys and assessments of the Fort St. Vrain facility, and preparation of documentation to allow termination. of the NRC license and unrestricted release of. the Fort St. Vrain site will be addressed in the Proposed Decommissioning Pl an.
m .
5-2
PREL'IMINARY '
DECOMMISSIONING-PLAN SECTION 5 I REFERENCES FOR SECTION 5 5.1 USAEC . Regulatory Guide 1.86, " Termination of Operating; Licenses for Nuclear Reactors," June 1974. ,
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5-3 i
_ _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ . - _ . . - - - _ _ - _ - _ . . . - _ . .1
PRELIMINARY DECOMMISSIONING PLAN SECTION 6 SECTION 6 FINANCIAL PLAN 6.1 NRC CRITERIA In accordance with 10 CFR 50.75 (b) and (c), a calculation of the minimum requirement to demonstrate reasonable assurance of funds for decommissioning must be made. However, there are no provisions in the rule for calculating a minimum required amount for Fort St.
Vrain as an HTGR. Therefore, PSC has made a site specific cost estimate in order to determine the appropriate certification amount for assuring the availability of adequate funds for the completion of decommissioning through 2046 as prescribed by 10 CFR 50.75 (b). The current PSC estimate, as noted in Table 3-3, is $80.9 million (1989 dollars).
PSC has chosen SAFSTOR as its decommissioning option, which delays completion of decommissioning by including a period of long-term storage. An external sinking fund, as described in 10 CFR 50.75 (e.l.ii), has been established in which to set aside funds on a periodic basis. PSC believes this approach adequately protects the -
public interest while at the same time maintains the financial ;
integrity of the company. Further description of the funding plan is presented in the following sections.
6.2 CORPORATE FINANCIAL HISTORY WITH RESPECT TO FORT ST. VRAIN On September 24, 1986, as a result of negotiations regarding Fort St. Vrain with the CPUC and consumer groups, PSC entered into a settlement agreement with the CPUC and certain other parties, including the consumer groups. The settlement agreement includes ,
removal from rate base of PSC's investment in Fort St. Vrain assets and recovery over five years of $11.5 million of decommissioning costs. As a result of these actions agreed to by PSC, all legal and administrative proceedings relative to the ratemaking treatment of Fort St. Vrain were terminated. The removal of Fort St. Vrain from regulatory rate base has left PSC shareholders responsible for the majority of costs of decommissioning.
6.3 DECOMMISSIONING FUNDING
As described in Section 3.8, " Decommissioning Costs", the cost estimate for decommissioning is $80.9 million in 1989 dollars.
The costs for decommissioning were escalated to the estimated dates ;
of expenditure in order to determine the level of funding that will be necessary over the life of the decommissioning project. The escalation factors used are the long-term projections of the CPIU (Consumer Price Index for All Urban Consumers) as. forecasted by Data Resources Inc. (DRI) with an average of the DRI escalation rates applied beyond the last year of the DRI forecast. The current escalation factors range from 4.9% to 5.6% over the 6-1
i l
l PRELIMINARY' DECOMMISSIONING PLAN SECTION 6 l
i' period of the decommissioning project. The escalator coupled with the 25% contingency (see Section 3.8) should be sufficient to cover the potential for future costs to be realized. The site specific cost estimate escalated to the dates of expenditure is included as Table 6-1. The escalated costs of decommissioning will be updated on a periodic basis as necessary using the current forecast for escalation rates provided by DRI or a similar forecast service, taking into account NRC guidance.
TABLE 6-1 SITE SPECIFIC COST ESTIMATE (Costs Reflect Dollars Escalated to Expected Year of Expenditure)
Dollars in Thousands Component Removal Period (1988 - 1993) $ 22,558 SAFSTOR Period (1994 - 2043) 140,486 Dismantlement Period (2044 - 2046) 615.067
$ 778,111 Future value equivalent to $80.9 Million (1989 dollars) site specific cost estimate.
1 The funding plan for decommissioning has been developed whereby funding will be derived from four sources: recovery of allowed decommissioning costs from customers; annual contributions to the trust fund of $1.76 million per year made from cash operations of PSC; earnings on the trust fund; and increased after-tax cash flow resulting from tax deductions associated with the decommissioning expenditures.
In addition to the $11.5 million of decommissioning funds currently being recovered from customers, $6.2 million (including accrued interest) had been accumulated through the composite depreciation rates in effect prior to the settlement (See Section 6.3). The total balance including accrued interest in the trust fund at December 31, 1988 was $12.6 million. A total of approximately
$6.5 million remains to be recovered over the next 3 years.
6-2
PRELIMINARY DECOMMISSIONING PLAN SECTION 6 In the period beginning in 1989 and continuing through 2043, annual payments of $1.76 million will be made to the trust account to supplement the existing trust balance and ongoing recovery of the remaining $6.5 million from customers. The annual payments will continue until the decontamination and dismantlement period beginning in 2044. The annual payments to the trust fund will be drawn from cash operations of the company.
The funds currently in place and those funds to be placed in the trust fund, together with earnings on the account balance, are sufficient to cover the after-tax decommissioning expenses. The estimated earnings on the average trust fund balance were derived assuming a conservative after-tax return on secure investment instruments of 5.1%. The return factor will be updated on a periodic basis to reflect market conditions.
PSC plans to make use of tax deductions associated with for the annual decommissioning expenditures. Equivalent monies derived from increased after-tax income will be applied to that year's decommissioning expenditures as a direct result of the tax deduction.
The net decommissioning expenditures (after application of income resulting from tax deduction) will be withdrawn from the trust fund. -
The estimated tax rate ranges from 37.3% to 37.63% representing a blended state and federal tax rate.
The funds derived from the tax deduction are: an integral part of the funding plan and should not be considcred a form of internal funding. The tax deductions to be taken for decommissioning costs will be the direct result of the actual cash expenditures made in a particular year. Therefore, the after-tax amount of the annual expenditure for decommissioning will be withdrawn from the fund in the year those decommissioning dollars are spent. Funding before-tax decommissioning expenses would result in a gross overfunding of the costs for decommissioning. Overfanding could put an undue burden on PSC shareholders in light of the inability to recover decommissioning costs through the regulatory process (see Section 6.3).
To provide insight into the effect which the four funding sources have on the projected trust fund, the annual computations and resulting balances are presented in Table 6-2.
PSC's trust funds are administered externally by a professional agent. Copies of the existing trust agreements are provided as Appendix A. PSC is currently interviewing prospective trustees and investment managers to determine if it is possible to upgrade existing services. Any new contract or agreement which results from this process will meet the various requirements of the decommissioning rule and will be provided to the NRC.
PSC will appropriately consider any future regulatory guidance in terms of periodically updating the cost estimate and funding plan. Updates will include a review of decommissioning plan 6-3
PRELIMINARY DECOMMISSIONING PLAN SECTION 6 activities, cost estimates for planned decommissioning activities, escalation rates to be applied to those costs, earnings on the trust account, tax rates and availability of tax benefits, and the level of annual contributions to the trust fund. The review of the decommissioning and funding plans will be performed by PSC's Nuclear l Operations Division in conjunction with the Finance Division to ensure'that the most realistic estimate is currently available and to ensure adequate funding based on that estimate.
6.4 FINANCIAL ASPECTS OF DEFUELING PLAN Although not directly related to decommissioning activities as defined in the Decommissioning Final Rule, PSC has committed to submit information concerning PSC's financial plans to support the defueling of Fort St. Vrain. The majority of these fuel related costs will occur during the initial Defueling Period of 1990-1993.
PSC may fund these costs through a combination of cash from general operations, debt facilities and tax benefits.
The analysis of funding needs for the fuel related activities i at Fort St. Vrain employed DRI escalation factors as described in the previous decommissioning funding section. -
6-4 ;
l l
i
_ - _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ - - - - - _ - - _ _ _ _ _ - - _ - - _ _ l
PRELIMINARY DECOMMISSIONING PLAN SECTION 6 TABLE 6-2 DECOP911SSIONING TRUST FUND ESTIMATE (as of May 1989)
The attached table is a presentation of the schedule of annual transactions projected for the Fort St. Vrain decommissioning trust fund. The table includes the following information:
Section A:
- 1. Estimated decommissioning expenditure escalated to the estimated year of expenditure as given.
- 2. Estimated decommissioning expenditure for the given year after the tax deduction has been applied.
Section B:
- 1. Estimated balance of the trust fund at the beginning of the -
given year.
- 2. Estimated after-tax decommissioning expenditure to be withdrawn from the trust fund (see Section A.2 above) in the given year.
- 3. Remaining amount to be recovered from customers for decommissioning (see Section 6.3 of the Financial Plan) to be collected in the given year.
- 4. Annual fixed contribution to fund by PSC (see Section 6.3, Financial Plan).
- 5. Earnings on average trust fund balance (see Section 6.3, Financial Plan).
- 6. Trust fund balance at end of given year after reflecting effect of contributions and withdrawals as described above.
(Results may be off slightly due to rounding.)
6-5
PRELIMINARY DECOMMISSIONING PLAN SECTION 6 E ! ! 4 s
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DECOMMISSIONING PLAN SECTION 6 )
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6-11
PRELIMINARY DECOMMISSIONING PLAN APPENDIX A EXTERNAL TRUST FUND CON 1Eit,T h.__--____________ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - _ _ - - _ _ - - -
1 l
l-TRUST AGTEEMDC
' nils TRUST AGTEEMDC, the ("Agreerent"), entered into this F/4 5 f" day of N4/ ,1981, by and betvoen Pt0LIC SctwICC COMPANY OF COIDRNV, a c3rpration ortjanizeri nncl existi/rj under the laws of the State of Colorado (" Grantor"), and 'nE AMCRICAN NATIONAL BANK OF DDWER, a national banking association, the (" Trustee').
WITNESSETH:
WHEREAS, The Public Utilities Q:rmission of the State of Oslorado ("PUC"), an agency of the State of Cbloracb, is authorized and enToweral to establish certain Wies, orders, and regulations applicable to the Grantor and WHEMAS, to provide assurave that funds will be available when needed for dummlesioning Graattor's Pbrt St. Vrain Nuclear Generatify]
Station (decamissioning cost), the PLC did order in Decision No. C30-2346, ~
issued December 12, 1980, that the Grantor, a:nmencing with the first calendar quarter of 1981, subsegtut to the effective date of the PUC's Decision and Order, deposit with an independent trustee on or before the end of the tronth subesguset to the end of each calendar quarter a stated arount of snoney to establish a decrmmiasioning fund; and Wt said independent trustee be responsible for the investrnent of the arrount a) deposited and render reports to rennter on the status of said furs $ no less frequently than annually: and, further, that the release and dis-position of the arount so deposited with the ir4 4.id. trustee is to be subject to further order by the PLC: and WHEREAS, Grantor has challenged the validity of this Order but desires to act in accordance with its terme unless and until it is in-validated e
by a court having jurisdiction, and therefore has agreed to establish a trust to provide for the decommissioning cost of the facil-ities identified he_ % : and
%4tEREAS, the Crantor, acting through its duly authorized officers, has selected the Trustee to be the Trustas under this LJree-ment, and the Trustee is willing to act as Trustee.
im, THETCDPE, the Granter and the Trustoc a] rec as follows; Sir?I0tl I Definitions As Used In 1his Agreenent 1.
'Ihe term Fiduciary treans any person who exercises any pwer of control, managernent, or disposition, or renders investment advice for a fee or other carpensation, directly or indirectly, with respect to any nonies or other property of this trust fund, or has any autMrity or responsibility to do so, or who has any authority or re-sponsibility in the a&unistration of this trust fund.
2.
The term Grantor means Public Service Q:ripany of Cblorado and any successors or assigns of the Grantor.
e 3.
The term Trustee means The American National Bank of Denver and any successor trustee.
StrTION II '-
Identification Of Facility And Cbst Estimates
'Ihis *Jreement pertains to the adjusted doo:mmissioning cost estimates, or portions thereof, related ,to the Grantor's Fbrt St. Vrain t& clear Generating Station for which financial ast.urance is demonstrated by this Agreement.
SIrTIONIIJ !
Establishment Of '!he Fund
'Ihe Grantor and the Trustee hereby establish a trust fund (the -
" Fund") for the benefit of the deoarmdssioning of the Grantor's Fbrt St.
Vrain Nuclear Generating Station. 'Ihe Grantor and the Trustee intend that no third party have access to the Fund except as herein provided.
'Ihe Fund is to be established and funded as provided and described 1
herein, which manner is acceptable to the Trustee. Such p w i.y and anounts as are to be depsited with the Trustee, together with all 1
earnings and profits thereon, less any payments or distributions made by the Trustee pursuant to this Agreernent, are referred to as the Fund.
2-C_______-_----_-----------~-- - ---
Bc Tbnd will he hold by the "Yustec. El MLt. as hercinaf tcr pmvukx1.
We Trustee undertakes no responsibility for the arount or adequacy of, nor any duty to collect from the Grantor, any payments to discha.7c any liabilities of the Grantor established by the PUC.
ArrIm IV Payment Fbr Decommissioning Cost me Trustee will rake such payments from the Fund as the Grantor, acting pursuant to order of the PUC, may direct in writing to provide for the payment of the daccrmtissioning cost of the facility covered by this Agreemant or the disposition of any balance renes.Ang after the payment of such mst.
SECTION V Payments Cbnprising me Tbnd Paymants made to the Trustee for the Fund will consist of funds'in an anctat equal to that portion of the depreciation allowaxe permitted in connection with Grantor's Fbrt St. Vrain Nuclear Generating Station that tr(Pic voy deternine frem time to time is necessary to provide for deocimissiocing costs, said anounts to be deposited acnthly, this being the basir, on Aich Grantor bills its customers.
SterIm VI I
Trestae w int Trustee wil3 invest arvi reinvest the principal and income of the Fund and heap the fund invested as a single fund, without distinction between ptincipal and inmme. In investing, reinvesting, purchasing, acquirim, esechsu1ging, selling and maneging the Fund, the Trustee or any other Fideclary will dischsr9e his duties with respect to the truet fund solely in the interest of and for the benefit of this trust fund, and with the care, skill, prudence and diligence under the ciretanstances then prevailing which per mns of prudence acting in a like capecity and familiar with su:h matters would use in the conduct of an enterprise of
a inc character and with Olle a2ms. W2 thin the Itnutnaons of the forogoing standard, the 'Prustee is authorized to acquire every kind of property, real, personal or mixed, and to make every type of investrent, specifically including, but rot limited to, corron trust funds e* tin-istered by the Trustee, tax exmpt obligations, reney market funds, certificates of deposit, corporate obligations and securities of every kind, preferred or cormon stocks, and interests in investment trusts And nutual funds that men of pru:bnce, discretion and intelligence wuld use in the conduct of an enterprise of a like character, except that:
a 1. Securities or any obligations of the Grantor, or any successor to the Grantor, or any of their affiliates, will not be ac-quired or held.
2.
'1he Trustee is authorized to invest the Fund in time or demand deoosits of the Trustee, tu the extent insured by an agency of the Tederal or State Coverviment.
3.
The Trustee is authorized to hold cash amiting invest 2nent -
or distribution uninvested for a reasonable time and without liability for the payment of interest thereon.
StrTION VII bcoress Powers of Trustee Without in any way limid.ng the rowers and discretions con-ferred upon the Trustee by the other previsions of this / gree'ent or by law, the Trustee is expressly authorized and enpowered as follows:
- 1. 're hold, retain, invest, reinvest, and manage without diversification as te kind, arount or risk of nonproductivity in realty or personalty, and without limitation by statute or rule of law; par-tition, sell, exchange, grant, convey, delivar, assipt, transfer, lease, option, trortgage, pledge, abantbn, borrow, loan, contracte distribae in cash or kind or partly in each at fair market value on the date of distribution and without requirirg pro rata distribution of specific assets, and without requirim pro rata allocation of the tax bases of such assets: hold in nominee form, tontinue businesses, carry out agree-
ments, deal with Atself. other fiduciaries and business crgant .attens 2n which the Trustcc nny have an interest, establish reserves,- release purs, and abandon, settle, contest, oonproruse, or otherwise adjust.
all clafans in . favor of or against the Fund.
- 2. 'me Trustee my also exercise'all the powers in the Cblorado Fiduciaries' Powers Act as amWed after the date of this 7qrocmont.
SB: TION VIII Tares And Deenses All taxas of any kind that may he assessed oi levied against or in respect of the Fund, and all brokerage commissions incurred by the Fund, will be paid firen the Fund. All taxes of any kind incurred by the Grantor due to inclusion of Fund incme on the Grantor's tax returns will be reimbursed to the Grantor frem the Fund upon certification of the proper arrount by the Grantor. All tax returns or information returns required pursuant to any taxes assesad or levied apinst or in respect '
of the Fund will be prepared by the Trustee or at ti. < Trustee's direc-t2.on.
All other expenses incun mi by the Trustee in connection with the -
< administration bf this trust, including fees for legal services rendered to the Trustee in respect of the FG1d, the cogensation bf the Tructee to the extent tot paid directly by the Grantor, preparation of tax !
returns, and all other proper charges and disbursements of the Tnastee will be paid from the Furd.
SECTION IX O.narterly valuation Periodic reports shall be rardered by tfm Trustee to the Grantor stowing all of the reculpts, disbursements, secpenses, and dis-posidons durirs; the period &nd arsets then held as the principal of the Furd, which reports shall be recaered quarterly, within thirty (30) days of the and of each calendar quarter. Any securities in the Fund will be valued at market value as of to nere than thirty (30) days prior to the date of the statement.
LL.__--_. - - - - - - . . - _ - - . . . - - . - . _
1 At'rlm x '
/\dvice Of Counsel The Trustee may frcm time to time consult with counsel, who may be counsel to the Grantor, with respect to any questions arising as to the construction of this tv]rconent or any action to bc; taken herc-under.
We Trustcc will be fully protteted to the extht permitted by law in acting up3n the advice of counsel.
SECTICN XI Trustee compensation the Trestee will be entitled to reasonable ocupensation for its services, as agreed upon in writing fran time to time with the Grantor.
i SECTICH XII Soccessor. Trustee .
Upon the written hgrea'ent of the Grantor, the Trustec, and the PtC, the Trustee may resign or the Grantor may replace the Trustee.
In either event, or arould the Trustee for arry reason fail to qt:411fy or cease to act as Trustee, the Crantor with the approval of the PLC will appoint a successor trusteet who will have the same powers and duties as tlnse conferred won the Trustee hereurder; provided, however, that the .
Grantor shall have no right or power to become a trustee and to pro-vision of this Agreement shall be construed so as to create any right or pw in the Grantor to so act as the trustee. Upon acceptance of the appointment by the successor trustee, the Trustee will assign, transfer, and pay over to the successor trustee the funds and properties thren constituting the Fund. If for any reason the Stantor cannot or cbes tot act in the event of the resignation of the Trustee, Trustee may apply to a court of ccat1petent jurisdiction for the appointment of a successor trustee or for instructions. the successor trustee and the date on which it will asstate administration of the trust will be specified in writing and 'sent to the Grantor, PLC, ard the present and successor
-G-
Wustocs by certifiel nuii tt:a (10) obys tcfore such cimye luuta i
effcetivo. Any expensos incurred by the Trustoc as a result of .any of )
the acts mntemplated by this Soetion will be paid as provided in Sec-tion VIII.
V b
i SDCTIm XIII f
Instructions 1e 'Ihe Trustec- 1 j
)
All orders, requests, and instructions by the Grantor to the li Trustee will be in writing, signed by proper officers of the Grantor-and, if appropriate, accenpanied by relevant orders of the PLC. Trustec will be fully protected in acting without inquiry in accordance with the Grantor's orders, requescs, and instructions. The Trustee will have no duty to act in the absence of such orders, requests, and instructions from the Grantor except as provided herein.
SECTION XIV Anandment of Agreenant This Agreement nay be amended only by an instrument in writing,
~
executed by the Grantor, Trustee and approved by the PLC.
SECTION XV Irrevocability And Termination Subject to the right of the parties to amend this Agreement as provided in Section XIV, this Agreement will be irrevomble and will continue until terminated by payment of decommissioning cost from the rund as provided in Section IV hereof, which event is contemplated to occur approximately thirty (30) years from date hereof, or upon the written agreement of the Grantor, the Trustee and the PLC, or upon the written order of the Ptr. Upon termination of the Ibnd, any renainirg trust pety in excess of decommissioning expenses contatplated by the terms of this Ibnd will revert to the Grantor subject to the order of the PLC; and final closirg of this trust acmunt shall be' subject to written approval from the PUC.
A&";;oa en
' Imnun aty retd Irtdannir icat 2cn The Trustee will not incur persorn1 liability of any nature in mnnection with any act or ort.ission made in good faith in the adntinis-trat. ion of this trust,- or in carrying out any direction by the Crantor issued in accordance with this Agreenent. The Trustee will be indemified h and. saved harmless by the Grantor or from the trust Nrd, or both, from and against any personal liability to which the Trustee may be subjected by reason of any act or conduct in its official capacity.
SEETICN XVII Cteic3 Of Iaw
'Ihis Agreenent will be administered, construed ard enforced according to the laws of the State of Colorado. !
SECTION XVIII Interpretation As used in this Agreenant, words in the singular include the plural and words in the plural include the singular. 'Ihe descriptive headirgs for each section of this Agreement will not affect the inter-
- pretation or legal efficacy of this Agreement.
IN WI1 NESS WHERET, the parties have caused this Agreament to be executed by their respective officers duly authorized and their corporate seals to be hereunto affixed ard ettested as of the date first alzwe written.
PUBLIC Str(VI2 CCNPANY T COIDRADO By 4/44
Title:
/LCL kM/M ATITST:
- t (S'EAL)
- -B-
'1".
':t iv;; T ::u:,:L n';N Cl' . :i ::.u . .
D_ Y sE& r. & A -
Tide-[ rA, Chi b W
- { SPAL)
-l 1
STATE OF COI.OPADO ) i
) ss.
CITY AND COfJt7tY OF DDNER )
ne foregoing TRLST AGREEMENT was acknowledged before me this f/Aff day of MAY __ _
, 1981, by _ l* W 80MNS and ~;D 3. N a C A. as the Grantor, and G. N/C# DEL. -
////L t.//J and -
as the Trustee.
Witness my hand and official saal.
14y Comnission Dcpires: 4,8C-fS .
J
~
tbtary Public / .
l l
- ' . e-w 1983 AL APR TRUST AGREEMENT THIS TRUST AGREEMENT, the (" Agreement"), entered into effective as of the 1st day of January, 1981 , by and between PUBLIC SERVICE COMPANY OF COLORADO, a corporation organized and existing under the laws of the State of Colorado (the " Company"),
and FIRST INTERSTATE BANK OF DENVER, a national banking association, ( the " Trustee") .
W I T N E S S E T H:
WHEREAS, pursuant to section 466A of the Internal Revenue Code of 1986 (" Code"), certain federal income tax benefits are available to the Company by creating and making contributions to -
qualified nuclear decommissioning reserve funds associated with the Company's ownership of Fort St. Vrain Nuclear Generating Station; and WHEREAS, the Company wishes to establish a qualified nuclear decommissioning reserve fund to hold monies for decommissioning
' Fort St. Vrain Nuclear Generating Station; and WHEREAS, the Company wishes to establish a Fund for the investment of the assets of the qualified nuclear decommissioning reserve funds for Fort St. Vrain Nuclear Generating Station; and WHEREAS, the Company, acting through its duly authorized officers, has selected the Trustee to be the Trustee under this Agreement, and the Trustee is willing to act as Trustee.
+
- - - - - - - _ _ _ _ . . . _ _ _ _ _ _ . . _ _ _ ~ _ _ _ ____
h i NOW, ~ THEREFORE, the Company and. the Trustee agree as fol'-
l-~ ,
lows:
SECTION I Definitions As Used in This Agreement l 1. The term Fiduciary means any person who ' exercises any l
power of control, management, or disposition, or renders invest-ment advice' for a fee or other compensation, directly or indi-rectly, with respect to any monies or other property of this trust ' fund, or has any authority or responsibility to do so, or who has any authority or responsibility in the administration of this trust fund.
- 2. The term Company means Public Service Company of ~'
Colorado and any successors or assigns of the Company.
- 3. The ' te rm Trustee - means the First Interstate Bank of Denver and any successor trustee.
- 4. " Code" shall mean the Internal Revenue Code of 1986, as the same may be amended from time to time.
SECTION ll Identification Of Facility And Cost Estimates This Agreement pertains to the adjusted decommissioning cost estimates, or portions thereof, related to the Company's Fort St.
Vrain Nuclear Generating Station for which financial assurance is demonstrated by this Agreement.
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SECTION lll Establishment Of The Fund i
The Company and the ' Trustee hereby establish a trust fund (the " Fund") for the benefit of the decommissioning of the
. Company's Fort St. Vrain Nuclear Generating Station. The Company and the Trustee intend that no third party have access to the Fund except as herein provided. The Fund is to be established and funded as provided and described herein, which manner is acceptable to the Trustee. Such property and amounts as are to be deposited with the Trustee, together with all earnings and profits thereon, less any payments or distributions ' made by the Trustee pursuant to this Agreement, are referred to as the Fund. -
The Fund will be held by the Trustee, IN TRUST, as hereinafter provided. The Trustee undertakes no responsibility for the
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amount or adequacy of, nor any duty to collect from the Company, any payments to discharge any liabilities of the Company estab-lished by any governmental authority.
SECTION IV Payment For Decommissioning Cost The Trustee will make such payments from the Fund as the Company may direct in writing to provide for the payment of the decommissioning cost of the facility covered by this Agreement or the disposition of any balance remaining after the payment of such cost.
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SECTION V Payments Comprising The Fund Payments made to the Trustee for the Fund will consist of funds in an amount equal to that portion of the depreciation allowance permitted in connection with Company's Fort St. Vrain .
i Nuclear Generating Station that may be determined from time to time is necessary to provide for decommissioning costs, said i
amounts to be deposited monthly, this being the basis ~ on which ]
i Company bills its customers.
SECTION VI Trustee Management l
Trustee will invest and reinvest the principal and income of' the Fund and keep the Fund invested as a single fund. In investing, reinvesting, purchasing, acquiring, exchanging, selling and nanaging the Fund, the Trustee or any other Fiduciary will discharge his duties with respect to the trust fund solely in the intereat of and for the benefit of this trust fund, and with the care, skill, prudence and diligence under the circumstances then prevailing which persons of prudence acting in a like capacity and familiar with such matters would use in the conduct of an enterprise of a like character and with li'ke aims.
Within the limitations of the foregoing standard and pursuant to the requirements of Section 468A of the Internal Revenue Code of 1986, the Trustem is authorized to sell, exchange, partition, or otherwise dispose of all or any part of the Fund at public or
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I private sale, without prior application to, or approval by, or order of any court, upon such terms and in such manner and at such prices as the Trustee shall determine; to modify, renew or extend bonds, notes or other obligations or any installment of principal thereof or any interest due thereon and to waive any defaults in the performance of the terms and conditions thereof; 1
l and to execute and deliver any and all bills of sale, assignments, bonds or other instruments in connection with these powers, all at such times, in such manner and upon such terms and conditions as the Trustee may deem expedient to accomplish the purposes of this Fund.
The Trustee is authorized to hold cash awaiting investment ~
or distribution uninvested for a reasonable time and without liability for the payment of interest thereon.
SECTION Vil Express Powers Of Trustee Without in any way limiting the powers and discretions conferred upon the Trustee by the other provisions of this Agreement or by law, the Trustee is expressly authorized and empowered as follows:
- 1. To retain, manage, invest and reinvest all or part of the Fund, including any undistributed income therefrom; provided, however, that no such investment or reinvestment of the Fund may i be made by the Trustee unless such investment is permitted to be l
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made by Code sections. 501(c)(21)(B)(ii) and 468A(e)(4)(C), the I I
l regulations thereunder, and any applicable successor provisions. I
- 2. To renew or extend the time of payment of any obligation, secured or unsecured, payable to or by this Fund, for as long a period or periods of time and on such terms as the Trustee shall determine, and to adjust, settle, compromise, and arbitrate claims or demands in favor of or against this Fund, including claims for taxes.
- 3. To hold any stocks, bonds, securities, and/or other property in the name of a nominee, in a street name, or by other title-holding device, without indication of trust.
- 4. To borrow money in such amounts and upon such terms as ~
the Company may authori::e in writing as necessary to carry out the purposes of this Fund, and to pledge any securities or other e
property for the repayment of any such loan as the Company may direct.
- 5. The Trustee may also exercise all the powers in the Colorado Fiduciaries' Power Act as amended after the date of this Agreement.
SECTION VIII Taxes and Expenses All taxes of any kind that may be assessed or levied against or in respect of the Fund, and all brokerage commissions incurred by the Fund, will be paid from the Fund. All taxes of any kind incurred by the Company due to includion of Fund income on the
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$ 4.
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l Company's tax- returns will be reimbursed to the Company from the 1 _
Fund upon certification of the proper amount by the Company. All tax returns or information returns required pursuant to any. taxes J assessed _ or: levied against or in respect of .the Fund will be prepared by._the Trustee or at the Trustee's direction. All other l
- l. expenses incurred by the Trustee in connection with the adminis-tration of this trust, including fees for legal services rendered to the Trustee - in respect of the Fund, the compensation of the Trustee.to the extent nce paid directly.by the Company, prepara-tion of tax returns, and all other proper charges and disburse-ments of the Trustee will be paid from the Fund.
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SECTION IX Quarterly Valuation Periodic . reports shall be rendered by the Trustee to the Company showing all of the receipts, disbursements, expenses, and dispositions during the period and assets then held as the principal of the Fund, which reports shall be rendered quarterly, within thirty (30) days of the end of each calendar quarter. Any securities in the Fund will be valued at market value as of no more than thirty (30) days prior to the date of the statement.
SECTION X Advice Of Counsel The Trustee may from time to time consult with counsel, who may be counsel to the Company, with respect to any questions I
_m-______._m_______ .m_ _ _ _ . _ _ _ _ ______m_ _ _ _ _ _
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arising as to the construction of this Agreement or any action to be taken hereunder. .The Trustee will be fully protected to the extent permitted by law in acting upon the advice of counsel.
SECTION XI Trustee Compensation The Trustee will be entitled to reasonable compensation for its services, as agreed upon in writing from time to time with the Company.
SECTION Xll successor Trustee Upon the written agreement of the Company and the Trustee, the Trustee may resign or the Company may replace the ~
Trustee. In either. event, or should the Trustee for any reason fail to qualify or cease to act as Trustee, the Company will appoint a successor trustee who will have. the same powers and.
duties as those conferred upon the Trustee hereunder; provided, however, that the company shall have no right or power to become .
a trustee and no provision of this Agreement shall be construed ,
so as to create any right or power in the Company to so act as the trustee. Upon acceptance of the appointment by the successor trustee, the Trustee will assign, transfer, and pay over to the sucs +ssor trustee the funds and properties then constituting the Fund. If for any reason the Company cannot or does not act in the evens of the resignation of the Trustee, Trustee may apply to a court of competent jurisdiction for the appointment of a
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F e I-
. successor trustee or for instructions. The successor trustee and the date on which it will assume administration of the trust will be specified' in writing and sent . to the Company, and the present
- l. and successor Trustees by certified mail ten (10) days before i
such change becomes effective. Any expenses incurred by the d Trustee as a result of any of the acts contemplated by this Section will be paid as provided in Section VIII.
( SECTION Xill Instructions To The Trustee }
All orders, requests, and instructions by the Company to the Trustee will be in writing, signed by proper officers of the Company and, if appropriate, accompanied by relevant orders of -
governmental authority. Trustee will be fully protected in acting without inquiry in accordance with the Company's orders, 1 1: requests, and instructions. The Trustee will have no duty to act k k
in the absence of such orders, requests, and instructions from j l
the Company except as provided herein. I SECTION "lV Amendment Of Agreement This Agreement may be amended only by an instrument in writing, executed by the Company and Trustee.
SECTION XV Irrevocability And Termination Subject to the right of the parties to amend this Agreement as provided in Section XIV, this Agreement will be irrevocable
and will continue until terminated by payment of decommissioning cost from the Fund as provided in Section IV hereof, which event 1
is contemplated to occur approximately thirty (30) years from date hereof, or upon the written agreement of the Company and the i
Trustee. Upon termination of the Fund, any remaining trust property in excess of decommissioning expenses contemplated by the terms of this Fund will revert to the Company subject to the order of appropriate governmental authority; and final closing of this trust account shall be subject to written approval from the appropriate governmental authority, if any.
SECTION XVI immunity And Indemnification
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The Trustee will not incur liability of any nature in connection with any act or omission made in good f aith in the administration of this trust, or in carrying out any direction by the Company issued in accordance with this Agreement. The Trustee will be indemnified and saved harmless by the Company or from the trust Fund, or both, from and against any personal liability to which the Trustee may be subjected by reason of any act or conduct in its official capacity.
SECTION XVil Choice Of Law This Agreement will be administered, construed and enforced according to the laws of the State of Colorado.
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SECTION XVill' As used in this Agreement, words in the singular include the plural and words in the plural . include the singular. The de-scriptive headings for each section of this Agreement will not affect the interpretation or legal efficacy of this Agreement.
IN WITNESS WHEREOF, the parties have caused this Agreement to be executed by their respective officers duly authorized and their corporate seals to be hereunto affixed and attested as of the date first above written.
PUBLIC SERVICE COMPANY OF COLORADO By - --
Title:
__ Md6 FAu ATTEST:
M kr See ary
[ SEAL)
FIRST I ' ERSTATE BANK OF DENVER By:
VICE PT.::::::i t.::3 TT"ST OfflCER
Title:
ATTEST:
[t4h[
vi " 'fe $w 7"cYjl
[ SEAL) "S&O r 7~ El Y-Y- 8 8
STATE OF COLORADO )
, )ss.
CITY & COUNTY OF DENVER )
Thp foregoing TRYST AGREEMENT was acknowled
- this E day of /ldtL , 1981, by R,ch and ded before
. h!,. n o me as Vice President"ahd by Tam ec A f)1c d eth.y as Secre'tary of Public Service Company of Colorado.
Witness my hand and official seal.
My Commission Expires: 4 o70 -k .
0 97t thf/f4RM_
Notary Public STATE OF COLORADO )
)ss.
. CITY & COUNTY OF DENVER )
Thykforegoing fTRUST AGREEMENA was knowledged before me.
thip_6 day of (lom! , 198 f , by aj:;d K u /cJ l C. ['41.
asV6 Pat,1 % i n M Opb s and E. C. M ad as
[hathM (a,fw /*
of First Interstate Bink of Denver.
Witness my hand and official seal.
My Commission Expires li k /
/ '
9 - &r Notdry Public ' '
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