ML20086S247
| ML20086S247 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 02/09/1995 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20086S239 | List: |
| References | |
| FOIA-94-468 PROC-950209, NUDOCS 9508010157 | |
| Download: ML20086S247 (232) | |
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ns Decommissioning Plan for the Fort St. Vrain Nuclear Generating Station Public Service Company of Colorado 9508010157 950209
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SECTION 1
SUMMARY
OF PLAN
1.1 DESCRIPTION
OF DECOMMISSIONING PLAN AND DECOMMISSIONING ALTERNATIVE 1.1.1 Introduction By letter to the Nuclear Regulatory Commission (NRC) dated December 5,1988 (Ref.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." At that time, PSC began decommissioning planning to suppon premature decommissioning, resulting in submittal of the Preliminary Decommissioning Plan to the NRC on June 30,1989. (Ref. 2)
This Proposed Decommissioning Plan is submitted by PSC in accordance with the requirement of 10 CFR 50.82(a), which requires submittal of the Proposed Decommissioning Plan "within two years following permanent cessation of operations." PSC previously provided a target date of October 31,1990, for submittal of the Proposed Decommissioning Plan.
The Proposed Decommissioning Plan represents a departure from PSC's Preliminary Decommissioning Plan (Ref. 2) in that, after consideration of financial risks, regulatory environment, and uncertainty M other issues, PSC has selected the DECON alternative for immediate dismantlement and decommissioning of Fon St. Vrain.
Through a competitive bid process, PSC has selected a team headed by the Westinghouse Electric Corporation to carry out the decommissioning of Fort St. Vrain on a fixed price basis.
Coincident with decommissioning, the Fort St. Vrain plant may be converted to a fossil-fueled j
facility (See Section 5.5).
l j
1.1.2 Background Fort St. Vrain was shutdown on August 18, 1989. On August 29, 1989, the PSC Board of Directors reviewed and confirmed the Executive Management decision that Fort St. Vrain would not be restarted, and that PSC would pursue the decommissioning of Fort St. Vrain.
The decision to permanently shut down and decommission Fort St. Vrain was based on related j
technical and financial considerations. Problems were identified with the control rod drive assemblies and the steam generator steam ring headers that presented significant technical obstacles which could be overcome, but at significant cost in dollars and time to PSC.
Additionally, due to the uniqueness of the one-of-a-kind High Temperature Gas-Cooled Reactor (HTGR) fuel cycle, the cost to purchase new fuel was prohibitive. This, in conjunction with low 1.1-1
1 8
- 3 DECOMMISSIONING PLAN REV1
- SECTION 1 plant availability and cutir.spuiidingly high operating costs, made continued operation of Fon l
St. Vrain imprudent.
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-i Coupled with these_ technical and fuel cycle considentions, Fon St. Vrain had previously been l
removed from the rate base as a result of a 1986 Settlement' Agreement between PSC, the Colorado Public Utilities Commission (CPUC), the Office of Consumer Counsel (OCC) and other panies. With the exception oflimited funds to be collected for decommissioning, the j
. removal of Fort St. Vrain from the regulatory rate base left PSC shareholders responsible for j
further operating and decommissioning costs of Fon St. Vram.
1.1.3 Contents of the Pronosed Decommissionine Plan i
'Ihe Proposed Decommissioning Plan has been prepared to be' responsive to the requirements of.
10 CFR 50.82(b) and the guidance of Draft Regulatory Guide DG-1005 " Standard Format and l
Content for Decommissioning Plans for Nuclear Reactors" (Ref. 3). The following is a brief summary of the sections contained within this plan.
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Section Description i
1
" Summary of Plan" provides a brief description of the proposed plan and -
background information related to the decision to decommission Fort St. Vrain.
Information is provided to describe the major activities involved in. the dismantlement and decommissioning of Fort St. Vrain, and the projected project schedule. The cost to decommission Fort St. Vrain is identified, as well as status of the availability of funding.
Details are provided in Section 1.4 on implementation and administration of the proposed plan. Section 1.5 describes the controls which will be effective during the transition period prior to approval of the Proposed Decommissioning Plan.
2
" Choice of Decommissioning Alternative and Description of Activities" identifies the selected decommissioning alternative. Section 2.2 provides a description of Fon St. Vrain and identifies major site factors, and identifies contam' ated or m
activated structures and components which will be removed _ during decommissioning.
The major decommissioning activities and schedule are provided in Section 2.3. Organizational structures are provided for the PSC organization (Section 2.4) and the selected contractor (the Westinghouse team, Section 2.6). Decommissioning training requirements are identified in Section 2.5.
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REV_1 DECOMMISSIONING PLAN SECTION 1 P
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" Protection of Occupational and Public Health and Safety" describes the."as-is"
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radiological status of the Fort St. Vrain facility (Section 3.1).
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decommissioning radiation pAdon organization and exposure estimates are described in Section 3.2, and proposed metiods of managing radioactive waste, including offsite transportation and disposal are discussed in Section 3.3. The
- analysis of postulated bounding decommissioning necWnts is provided in Section -
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3.4.
4
" Final Radiation Survey Plan" provides the purpose, criteria, and methodology' 4
that 'will be used to. formulate the ' final radiation survey plan, including instrumentation, documentation and quality assurance mquirements, and eventual site closure.
5
" Decommissioning Fixed Price Contract and Funding Plan" provides a description 1
of the decommissioning fixed price contract, major assumptions and bases used to derive the decommissioning cost, and status of deconunigioning funding.
Provisions are also identified for updating both the decommissioning cost and the.
funding plan.
6
" Decommissioning Technical and Environmental Specifications" provides the methodology and philosophy that was used to develop'the decommissioning technical specifications.
7
" Decommissioning Quality Assurance Plan" provides the QA plan'which will be effective during decommissioning.
8
" Decommissioning Access Control Plan" identifies, those access control-1 requirements to be administered during the decommissioning process once all-spent fuel has been removed from the Protected Area. 'Ihis access control plan will replace the existing physical security plan during the. decommissioning
- j period.
Appendix I, " Westinghouse Team Scope of Work", provides a detailed description of the proposed Westinghouse team decommissioning and dismantlement activities. Appendix II, " Fort St. Vrain Activation Analysis", provides the results of the analysis to identify activation levels and isotopes for Fort St. Vrain components.
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SECTION 1 1
l 1.2 -
MAJOR TASKS, SCHEDULES AND ACTIVITIES 1.2.1 DWotion of Maior ActivWe
~The major dismantlement and decontamination activities to be performed' dunng l
decommissioning are described in detail in Section 2.3. The decommissioning project is' divided into three major work areas.
1.
Decontamination and dismantlement of the PCRV.
2.
Decontamination and dismantlement of the contaminated balance of plant (BOP).
systems.
3.
Site cleanup and final site radiation survey.
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Site cleanup is described in Section 2.3 and the final site radiation survey is described in Section 4.
1.2.2 Final Release Criteria The release of the site, facilities and materials will be based on proper application of release criteria for surface contamination, soil / water concentrations and exposure rates.
Final site release criteria are fully identified in Section 4.2 of this plan.
1.2.3 Decontamination and' Dismantlement of the PCRV The major decommissioning task is the dismantlement and decontamination of the radioactive I
portions of the Prestressed Concrete Reactor Vessel (PCRV).
Section 2.3 provides a comprehensive description of the steps necessary to dismantle and decontaminate the PCRV.
PCRV dismantlement activities will begin only after all irradiated fuel has been removed from the Reactor Building.
PSC and the Westinghouse team have evaluated technical options available for dismantling radioactive portions of the PCRV, and a decision has been made that the best technical approach is to flood the PCRV with water, and perform the majority of dismantlement activities submerged. This will allow the most direct access to highly radioactive portions of the PCRV, while affording the maximum shielding benefit.
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DECOMMISSIONING PLAN REV1 SECTION 1 1.2.4 Decontamination and Dismantlement of Contaminated Balance of Plant Systems For the purposes of this Proposed Decommissioning Plan, balance of plant systems refer to those contaminated or potentially contaminated plant systems outside the PCRV. Decontamination and dismantlement of contaminated or potentially contaminated balance of plant systems will be performed by one of the following approaches: (1) decontamination in place, (2) removal and decontamination, or (3) removal and disposal as radioactive waste.
Systems which are contaminated or potentially contaminated above releasable limits requiring decontamination and dismantlement are described in Section 2.3.
1.2.5 Schedule for Decommissionine Activities The schedule for decommissioning activities is provided in Section 2.3.5 and Figure ".3-15.
The following is a brief description of the two phases of the Fort St. Vrain Decomt.asioning Project:
Phase I Decommissionine Plannine Phase consists ofinitial-sita, characterization, preparation of work scope planning, work specifications and procedures, and equipment and material staging.
There will be NO physical decommissioning activities performed as part of this planning phase, although some component removal and disposal activities may occur prior to commencement of Phase II (described below) as described in Section 1.5 of this plan.
Phase II Decontamination and Dismantlement Phase. with an estimated duration of 39 months.
Actual dismantlement, decontamination, and physical decommissioning activities will occur as part of this phase. The actual physical decommissioning activities are scheduled to commence after:(1)
NRC approval of the Proposed Decommissioning Plan, and (2) removal of all irradiated fuel from the Reactor Building.
It is important to note that Phase I and Phase 11 activities are not conducted in series. These two phases have considerable overlap. Further detailed descriptions of the work scope to be performed in each project phase are provided in Appendix I of this plan.
Some component removal activities will be conducted prior to commencement of the Decontamination and Dismantlement Phase, as described in Section 1.5. Decommissioning of Fort St. Vrain, including site cleanup and final site radiation survey, is expected to be completed by October 1995.
1.2-2
REV 2 DECOMMISSIONING PLAN SECTION 2 2.2 FACILITY DESCRIPTION 2.2.1 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.
During the plant operation, approximately one mile square within the site area was designated as the exclusion at:a, and the licensee maintained complete control over this area. The completed facility is shown in Figure 2.2-1.
The basic installation consists of a Reactor Building, a Turbine Building, cooling towers, and an electrical switchyard.
2.2.1.1
' Reactor Building The Reactor Building (Figures 2.2-2 and 2.2-3) houses the prestressed concrete reactor vessel (PCRV), fuel handling area, fuel storage wells (FSWs), fuel shipment preparation facilities, decontamination and radioactive liquid and gas waste processing equipment, and most 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 Reactor Building ventilation exhaust filter system is designed to filter the Reactor Building atmosphere prior to release to the vent stack during both normal and most accident conditions during decommissioning. The Reactor Building is maintained in a subatmospheric condition to ensure that all air leakage will be inward and to minimize unfiltered fission product release from the building. The ventilation system was designed to maintain a subatmospheric condition approximately 1/4-inch water gauge negative. In actual practice, the Reactor Building pressure is normally 0.15 to 0.20 inches water gauge negative, depending on building activities and ventilation system configuration.
The Reactor Building overpressure protection system consists of 94 louvered panels, 2
4-feet by 8%-feet, each of which provides 12.02 ft of free flow area for a total of 2
1130 ft of free flow area when fully opened. The louvers are opened by spring pressure and closed (or held closed) by air pressure acting through a pneumatic cylinder. Subatmospheric conditions can be maintained with several louver banks l
2.2-1
DECOMMISSIONING PLAN REV 2 SECTION 2 open. The overpressure protection system louvers may be opened on a controlled basis for various reasons (e.g., to provide extra ventilation cooling during hot weather). The louvers must be closed whenever Reactor Building integrity is required.
The PCRV and nuclear steam supply system (NSSS) are located in the west portion of the Reactor Building. The east portion of the Reactor Building houses auxiliary and suppon systems and facilities such as the FSWs, the hot service facility (HSF),
the equipment storage wells (ESWs), storage and laydown areas for various pieces of equipment, radioactive gas and liquid waste storage facilities, and the loading pons for the spent fuel shipping casks (SFSC). The basement area of the Reactor Building contains the building sump / keyway.
The volume of the sump / keyway is approximately 44,600 cubic feet.
The Fon St. Vrain Reactor Building is presently designed to withstand the Design Basis Eanhquake (DBE) of 0.10 g horizontal ground acceleration at the site without unsafe damage or failure to function. During decommissioning, the Regto(Building will continue to be required to perform its confinement function following a seismic event.
The decommissioning of Fon St. Vrain will not involve any major modifications to the Fon St. Vrain Reactor Building without verification of the seismic qualification.
Other than the Reactor Building, no additional seismic analysis of individual decommissioning tasks and removal activities will be required.
The Reactor Building overhead crane is located inside the Reactor Building, over the refueling floor. The Reactor Building crane is the means by which heavy lifting operations and maintenance are performed on the refueling floor of the Reactor Building. The design of the overhead crane conforms to Class "D" crane type specified in the Electric Overhead Crane Institute (EOCI) Specification No. 61 and the AISC Specification " Designs, Fabrication, and Erection of Structural Steel for Buildings" adopted November 30,1961. All structural steelis ASTM A-36 or better.
The crane capacity has been upgnded from 160 tons to a revised capacity of 170 tons. The crane trolley main hook has a capacity of 50 tons and the auxiliary crane hook has a capacity of 17.5 tons.
In order to meet the requirements of the EOCI and AISC specifications, the building girders and crane rails are designed for 125% of the rated load, and the crane bridge girders are designed for an impact loading of not less than 10% of the lifting forces 2.2-2
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t REV 2 DECOMMISSIONING PLAN SECTION 2 required for 125% of the rated load. The hoisting cable at the main hoist consists of 12 pans of 1-3/8 inch diameter, 6-strand,37-wire improved plow steel crane rope with a rating of 14.8 tons per part, for a total capacity of 177 tons. The breaking strength of the hoisting cable arrangement is 775 tons.
2.2.1.2 Turbine Building The Turbine Building (Figures 2.2-2 and 2.2-3) houses the turbine generator with condensing, feedwater, and other auxiliary systems. Included in the Turbine Building is an auxiliary bay area housing the reactor plant ventilation equipment, the controlled personnel access to the Reactor Building, and an area housing the control room and miscellaneous electrical services. The Turbine Building also houses a service and office area which provides space for miscellaneous shops, auxiliary steam system components, and administrative offices.
2.2.1.3 Fuel Storage Building The Fort St. Vrain Fuel Storage Building is a single level concrete structure located east of the Reactor Building (see Figure 2.2-4). The building is constructed of prestressed concrete panels and twin tees, and is designed to withstand a 202 mph tornado wind and can withstand the design basis tornado missile. This building will be used for decommissioning support.
2.2.2 Prestressed Concrete Reactor Vessel (PCRV) and Internal Comoonents The PCRV (Figures 2.2-5 and 2.2-6), which contains the NSSS, 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 Fort St. Vrain systems associated with the PCRV are as follows:
System 11 PCRV and Internal Components System 12 Control Rod Drive and Orifice Assembly System 17& Reactor Reflector and Defueling Elements System 18 System 21 Helium Circulators System 22 Steam Generators ai System 23 Helium Purification System i
2.2-3 h
DECOMMISSIONING PLAN REV 2 SECTION 2 These systems make up the primary reactor vessel and internal core components located within the PCRV. These systems and components are discussed further in this section and in Section 2.3.
t Portions of the PCRV concrete and rebar are expected to remain activated due to direct irradiation from the reactor core. Highly radioactive components will remain l
inside the PCRV until removed during PCRV decontamination and dismantlement.
Physically, the 15-1/2 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 i
cavity. The liner is anchored to the concrete at frequent intervals. A core support floor (CSF) is provided within the PCRV in the form of a reinforced 5 foot thick concrete disk with a 3/4-inch carbon steel outer liner, supported by 12 steel core support floor columns from the bottom of the PCRV cavity.
Longitudinal, circumferential and top and bottom crosshead prestressing tendons (448 total) are located in conduits embedded in the PCRV concrete.
Tendons are positioned both circumferentially and vertically along the PCRV side walls. There are also tendons across the top and bottom heads in a criss-cross arrangement.
The reactor core arrangement within the PCRV is shown in Figure 2.2-7. The top layer of the core arrangement consisted of hexagonal shaped metal clad reflector blocks (MCRBs) with openings for 37 control rod pairs. The MCRBs provided an inlet plenum for the reactor coolant to the active core. Region constraint devices (RCDs) were located on top of the MCRBs and mechanically interlocked the top layer (not shown on Figure 2.2-7). Hexagonal top reflector elements with coolant channels are located directly below the MCRBs and above the active core region.
The active core was divided into 37 regions and consisted of 1482 fuel elements.
Individual fuel elements were hexagonal in cross section and aligned with the coolant channels from the reflector elements and MCRBs. During reactor defueling, the fuel elements are being replaced with defueling elements of identical shape and size.
Hexagonal reflector elements are also located to the sides of and below the active core region. Many of the bottom reflector elements contain boronated graphite in Hastelloy cans.
2.2-4
REV 2 DECOMMISSIONING PLAN SECTION 2 Radially outside of and immediately adjacent to the top, side and bottom hexagonal reflector elements are the large irregular-shaped side reflector blocks. Between the side reflector blocks and the core barrel are the boronated side reflector spacer blocks that contain boronated steel pins and were used for shielding.
The core barrel is a steel cylinder approximately 27 feet 4 inches inside diameter and 29 feet high. The core barrel has 12 upper outer keys and 12 lower outer keys which center the core barrel to the PCRV liner. The lower three feet of the inside surface of the core barrelis insulated. In addition, there are seven thermocouple penetrations located about four feet above the bottom of the core barrel that are between the PCRV liner and the core barrel.
Immediately outboard of the core barrelis 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. See Figure 2.2-8 for a general arrangement of the thermal barriers.
j Below the core region containing the defueling elements, the CSF will bear the weight of the defueling elements and reflectors through the core suppon posts and the core support blocks. 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 supponed from the bottom head of the PCRV with 12 core suppon floor columns (See Figure 2.2-9). The CSF is a complex component that includes the following features:
1.
The CSF is a 29-foot in diameter, 5-foot thick concrete disk, clad with 3/4-inch plate steel, weighing approximately 270 tons.
2.
There are 12 conical penetrations which discharged the hot helium gas from the reactor to the steam generators via 12 inlet ducts.
3.
The CSF is supponed by 12 steel columns that are located near the CSF periphery that are welded to the cladding plate.
4.
Within each of the 12 CSF suppon columns is an array of cooling tubes and instrumentation tubes.
5.
All surfaces of the CSF are insulated.
6.
There is a monorail spider consisting of twelve heavy structural steel beams in a radial arrangement on the bottom side of the CSF, that were used to position the steam generators during construction.
2.2-5
DECOMMISSIONING PLAN REV 2 i
SECTION 2 The lower plenum is below the CSF and houses the steam generator modules (12),
circulator diffusers (4), circulators (4) the CSF support columns (12) and the lower floor. A number of instrument and equipment penetrations and wells exist in the PCRV heads and sidewalls.
2.2.3 Balance of Plant Contaminated Comoonents The systems identified below are considered to be the potentially contaminated balance of plant systems outside of the PCRV at Fort St. Vrain. Decontamination and dismantlement of these BOP systems are discussed in Section 2.3.4:
i System 13 Fuel Handling Equipment System 14 Fuel Storage Facility System 16 Auxiliary Equipment System 21 Helium Circulator Auxiliaries
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System 23 Helium Purification Auxiliaries System 24 Helium Storage System System 46 Reactor Plant Cooling Water System System 47 Purification Cooling Water System System 61 Decontamination System System 62 Radioactive Liquid Waste System System 63 Radioactive Gas Waste System System 72 Reactor Building Drain System System 73 Reactor Building Ventilation System i
System 93 Instrumentation and Controls System 15, fuel and reflector shipping equipment, consists primarily of the shipping casks, truck-trailers, spent fuel container, and cask lifting apparatus and is not a part of the decommissioning project. These equipment items will be retained under their separate 10 CFR 71 license or will be disposed of at some time in the future.
A brief summary of the major components in each of the above balance of plant contaminated systems is as follows:
2.2.3.1 System 13 - Fuel Handling Eauipment The fuel handling equipment that remains contaminated includes the fuel handling machine (FHM, Figure 2.2-10), five reactor isolation valves (Figure 2.2-11) and two refueling sleeves (Figure 2.2-12).
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REV1 DECOMMISSIONING PLAN SECTION 2 NELSON TYPE THREADED STUD k
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DECOMMISSIONING PLAN REV 1 f
SECTION 2 1
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REV 2 DECOMMISSIONING PLAN SECTION 2 2.3 DECOMMISSIONING ACTIVITIES AND PLANNING
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2.3.1 Idroduction Decommissioning of Fort St. Vrain includes the dismantlement, decontamination and disposal c,f radioactively contaminated or potentially contaminated material and components within the PCRV, and in contaminated or potentially contaminated balance of plant systems, and on the remaining site, followed by the final radiation survey. Some of the activities described in this section will be performed prior to approval of the Proposed Decommissioning Plan, and are considered plant closure activities in preparation for decommissioning. Section 2.2 provided a description of the facility and site characteristics. The activated and contaminated portions of Fort St. Vrain which will be decontaminated, dismantled and removed during the decommissioning process are identified in Sections 2.2,2.3 and 3.1. The specific tasks to be performed to accomplish this goal are discussed in this section. Although personnel conducting the dismantling activities will be exposed to radiation above background levels, the dismantling and decontamination activities, have been developed to limit exposure to and control radioactive material in order to maintain occupational doses As Low As Reasonably Achievable (ALARA).
Exposure estimates to accomplish the individual tasks and overall project are also provided.
To accomplish the decommissioning of Fort St. Vrain, substantial portions of the existing plant will be dismantled and removed. However, Reador and Turbine Building components and structures which are not radioactive above releasable limits will remain.
The decommissioning project is divided into three major work areas:
1.
Decontamination and dismantlement of the PCRV.
2.
Decontamination and dismantlement of the contaminated or potentially contaminated balance of plant systems.
3.
Site cleanup and final site radiation survey.
Site cleanup involves pre-and post-decommissioning surveys of the site, and the radiological decontamination necessary to meet the regulatory guidelines to allow release for unrestricted use. These activities are discussed in detailin Section 4 and are not addressed in this section.
2.3 1 m '
N DECOMMISSIONING PLAN REV 2 SECTION 2 PCRV Decontamination and Dismantlement Activities The following are the major activities involved in dismantling and removing the radioactive portions of the PCRV. These activities will be discussed in further detail in the following Section 2.3.3:
1.
Initial PCRV Preparation
)
2.
Removal of the Helium Circulators 3.
Steam Generator Disassembly
- a. Initial Preparation
- b. Renoval of Steam Generator Secondary Assembly 4.
Removal of Activated Components using the ATC and FHM 5.
Detensioning and Removal of Pretensioned Tendons 6.
Flooding of the PCRV 7.
PCRV Top Head Concrete and Liner Removal 8.
Dismantling PCRV Core Components 9.
Removing the Core Barrel 10.
Removal of the Core Support Floor 11.
Disassembling the PCRV Lower Plenum 12.
Final Dismantling, Decontamination, and Cleanup Activities A technical evaluation is provided in Section 2.3.2 which provides the basis for the technical approach selected to decontaminate and dismantle the PCRV. A brief description is also provided to identify various techniques which were considered for removal of the PCRV activated concrete.
Balance of Plant System Decontamination and Dismantlement Activities The balance of plant systems that are contaminated or potentially contaminated above releasable limits and will require decontamination or dismantlement are identified in Section 2.2.3.
Work activities associated with these systems are discussed in paragraph 2.3.4 of this section.
2.3.2 Technical Approach Selection 2.3.2.1 Ootions Considered for Removal of the PCRV Key elements of the decommissioning plan include the techniques to be used to remove the internal components from the PCRV and to remove the activated concrete 2.3-2
REV 2 DECOMMISSIONING PLAN SECTION 2 from the PCRV structure. This technical approach is based on filling the PCRV with water for shielding while internal components are being removed and using diamond-wire cutting to remove the activated concre'.c from the PCRV structure.
These metnods provide the decommissioning project with the optimum schedule, cost, ALARA, risk, and safety considerations for decommissioning the PCRV. A detailed description of the PCRV disassembly techniqu2s and the basis for selecting them are de chbed below.
Two basic methods to disassemble the PCRV were considered: (1) in-air (dry) disassembly, and (2) filling the PCRV with water to provide shielding. Two possible meiocis of in-air dismantlement were also evaluated, considering factors of ALARA, safety, risks, schedule and cost. The two in-air methods evaluated were fully remote disassembly through the refueling penetrations in the top head, and partidly remote disassembly from a massive shielded work platform with the top head removed. The following paragraphs provide an evaluation of each method, discussion of advantages.
and disadvantages, and a determination of its acceptability.
2.3.2.1.1 E' illy Remote. In-Air Disassembiv:
The fully remote, in-air approach to the PCRV disassembly relied upon the extensive use of complex remote tooling and resultant limited view of dismantlement operations, which would produce less than predictable results. Although use of remote operations would potentially result in the best ALARA and safety records, all activities would be performed with highly specialized robots. Therefore, the risk of failure or project delays would be greater due to potential breakdowns or delays, lack of reliable backup techniques, and Ir.:k of adequate contingency plans. Design, fabdcation and testing of specialized robotics would also have to occur in a relatively short period of time, which could cause unnecessary delays in the project schedule.
Additionally, removal of the CSF would be extremely difficult, since it is too massive (270 tons) for practical remote removal.
2.3.2.1.2 Partially Remote. In-Air Disassembiv:
Partially remote in-air disassembly of the PCRV relied upon a massive shielded work platform that would be required to protect workers from radiation exposure during disassembly.
Access ports would be required in this platform through which hand-held, pole-type tools could be inserted to perform the disassembly when the platform is properly indexed over the work location. Using this approach, radiation exposure would be increased because of the extended stay times resulting from 2.3-3
DECOMMISSIONING PLAN REV 2 SECTION 2 restricted tool access. Removal of the top head and the top of the PCRV liner for installation of the work platform would be difficult because of high radiation levels and would probably require remote operations.
2.3.2.1.3 Floodine the PCRV:
The final approach evaluated was to flood the PCRV cavity with water. This selected approach will provide optimum shielding and contamination control and will allow the PCRV disassembly to be completed with optimum balance of schedule, cost, ALARA exposure and minimum risks. Additionally, there is an inherent added measure of safety due to the passive nature of the water for shielding and contamination control. Dismantling operations are greatly simplified by "line of sight" manipulations as a result of direct viewing of the entire cavity.
2.3.2.1.4 Conclusion 1
The evaluation of the two dry" (in-air) approaches against the " wet" approach for i
PCRV disassembly favors filling the PCRV with water for shielding during disassembly. Additionally, it is noted that the " dry" techniques are not completely dry, since large volumes of water are required for any abrasive process used to cut i
the activated concrete inside the PCRV. Therefore, water would be introduced into the PCRV in each of the " dry" dismantlement options considered.
2.3.2.2 Techniques Considered for Removal of PCRV Activated Concrete Diamond wire cutting and abrasive water-jet cutting were evaluated for removing activated concrete from the PCRV walls. Diamond wire cutting was chosen as the method for cutting most of the concrete into sections because this proven technology lends itself well to the PCRV concrete removal activities.
Abrasive water-jet cutting was determined to be feasible for much of the concrete i
cutting but has been minimized to limit the production of contaminated abrasive waste i
and because of related ALARA considerations. The abrasive water-jet is presently being considered for one application, cutting of the CSF.
i The follov:ing techniques were also evaluated and were determined to.;e less
)
desirable for the following reasons: (1) Expanding grout and explosives could be used to break apart the PCRV concrete, but were less desirable because of the heavy
')
reinforcement of the concrete and the presence of the PCRV liner on the face of the l
2.3-4
. REV 2 DECOMMISSIONING PLAN SECTION 2 concrete; (2) Thermal techniques were evaluated but were less desirable due to tool positioning difficulties, which could cause cost and schedule concerns; (3) Mechanical impact was evaluated but was less desirable due to structural considerations (with the exception of removal of portions of the lowest concrete in the top head).
2.3.3 PCRV Dismantlement and Decontamination 2.3.3.1 Overview of PCRV Dismantlement Activities The major decommissioning task is the dismantlement and decontamination of the radioactive portions of the PCRV A description of the PCRV is Irovided in Section 2.2 and illustrated on Figure 2.2-6. It should be noted that the steps identifico in the following paragraphs represent preliminary planning -and may change during the detailed engineering and work development that will occur during the planning phase.
This section provides a description of the expected steps necessary to dismantle and decontaminate the PCRV. Initial dismantlement of the PCRV willincipde removal of selected PCRV internal components and removal of portions of the steam generators. The selected internal PCRV components will be removed from the upper portion of the PCRV using the Fuel Handling Machine (FHM) and Auxiliary Transfer Cask (ATC). These components may include the 37 control rod metal clad reflector blocks (MCRBs), all but six of the 276 non-control rod hexagonal MCRBs, and certain helium purification components. Simultaneously, the non-contaminated portion of the steam generators (also called the steam generator secondary assemblies) will be removed from the lower portion of the PCRV to provide access for detachment of the contaminated steam generator primary assemblies (See Figure 2.2-6).
To facilitate the removal of the remaining reactor core components, the reactor cavity will be flooded with water. As discussed in Section 2.3.2, flooding the PCRV will provide shielding for the workers associated with PCRV dismantlement activities.
After the steam generator secondary assemblies are removed from the bottom of the PCRV, the PCRV bottom head and side wall penetrations will be sealed, the PCRV Shield Water System will be connected, and the PCRV will be flooded.
To gain entry to the PCRV cavity, a plug of concrete will be removed from the top of the PCRV. Selected PCRV prestressing tendons (See Figure 2.2-6) will be either (1) detensioned and removed, or (2) detensioned and left in-place. The top head plug will be cut into sections of appropriate size such that the weight and dimensions will 2.3-5 er
J 1
DECOMMISSIONING PLAN REV 2 SECTION 2 allow them to be handled with the Reactor Building crane and permit them to be moved out of the building. After the majority of the concrete has been removed from the PCRV top head, the 3/4-inch steel PCRV liner plate will be cut and removed with the remaining concrete, together with the top head liner insulation. A detailed discussion of this activity is provided in Section 2.3.3.7.
Once access is gained to the PCRV cavity, a work platform will be installed at the approximate elevation of the top of the PCRV where the liner and concrete have been removed. Working from this platform, workers will remove core components, including the remaining MCRBs, defueling elements, hexagonal reflector blocks, large side reflector blocks, side spacer blocks, core support blocks and core support posts. This activity is described in Section 2.3.3.8.
Once the core internals have been removed, the core barrel (a large carbon steel cylinder) will be removed by cutting it into pieces sized to fit in radwaste containers.
(See Section 2.3.3.9)
Following removal of the core barrel, the PCRV water level will be lowered and the CSF insulation removed, in preparation for removal of the CSF. The CSF is a 29-foot diameter, 5-foot thick disk of reinforced concrete within a 3/4-inch steel casing weighing approximately 270 tons. The CSF will be detached from the twelve CSF columns and the twelve steam generator inlet ducts and lifted with a hydraulic jacking system to the PCRV top head region. Thejacking system will then lower the CSF onto supports on the ledge in the cavity where the PCRV top head was removed. Once supported, the CSF will be sectioned into segments small enough for handling by the Reactor Building crane. This activity is discussed in Section i
2.3.3.10.
Once the CSF is removed, the PCRV lower plenum is exposed and the helium circulator diffusers and steam generator primary modules can be removed. These activities are discussed in Section 2.3.3.11.
The removal of the steam generator primary assemblies completes the removal of the major PCRV radioactive components. Remaining radioactive components include the activated " beltline concrete" around the reactor core region, the PCRV liner, liner insulation and insulation cover plates, and the PCRV lower floor with its supports.
The activated beltline concrete is the PCRV region that was aJjacent to the reactor core. It is estimated that this activated region is defined by a cylinder with an 18 to 24 inch wall thickness and a height of 40 feet. This section of PCRV sidewall will 2.3-6 I
i 1
e
REV 2 DECOMMISSIONING PLAN SECTION 2 be removed by cutting and removing vertical segments. The activated liner plate, insulation and cover plates will be removed with the concrete. These activities are discussed in Section 2.3.3.12.
In the lower portion of the PCRV cavity (below the CSF), the insulation and insulation cover plates will be removed from the PCRV liner. The lower floor and all support members, insulation and other components will be removed, and the exposed PCRV liner will be surveyed and decontaminated as appropriate. These activities are also discussed in Section 2.3.3.12.
2.3.3.2 Initial PCRV Preparation Initial tasks to be completed in preparation for dismantling the PCRV will include acquiring tooling, setting up training mockups, installation of the PCRV Shield Water System, and craft personnel training in accordance with Section 2.6.
Preparation activities include any modification:, or revisions to existingJacilities and equipment and installation of new facilities and equipment that would be necessary for their use in supporting the decommissioning operations. No major facility modifications are required that will affect the safety of the facility.
l The refueling deck equipment hatch and truck bay door may be enlarged to allow passage of larger items. The Reactor Building crane may also be re-reeved to i
provide additional vertical travel which will allow the 170-ton main hook to travel from the refueling floor to ground level. This re-reeved configuration is consistent with the crane configuration used during original plant construction and may be j
necessary to provide the lifting capacity to lift heavy loads, such as large concrete sections, when components are removed from within the PCRV.
The need for extensive waste handling facilities in addition to those already present has been minimized by proper sequencing of the dismantlement activities and by proper management of the radioactive waste program, as described in Section 3.3.
Off-site facilities will be utilized when necessary and practical for waste processing and final packaging. Proper task planning and sequencing will aid in minimizing accumulation of radioactive waste on site.
A self-contained mobile laundry facility to clean all contaminated protective clothing will be utilized. The PCRV Shield Water System, installed to maintain water purity in the flooded PCRV, will be discussed in Section 2.3.3.6.
2.3-7
DECOMMISSIONING PLAN REV 2 i
SECTION 2 Following helium circulator machine assembly removal, as identified in Section 1.5.2 (See Figure 2.3-1), the diffuser shutoff valve assemblies will be disconnected from the PCRV penetrations and the penetrations will then be sealed by installing a closure fixture designed to withstand pressure when the PCRV is flooded with water.
2.3.3.3 Steam Generator DisassembLv 2.3.3.3.1 Initial Steam Generator Disassembly 1
Each of the twelve steam generators consists of a primary assembly and a secondary assembly (Figure 2.3-2). The primary assembly is located within the PCRV lower plenum and the secondary assembly is located beneath the primary assembly inside a PCRV bottom head steam generator penetration.
The primary assembly is l
contaminated and the secondary assembly is not expected to be contaminated.
However, in order to remove the prirrary assembly, the secondary assembly must first be removed from beneath the PCRV.
The removal of the insulation from the steam ge.nerator secondary side piping will be limited to the sections of feedwater, main steam, hot reheat, and cold reheat piping that need to be severed for the steam generator secondary side removal. Prior to removal, the insulation will be tested for asbestos content. If asbestos is present, appropriate controls will be implemented for removal of the insulation. Following the removal of the insulation, the main steam, feedwater, hot reheat and cold reheat piping will be cut which will allow the secondary side of the twelve (12) steam generators to be removed.
2.3.3.3.2 Egmoval of Steam Generator Secondary Assembly Removal of the steam generator secondary assemblies (See Figure 2.3-2) will be accomplished in the reverse of the original construction insta'!ation sequence. The steam generator secondary assemblies are expected to be free of contamination.
The Marmon clamp (See Figure 2.3-2) will be removed from the lower end of the steam generator secondary assembly. This will allow withdrawal of the hot reheat piping from the steam generators. Because of the length of the hot reheat pipe, it will be severed into several sections as it is being withdrawn from the steam generators.
l 2.3-8
REV 2 DECOMMISSIONING PLAN SECTION 2 The cold reheat pipe will then be severed by remote operations at the threaded connection below the primary closure dome. Severing this connection remotely will make it unnecessary to send an individual inside the cold reheat pipe, as was done during installation. After the top of the cold reheat pipe has been cut, the lower reheat nozzle assembly will be cut free of the steam generator secondary assembly at an elevation below the feedwater ring header. This will allow the withdrawal of the cold reheat pipe from the steam generator for disposal.
After the cold reheat piping has been removed, the 40 feedwater, instrument, and steam tubes will be cut remotely below the primary closure dome. The steam generator secondary assembly will then be rigged for lowering. The secondary closure weld will be cut and the steam generator secondary assembly will be lowered out of the PCRV penetration liner. The Rucker machine, which is a large turntable designed to handle heavy loads under the PCRV, will be used to handle the steam generator secondary assemblies in the reverse order of the installation operations.
In order to detach the primary assembly from the penetration liner, the. final step will be to remove the 36 nuts that attach the primary module to the penetration liner flange. The steam generator primary assembly is stabilized by the steam generator shroud connection to the lower floor and the helium duct connection to the CSF.
After each of the twelve steam generator primary assemblies is detached from its respective penetration, it will then be removed through the top of the PCRV after the 3
CSF is removed. This is discussed further in Section 2.3.3.11.
1 When cutting operations have been completed, the interior of the penetration liner may be sprayed with a strippable coating to ease future decontamination operations.
J A new secondary closure plate will be welded in place to seal the penetration liner in preparation for flooding the PCRV.
In parallel with the removal of the steam generator secondary assemblies, the PCRV lower plenum (See Figure 2.3-3) will be entered through the PCRV bottom head access penetration after removal of the shield plug. A radiological survey of this area will be performed to determine radiation levels and major contributors in this area.
Still photographs and video recordings will also be made to assist in mockup design and training for eventual dismantlement of the PCRV lower plenum.
2.3-9
DECOMMISSIONING PLAN REV 2 SECTION 2 2.3.3.4 Removal of Activated Comoonents Usine the ATC and FHM Selected activated components will be removed from the PCRV using the ATC and the FHM. Use of this equipment will provide shielding while transferring highly radioactive components from the PCRV to shipping casks with minimal personnel exposure. The 37 control rod MCRBs and all but six of the 276 non-control rod hexagonal MCRBs will be removed from the PCRV by the FHM. Removal of the highly radioactive components with the FHM is the preferred method to maintain personnel exposure ALARA.
However, use of the FHM for this purpose is dependent on its operability, and its availability has not been relied upon as the basis for removal of these components. Removal of certain componeats in the helium purification wells and penetrations, and placement of the refueling sleeve, will be performed by the ATC.
As identified in Section 1.5.2, the RCDs, CRDOAs, and high temperature helium purification equipment may have been previously removed.
2.3.3.5 D_etensionine and Removal of Pretensioned Tendons 2.3.3.5.1 Tendon Removal Concurrent with operations discussed in Sections 2.3.3.2 through 2.3.3.4 is the detensioning of selected tendons in the PCRV. The PCRV has a total of 448 prestressing tendons made up of vertical, circumferential, and top and bottom cross head tendons (See Figure 2.2-6). The following identifies the number of tendons planned to be detensioned and the number of tendons planned to be both detensioned and removed for each of the various tendon types. The exact tendons affected could change somewhat based on conditions encountered during decommissioning.
Number of Number to No. of Tendons Type of Tendon Tendons Be Dete'nsioned To Be Removed Vertical 90 90 90 Circumferential 310 70 24 Top Cross Head 24 24 24 Bottom Cross Head 24 0
0 l TOTALS:
448 184 138 2.3-10 l
REV 2 DECOMMISSIONING PLAN SECTION 2 Temporary scaffolding will be erected to facilitate tendon removal. Tendons will be detensioned by cutting or grinding individual tendon wire buttonheads.
2.3.3.5.2 Vessel Integrity The modified PCRV structure was evaluated for the loadings produced by the dead weight of the PCRV structure and components, the lifting operations of the CSF, and a design basis seismic event. The structural evaluation considered the detensioning and removal of all 24 top cross head tendons, all 90 vertical tendons, and all circumferential tendons from group 9 through group 19.
In addition, it was conservatively assumed that all circumferential tendons (inner, middle and outer) were detensioned even though it is planned to only detension the inner and middle i
tendons in the top head and the inner tendons in the belt line region.
To evaluate this modified structure, a simplified free-body lumped mass model fixed at the basement floor of the PCRV structure was developed for analysis with the STAAD-III/ISDS (Ref. 5) computer code. Inputs to the analysis included NRC Regulatory Guide 1.60 " Design Response Spectra for Seismic Design of Nuclear Power Plants" (Ref. 6) design response spectra normalized to the Fort St. Vrain specific " double design earthquake" (referred to as the Design Basis Earthquake (DBE)) ground motions with NRC Regulatory Guide 1.61 " Damping Values for Seismic Design of Nuclear Power Plants" (Ref. 7) damping values. Specifically, the FSV Operating Basis Earthquake (OBE) ground motions of 0.05g horizontal and i
0.033g vertical accelerations and DBE ground motions of 0.10g horizontal and 0.067g vertical accelerations were used in the analysis. In addition, the concrete has a compressive strength of f'c = 6000 psi and the reinforcing steel was conservatively assumed to be Grade 40 steel with a F = 40,000 psi, although it is Grade 60 or y
better. Tne damping values 32e 2% horizontal and vertical for the OBE, and 2%
venical and 5 % horizontal for the DBE.
The modified structure was conservatively evaluated for the loadings produced by the dead weight of the PCRV structure (assuming the PCRV is flooded with water),
PCRV internal components, the lifting operations of the CSF, and OBE and DBE events. The resulting forces and rnoments in each of the individual cross sections of the PCRV were used to develop concrete compressive and reinforcing steel tensile stresses.
In the development of the concrete and reinforcing steel stresses, all vertical, cross-head and circumferential pre-stressing tendons were considered detensioned.
2.3-11 e-I
}
L DECOMMISSIONING PLAN
' REV 2 -
' SECTION 2-
.l The resulting concrete compressive and reinforcing steel tensile stresses are provided E
. below:
i i
CONCRETE l
REINFORCING '
~ CROSS LOADING COMPRESSIVE STEEL TENSILE SECTION STRESS (psi) q STRESS (psi)
Top Head DW 16.7 0.0 Lift -
7.7 0.0 -
OBE Seismic' 22.0 10.2 DBE Seismic 8-25.1 23.7-J L
Belt Line DW 134.6 0.0 W
Region Lift 6.1 O.0 OBE Seismic 298.9 814.8 DBE Seismic' 381.5 1.268.0
')
8
~ '
Includes the effects of Dead Weight (DW).
l 1
To determine the margins of safety, the worst case load combinations of dead weight, lift loads, and OBE and DBE seismic events were considered using the following equations, consistent with the " Building Code Requirements for Reinforced Concrete" (Ref. 8):
Equation-1:
U = 0.75 (1.4 DW + 1.7 Lift + 1.87 OBE Seismic) i Equation-2:
U = 0.75 (1.4 DW + 1.7 Lift + 1.4 DBE Seismic)
The total combined concrete compressive stress was compared to the " Limit Condition 2" compressive stress allowable of 0.85 f'c or 5100 psi as outlined by Section E.1.2.6.2 of the FSV FSAR. The reinforcing steel allowable tensile stress was considered to be 90% of the yield stress of Grade 40 steel, or 36,000 psi. The margins of safety, where margin equals (allo' able stress)/(combined actual stress),
w are summanzed below:
CONCRETE REINFORCING STEEL SECTION EQ-1 EQ-2 EQ-1 EQ-2 Top Head 87.6 95.0 2517 1447 Belt Line Region 8.97 9.28 31.5 27.0 2.3-12 1
....I
REV 2 DECOMMISSIONING PLAN SECTION 2 In summary, the structural evaluation has determined that the resulting concrete compressive and reinforcing steel tensile stresses are well within allowable limits.
Furthermore, adequate margin of safety exists for all loading conditions specified.
The potential for cracking of concrete in the modified top head and beltline regions has been reviewed and, considering the relatively low tensile stress in a conservative number of reinforcing bars, cracking due to tension in the concrete is not expected.
2.3.3.6 Floodine of the PCRV 2.3.3.6.1 Precaration for Flooding the PCRV Once operations described in Sections 2.3.3.2 through 2.3.3.4 have been completed, activities may proceed to install the PCRV Shield Water System and flood the PCRV.
A network of PCRV liner cooling tubes (System 46) and the tendon tubes within the PCRV concrete wall creates a potential pathway for water leakage and the spread of contamination during the cutting of the PCRV concrete. To block these potentialleak paths and prevent the spread of contamination, the liner cooling tubes and selected tendon tubes will be scaled with grout or other suitable sealing methods.
Before flooding the vessel, all PCRV penetrations that are below the PCRV waterline j
and have had their instrumentation removed (including instrument penetration internal l
components and other items ) will be sealed. These penetrations will be sealed by either one or a combination of the following: cutting and capping just outside the j
PCRV or by installation of bolted and gasketted blind flanges. Where weldmg is 1
utilized, the welds will be non-destructively tested per applicable codes.
l Some of the instrumentation was removed from PCRV penetrations prior to flooding the PCRV. High dose rates encountered on the core outlet thermocouple assemblies precluded their complete removal prior to PCRV flooding, and it was determined that underwater removal was preferable to limit occupational radiation exposures.
Therefore, preparations for flooding the PCRV included installation of a push rod assembly and redundant wiper seals in each core outlet thermocouple penetration, which will permit underwater removal of the thermocouple assemblies, as described i
in Section 2.3.3.8.6 (Ref.16).
Two PCRV low point penetrations (the bottom head access penetration and one helium circulator penetration) will be sealed with specially designed cover plates before the PCRV is flooded. These closures will provide alternate suction and fill paths for the PCRV Shield Water System described below. The welded connections 2.3-13
DECOMMISSIONING PLAN REV 2 SECTION 2 will be non-destructively tested. After verifying that the steam generator and helium circulator penetrations are sealed, the PCRV Shield Water System will be installed.
2.3.3.6.2 Exoected Conditions Within the Flooded PCRV 1.
Radionuclides:
The radionuclides of concern that will be encountered during dismantlement operations inside the PCRV have been previously identified in the activation analysis provided as Appendix II and are summarized in Table 3.1-2. A fraction of each of these radionuclides is expected to teach into the water from the graphite when the PCRV is flooded.
The principal radionuclides of concern are tritium, Co-60, Fe-55 and Cs-137. Of these, Co-60 is expected to provide the majority of the whole body exposure to occupational workers as a result of dismantlement operations. These radionuclides will appear in particulate and ionic form, and the PCRV Shield Water System will be designed to remove the principal radionuclides.
j Although not a major contributor to whole body exposures, the other major radionuclide of concern is tritium.
Since the tritium cannot be removed by 1
processing through filters or demineralizers, it will be processed and released using j
liquid effluent discharge operations in accordance with 10 CFR 20 limits. The j
maximum tritium concentration shall not exceed the limit specified in the Decommissioning Technical Specifications.
2.
Particulates: During PCRV dismantlement evolutions, debris will be generated from handling graphite blocks, concrete cutting operations, insulation, and dross from underwater cutting operations. Various size particles of debris are expected to be generated from the various cutting methods to be employed during PCRV dismantlement operations, including diamond wire cutting (PCRV top head), oxy-acetylene cutting, thermitic rod cutting, and underwater plasma are cutting. Large particles will settle downward and remain in the PCRV. Smaller i
particles will be circulated with the water and will be removed by the PCRV Shield Water System. Suitable provisions will be included in the system design to collect this debris and prevent it from damaging system components.
Some graphite dust is expected to become waterborne after the PCRV is flooded. The need to filter this graphite has been incorporated into the design and filter sizing of the PCRV Shield Water System. The possibility of breakdown of the Kaowool insulation (described in Section 2.2.2 and Figure 2.2-8), attached to the PCRV liner 2.3-14
a l
l REV 2 DECOMMISSIONING PLAN i
SECTION 2 immediately outboard of the core barrel, has also been considered. However, based on information from the manufacturer, this insulation is not expected to break down when immersed in water and therefore will not be a factor in the design of the system filtration trains.
2.3.3.6.3 PCRV Shield Water System - Design Considerations The primary function of the system is to provide water shielding to minimize 1
personnel exposure during dismantlement operations internal to the PCRV. The system will also be designed to provide a means to meet 10 CFR 20 discharge limits for the radionuclides identified above and ensure compliance with 10 CFR 50 Appendix I guidance for radioactive liquid waste discharges to unrestricted areas.
j Specifically, the system design will provide:
J (1) an acceptable method to reduce tritium inventory by liquid effluent discharge operations.
(2) an acceptable radioactive liquid waste processing path.to reduce the concentrations of fission and activation products for discharge to unrestricted areas, as well as control radionuclide concentrations in the i
PCRV water inventory to maintain occupational exposures in the work j
area ALARA.
1 In addition to these regulatory criteria, the system will also be designed to meet the following non-regulatory considerations:
(1) maintain acceptable water clarity to conduct undenvater dismantlement operations.
(2) minimize corrosion and biological fouling by suitable chemistry control.
(3) provide a means of initial fill of the PCRV, as well as the ability for makeup with demineralized water to compensate for system losses due to effluent discharges and evaporation.
The recommendations of Regulatory Guide 1.143 " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants" (Ref. 9) will be used in system design.
This system will be designed to maintain occupational radiation exposures within 10 CFR 20 regulatory limits and as low as is reasonably achievable (ALARA). The recommendations of Regulatory Guide 8.8 "Information Relevant to Ensuring That 2.3-15
DECOMMISSIONING PLAN REV 2 SECTION 2 Occupational Radiation Exposures at Nuclear Power Stations Will Be ALARA" (Ref.
- 10) will also be incorporated (to the degree applicable) into system design.
2.3.3.6.4 PCRV Shield Water Svstem - General Desien Information The PCRV Shield Water System is shown in Figures 2.3-4. The system will consist of two parallel trains of equipment, each sized for 500 gpm (or 50%) of the total flow. This total design flow rate (1000 gpm) will provide a turnover rate of approximately five PCRV volumes per day. Based on discussions with personnel involved in the TMI cleanup, this rate is considered adequate. The system will be designed to allow a complete train or individual components to be removed from service for preventive or corrective maintenance. Provisions will be included for the addition of another complete train if required. The trains are cross-connected to permit the pumps and filters to be used interchangeably between the two trains.
Maximum flexibility will be designed into the system to minimize the impact of individual component failure on system availability. Sufficient valves,,byp, asses and interconnecting piping will be utilized to allow continued system operation during scheduled maintenance or in the event of a component failure. Remote-indicating radiation detectors will be used to monitor dose rates on components in high radiation areas, such as at strainers, filters and demineralizers.
A.
Filtration Trains:
The purpose of the system filtration trains will be to maintain PCRV water clarity by removing suspended solids and particulate matter, and to reduce concentrations of suspended radioactive pa ticulates. In order to maintain optimum water clarity, suction of the PCRV Water Shield System will be taken from the bottom of the PCRV and clarified wner will be returned to the top of the vessel. The system will have two filtration tra:ns consisting of the following components:
l 1.
Clarifyine Pumo Suction Strainer:
A strainer will be installed in the suction line of each clarifying pump to prevent equipment damage due to large particulate debris. The strainers will be duplex type, and provisions for radiation monitoring of these filters will be included in the design.
2.
Clarifyine Pumo: Each train will have one clarifying pump.
The pumps will be horizontal, centrifugal process pumps. Each pump will 2.3-16
REV 2 DECOMMISSIONING PLAN SECTION 2 have a capacity of 500 gpm through the associated train of equipment and will return the clarified water to the PCRV.
3.
Filter Trains: Each filter train will consist of two filters, with the filters mounted in a series arrangement. Bypasses will be provided to allow each filter to be operated individually or in series with other filters.
The micron sizing of filter elements will be determined by on-going decommissioning operations. Filter element change out requirements will be based on a maximum differential pressure of 20 psid or a maximum radiation reading of 1 R/hr. It is expected that approximately five micron filters will be effective in removing particulates after the PCRV is flooded. Provisions will be included for shielding and radiation monitoring.
B.
Demineralizer Train:
The system will also be equipped with a sidestream demineralizer train. The purpose of the demineralizer train will be to reduce concentrations of dissolved.zadionuclides (specifically Co-60, Fe-55 and Cs-137) to levels that will allow discharge to unrestricted areas, as well as reduce concentrations in the PCRV to minimize radiation exposure to occupationally exposed personnel. The demineralizer train will consist of the following components:
1.
Demineralizers: The demineralizers enable removal of the principle dissolved radionuclides of concern (Co", Fe", Cs+) from the shield water, by means of ion exchange resin.
There are 8 roughing demineralizer vessels, arranged in parallel, and 2 polishing demineralizer vessels, arranged in series. The polishing demineralizers are used to further process water being routed to the radioactive liquid waste system (System 62).
During normal operation, 4 roughing demineralizer vessels will be in service, with a flow of approximately 50 gpm to each vessel. If the shield water contzins a high level of dissolved radioactive nuclides, a design flow rate of up to 400 gpm can be established through all 8 roughing demineralizers. A return line directs effluent from the roughing demineralizers back to the top of the PCRV.
A portion of the water discharged from the roughing demineralizers can be routed through the polishing demineralizers to the liquid l
waste system. Normal flow through the two polishing demineralizers is 10 to 50 gpm.
As water is discharged from the shield water system via the polishing demineralizers, an equivalent volume of demineralized water must be added to maintain PCRV level. Shieldmg is provided to protect personnel i
i 2.3-17
g.
u r
i
. DECOMMISSIONING PLAN REV 2 SECTION 2 o
working with the demineralizer system, and to keep general area radiation levels within acceptable limits. Remote radiation monitors are installed.
2.
Resin Fines Filter: One cage-type filter will be provided to i
prevent the loss of resin fines from the demineralizer and possible discharge
.j into the PCRV. This filter will be designed for a minimum capacity of 100 i
gpm and to retain 98% of all particles greater than 5 micron at 15 psid.
i C.
Chemical Addition Train:
i The system will also include a chemical addition train. The purpose of the chemical t
addition train will be to minimize corrosion by suitable chemistry control within the PCR.V system and to minimize biological fouling. The chemical addition train will
- consist of the following components:
\\
1.
Chemical AdditionTanks: Two 100-gallon chemicaladdition l
tanks will be included. For the initial PCRV shield water chemistry control l
program, described in Section 2.3.3.6.5, one tank was used to contain Calgon
]
LCS-20 (corrosion inhibitor) and the other hydrogen peroxide (biocide). The tanks will be used to add chemicals to the system for the maintenance of proper chemistry and to control biological fouling.
2.
Chemical Addition Pumos: Two chemical addition pumps l
will be included.
D.
Submersible Skimmer and Vacuum System:
Consistent with experience gained conducting underwater operations at TMI, a submersible skimmer and vacuum subsystem has been included in the system design to maintain adequate surface visibility, to reduce surface contamination and to provide vacuuming capability.
E.
System Controls:
j i
The system incorporates remote manual controls on a central control panel located in a non-congested area on level 1 of the Reactor Building. All major operations of the system, including flow adjustments and valve positioning, will be performed from
)
I the panel. Pumps will be controlled both locally and remotely from the central control panel. Alarms for either high differential pressure or high radiation will 2.3-18 i
i I
REV 2 DECOMMISSIONING PLAN SECTION 2 notify the operator of the need to replace filter elements or the demineralizer resins.
PCRV water level indication and PCRV high/ low water level alarms will be included in the design to facilitate system operation and control.
F.
Other General Desien Information:
The filters and demineralizers will be shielded to reduce radiation fields in the immediate vicinity of these components during operation. The system will have isolation valves at the outlets from the PCRV to the suction of the PCRV Shield Water System to allow isolation if a problem should develop in the system. There will also be other valves to allow isolation of portions of the system for maintenance or repair. A connection will also be provided for an additional train, if necessary.
The major components of the system will be prefabricated on skids with drip pans to contain potential leakage, and will be installed in low occupancy areas of the Reactor Building to minimize personnel exposure. Skids that include system filters will be shielded. The operating controls and chemical addition skids will be located on level 1 of the Reactor Building. The skids will be interconnected with other skids as well as with the suction piping from the bottom of the PCRV and the retum piping to the top of the PCRV. Drains from the PCRV Shield Water System will be directed to the existing Fort St. Vrain Radioactive Liquid Waste System (System 62) to permit the use of the existing effluent discharge paths and radioactivity monitoring and controls.
2.3.3.6.5 PCRV Shield Water System - Ooeration A.
Initial Fill of the PCRV After system installation and check out, the first operation of the system will be during the initial fill of the PCRV. As opposed to normal system operation, the initial fill will be from the bottom of the PCRV via the suction piping. The initial fill will be accomplished prior to the final cutting and removal of the PCRV top head concrete, and prior to gaining access into the PCRV internal cavity. Demineralized water for the initial PCRV fill (estimated to require approximately 325,000 gallons) will be from the portable demineralized water makeup system, which is described in a following section.
As the PCRV is being filled, the displaced air and gas will be passed through a portable HEPA filter system attached to a penetration in the PCRV head. Using 2.3-19
DECOMMISSIONING PLAN REV 2 SECTION 2 temporary ventilation ducting, the displaced air from the HEPA filter will then be routed to the existing Reactor Building Ventilation Exhaust System (System 73). The PCRV water will be sampled and analyzed daily during initial filling to verify that the tritium concentration remains below the Technical Specification limits.
In addition, the air displaced during the filling process will be sampled for tritium and analyzed daily until the filling operation is completed to ensure that discharges will be in compliance with the Fort St. Vrain Offsite Dose Calculation Manual (ODCM Ref.11). The PCRV will be inspected for leaks during the initial fill process. The fill operation is expected to take several days. After the PCRV has been filled, the air will be sampled and analyzed for tritium at the same frequency required by Technical Specification LC 3.4 for water until the air in the PCRV has been adequately characterized.
The Decommissioning Technical Specifications require that the PCRV water be sampled and analyzed daily for tritium concentration during the initial fill of the PCRV. Following the initial fill, sample frequency may be reduced to weekly after the tritium concentration has decreased to less than 0.1 Ci/cc.
i Lintts.]! ave been established in the Decommissioning Technical Specifications to assure that tritium activity concentrations in the PCRV Shield Water System will not exceed those postulated in the decommissioning accident analyses.
B.
Normal System Ooeration Once the PCRV has been filled, the PCRV Shield Water System lineup will be restored to take a suction from the bottom of the PCRV, with the return flow to the top of the PCRV. The system will be operated to establish and maintain water clarity, water chemistry and to minimize waterborne concentrations of radionuclides.
The demineralizer will be placed into service as required. Filters and demineralizers will be monitored for differential pressure and radiation levels to determine when replacement is required. It is expected that approximately 60 days will be available to operate the system to establish water clarity and reduce radionuclide concentrations before the PCRV top head is removed.
1.
System Recirculation to the PCRV:
The normal operational mode of the system will be with both trains processing PCRV water at a total flow rate of approximately 1000 gpm. Both clarifying pumps will take suction from the bottom of the PCRV (total flow rate - 1000 gpm) and will process the water through two parallel trains of filters. During system operation, water flows from the flooded PCRV through duplex 2.3-20
REV 2 DECOMMISSIONING PLAN SECTION 2 strainers to the suction of the clarifying pumps. From the pumps, the water is processed through the filters before returning to the PCRV. A minimum side stream flow of 20% of the water flow rate will be processed through the demineralizers for removal of radionuclides, namely Co-60, Fe-55 and Cs-137.
The filter trains will be set up in a series arrangement depending on the turbidity conditions in the PCRV. Initially, the system will be operated without filter elements installed.
Samples will be taken to characterize particulates and determine the appropriate filter micron sizes. As water clarity dictates, filter elements and filter alignment will be changed as needed to support ongoing operating conditions. The clarified water will be returned to the top of the PCRV.
2.
System Processing Via The Demineralizer:
A minimum :i& stream flow of aporoximately 20% of the total flow (200 gpm) will te taken from the return line downstream of the filtration units and pro;essed through demineralizers. The flow will be adjusted as required to maintain acceptable radiation levels on the work platform-to minimize personnel exposure. Effluent from the demineralizer train can also be routed back to the system return lines for recirculation to the PCRV or, after sampling, routed to the effluent discharge connections as described in the following paragraph. Suitable provisions will also be provided for additional demineralizer capacity as required.
3.
Discharee Via Radioactive Liouid Waste System (System 62d1 All liquid waste from the Fort St. Vrain decommissioning will be routed through the existing radioactive liquid waste system (System 62) discharge line i
that was utilized during normal plant operations. Further details of this system tre provided in Figure 2.2-23 and Section 2.2.3.10. Discharges will also be performed in compliance with the NPDES permit in effect at that time.
4.
Samoline Ooerations:
i Initially, any releases will be batch mode releases. Prior to liquid effluent discharge operations, representative samples obtained from the PCRV Shield Water System will be analyzed for principal radionuclides to ensure that the concentrations of radionuclides discharged to the environment do not exceed the values specified in 10 CFR 20. Samples will also be taken to verify maintenance of suitable water chemistry, water clarity and radionuclide concentrations. Sample locations will be included (see Figure 2.3-4) at the 2.3-21
DECOMMISSIONING PLAN REV 2 SECTION 2 suction line from the PCRV, at the outlet of the filter trains, and a: the outlet of the demineralizers.
5.
Chemistry Controh l
Two different chemistry control programs have been used in the PCRV shield j
water to minimize corrosion and maintain acceptable water clarity. Calgon l
LCS-20 was initially selected as the water treatment program for the PCRV l
shield water system. Calgon LCS-20 consists of a solution of sodium nitrite, l
sodium tetraborate and sodium hydroxide. The Calgon LCS-20 program used l
sodium hydroxide to maintain a basic pH (approximately 9 to 10), with nitrites l
to scavenge any oxygen in the shield water, thus minimizing corrosion of the l
carbon steel PCRV liner and liner insulation cover plates. Laboratory tests l
demonstrated that this water chemistry would be effective in inhibiting carbon l
steel corrosion. A polymerization compound was also added to resolve a l
colloidal suspension problem caused by entry of some concrete cutting slurry l
into the shield water. This water treatment proved adequate during removal l
of the graphite core components (including the core support blocks and posts),
l core barrel removal and removal of core support floor upper insulation.
1 l
Water clarity problems were encountered during core support floor upper l
insulation removal and steam generator inlet duct cutting operations, which l
were being performed by undenvater divers, as described in Section 3.2.12.
l During diving operations in the PCRV, air used by the divers continuously l
bubbled to the surface, depleting nitrites which react with oxygen to form l
nitrates. This necessitated the addition of substantial amounts of chemicals to l
maintain the desired nitrite concentration, resulting in formation of a complex l
colloid that caused deterioration in water clarity.
l l
The shield water chemistry control program was revised to end the LCS-20 l
water treatment, and permit addition of aluminum sulfate (" alum") to address l
the colloidal suspension problem and restore visibility. The revised shield l
water chemistry control program is more compatible with oxygenated water l
produced by underwater diving operations, using a neutral pH, without the l
addition of oxygen scavenging chemicals. The pH is maintained by periodic l
additions of sulfuric acid. A polymer (s) is added to flocculate materials, l
enhancing their removal. Laboratory tests determined that this water treatment i
l program would be effective in re-establishing water clarity, with a pH of I
l approximately 6.5 optimal, and that this program had significant benefits over l
use of other alternative water treatment programs. It is possible for some 2.3-22
?$ g '
, i REV 2 I
~ DECOMMISSIONING PLAN L
' SECTION 2 corrosion of carbon steel in the PCRV to occur using the second' water i
chemistry control piogram. However, it is considered that corrosion will be
-l
. minimal in the relatively short decommissioning. time frame, and a small~
l-amount of rust in the water will not significantly reduce visibility.
l Hydrogen peroxide will be added to the Shield Water system in batches, as necessary, to control biological fouling. A system will be provided for the -
l; purpose of adding chemicals to the PCRV shield water, consisting of two 100- '
j.
gallon tanks and two chemical addition pumps. The system will initially be -
used to add chemicals during the initial fill of the PCRV. The~ system will
-l-then be used to batch feed chemicals as required until tritium-levels are' reduced to a level where continuous effluent discharge operations are acceptable. To support a continuous discharge 6peration, it will be necessary.
to continuously feed chemicals.
- 6.
System Interfaces-The PCRV Shield Water System will interface with and require support from the existing site systems:
(1) Portable Demineralized Water Makeup System.
The PCRV Shield Water System will require a supply of demineralized water at a flow rate of up to 50 gpm at 100 psig. Demineralized water is required for system makeup, replacement of water removed by effluent discharge operations, chemical additions, 'and to replace evaporative losses.
The demineralized water supplied to the PCRV Shield Water System must meet typicalindustry standards for oxygenated, deionized water. The demineralized water for the initial PCRV fill and for makeup due to subsequent effluent-discharge operations will be from the portable demineralized water makeup system.
(2) Radioactive Liquid Waste System (System 62)
Tritium inventories will be initially controlled and subsequently reduced using effluent discharge operations, as necessary. The discharge from the PCRV l
Shield Water System will initially be to either the existing plant liquid waste.
l holdup and monitoring tanks for processing and subsequent discharge or, to the l
~
Reactor Building Sump (RBS), as described in Sections 2.2.3.10,3.3.2.2 and
.l 3.3.2.3.-
As tritium levels are reduced below the 10CFR20 MPC, the. 'l discharge may be permitted to flow directly from the PCRV to the radioactive l~
r liquid effluent discharge line, as discussed in Section 3.3.2.2.
l 2.3-23
+ - - -
y<
gf
+
1
~'
y-m DECOMMISSIONING PLAN :
REV 2 1
SECTION 2 (3) Electrical Power Tie-ins to the site supplied power of 480 VAC 3-phase will be required. The -
skids will ' be ' pre-wired and. transformersL provided to ' facilitate interconnections.'
(4). Compressed Air System-A source of dry compressed air at a nominal value of 90 psig is required to.
support system operations, particularly..for dewatering the filters and demineralizers.
a (5)' Heating, Ventilation and Air Conditioning,
j The PCRV Shield Water System will be located within the Reactor Building.
- to provide the required environmental conditions. No special or additional environmental conditions are required. The PCRV will require a temporary connection to the Reactor Building ventilation exhaust system to accommodate the displaced air during the initial filling of the PCRV.
~'
C.
System Maintenance j
Methods for handling the replacement of radioactive strainers, filter elements and the
- hange out of demineralizer resins will be designed for ease of replacement and will i
c
' incorporate ALARA concepts, consistent with the recommendations of Regulatory Guide 8.8 (Ref.10). These components will be shielded as necessary to. minimize occupational radiation exposures.
The system will have sufficient interconnecting piping and isolation valves to allow repair or maintenance on a portion of the system while the remainder of the system continues in operation. In the unlikely event of a leak within the system,' the entire i
system will be isolable from the PCRV, or that portion of the system with the leak 3
will be isolated. The strainers, filters, and demineralizers will be designed to minimize exposure during maintenance. The strainers will be provided with inserts for ease of handling during replacement. The filters will be provided with vents and '
! drains, and filter cartridges will be removed into shielded containers, if necessary, l
to minimize exposure. Filters and demineralizers will be shielded and strainers, filters and demineralizers provided with radiation monitors.
i D.
System Removal t
The system will be used to support ongoing decommissioning operations. When the 2.3-24
REV 2 DECOMMISSIONING PLAN SECTION 2 system is no longer required, it will be dismantled and treated as contaminated BOP equipment and piping. The system will be surveyed and decontaminated or disposed of as radioactive waste, as necessary.
2.3.3.7 PCRV Too Head Concrete and Liner Removal The PCRV top head concrete and top carbon steel liner will be removed in two phases:
Removal of 12 large sections of PCRV top head concrete.
Removal of a final approximately 1 inch thick layer of activated concrete and the top of the carbon steel PCRV liner.
It is planned that the PCRV top head will be cut using diamond wire techniques, and removed in sections that can be handled with the Reactor Building crane. These sections will be cut so as to leave a thin horizontal layer of concrete above the PCRV liner. The remaining layer of activated concrete and liner will-be removed by thermal methods (oxy-lance). This sequence is performed in this manner to prevent inadvertently breaching the PCRV liner and minimize exposure of equipment and personnel to radioactive material.
Prerequisite activities that are necessary to begin removal of the top head concrete include the following:
1.
Detensioning and removal of selected tendons as discussed in Section 2.3.3.5.1.
2.
Removal of selected highly radioactive components (control rod elements and metal clad reflector blocks) from the reactor core with the Fuel Handling Machine (FHM) as described in WBS Nos. 2.3.1.8.2 and 2.3.3.4 of the Decommissioning Cost Estimate and discussed in Sections 1.5.2 and 2.3.3.4.
3.
Plugging the PCRV cooling tubes, top head penetrations, and tendon conduits as discussed in Section 2.3.3.6.1 and described in WBS No.
2.3.2.3 of the Decommissioning Cost Estimate.
4.
Removal of helium purification equipment from PCRV top head wells using the Auxiliary Transfer Cask (ATC) as described in WBS No.
2.3.1.9 of the Decommissioning Cost Estimate and in Section 2.3.3.4.
2.3-25 l
l
I DECOMMISSIONING PLAN REV 2 SECTION 2 5.
Sealing of PCRV penetrations which are below the PCRV waterline and have had their instrumentation removed as discussed in Section 2.3.3.6.1.
6.
Removal of interfering piping, instrumentation, and electrical components.
7.
Flooding the PCRV prior to liner removal and acquiring access to the PCRV internal cavity as described in Section 2.3.3.6.
Plugging of the cooling tubes is a necessary requirement to mitigate the spread of contamination from the diamond wire cutting operation.
The refueling, high temperature filter adsorber, and access penetrations in the top head in the path of the diamond wire saw will be plugged prior to diamond wire cutting operations to limit the amount of cutting slurry entering the PCRV cavity. Certain penetrations will be designated for use to draw air from the cavity and provide a negative pressure in the cavity. This air will be exhausted to the Reactor Building Ventilation System (System 73) for discharge and will be monitored for concentrations of tritium and other radionuclides.
The first phase, removal of 12 large sections of top head concrete, consists of the following major activities:
1.
Seal the top head penetrations to prevent debris from entering the PCRV.
2.
Set up the core drilling machines on the external wall o' the PCRV to f
create six horizontal core drilled holes. (Figure 2.3-5).
3.
Thread the diamond wire through the intersection points of the cored holes to make a loop to allow cutting of the concrete (Figure 2.3-6).
4.
Make 12 venical core drilled holes to intersect with the horizontal cored holes (Figure 2.3-7). The first two vertical holes will be inclined to facilitate removal of the first concrete segment.
5.
Make the horizontal cut, then make the 12 vertical sectioning cuts and the 12 vertical back cuts using the diamond wire method (Figure 2.3-8).
6.
Rig the sections for removal.
'Ihe concrete sections will be cut using the diamond wire cutting process, consisting of a wire with collars containing a diamond-matrix and made to length for each individual cut, and a hydraulic pulley drive system to circulate the wire. The diamond wire is routed to envelop the cut area and then returned to a drive wheel on the drive system. The wheel rotates and pulls the wire through the cut areas.
2.3-26
REV 2 DECOMMISSIONING PLAN SECTION 2 I
i Hydraulic cylinders control the tension of the wire. Once the cut is started, the tension and drive wheel speed are adjusted to optimize cutting efficiency.
The diamond wire cutting process will utilize appropriate radiological engineering controls to contain the cutting slurry and control airborne radioactivity. The diamond wire saw uses water as a coolant and lubricant for the cutting process. However, the coolant water is independent of the PCRV Shield Water System. A water collection system will collect the cutting slurry, decant the slurry and recycle the water. In addition, airborne and loose surface contamination control will be achieved by containing the diamond wire path and drive units (s) in a containment tent (s) served by HEPA ventilation.
Removal of 12 sections of tophead concrete, of which the lower portions of each section may be activated, will be accomplished utilizing the Reactor Building crane.
This will leave a thin layer of activated concrete covering the PCRV liner. Specially engineered lifting attachments will be used to safely handle the heavy components, consistent with the requirements of 29 CFR 1926 and applicable ANSI standards.
The concrete sections will be moved to a containment tent on the refueling floor where each large wedge-shaped segment will be cut into three approximately equai size smaller wedges using the diamond wire cutting methodology to make two horizontal cuts. Whereas the 12 large segments removed from the PCRV will each weigh approximately 110 tons, and will require use of the Reactor Building crane's 170 ton hoist for removal from the PCRV and transport to a containment tent, the smaller wedges will weigh approximately 37 tons. The smaller wedges will be handled with the 50 ton hoist of the Reactor Building crane, which has the capability to extend down to grade level in the Reactor Building truck bay. Section 3.4.3 evaluates the postulated drop of one of these wedges while it is being lowered down the truck bay, assuming the wedge was located adjacent to the top head liner while it was in the PCRV (most highly activated concrete).
The first phase activity will account for the majority of the effort that will be spent to remove the top head concrete and liner. Occupational exposure is expected to be negligible during the core boring and concrete cutting operations on the top head due to the relatively low radiation fields external to the PCRV.
The second phase consists of removal of the final thin (approximately 1 inch thick) concrete layer and the top of the PCRV carbon steel liner. In this activity, the PCRV liner plate and adjoining concrete will be sectioned using a long handled (10 ft. initial length) thermal torch (" oxy-lance"). Concrete near the circumference of the steel liner will be sectic. ed with a circular concrete saw. The concrete saw is a hydraulic-2.3-27
~
i DECOMMISSIONING PLAN REV 2 SECTION 2 driven, track-mounted, manually-operated saw. The concrete saw will be used to score the cut lines around the circumference. Jack hammers will then be used to chip away the concrete around the circumference to expose the carbon steel liner.
Prior to sectioning of the top head liner, the Decommissioning Rotary Work Platform (DRWP) will be installed on the ledge formed by the horizontal concrete cut (Figure 2.3-10). This platform will be a rotating platform with openings to provide access to all sections of the PCRV. The rotating platform is approximately 36 feet in diameter and rides on a circular track (rails) set 15 feet below the top of the PCRV.
A stationary platform will fill in the voids between the circular rotating platform and the hexagonal opening cut in the PCRV. The Airborne Contamination Control System (ACCS) provides a flow path for air to move from the refueling floor down through the tool slots of the DRWP, then upward through exhaust ducts from the stationary work platform over the PCRV wall to the Reactor Building Exhaust system. The Reactor Building exhaust fans p ovide the air flow for the ACCS.
ACCS ducting and shield water system piping pass through openings in the stationary platform. Openings or tool slots in the rotating platform provide access to in-vessel components. A wiper seal between the rotating and stationary platforms minimizes air flow between the two platforms so that most of the air flow is downward through the tool slots. The tool slot openings are sized to ensure a downward flow of air.
The ACCS provides for control of contamination during cutting of the PCRV top head liner. Smoke and airborne activity generated by the liner cutting operations will be pulled down through the DRWP tool slots, pass through a filter in the ACCS ducting, then through the Reactor Building ventilation exhaust system, including the System 73 HEPA filters and activity monitors, before being released from the Reactor Building. Therefore, workers performing the liner cutting operations will not be exposed to significant levels of airborne activity. Cutting operations will cease in the event ofloss of ACCS ventilation flow.
The thermal torch (oxy-lance) will be used to cut through the thin layer of concrete remaining, the liner, insulation and cover plates to free sections for removal (see Figure 2.3-9). The layout and sequencing of cuts will take into consideration the structural stability of the liner during the disassembly process.
The concrete / liner / insulation segments will be removed and packaged for disposal.
Of the two phases, the second phase represents the greatest potential for personnel exposure. The PCRV liner plate and the remaining thin layer of activated concrete will be uncovered as the segments of the top head concrete are removed. The PCRV liner plate is estimated to have radiation levels of up to 600 mrem /hr on contact.
This estimated exposure rate is a conservative interpretation of information provided 2.3-28
~.
n u
REV 2 DECOMMISSIONING PLAN SECTION 2 -
I in the activation analysis for the bottom side of the cover plate, insulation, liner plate '
aiid activated concrete. Shielding for the workers will be utilized as appropriate for the close operations such as installing the saw tracks, operation of the saw' and.
4 M
thermal cutting.
2.3.3.8 Dismantling PCRV Core Comoonents
=2.3.3.8.1 General'Descriotion - Graohite Blocks Following the removal of the PCRV top head and installation of the work platform, PCRV core components will be removed. These activities will include the removal of various types of graphite blocks and other reactor internals within the core barrel down to the CSFl A listing of the types of graphite blocks that will be removed during decommissioning activities is identified in Table 2.3-1.
The top layers of blocks (i.e., metal-clad reflectors and some hex reflector blocks j
without Hastelloy cans) will be removed using the Fuel Handling Machine (FHM) and current plant methods to transfer the components from the PCRV~to a shipping q
container, as discussed in Section 2.3.3.4 and WBS No. 2.3.1.8.2 of the -
Decommissioning Cost Estimate.
Use of the FHM will provide the necessary shielding and containment while transferring components from the PCRV to the shipping containers with minimal personnel exposure. The FHM may be used to remove blocks other than those on the top layer of the core, including defueling_-
elements and hexagonal graphite reflector blocks. However, use of the FHM for this
]
purpose is dependent on its operability, and its availability has not been relied upon as the basis for removal of these components. The following sections provide a discussion for removal of these graphite blocks using manually operated tools.
2.3.3.8.2 General Arrangement of Work Area for Graohite Block Removal
.The arrangement of the work area that will be typically used for removal of all types of graphite blocks is provided in Figure 2.3-10. The PCRV will have been filled with shield water to a level approximately 4 feet above the graphite blocks, but below -
the top of the PCRV liner. Suitable controls will be implemented to prevent water from splashing or the water level from approaching the exposed concrete above the top of the PCRV liner.
These controls are necessary to prevent potential contamination of the concrete.
The Work Platform will have been installed on the ledge at the bottom of the hex ~
opening in the PCRV. The Work Platform will be designed with the capability of I
rotating to provide access to all areas of the core. It will have two access openings 1
2.3-29
-s-
,,,-,---g y
e
DECOMMISSIONING PLAN REV 2
' SECTION 2 to allow insertion and removal of tools and components, which will permit operations to proceed in parallel. A floor will be installed isetwcen the platform and the walls of the PCRV at the level of the Work Platform. Them will be three jib cranes installed on the refueling floor level to service the access openings in the platform.
The Reactor Building crane will also be available to service the platform and the remainder of the refueling floor area.
A ventilation system will be installed to provide control of airborne contamination, including tritium. Air will be drawn from the refueling floor to the Work Platform, down through the access openings in the platform, and then exhausted to the Reactor Building Ventilation (exhaust) System (System 73) for discharge. The discharges from the Reactor Building Ventilation System will be monitored in accordance with the FSV Offsite Dose Calculation Manual (ODCM) (Ref.11). The airflow from uncontaminated areas to contaminated areas through the Wo'rk Platform will minimize personnel exposure to airbome contamination.
The area on the Work Platform will be quite large, approximately 43 feet across the corners of the hexagonal opening. This will provide the capability to movepersonnel on the Work Platform to a considerable distance away from an operation if a significant radiation field is encountered. During dismantlement operations, workers on the work platform will be protected from direct radiation and airborne contamination during removal of core components from the open PCRV. Radiation protection features include:
Core dismantlement will be performed underwater, shielding workers and minimizing airborne particulate radioactivity.
PCRV Shield Water System will strip soluble radionuclides from the j
shield water. Tritium inventory control is discussed in Section 3.3.2.3 i
of this plan.
The ventilation system will ensure a positive downward flow of air over the workers. Exhaust ducts under the work platform will carry air through a filter, then to the existing plant ventilation system.
J Procedures and equipment for core dismantlement and operation of the work platform will be provided to minimize radiation exposure to workers.
All work will be performed in accordance with approved Radiation i
Work Permits.
4 2.3-30
7; 4
REV'2 ~
DECOMMISSIONING PLAN' SECTION 2 -
2.3.3.8.3 Special Considerations During Graohite Block Re:noval~
p
_ The graphite block removal. tasks represent a significant portion (22%) of the:
project's total person-Rem estimate. Due to the repetitive nature of the tasks, even small successful reduction measures will result in a significant savings of cumulative exposure. Although this process will benefit from additional future reviews and -
improvements, the following considerations are being taken to reduce personnel L__
exposures for this series ofjobs:
L.
Use of the PCRV Shield Water System.
Use of the Work Platform will improve worker efficiency and safety.
Utilization of a ventilation system to move evaporated tritium and other
. airborne contaminants away from the work, area under and around the.
platform.
Use of long handled tools and submerged staging areas to perform potential high exposure activities underwater.
Use of temporary shielding as appropriate to maintain exposures ALARA.
Installation of additional area radiation monitors (ARMS)'widt local' alarm features to detect unexpected dose rates around the platform work areas.
I Audio-visual communication equipment to support remote surveillance of.
i activities and equipment operations.
Use of shielding bell (s)..
i 2.3.3.8.4 Prereauisites for Graohite Block Removal l
Prerequisite activities that are necessary to begin removal of the graphite blocks include the following:
l 1.
Flooding the PCRV with shield water as described in Section 2.3.3.6.
2.
Removal of the top head concrete and liner as oiscussed in Section 2.3.3.7.
3.
Installing the PCRV work platform as discussed in Section 2.3.3.7.
4 Radiological survey of the work area and installation of temporary shielding if necessary for ALARA purposes, l
2.3.3.8.5 General Graohite Block Removal Secuence The following is the general sequence of operations that will be used for removal of all graphite blocks:
1 2.3-31 O
DECOMMISSIONING PLAN REV 2 SECTION 2 1.
Removing 5.
Unloading 2.
Staging 6.
Packaging 3.
Loading 4.
Transferring Since this general sequence of operations will be used for the removal of all types of graphite blocks, the discussion of the six steps provided below are applicable to the removal of the defueling elements, hex reflectors (with and without hastelloy cans),
the large permanent reflector blocks, side spacer blocks with boronated pins, and core support blocks and posts. Relative locations of the graphite blocks within the core area around the circumference of the core, are shown in Figure 2.3-11. The activities that are specific for one type of block are discussed in the subsection following this general discussion.
(1) Removing:
The blocks will be lifted from their position in the PCRV core area, and placed in a transfer basket that is below the surface of the water (see Frgure 2.3-12).
This will be accomplished using remotely engaged long handled tools (LHT's) attached to an overhead jib crane that is operated by personnel on the Work Platform. The workers will be working from the Work Platform that will be installed over the flooded PCRV. The tool for handling hex reflectors, defueling elements, and large permanent reflector blocks will be an expanding collet type similar to that used in the Fuel Handling Machine (FHM). The er.d of the tool will be inserted into the reverse counterbored hole in the top of the block with the end of the tool retracted. The end of the tool will then be expanded in the larger diameter in the lower ponion of the hole and the block will be lifted.
The side spacers will primarily be handled by attaching a lifting bail to the top of the block using the existing threaded holes in the graphite block. Some of the side spacers may be lifted by means of a bucket device. The hex shaped core support blocks will be handled using a three pronged gripper tool. The prongs will be inserted into three of me six holes in the core support block and then contracted to grapple the block. The irregular shaped perimeter Door blocks will be handled using a fork lift type tool since there are no holes in the block which can be used for grappling. The core support block posts will be handled by a tool which slips over the top of the post and is contracted to grip the post.
(2) Stacine:
After removal from the PCRV core area and while still submerged, the blocks will be lifted and placed in a transfer basket located on a transfer stand attached 2.3-32
REV 2 DECOMMISSIONING PLAN SECTION 2 to the work platform (see Figures 2.3-12). During this operation, the block will n
remain submerged underwater. The LHT will be disengaged and removed, leaving the block temporarily stored in the basket.
t (3) Loadine:
A shield bell will then be lowered into position over the transfer basket and I
seated on the work platform, as shown in Figure 2.3-12. A grappling tool will be lowered from the inside of the shielding bell, engage the transfer basket, and lift the basket into the shield bell. The shield bell guide pins in the transfer stand will provide the necessary alignment for engagemer.t of the tool. The actual raising of the basket will be accomplished in a few minutes. After the basket has been raised into the shield bell, water contained in the basket and blocks will drain back into the PCRV through hol~es in the bottom and sides of I
the basket.
The shield bell will then be lifted to just above the floor of the Work Platform, and a catch pan built into the shield bell bottom (see Figure 2.3-13) will be installed under the shield bell to cortain possible drippings of contaminated water. The shield bell bottom will be strong enough to retain the basket in the shield bell in the unlikely event that the grappling mechanism should fail. The bottom will also provide shielding at the bottom of the shield bell. During loading operations, radiation levels in the immediate vicinity of the shield bell will be closely monitored and personnel access to the affected area will be limited by administrative procedural controls.
The expected dose rates on the Work Platform, both with and without the shielding bell, are shown in Figure 2.3-15 for the large permanent side reflector block, in Figure 2.3-16 for the hex reflector blocks without hastelloy cans, in Figure 2.3-17 for the hex reflector blocks with hastelloy cans, and in Figure 2.3-18 for the side spacers without boronated pins.
(4) Transfer:
As the shield bell is moved from the work platform (using the Reactor Building crane) to the packaging area, nonesse.ntial personnel will be required to stay clear of the area to create a clear path for movement of the load.
2.3-33
DECOMMISSIONING PLAN REV 2 SECTION 2 (5) Unloading:
Unloading of the shield bell into a liner contained in a storage / shipping cask will be accomplished by removing the shield bell bottom and lowering the basket from the shield bell into the liner (see Figure 2.3-14). An alignment fixture will be used as necessary, to align the shield bell to lower the basket into the liner.
While the basket is being unloaded, any water accumulated in the catch pan will be removed and disposed of or returned to the PCRV.
(6) Packacine:
After the basket has been loaded into the storage / shipping container, water absorbing material will be added in sufficient quantity to absorb any incidental water. The liner top will then be secured to the top of the liner. If the basket has been loaded into a shipping cask, the top will be installed and secured for shipment. If the basket has been placed in a storage cask, the top will be placed on the cask to provide shielding if necessary until ready for shipment. When ready for shipment, the top will be removed from the storage cask-and the liner will be transferred to a shipping cask using an overhead crane. Any necessary shielding will be provided during the transfer of the liner from the storage cask to the shipping cask.
2.3.3.8.6 Descriotion of Activities Snecific to Block Tvoe A.
Defueline Elements Removal of the defueling elements not removed with the FHM will be handled as described in the six steps described in the previous subsection. The defueling elements are not activated and were uncontaminated when installed iri the core.
However, cross-contamination is expected to have occurred during reactor flooding and suitable contamination control procedures will be implemented to handle the defueling elements. Due to the very low radiation levels (< 1 mR/hr), it will not be necessary to stage the defueling elements and they may not be loaded into a shield bell for transfer to the shipping container.
B.
Large Side Reflector Blocks A typical large side reflector block is shown in Figure 2.3-19. There are 312 large side reflectors, ranging from approximately 522 tc 2030 lbs. each, around the circumference of the core as shown in Figure 2.3-11. The initial task is to remove the 24 upper reflector keys, which must be accomplished in order to remove the side 2.3-34 i
e op
REV 2 DECOMMISSIONING PLAN SECTION 2 reflector blocks. The keys will be detached by removing the five nuts per key or by thermally cutting the keys. The large side reflector blocks will then be removed and processed using the general steps described above. The large side reflector blocks will be handled using a dual collet tool inserted into the reverse counterbored holes.
C.
Hex Reflector Blocks Without Hastellov Cans Removal of the bottom, side and top hex reflector blocks without hastelloy cans not removed with the FHM will be handled as described in the eight steps described in the previous subsection. The position of the hex reflector blocks without hastelloy cans in the core is shown in Figure 2.3-11.
D.
Hexaeonal Grachite Blocks With Hastellov Cans Each Hastelloy can hex reflector block has 0.531-inch diameter holes to accommodate Hastelloy cans. There are 270 hex reflector blocks that contain 72 Hastelloy cans and 2 hex reflectors that contain only 4 Hastelloy cans each. The Hastefioy cans are 0.51 inches in diameter, approximately 8 inches long, and contain boronated graphite.
The location of Hastelloy can hex reflector blocks in the PCRV is shown in Figure 2.3-11. Table 2.3-1 indicates that the hex reflector blocks with Hastelloy cans have one of the highest radiation levels (300 R/hr) of those irradiated components to be removed from the PCRV with manually operated tools.
Removal of the hex reflector blocks with Hastelloy cans not removed with the FHM will also be handled as described in the six steps described in the previous subsection.
The Hastelloy cans in the hex reflector blocks are not expected to fall out of the block.
Therefore, removal of the Hastelloy cans using a dumping or tipping operation will not be attempted. Removing the Hastelloy cans from the graphite blocks would require the use of some mechanical method (broaching, cutting, pressing, and/or crushing). After considering the methods for removing the Hastelloy cans and comparing them to the alternative of leaving the Hastelloy cans in the blocks, it was decided that the Hastelloy cans would be left in the blocks.12aving the Hastelloy cans in the blocks eliminates the need for special equipment to remove the Hastelloy cans, simplifies the process of block removal, and will minimize personnel exposure.
j l
E.
Side Soacer Blocks With Boronated Pins The side spacer blocks with boronated pins are shown in Figure 2.3-20. There are 1152 boronated side spacer blocks weighing approximately 100 - 150 lbs each. Their 2.3-35 i
DECOMMISSIONING PLAN REV 2 SECTION 2 location in the core is shown in Figure 2.3-11 and dimensions of the pins are shown in Figure 2.3-21. Removal of the side spacer blocks will be handled as described in the six steps described in the previous section. Based on expected radiation levels, the pins and blocks may be shipped in the same type of shipping container.
Therefore the pins will remain in the blocks during removal and shipping. The blocks will be lifted vertically from their position in the PCRV and placed in a vertical position in the transfer basket. The drainage holes in the basket will be sized to retain any pins which might fall out of the blocks during transfer from the PCRV to the shipping cask.
F.
Core Suoport Blocks and Posts There are 37 hex Core Support Blocks (CSB) and 24 irregular shaped perimeter CSBs. The CSBs are supported by 183 posts. Removal of the CSBs will be handled as described in the six steps described in the previous section.
The defueling blocks and hex reflectors must be removed prior to removal oithe hex CSBs. Thermocouple assemblies run from the north west to south east through the CSBs. These assemblies may be removed by pulling them out of penetrations in the north west side of the PCRV or may be removed by personnel on the rotary work platform using long handled tools. The assemblies must be removed from a given CSB prior to its removal. In accordance with the Decommissioning Technical Specifications, blind flanges for the seven core outlet thermocouple penetrations may be removed, one at a time, during underwater removal of the thermocouple assemblies. During this time, PCRV shield wa:er leakage will be prevented by I
redundant seals on the thermocouple removal tools. The assemblies will be placed in a container or basket under water for movement to the radwaste area for packaging and disposal in a manner similar to that described above.
1 The large side reflectors and side spacer blocks sit on top of the perimeter CSBs.
The respective blocks and the lower reflector key must be removed prior to removal of a perimeter floor block. The 24 lower keys, which are made of Hastelloy X, will have estimated radiation levels of 10 R/hr at 1 meter. The lower reflector keys will j
be placed in a container or basket under water for movement to the radwaste area for packaging and disposal in a manner similar to that described above.
2.3.3.9 Removine the Core Barrel The following prerequisites must be completed to begin dismantlement of the core barrel:
2.3-36
REV 2 DECOMMISSIONING PLAN SECTION 2 1.
The PCRV top head concrete has been removed.
2.
The PCRV has been flooded above the core barrel with shield water and water clarity has been established as discussed in Section 2.3.3.6.
3.
Reactor core graphite blocks have been removed from the PCRV to a level low enough to permit the cutting of a core barrel section.
The core barrel and core barrel keys will be segmented underwater using remotely operated cutting equipment after the graphite core components are rev.oved.
However, if radiological surveys in the core barrel indicate that actual radiation and contamination levels are low, the PCRV water level will be progressively lowered and the core barrel and outer keys will be thermally cut above the water line.
Exhaust hoods, powered by HEPA-filtered air handlers, will be positioned at the water surface or, if the cut is performed dry, in close proximity to the cut. These exhaust hoods will capture the majority of the fumes at their source. While cutting of the core barrel above the water line appears to have a schedule advantage over the underwater cutting, it will only be considered if it can be justified by an ALARA 1
review.
1 i
If the removal of graphite core components is interrupted due to a shortage of shipping casks, work would commence cutting the core barrel underwater using remotely operated cutting equipment as the core barrel is exposed with the removal of successive layers of graphite core components. This is not expected to affect i
safety, occupational exposure or cause an undue schedule delay, i
With either cutting alternative (i.e., underwater or above the water line), the major activities for removing the core barrel are as follows:
1.
Rigging the core barrel sections for removal.
2.
Making horizontal and vertical cuts in the core barrel to segment it into sections suitable for handling.
3.
Removing the core barrel segments out of the PCRV.
4.
Progressively removing the outer keys and thermocouple expansion joint assembly that is between the PCRV liner and the core barrel.
]
The cutting of the core barrel will be performed with the Work Platform in place.
For underwater cutting, a mast or a remotely positioned track-mounted cutting tool will be operated from the Work Platform to make the vertical cuts around the core barrel. When the vertical cuts are complete, rigging will be attached to the core barrel segments prior to making the horizontal cuts. The horizontal cut will then be made and the core barrel segment removed. The jib cranes will be used to lift the segments to awaiting IEA boxes positioned adjacent to the opening on the work 2.3-37
DECOMMISSIONING PLAN REV 2 SECTION 2 platform. The cut pattern will be predetermined based upon the size of LSA containers selected and the features of the remote cutting system. For disposal and cost estimating purposes, it was assumed that the segments were 7.5 feet high X 3.5 feet wide for a 4 foot X 8 foot LSA box and cutting was performed by a sequence of vertical cuts followed by horizontal cuts. However, ifit is determined that larger pieces can be packaged, a reduction of time and exposure will be achieved. This process will continue down the entire length of the core barrel until approximately two feet of core barrel remains above the silica blocks. Removal of the lower portion of the core barrel will be coordinated with the removal of the silica insulation j that is on top of the CSF. Underwater divers were used to remove most of the l insulation on top of the CSF. Additionally, divers removed insulation from the inside l of the lower portion of the core barrel, and removed the core barrel bottom remnant.
l Diving operations in the PCRV are discussed in Section 3.2.12.
The core barrel sections will be surveyed as they break the water to determine exposure rates before being handled. The segments are expected to have a contact dose rate of 40 mR/hr. Loose contamination is expected to be moderate (1Q0,000 -
2 300,000 dpm/100 cm ). Loose surface contamination from pieces removed from the water will be controlled by a combination of pressure washing, rinsing with clean water, wet vacuuming and swabbing. These measures will control the spread of contamination and minimize potential for airborne contamination.
2.3.3.10 Removal of the Core Sunoort Floor (CSF) i The radiological conditions expected at this time are based on two sources, the PCRV cavity walls and tiie CSF. The cavity wall source consists of the fixed contamination j
on the wall and activated cover plate, insulation, liner plate and concrete. The dose j
rate from this source is estimated to be 30 mR/hr at any point within the PCRV. The CSF, as a radioactive source, consists of the surface contamination and the activated insulation on the top of the CSF, the activated CSF cladding plate, and the activated concrete. The dose rate contribution from the CSF is expected to be 400 mR/hr on contact with the insulation in place. Removal of the insulation from the top of the CSF, which contains various components and retaining devices made ofInconel, will reduce the exposure rate to approximately 360 mR/hr.
2.3.3.10.1 Removal of CSF Silica Blocks. Cover Plates and Insulation 1
l The insulation on top of the CSF consisted of several layers of silica blocks and a l layer of Kaowool (see Figure 3.1-30). The top layer of insulation consisted of dense l cast fused silica blocks, approximately 3 inches thick. Under the top layer were two l additional layers of lower density silica foam blocks, each layer also approximately 2.3-38
PEV 2 DECOMMISSIONING PLAN i
SECTION 2 3 inches thick. The bottom silica layer rested on cover plates, under which was a l
)
layer of Kaowool insulation, approximately 2 3/4 inches thick. The cover plates l
l were pressed down onto the Kaowool by anchor bolts attached to the 3/4-inch-thick l
CSF carbon steel casing. The cover plates and anchor bolts were carbon steel,
[
j except for a cover plate ring around each of the 12 steam generator inlet ducts, where l
the cover plates and anchor bolts were Inconel, to accommodate higher temperatures.
l l
Removal of the CSF upper insulation commenced while the core reflector blocks and l
side spacer blocks were still being removed, before removal of the entire core barrel, l
in areas above the CSF where all reflector blocks, core support blocks and posts had l
already been removed.
In these areas, insulation could be removed without l
destabilizing remaining columns of side reflector blocks and/or side spacer blocks.
l i
While the insulation was being removed, the PCRV shield water level was near the l
top of the PCRV liner, to minimize dose rates to personnel on the decommissioning l
rotary work platform (DRWP) from PCRV internal components. Most of the upper l
layer of dense fused silica blocks, and some of the silica foam blocks in the second l
layer, were removed with long handled tools. The remaining silica 'olocks were l
removed by divers, which was much more efficient. Diving operations iftthe PCRV l
)
are discussed in Section 3.2.12. The silica blocks around the 12 steam generator l
inlet ducts were glued together, and these were removed by divers usingjackhammers j
to break up the blocks, as required. The divers loaded the insulation material l
removed from above the CSF into baskets, underwater, using shovels and vacuum l
equipment. The baskets were then lifted by the jib cranes to the DRWP and l
j packaged in shipping containers. Divers and remote tooling were used to remove the j
Inconel sleeves and the alumina pads and dense cast fused silica discs within the j
sleeves that supported the core support post seats (Figure 3.1-30). Divers also l
removed the cover plates and bottom layer of Kaowool that was under the silica l
j blocks, exposing the top CSF casing. The bottom several feet of the core barrel was j
j removed in a manner similar to that described in Section 2.3.3.9, with the assistance l
of divers.
l 2.3.3.10.2 Removal of the Core Suncort Floor The prerequisites necessary to begin removal of the CSF are as follows:
l l
1.
All core components, including the core support blocks and posts, have l
l been removed from the PCRV, l
2.
The core barrel has been removed to within at least a few feet of the CSF.
l 3.
If the PCRV shield water level has been lowered, then loose contamination l
will have been removed from, or stabilized on, the interior walls l
(insulation cover plate) of the PCRV, as practical.
l 2.3-39
DECOMMISSIONING PLAN REV 2 SECTION 2 l
4.
The DRWP is removed prior to lifting the CSF to the upper PCRV area.
The CSF is a large disc approximately 29 feet in diameter by 5 feet thick and weighing 270 tons.
As noted in Section 2.2.2, the following items must be disconnected to allow removal of the CSF from the PCRV:
(1) 12 steam generator helium inlet ducts.
(2) 12 steel support columns, located near the CSF periphery, that are welded to the cladding plate and contain an array of cooling tubes and l
instrumentation tubes.
There is also a monorail spider consisting of twelve heavy structural steel beams on the bottom side of the CSF, that were used to position the. steam generators during i
construction.
j i
Since the existing Reactor Building crane has a capacity limit of 170 tons, the CSF will be jacked-up, sectioned, surveyed and removed in sections that can_be lifted by l the Reactor Building crane. Due to the tight clearance between the CSF and the PCRV cavity walls, it is necessary to raise the CSF to the upper PCRV region in order to provide access to the sides of the CSF for cutting and sectioning. The major activities that will be performed to cut and remove the CSF include the following:
1.
Raisine the CSF l
Based on the results of radiation surveys, steel shield plates have been placed on l
top of the CSF while it is underwater to reduce radiation dose rates to acceptable l
levels when the CSF is lifted out of the water, since personnel will need to l
l access the CSF for diamond wire cutting operations. Gaps between the steel l
shield plates should enable diamond wire cutting of the CSF to proceed without l
the need for cutting the steel shield plates. The added weight of these shield l
plates has been considered in the design of the CSF lifting system.
l Prior to lifting the CSF, workers will require access to the areas immediately l
above and below the CSF inside the PCRV to perform the cutting of the steam generator ducts and CSF columns, and to attach the lifting cables to the CSF.
l These activities will be accomplished using either diving operations or workers l
in a man-basket suspended from a crane. The method selected will minimize the I
time that will be spent in the radiation field and minimize the resultant exposure.
1 1
2.3-40 1
- ----e
~<
s j'
w m
~
- REV 2 '
. DECOMMISSIONING PLAN SECTION 2 i
u, The use of the man-basket will comply with the requirements of 29 CFR q
1926.550(g) and will also be coordinated with the containments that will be in place during the various phases of the work. Controls associated with diving
.l.
j operations are described in Section 3.2.12.
l t
Unless dose rates are determined to be significantly below' those estimated, it i
will be necessary to disconnect the steam generator penetrations and the CSF
- support columns using underwater cutting. Therefore, PCRV water level will
. be maintained above the' top of the CSF to provide adequate shielding during l-l performance of these activities.
Underwater ' cutting, in combination with
.3 exhaust hoods and respiratory ' protection, will provide. a suitably; safe j
environment to workers suspended in a man-basket. The DRWP may remain l
j in place during severance of the steam generators and CSF columns from the l
]
CSF. A localized containment may be used to prevent the spread of airborne -
l contamination to other areas or workers.
Stress analysis of the CSF columns will be utilized to determine the number of columns required to support the CSF at this stage of dismantlefhedt The CSF
-l' l
lifting system will be installed on top of the PCRV, and will be supported by the
-l l
PCRV. Heavy, wide-flanged beams will be placed across the top of the PCRV, l
1 with hydraulicjacks supported by the beams,' as shown in Figure 2.3-22. Cables l-of the strand jacking system will be routed through four of the CSF steam l
l generator penetrations and connected to lifting toggle beams positioned -
l' horizontally under the CSF. The CSF strand jacking system will be tested to j.
l 110% of the estimated weight of the CSF lift, prior to cutting the last several l
1 CSF columns. After the CSF has been cut free from the steel support columns, l
{
the CSF will be lifted and supported inside the PCRV. ' The CSF will be raised.
to the PCRV top head region using the strand jacking system, which uses.
l multiple cables attached both to the lifting toggle beams positioned under the l
j
. CSF and to the jacking stations that have been established.on top of the PCRV -
)
(see Figure 2.3-22). After raising the CSF, supports will be installed on the.
)
PCRV ledge where the PCRV top head was previously cut and removed, and the l
CSF will then be lowered onto these supports.
2.
Senmenting the CSF With the CSF supported in the top head area, the CSF will be cut in half by - l means of diamond' wire cutting operations.'
Prior to initiation of cutting l-j activities, a slurry catch pan will be installed below the CSF to minimize slurry l
entering the shield water. Radiological containments may also be constructed
.l
-l if determined necessary.
2.3-41
DECOMMISSIONING PLAN REV 2 SECTION 2 l
After the CSF is cut into two half-moon sections. each of these sections will be l
transferred by the Reactor Building crane from the PCRV cavity to a CSF l
segmenting area established on the refueling floor, for further segmenting.
Segmenting the CSF will be performed using the diamond wire cutting operation. The primary work area for the segmenting activity will be around the perimeter of the CSF. This will keep the workers away from the top of the CSF which is the significant source of radiation exposure. The diamond wire cutting process is adequate to segment the CSF and the monorail spider located under l
the CSF, eliminating the need to remove this monorail,eparately. However, the l
monorail spider may be removed by divers prior to lifting and segmenting the l
CSF, if evaluation determines this method is more efficient. Individual segments j
of the CSF will be packaged for transport either on the refueling floor, or in the l
Reactor Building truck bay. These CSF segments will be lowered by the l
Reactor Building crane down the truck bay to a trailer, where the segments will l
be transferred to a temporary staging area and/or shipped to the disposal site.
2.3.3.11 Disassembline the PCRV Lower Plenum l During prior operations, the helium ducts connecting the CSF to the 12 steam l generators were severed and the CSF was removed from the PCRV. Removal of the CSF will make lower plenum components accessible, including the steam generator primary assemblies, the helium diffusers, the CSF support columns, the lower floor, the lower plenum insulation and other miscellaneous components (see Figure 2.2-5).
The following prerequisites should be completed prior to beginning~ removal of the helium circulator diffuser assemblies and the steam generator primary modules:
1.
The helium circulator and steam generator secondary assemblies outside the PCRV have been removed.
2.
The steam generators have been disconnected from the PCRV penetration flanges.
3.
The CSF has been removed from the PCRV.
2.3.3.11.1 Steam Generator Primary Assemblim Each steam generator primary module is approximately 6 feet in diameter, 26 feet in height and weighs 65,000 pounds. The radiation source is primarily attributable to plateout contamination with a minor contribution due to activation. The uppermost portion of the primary steam generator nearest the outlet of the reactor is estimated to have a contact dose rate of 700 mR/hr. The lower portion of the primary steam generator is estimated to have a contact dose rate of 50 mR/hr. Localized hot spots on the generators are estimated to be up to 2 R/hr on contact.
2.3-42
REV 2 DECOMMISSIONING PLAN SECTION 2 L Disconnectine the Steam Generator Primary Assemblies The primary modules will have been structurally disconnected from the PCRV penetration flanges during the activity that removed the uncontaminated steam generator secondary assemblies.
That task (Section 2.3.3.3.2) will have been accomplished in an uncontaminated environment with a low radiation field. The primary modules remain connected to the PCRV internals by the connection of the steam generator shrouds to the plenum floor in the lower portion of the PCRV and must be disconnected to allow the lifting of the primary modules from the PCRV.
The separation of the steam generator primary modules from the plenum floor is the most complex task associated with the primary steam generator removal tasks. The steam generator helium inlet ducts, which were cut from the top and bottom of the l
CSF, will be removed frem atop the steam geneator primary assemblies.
The steam generator primary assemblies will be rigged (to the Reactor Building crane using standard rigging techniques and devices) to secure them before the final unclamping or severance cut. If necessary, the PCRV water level will be maintained l
above the top of the primary module to reduce radiation levels Whil5'it is being separated from the plenum floor. The steam generator will be disconnected from the plenum floor using either remotely-operated cutting equipment or locally-operated l
cutting equipment if conditions permit, by unclamping or cutting the clamp or lower l
seal at the connection of the steam generator shroud to the !ower floor. Any remaining instrumentation or connections between the steam generators and the lower plenum will be severed remotely, or locally, if conditions permit.
l Due to the expected radiation levels associated with the steam generator primary modules and the limited access in the area of the joint between the primary modules 1
and the lower plenum floor, there is no simple means of making the separation. An alternative method for separating the steam generators from the plenum floor requiring similar precautions and effort is to cut the plenum floor around the l
attachment location, using a thermal cutting means. The method to be used will be l
based on an evaluation of the performance characteristics of both methods in the limited access in which it will be used.
Either of these methods would be utilized underwater to derive the benefit of the water for shielding the workers from radiation. Fumes from cutting and any potential
]
airborne contamination will be collected by an exhaust hood at the surface of the j
water. However, if radiological surveys of the primary modules indicate that actual radiation and contamination levels are low, the PCRV water level may be lowered 2.3-43
- DECOMMISSIONING PLAN REV 2 SECTION 2 l to obtain more direct control of the unclamping or severance cut during the separation of the primary module from the lower plenum floor. Lowering the PCRV water j
level will only be considered if it can be justified by an ALARA review.
2.
Removine the Steam Generator From the PCRV Upon reparation from the lower plenum floor, the steam generator will be removed l from the PCRV cavity by the Reactor Building crane. Prior to, or as the primary l modules are lifted from the PCRV, the outer shroud and tube outer surfaces will be washed down to remove as much contamination and cutting debris as possible, and j
l will be allowed to drain as necessary over the PCRV cavity. Each of the twelve l steam generator primary modules will be lowered into a cylindrical steel shipping l container, whose inner diameter is slightly greater than the outer diameter of the l module. The shipping containers will have side walls approximately 0.5 inch thick, l which is calculated to provide adequate shielding. The shipping containers will be l positioned on the horizontal ledge of the PCRV, where the decommissioning rotary l work platform was previously situated. Each shipping container will be secured, l oriented vertically, to ensure that the container remains in place during' ins'ertion of l a steam generator primary module and during the time that the Reactor Building l crane is disconnected from the module. During the movements of the modules, radiation protection personnel will ensure that distance is maintained between the workers and the source to keep exposures ALARA.
l After a steam generator primary module has been lowered into its shipping container, l absorbent material will be added as necessary, and the shipping container lid will be l bolted onto a flange sealing the package. The weight of the steam generator primary l module, shipping container and absorbent material is estimated to be less than 90,000 l lbs., within the rating of the Reactor Building crane's 50 ton hook that will be used l to lower it down the truck bay. Each loaded steam generator shipping container will l be placed onto a trailer in the truck bay, then moved to an on-site location for l temporary staging and/or shipped by tractor-trailer rig to the disposal site. Postulated l drop of a steam generator primary module in the truck bay is evaluated in Section l 3.4.10.
2.3.3.11.2 Helium Diffuser and Shutoff Valve Assemblies The helium diffuser assemblies will have been detached from the PCRV penetranon; after removal of components from within each penetration. The helium diffuser and shutoff valve assemblies will be removed using techniques similar to those described above for removal of the steam generator primary assemblies. Radiation levels on the helium diffuser assemblies are expected to be much lower than those experienced 2.3-44
REV 2 DECOMMISSIONING PLAN SECTION 2 on the steam generator primary assemblies. The helium diffuser and shutoff valve assemblics will be disconnected by cutting the clamp at the connection of the diffuser i
to the lower floor. An alternative method for separating the helium diffuser and l
shutoff valve assemblies from the lower floor is to cut the lower floor around the l
attachment locations. The assemblies will be rigged to the Reactor Building crane, l
removed and transferred to the waste handling area for processing and disposal. Due to the lower radiation levels, no special shipping containers will be required.
2.3.3.11.3 Remainine Components With the steam generators and helium diffusers and shutoff valves removed, all of the significant radiation sources in the PCRV will have been removed. This will allow the PCRV vessel to be totally drained. The work remaining in the lower plenum includes the removal of the CSF support columns, the lower floor and other miscellaneous lower floor area components, and the insulation and insulation cover plates on the PCRV liner and penetrations. These features will be removed utilizing hands-on tools and will be processed for disposal. The Kaowool insulation removed in this activity will most likely require removal of the absorbed ' water to assure compliance with shipping and disposal regulations. The removal of the absorbed water will initially be accomplished by pressing or squeezing the wet Kaowool, or other suitable drying techniques as required. All of the remaining components in the lower plenum will be removed and transferred to the waste handling area for processing and disposal.
During these fm' al dismantling activities, the dose rates inside the PCRV lower plenum will be significantly lower than during previous operations since the largest radiation source, the steam generators, will have been removed. It is estimated that the general area radiation level will be low enough to allow activities to be performed in the lower plenum manually, which will increase productivity and still be ALARA acceptable.
2.3.3.12 ELD;tl Dismantline. Decontamination. and Cleanun Activities The following activities are included in this task:
1.
Scoring and cutting the PCRV sidewall insulation and liner.
2.
Cutting and removing the activated concrete in the beltline region of the PCRV (See Figures 2.3-24 and 2.3-25).
i 3.
Removal and/or decontamination of all remaining contaminated concrete.
l 4
Decontaminating the PCRV lower plenum Imer.
l l
S.
Performing the final survey of the PCRV.
I l
l 2.3-45
DECOMMISSIONING PLAN REV 2 l
SECTION 2 6.
Demobilization and decontamination of the PCRV D/D tools and equipment.
7.
Disposal of the PCRV Shield Water System.
f The activated concrete will be removed in sectional units from the side walls of the
[
PCRV, with the attached liner and both layers of thermal insulation intact as part of each unit. Diamond wire cutting has been selected as the method to remove the j
activated Concrete sections.
The thermal insulation, steel cover plates and steel seal sheets will be cut, and the liner plate will be scored by thermal methods before the concrete is cut with the j
diamond wire technique. This prevents the diamond wire from entangling in the steel i
seal sheets and insulation.
Tendons which must be removed for access of the diamond wire will be detensioned i
and removed. Other tendons detensioned to relieve compressive stress on the kerf of the diamond wire cut will be left in place. This was discussed in Section,2.3.3.5.
Circumferential tendons at the elevations of the horizontal cuts will be removed to provide a path for the diamond wire. The diamond wire cuts will be made in two steps from opposite directions, making a complete cut underneath the activated belt line concrete as shown in Figure 2.3-24 j
The inner ring of vertical PCRV tendon tubes, located 32 inches from the PCRV sidewall, are suitably positioned for removing the beltline activated concrete (see Figure 2.3-25). However, in the event that these tendon tubes prove to be unsuitable for the initiation of diamond wire cuts, new vertical holes will be core drilled.
Communications down the vertical tendon tubes or new core drilled holes, through the horizontal cut, and up through the PCRV interior will allow threading of the l
diamond wire for the radial cuts to be made. Sections of concrete, liner and insulation that are approximately 3 feet thick, 8 feet wide, and 40 feet long will be produced and rigged to the Reactor Building crane before the final back cut is made between every third tendon tube. These sections will be moved to a radwaste processing area for further cutting and preparation for disposal.
This cutting technique will remove a maximum depth of 32 inches at the tendon tube and a minimum depth of approximately 27 inches midway between the tendon tubes.
This minimum removal depth is adequate to meet the activation analysis requirements (plus uncertainties) described in Section 3.1.4.1.
2.3-46
REV 2 DECOMMISSIONING PLAN SECTION 2 The PCRV Shield Water System will be dismantled and decommissioned similar to balance of plant piping system. The system will be drained and the water processed as liquid waste as discussed in Section 3.3.2.2. The piping and components will be decontaminated, dismantled and packaged for disposal. The demineralizers will be the last items taken out of service. The demineralizer resins will be processed, packaged, and disposed of as radioactive waste. The demineralizers will be leased equipment, and will be decontaminated and packaged as necessary for return to the owner.
Following the removal of the activated beltline concrete, a final cleanup and decontamination of the entire PCRV cavity will be performed. Decontamination l
methods may include conventional wiping techniques, scabbling, scarifying, vacuum sand blast, or a hydrolaser method, depending on, the degree to which the contamination is fixed on the surface. A survey of the PCRV will be conducted to verify that free release criteria has been met.
As dismantlement activities proceed, guardrails, covers, barricades, caps, etc., will be placed as appropriate consistent with industrial safety considerattens. Upon completion of PCRV activities, a top head closure along with other appropriate penetration caps and guardrails will be installed in compliance with good industrial safety practices.
2.3.4 Contaminated BOP Svstem Dismantlement and Decontamination 2.3.4.1 Introduction The decontamination and dismantlement of contaminated or potentially contaminated balance of plant systems will be done by either (1) decontamination in place, (2) removal and decontamination, or (3) removal and disposal as radioactive waste.
Systems which are contaminated or potentially contaminated above releasable limits requiring decontamination or dismantlement include the following:
l 1.
System 13 - Fuel Handling Equipment 2.
System 14 - Fuel Storage Facility 3.
System 16 - Auxiliary Equipment 4.
System 21 - Helium Circulator Auxiliary Equipment 5.
System 23 - Helium Purification System 6.
System 24 - Helium Storage System 7.
System 46 - Reactor Plant Cooling Water System 8.
System 47 - Purification Cooling Water System 9.
System 61 - Decontamination System 2.3-47
DECOMMISSIONING PLAN REV 2 SECTION 2
- 10. System 62 - Radioactive Liquid Waste System
- 11. System 63 - Radioactive Gas Waste System
- 12. System 72 - Reactor Building Drain System
- 13. System 73 - Reactor Building Ventilation System
- 14. System 93 - Instrumentation & Controls Contaminated balance of plant decommissioning is scheduled to coincide with fluctuations in critical path PCRV activities to level project manpower and to minimize competition for use of plant equipment.
In general, contaminated or potentially contaminated piping, components, structures, walls and ductwork will be dealt with in the following manner.
Potentially contaminated items will be surveyed to determine acceptability for unrestricted free release or to determine the cleanup required for release. Verification that plant systems or structures may be released for unrestricted use will be provided by a j
comprehensive radiological assessment that provides statistically significant l
confidence levels for all plant systems. Since the plant systems cannot be altered for these detailed radiological surveys until the systems are no longer needed to meet NRC license requirements, the detailed surveys will be conducted during the l
implementation phase of the decommissioning project.
The results of these l
radiological assessments will be used to determine the workscope required for final removal of contaminated or potentially contaminated systems and components.
i The piping and equipment removal experience gained at the Shippingport Station Decommissioning Project demonstrated that contaminated or potentially contaminated piping and components can be quickly removed by plasma-are torch without compromising contamination controls when aided by portable HEPA filtered ventilation units. Because of the relatively small volume of contaminated piping at Fort St. Vrain, however, the cost and support requirements of plasma-are torch operations (setup, torch maintenance, and HEPA-filter changeout) may dictate the use of other methods, such as portable band saws, hydraulic shears, and alternate thermal cutting processes such as oxy-acetylene.
Piping will be cut into segments of approximately equal length. As piping is removed, the open ends will be covered and the piping segments will be placed in LSA containers (e.g., a 4-ft X 8-ft X 3-ft I
box). All piping, instrumentation, valves, and fittings can be packaged in this size of waste containers.
Piping will be removed by following controlled steps in accordance with project procedures and radiation work permits. System tagout procedures will be followed l
to de-energize pumps and other electrical equipment. Piping dead legs and traps will j
be drained of residual water. Piping released for removal will be positively marked j
2.3-.48
i
)
REV 2 DECOMMISSIONING PLAN SECTION 2 before being turned over for dismantling. Contamination controls and waste containers will be set up to suppon dismantling operations. Contamination controls will include saddle tap valves for draining residual water, drip containments to capture metal filings, HEPA vacuums, anti-contamination clothing, and respirators, as identified by the radiation work permits. Contamination control enclosures may be built where necessary to prevent spread of contamination.
Any potentially contaminated piping that is embedded in concrete will be separated from the rest of the piping system near the face of the concrete structure and internally surveyed with a detector probe inserted into the pipe. Embedded pipe that satisfies the release criteria identified in Section 4.2 will be capped, tagged, and abandoned in place. Piping that does not meet the release criteria will be internally decontaminated by scrub brush or pipe-turning tools, such as a boiler tube cleaner, i
and internally wiped with moist rags until it meets the release criteria. If it is embedded near the surface, the pipe may be removed from the concrete with a concrete coring tool.
2.3.4.2 System 13 - Fuel Handline Eauioment 4
The contaminated fuel handling equipment at Fon St. Vrain includes the fuel handling machine (FHM), five reactor isolation valves (Figure 2.2-11) and two refueling sleeves (Figure 2.2-12). However, the residual radiation and contamination levels for this equipment are low enough to allow manual disassembly on the operating floor.
i During decommissioning activities, the following System 13 components will be used:
- 1. The FHM will be used to remove MCRBs from the PCRV and ; place them into shipping containers or an interim storage area (such as the Fuel Storage Wells (FSW)). If placed into an interim storage area, the MCRBs will have to be removed by the FHM and placed directly into shipping containers. It may also be used to remove the Region Constraint Devices from their storage location and to place them into shipping containers. The ATC may be used for removal of helium purification components from the PCRV top head.
- 2. The Refueling Sleeves (H-1304) are required to guide the FHM arm while it is in the PCRV removing MCRBs, and when unloading MCRBs from the FHM into a shipping container in one port of the Hot Service Facility (HSF).
2.3-49
DECOMMISSIONING PLAN REV 2 SECTION 2
- 3. The Reactor Isolation Valves are required to adapt the FHM to the PCRV and to the facility (FSW, HSF, Fuel Loading Port, Regen Pit) into which it will be unloaded.
- 4. The System 13 Fuel Handling Purge System provides helium to operate internal actuators in the FHM. This helium is supplied from System 24, via System 13 piping and the FHM umbilical. If either System 24 or the System 13 Purge System is inoperable, pressurized air can be used to operate the actuators.
- 5. The HSF Adapter & Sleeve Assembly (Zook Sleeve) and the Modified Refueling Sleeve (S-1615) will only be used if the FHM is to interface with the HSF for loading shipping containers, and repair, maintenance, and interchange of grapple heads / manipulators.
- 6. The Fuel Loading Port and associated equipment may be used as an area for unloading MCRBs from the FHM.
- 7. The Core Servicing Manipulator (H-1603) and Core Service Vacuum Tool Assembly (H-1606) are FHM attachments which may be used for special operations within the PCRV and shipping containers.
- 8. The spare grapple (H-1301), spare mast camera (H-1601), and spare Z-drive pumps may be used in the event of a failure of the primary component, allowing repair without affecting MCRB removal operations.
- 9. The Shipping Cask Loading Seal Adapters (S-1604-250) may be used if the FSV shipping liner / casks are to be used.
The FHM will be externally surveyed and any loose contamination removed. It will then be disassembled into its component parts as necessary for decontamination or disposal. Sleeves will be attached as necessary to maintain a contamination envelope.
The body of the FHM will be decontaminated and will be left on the refueling deck if release for unrestricted use limits are achieved. If further disassembly is required for release, the lead shot will be removed and the body' will be segmented to segregate the contaminated material from the uncontaminated components. The contaminated scrap will be disposed of as described in Section 3.3 of this plan.
The reactor isolation valve exteriors will be surveyed and decontaminated by manual The valves will be removed from the operating floor and the lead shot means.
j removed. The shot is not expected to be contaminated or activated. The valve bodies will be disposed offsite according to Section 3.3 of this plan.
i 2.3-50
REV 2 DECOMMISSIONING PLAN SECTION 2 The refueling sleeves will be surveyed and decontaminated by manual means, then surveyed for release for unrestricted use. If they cannot be decontaminated, they will be disposed of as described in Section 3.3.
The purge vacuum system will be removed and disposed of as described in Section 3.3.
2.3.4.3 System 14 - Fuel Storage Facility The fuel storage facility consists of nine fuel storage wells (FSWs) constructed of carbon steel liners suspended in concrete pits.
The FSWs were used for storing new and irradiated fuel during normal plant operation and may be used to temporarily store MCRBs or graphite reflector blocks i
during decommission. All fuel will have been removed from the Reactor Building prior to initiation of decommissioning activities. The actual contamination levels in the FSWs will be determined after the fuel has been permanently removed.
i l
During decommissioning activities, the System 14 FSWs may be used as an mtenm j
storage area for MCRBs when they are removed from the PCRV with the FHM.
i When the FSWs are no longer needed, each of the nine inner storage wells will be decontaminated to the criteria for release for unrestricted use, surveyed, and the top access plugs replaced. The outer wells and the reactor plant water cooling system are not contaminated and no outer well decontamination or dismantling is expected j
to be required. The water cooling system piping at the bottom of each pit will be cut j
open for survey.
i Decontamination of the FSWs will be accomplished using a HEPA filtered vacuum.
Following vacuuming, the wells will be mechanically blasted with sand grit or cleaned using a hydrolaser. Spent sand will be collected in catchments placed at the bottom of the well. The well drain pipe will provide water drainage if hydrolaser operation is used. After sandblasting or hydrolasing, the five standoff plates at the bottom of the wells will be removed manually. This will provide access for the final release surveys.
Minor components will be shipped as radioactive waste rather than decontaminated.
the well plugs will be decontaminated and replaced and sealed after the release surveys have been completed.
1 2.3.4.4 System 16 - Auxiliarv Eouioment i
The auxiliary equipment consists of the Auxiliary Transfer Cask (ATC, (Figure 2.2-2.3-51
DECOMMISSIONING PLAN REV 2 SECTION 2 12), ten Equpment Storage Wells (ESWs), (Figure 2.2-14), the Hot Service Facility (HSF), (Figure 2.2-15), and two shielding adapters (Figure 2.2-16).
The ATC was used to transfer the control rod drive assemblies, refueling sleeves and the shield plugs to and from the ESWs. The ten ESWs are carbon steel structures embedded in concrete. They were used to store the control rod drives and the refueling sleeves. The HSF is constructed of concrete and steel shielding and was used for inspection, repair, maintenance, testing and decontamination work.
Figure 2.2-16 is a general layout of the location of the various fuel handling and storage system components and associated auxiliary equipment on the refueling floor.
During decommissioning activities, the following System 16 components will be used:
- 1. The ATC may be used for removing and installing Refueling Sleeves into the PCRV during MCRB removal. The ATC may also be used for removing shield plugs from the ESWs, removing helium purification components, as well as retrieving and storing the Refueling Sleeves in the ESWs. If RegiortConstraint Devices (RCDs) or Control Rod Drives and Orificing Assemblies (CRDOAs) are stored in the ESWs at the beginning of decommissioning, the ATC may be used to remove them.
- 2. The ESWs may be used as a shielded interim storage area for activated / contaminated components, such as Refueling Sleeves, long-handled tools, and core components. The ESWs may cantain CRDOAs and/or RCDs at the start of Decommissioning which will hr.e to be removed during Decommissionmg.
1 i
- 3. The Shield Adapters are required to adapt the ATC to the ESWs (for removal and storage of the Refueling Sleeves), to the PCRV (for insertion / removal of the Refueling Sleeves), or to the HSF, Regen Pit or Fuel Loading Port (for miscellaneous activities).
i
- 4. The HSF may be used as a multi-purpose area.
Uses include: a general dismantlement / decontamination area, an area for holding shipping containers as they are loaded by the FHM or other means, and/or a shielded interim storage area for activated / contaminated components.
All the components of the ATC above the top base (32 ft.11 in, above the operating floor) will be removed using the Reactor Building crane. A containment sleeve will be used to seal the contaminated ports in the cask and the hoist assembly floor as they 2.3-52
REV 2 DECOMMISSIONING PLAN SECTION 2 are separated. The hoist cover and lift extension will then be lowered to the operating floor and disassembled within a contamination control envelope. The components will either be packaged and shipped for burial or to a licensed facility for processing and final disposition, or decontaminated and released for unrestricted use.
The remaining structure of the ATC will be decontaminated on site. The internal bore will be decontaminated using mechanical means such as sand blasting or hydrolasing to the criteria for release for unrestricted use.
After internal decontamination, the crane will be used to lay the cask body over onto tiie floor for disassembly and decontamination of the bottom flange. When all surfaces meet the criteria for release for unrestricted use, it will be lifted by the crane and returned to storage on the operating floor.
The three shielding adapters will be decontaminated manually to the criteria for release for unrestricted use.
The ten ESWs are internally contaminated and will be decontaminated to the criteria for release for unrestricted use and abandoned in place. ContaminatEonTevels in the ESWs will be determined when they are no longer needed. After the plugs are removed, the ESWs will be vacuumed using a HEPA vacuum assembly similar to that for the FSWs. After vacuuming, the ESWs will be further cleaned using mechanical methods as necessary to reduce the contamination to the criteria for release for unrestricted use. After decontamination, the wells will be surveyed for release for unrestricted use. The top access plugs will be decontaminated, replaced and sealed.
Following final use of the HSF for decommissioning activities, all equipment (such as the manipulators and service platform sling) will be removed. This equipment may either be decontaminated onsite or packaged and shipped to a licensed facility for processing and final disposition.
The walls, floor, ceiling and remaining structural components will then be decontaminated by sandblasting or hydrolasing.
HEPA-filtered ventilation will be used to maintain a negative pressure in the HSF during decontaminations.
2.3.4.5 System 21 - Helium Circulator Auxiliaries The auxiliary equipment for System 21 was used to provide a supply of high-pressure water for the helium circulator bearing lubrication and a supply of purified buffer helium to prevent in-leakage of bearing water into the primary coolant helium. The I
l 2.3-53
DECOMMISSIONING PLAN REV 2 SECTION 2 major equipment items include buffer helium recirculators, heat exchangers, filters, pumps, helium dryers, chemical injection components, containment tanks, and compressors (see Figure 2.2-17).
Following the defueling of the reactor, the helium circulator system will no longer be used. It has no function in the decommissioning of the facility.
Contamination has been detected within the helium circulator auxiliary equipment.
Surveys will be performed during disassembly to determine the extent of the contamination.
2.3.4.6 System 23 - Helium Purification Auxiliaries The helium purification auxiliary equipment consists of two trains and was used to assist in purification of the helium used as the primary reactor coolant. Most of the contaminated major equipment items are located in the PCRV top head and include filters, adsorbers, heat exchangers, dryers and piping (see Figure 2.2-18).
~,
System 23 equipment is located in the top head in eight wells. This equipment will be surveyed and a determination made whether to remove the equipment with the ATC or by manual means. All System 23 equipment located in the PCRV top head will be disposed of as radioactive wastes. After the wells have been emptied, they will be surveyed and decontaminated as necessary.
The remainder of the helium purification system will be surveyed and decontaminated or removed as necessary.
2.3.4.7 System 24 - Helium Storace System The primary purpose of the helium storage system was to provide for both storage and transfer of helium from the reactor vessel to the storage tanks. In addition, the helium storage system was used in testing the control rod reserve shutdown system and for various FHM purging operations. The primary equipment items include a helium transfer compressor, storage tanks, oil absorber, and high-pressure helium supply tanks (see Figure 2.2-19).
Following the defueling of the reactor, the helium storage system will no longer be used. It has no function in the decommissioning of the facility.
The helium storage compressors have been found to be contaminated. This system, including the 108 helium storage bottles, will be surveyed during disassembly to 2.3-54 e
REV 2 DECOMMISSIONING PLAN SECTION 2 determine the extent of the contamination. The results of this survey will be used to determine decontamination or disposal requirements for specific components.
2.3.4.8 System 46 - Reactor Plant Cooling Water System l
l The reactor plant cooling water system (see Figure 2.2-20) provided cooling water for process heat removal from all auxiliary equipment in the reactor plant. Three loops were provided that formed the PCRV circuit (liner cooling tubes), the PCRV auxiliary circuit (closed loop for various systems / components) and the service water circuit (open loop for various systems / components). The major equipment items include surge tanks, pumps, demineralizers, filters, heat exchangers, chemical injection (tank and pump) and recondenser chiller.
Portions of the system external to the PCRV have been found to be contaminated.
The system will be surveyed during disassembly to determine the extent of the contamination. Cleanup or disposal requirements will be determined based on survey results.
The reactor plant cooling water system loop supplying the PCRV will not be used for cooling of plant components during decommissioning. It will be disconnected and isolated from the PCRV and from the FSWs before decommissioning of those systems occurs. Fifty percent of the PCRV cooling tubes will be cut and surveyed.
2.3.4.9 System 47 - Purification Cooline Water System The purification cooling water system (two loops) provided cooling water to the helium purification system heat exchangers. The major components are pumps, expansion tanks, exchangers and associated piping (see Figure 2.2-21).
This cooling water system has been found to be contaminated. The system will be surveyed during disassembly to determine the extent of the contamination. Cleanup or disposal requirements will be determined based on survey results.
The purification cooling water system will be isolated from the helium purification system before it is decommissioned. The purification cooling water system has no other use during the decommissioning.
2.3.4.10 System 61 - Decontamination System The decontamination system consists of a water heater, a drying air heater, a filter, pumps, a solution tank and a chemical injection system (see Figure 2.2-22).
2.3-55
DECOMMISSIONING PLAN REV 2 SECTION 2 The decontamination system will be surveyed to determine the extent and location of radioactive contamination following final system use. The decontamination system components are small, and will be removed and packaged in LSA shipping containers along with other contaminated components and piping. The decontamination solution tank may be removed in one piece for shipment, or segmented and packaged in LSA shipping containers.
2.3.4.11 System 62 - Radioactive Liould Waste System The major equipment items in the Radioactive Liquid Waste System include a waste sump (1000 gallon tank), pumps, filters, two 3000-gallon receiver tanks, two demineralizers, and a 3000-gallon waste monitor tank (see Figure 2.2-23).
The liquid waste system is expected to be used for its original function during decommissioning operations. Therefore, it will be one of the last systems to be decommissioned.
During decommissioning, System 62 piping and components can b'e used for collection, monitoring, and dispositioning of liquid effluent generated by decommissioning activities, mainly by the PCRV Shielding Water System.
Additionally, fluids may be processed from decommissioning activities which originate from the Helium Regeneration Pit Drains, Liquid Drain Tank (System 63),
Reactor Vent and Drain System Standpipe M-5 (System 72), Reactor Building Sump and Sump Pump (System 72), and the Reactor Plant Exhaust Filter housing drains (System 73). The latter sources are those normally encountered during reactor operation and shall be processed by established methods. All sources, including the PCRV shielding water, require the necessary piping and components that will be utilized to remain in service until no lenger needed.
In addition, effluent from the PCRV Shielding Water System will require the installation of a connection between the discharge of the PCRV Shielding Water System transfer pump and System 62 piping, and slight modification of valving to utilize System 62 as desired. This will permit pumping of shielding water directly into either of the Liquid Waste Receivers. Also, valves will be installed to permit shielding water to be pumped directly into the Liquid Waste Monitor Tank without travelling through the Liquid Waste Transfer Pumps, the Liquid Waste Demineralizers, and long lengths of piping.
A characterization survey of the radioactive liquid waste system will be performed l
when the system is no longer needed to determine the extent and location of radioactive contamination.
l 2.3-56
REV 2 DECOMMISSIONING PLAN SECTION 2 The contaminated radioactive liquid waste system components are small and include:
the two liquid transfer pumps, the two liquid waste sump pumps, the two liquid waste filters, and the two liquid waste demineralizers. The liquid waste monitor tank and the two liquid waste receivers may be decontaminated and abandoned in place, J
shipped as one piece containers, or segmented and packaged in LSA shipping 1
containers depending on the extent and location of radioactive contamination. The liquid waste sump will be considered for either (1) decontamination to free release i
levels and ab: ndonment, or (2) segmentation and packaging as LSA waste.
2.3.4.12 System 63 - Radioactive Gas Waste Svstem The major equipment items in this system include pre-filters, filters, exhaust blowers, tanks (vacuum, surge, and drain), and compressor (see Figure 2.2-24).
Following final use of the system, the radioactive gas waste system will be surveyed to determine the extent and location of radioactive contamination.
The large components such as the two gas waste surge tanks and the gas waste vacuum tank may be decontaminated and abandoned in place, shipped as one-piece units, or segmented for packaging and shipping. The other components are small enough to be shipped in LSA shipping containers with other contaminated piping.
Decontamination of these systems will be by manual mechanical methods depending on the levels of contamination found during the characterization survey. The system will not be used in the decommissioning of the plant.
2.3.4.13 System 72 - Reactor Buildine Drain SystCJD The major equipment items include drain tanks, sump, pumps, piping and filters (see Figure 2.2-25). Two gravity flow drains are provided to direct drainage from the Reactor Building equipment, pi :q and floor drains to either the radioactive liquid l
waste sump for potentially contaminated liquids or the Reactor Building sump for all other liquids. The drain system will continue to be used for its original function during much of the decommissioning work and will be one of the last systems to be decommissioned.
When no longer required to remain operational, the system will be surveyed, and a decontamination and decommissioning decision,will then be made. Contaminated piping or components will be either removed and shipped in LSA containers, or decontaminated to the criteria for release for unrestricted use and left in place.
2.3-57
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DECOMkIISSIONING PLAN
' REV 2 l
' SECTION 2-The portion of this system that drains to the Reactor Building' sump is not expected ~
l p
' to be contaminated. 'Ihe portion of the. system that drains to the radioactive liquid
)
waste sump is expected to be contaminated and~is included in dismantlement and j
removal plans.-
'j 2.3.4.14 System 73 - Pmtor Building Ventilation The Reactor Building HVAC system semces vanous areas of the Reactor Building -
'vith heated or cooled air. All. ventilation air, whether outdoor or recirculated, is a
filtered before distribution. In addition, the reactor plant HVAC system maintains.
j
' building' differential pressure control. As shown in Figure 2.2-26, this system i
consists of several-air handling units and filters. The only part of the system considered to contain possible contamination is the Reactor Building exhaust filters, HSF vent, and the analytical room vent.
~
The reactor plant exhaust filters are composed of banks of moisture separators and
HEPA filters.
During decommissioning, System 73 will be used to maintain the reactor building pressure subatmospheric for selected decommissioning operations as required by the Decommissioning Technical Specifications. The system consists of three trains, one
- of which will normally be in continuous operation. One train is sufficient to maintain j
the reactor building subatmospheric. The reactor plant exhaust filters will be
- periodically monitored and filter media changed, as necessary. Filter change out will
'l be based on excessive pressure drop across the _HEPA filters, or excessive leakage, as required by the Decommissioning Technical Specifications.
System 73 will also be used to support the Airborne Contamination Control System (ACCS) which draws air from under the Rotary Work Platform.' The ACCS is a -
temporary addition that will be used during decommissioning and ties into existing i
System 73 ductwork. The ACCS consists of several ducts that will pull air from under the platform. The ducts tie together forming a single duct that is routed to two roughing filters, in parallel, to remove particulate material. After the roughing filters the'ACCS ducting ties into the existing System 73 ducting, upstream of two of the three Reactor Building exhaust fans. Dampers are installed so that one or both of these two Reactor Building exhaust fans can be lined up to take a suction from the plenum under the Rotary Work Platform (ACCS) or from the Reactor Building (normal suction). If one HEPA filter / exhaust fan train in the Reactor Building exhaust system is not available due to changing of the filter media or other 2.3-58 i
M l
l
- i R E V 2' DECOMMISSIONING PLAN SECTION 2
-l i
maintenance, that train can be isolated and the air flow from the ACCS diverted to.
the alternate HEPA filter / exhaust fan train that is connected with the ACCS. The ACCS will be removed at the end of decommissioning.
The ventilation system has been found to be contaminated. This system will be
]
maintained during decommissioning to provide ventilation for decommissioning l
operations. Near the completion of decommissioning activities, surveys will be taken to determine the final disposition of the system.
i 2.3.4.15 System 93 - instrumentation and controls
.]
The instruments and tubing to be removed or decontaminated originate at PCRV j
These include thermometer penetrations, process and moisture 1
instruments, helium circulator instruments, and helium vent piping.
Moisture monitors will be removed during dismantling the PCRV.
All other instrument interfaces to contaminated or potentially contaminated gstems will be.
addressed when the respective system is decommissioned.
Those-System 93 1
components will be either removed or verified to be below the limits for release for I
unrestricted use. All systems are scheduled for inclusion in the charactenzation survey. Contaminated system components will be decontaminated or disposed of as LSA waste.
2.3.5 Decommissionine Schedule The individual tasks making up the decommissioning effort have been ' delineated using a work breakdown structure (WBS) approach. Figure 2.3-26 is a schedule of the major decommissioning tasks which includes PCRV and balance of plant system dismantling and decontamination, and site decommissioning. This schedule is used as the top-level view of the project milestones and detailed schedules. Throughout the project, dismantling the PCRV is the critical path activity,' with the BOP dismantling activities scheduled to coincide with periods of reduced PCRV efforts as a means of wcrkload leveling. During the planning phase, work will be directed toward characterizing the site, preparing the decommissioning plan, and planning and writing the procedures and specifications for the implementation phase.
The major activities and programs to be developed during the planning phase include:
1.
Initial site characterization 2.
Decommissioning planning 3.
Work specifications and procedures 2.3-59
DECOMMISSIONING PLAN REV 2 SECTION 2 4.
Quality assurance plan 5.
Radiation protection program 6.
Waste management plan 7.
Project performance and control The schedule depicts the planning phase occurring over an 18 month period, and the actual dismantlement and decontamination activity at the site occurring over a 39 month period.
2.3-60
REV1 DECOMMISSIONING PLAN SECTION 2 l
TABLE 2.31 ESTIMATED CONTACT DOSE RATES FOR GRAPHITE BIDCKS ESTIMATED i
I GRAPHITE BLOCK DESCRIPTION NO.OF CONTACT BLOCKS DOSE RATE 1)
Defueling Blocks 1,482
<1 mR/hr 2)
Hexagonal Reflector Blocks w/no 1,687 500 mR/hr Hastelloy Cans 3)
Large Permanent Reflectors 312
< 30 R/hr l
4)
Reflector Keys 24
< 100 mR/hr 5)
Side Spacer Blocks (a) with Boron Rods 1,152 30 R/hr (b) Boron Rods removed 1,152 -
<3 R/hr (c) Boron Rods 276,096 60 R/hr 1
6)
Bottom Reflector Blocks (a) with Hastelloy Cans 272 300 R/hr (b) Hastelloy Cans removed 272 500 mR/hr (c) Hastelloy Cans 19,448 10,000 R/hr 7)
Core Support Block Hastelloy Keys 24 10'"
R/hr 8)
Core Support Blocks (61) & Posts (183) 244 15 mR/hr 9)
MCRBs (Nonetrol rod) 6 300 R/hr Dose Rate at one meter i
1
PROPOSED DECOMMISSIONING PLAN SECTION 2 TABLE 2.3-2 PROJECTED PERSON-REM EXPOSURE FOR THE FORT ST. VRAIN DECOPMISSIONING PROJECT ESTIMATED WORK ACTIVITY
- PERSON PERSON HOURS HOURS 2.3 PCRV DISMANTLEMENT AND DECONTAMINATION Initial preparation / disassembly 11,310 7
Remove PCRV concrete top head 18,205 20 Dismantle PCRV core and core barrel 79,230 160 Remove core support floor 3,245 105 Remove steam generators and helium diffusers 1,750 25 D/D PCRV Lower plenum 4,293 29 Final PCRV dismantlement, decontamination and cleanup 12,075 20 SUB-TOTAL 130,735 366 2.4 CONTAMINATED SYSTEMS D/D (BOP)
Initial preparation / characterization 1,920
<1 Dismantle /decon operations 28,308 1
SUB-TOTAL 30,228 2
2.6 WASTE PREPARATION, PACKAGING, SHIPPING AND DISPOSAL Radioactive waste less than 50 mR/hr 21,036 22 Radioactive waste greater than 50 mR/hr to 500 mR/hr 2,510 19 Radioactive waste greater than 500 mR/hr 9,509 24 SUB-TOTAL 33,055 65 TOTAL 194,018 433
- Person-hours only for those tasks where the potential for measuring radiation exposures exists i
1
11 REVI DECOMMISSIONING PIAN SECTION 2 DIFFUSER TERIOR CAVITY
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DECOMMISSIONING PW REV1 SECTION 2 Refueling Floor TU1 Uft Rig M*
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Figure 2.3-23 (Deleted) l 1
-l REV1 DECOMMISSIONING PLAN SECTION 2 l
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i DECOMMISSIONING PLAN REV1 SECTION 2 Steo2 Steo 1 The snaced area The diamond wire saw #5 is then cut inserted inrougn tendon tuce i
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Figure 2.3-24 PCRV Beltline Concrete - Horizontal Cuts
DECOMMISSIONING PLAN REV1 SECTION 2 Step 2 Step t The snaced area The diamond wire saw is is then cut inserted inrough tencon tube b
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Figure 23-24 PCRV Beltline Concrete - Horizontal Cuts
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REV1 DECOMMISSIONING PLAN SECTION 2
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1
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4 i
REV1 DECOMMISSIONING PLAN SECTION 2
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(
n I
r DECOMMISSIONING Pl.AN REV1 i
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SECTION 8 DECOMMISSIONING ACCESS CONTROL PLAN 8.1 BASIS FOR ACCESS CONTROL PROGRAM i
The Fort St. Vrain Decommissioning Access Control Plan is based on 10 CFR l
20.105 and NRC Regulatory Guide 1.86 (Ref.1) which requires isolation of l
l radioactive materials remaining on site from public access to preclude inadvertent l
exposure to hazardous levels of radiation and administrative procedures for l
l notification and reporting abnormal occurrences.
l l
The Fort St. Vrain Decommissioning Access Control Plan provides adequate l
l industrial security measures to assure public health and safety are not endangered by l
inadvertent access to the decommissioning site, provides support to implement the l
Radiation Protection Plan and the Decommissioning Emergency Response Plan, l
minimizes corporate liability during decommissioning, and establishes controls and l
i procedures related to access to the decommissioning site. Movements by personnel l
within the decommissioning site and access control requirements for radiologically l
controlled areas are addressed in Section 3.2, Radiation Protection Program.
l 8.2 SITE ACCESS CONTROL ORGANIZATION The PSC Facility Support Manager is the designated site representative responsible l
for implementing the FSV Access Control Program. The PSC Facility Support l
Manager is also the designated site representative responsible for coordination with l
]
local law enforcement authorities. Familiarization briefings on access procedures, l
plant layout, and emergency arrangements will be made as requested by the local law l
enforcement agency.
l Access control personnel will be properly trained and demonstrate understanding of decommissioning area access control requirements and responsibilities.
Access control personnel will be unarmed and equipped for continuous onsite and offsite communications.
l 8-1 es w.,.,
..,y
e DECOMMISSIONING PLAN REV 2 SECTION 8 8.3 ACCESS CONTROL PHYSICAL SECURITY MEASURES l The decommissioning area will be surrounded by a physical barrier to inhibit l unauthorized access to restricted areas. This barrier shall be inspected at least l quarterly for deterioration or breaches and that applied locks and locking apparatus l are intact. Repairs to physical barriers and equipment will be accomplished in a l timely manner. All portals and gates associated with decommissioning area physical l barriers will be kept locked or continuously monitored by access control personnel l and all locks and keys will be controlled by access control personnel.
l Personnel access point will be located at the main plant entry for the l decommissioning area. Even though other access points may be established as l needed, this gate will generally serve as the primary personnel access portal.
l l Vehicle access gates will be established as needed to accommodate decommissioning l activities.
l Access for decommissioning workers will be authorized by project management. A l list ofindividuals granted authorized access will be prepared and maintained. Access l control personnel will verify authorized access for each individual before granting l
l passage through the physical barrier surrounding the decommissioning area. Visitor l
l access to the decommissioning area will be controlled in a manner consistent with the l Radiation Protection Plan and Emergency Response Plan.
l l Personnel accountability, including accountability for visitors, in the event of an l emergency will be maintained through the Access Control Program in support of l Emergency Response Plan requirements.
Access to the restricted area does not guarantee access to radiological controlled l areas. The radiation protection staff will administer the radiological controlled area access control program. Specific requirements that must be met prior to accessing radiologically controlled areas are identified in Section 3.2.
I 8-2 4
m REV 2 DECOMMISSIONING PLAN SECTION 8 8.4 COMMUNICATIONS l
r Telephone service will be available for access control personnel to contact local law l
enforcement authorities in the event of an abnormal occurrence and to make l
necessary notifications to decommissioning site management. Radio communications
-l j
between the FSV control room and access control personnel will also be provided to l
supplement telephone servier,, especially during emergencies.
l I
8.5 PROCEDURES Written procedures will be prepared and implemented to provide the access control i
personnel guidance for routine occurrences. Examples include:
l 1.
Personnel access control 2.
Vehicle access control 3.
Communications equipment and routine testing requirements 4.
Surveillance / inspection of decommissioning area physicaLbaniers I
-l' Written procedures will be prepared and implemented to provide the access control personnel guidance for abnormal occurrences. Examples include:
l 1.
Personnel disturbance l
2.
Acts or perceived threat of sabotage l
3.
Civil disturbance l
4.
Suspected or confirmed sabotage or intrusion attempt' j.
5.
Breached security area barrier l
6.
Unidentified person in security area l
7.
Site evacuation l'
The content of abnormal occurrence procedures will include (1) criteria for l
identifying abnormal conditions within the decommissioning area; (2) access control personnel actions; and.(3) required notifications. Guidance to respond to other l
abnormal occurrences such as fire, explosion, site radiological emergencies, and l
medical emergencies is contained in emergency plan procedures.
l
8.6 REFERENCES
FOR SECTION 8 i
1.
Regulatory Guide 1.86: " Termination of Operating License for Nuclear Reactors". June 1974.
8-3
c s
r.
M - 7l994 Mr. William H. Hamilton Highland Park Club oisTaisurion:
cwton 1650 Highland Park Drive Lake Wales, FL 33853 7
n k EveneWRel*
December 21,1993 r
s.n.-
Ho6mes Hug McBnde MeGeace u;eson Mr. Del Hock e.reene Chief Executive Officer R*io*i Public Service Company of Colorado 122517th Street, Suite 900 Denver, CO 80201
Subject:
Fort Saint Vrain (FSV) Defueling/ Decommissioning Oversight Committee
Dear Mr. Hock:
Enclosed is the report from the recent meeting at the Public Service Company of Colorado (PSCC) on December 2,1993. The purpose of the meeting was to continue 4
our review of the status and plans for decommissioning of the Fort Saint Vrain. facility.
The plant decommissioning work at this particular stage in the project is more of a repetitive type operation which will continue for the next few months as various pieces are removed from the vessel. Our particular objective for this meeting was to receive a status update of ongoing work tasks. As such, we completed our meeting in one day (versus our normal two day meeting). The length of this report reflects our shorter meeting.
As with our past several meetings, we started with a general tour of the reactor building.
As part of our tour, we inspected the work platform area above the PCRV and the shield water system. We also were shown a video of previous shield bell handling operations that have been used to transfer the graphite blocks from the PCRV to the shipping casks. In general, the equipment and systems looked very impressive. The engineering of the block removal tooling and handling was well done. In fact, we understand that the block removal process is proceeding at about twice the rate that was originally envisioned.
The clarity of the water in the PCRV is much improved. We caution, however, that this could be the result of the passive work activities currently on going in the vessel. Water clarity problems can probably be expected to reoccur during the upcoming cutting and dismantlement operations (e.g., cutting of the core barrel). As a last note from our tour, similar to previous tours, we thought the reactor building and work areas were cican and well maintained.
Following our tour, we heard presentations by PSCC personnel regarding the status of on going work and an update on the final site survey plan. The main point of discussion was the final site survey plan. Both the Oversight Committee and PSCC personnel recognize the importance of this document, which will describe how the final termination surveys will be implemented and provide acceptance criteria. We were concerned that bW
O Mr. Del Hock December 21,1993 the schedule for the plan seems to be slipping and that we have not seen many of the ic of concern is the methodology that will be technical details as of yet. One technical top (low energy) radionuclides. Dr. Woollam used to identify and quantify hard-to-detect indicated that this issue can be complex. The NRC has also indicated interest in this area. Based on our discussions with Mr. Crawford, we understand that PSCC is putting extra resources on this issue to ensure the survey plan is correct and ready in a timely manner. We agree with this action and look forward to hearing more as the plan is developed. PSCC is currently preparing a letter to the NRC on the proposed methodology for dealing with the hard-to-detect radionuclides. It was agreed that the Oversight Committee will review this letter before it is submitted to the NRC.
Following our meeting, we summarized our discussions, including our concerns on the final survey plan, with Messrs. McCarter, Crawford, Warembourg, Calton, and Mrs. Fisher. The next Oversight Committee meeting is scheduled for April 13 and 14, 1994. We look forward to this meeting.
Please call if you have any questions or comments on this letter or the enclosed report.
Sincerely, p
tilliam H. Hamilton Oversight Committee Chairman Enclosure cc:
P. McCarter C. Crawford D. Warembourg M. Fisher Oversight Committee Members
J
's Enclosure to Letter Dated December 21,1993 MEETING REPORT l
Date:
December 2,1993 1
Place:
Fort Saint Vrain Site
Subject:
Oversight Review of Decommissioning Plans for the Fort Saint Vrain (FSV) Facility
Participants:
Oversicht Committee W. H. Hamilton, Consulting Engineer - Chairman of Committee i
P. Woollam, Nuclear Electric PLC l
N. Cole, MPR Associates B. Lipford, MPR Associates F. Crimi, Lockheed Public Service Co. of Colorado (PSCC)
P. McCarter, PSCC Senior VP Electric Operations C. Crawford, VP Electric Production D. Warembourg, Program Mgr. Defueling/ Decommissioning M. Fisher, Assistant Program Mgr.
T. Borst, PSC S. Chesnutt, PSC ;
M. Holmes, PSC M. McBride, PSC J. McCauley, PSC l
H. O'Hagan, PSC G. Schmalz, PSC PSCC Subcontractors C. Calton, Westinghouse V. Likar, Westinghouse E. Parson, Westinghouse - SEG B. Hug, MK Ferguson D. Sexton, Westinghouse - SEG T. Dieter, MK Ferguson T. Howard, Westinghouse R. Kvasager, MK Ferguson 4
1
t
. Q; I
SUMMARY
This was the nineteenth meeting of the Oversight Committee. The purpose of the meeting was to review the status and plans for decommissioning the Fort Saint Vrain
' facility.
Handouts presented to meeting attendees are provided as Attachment I to this report and are listed below:
i Descrintion of Handouts Oversight Committee Agenda for December 2,1993 FSV Status and Specific Issues f
Final Survey Plan Update s
Our comments during the meeting are presented in the forwarding letter of this report.
l The proposed dates for the next meetings in 1994 are provided below:
i Meeting Dates in 1994 (Alternate Dates) i March 23,24 April 13,14 June 22,23 July 6,7 i
September 21,22 September 28,29 December 14,15 December 7,8 j
l 1
i
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- to i
Meeting Report dated i
- December 21,1993 l
f
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HANDOUTS FROM PSCC MEETING ON DECEMBER 2,1993 i
e
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1 I
i
OVERSIGHT COMMITTEE MEETING December 2,1993 FORT M
ST. VRAIN W
,,it,,
,6.
nunum i646ii eie 1:llt i
4
- r
- 1.,
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F OVERSIGHT' COMMITTEE MEETING AGENDA
)
-TNURSDAY - DECEMBER 2, 1993 0730 - 0830 Travel to Fort St. Vrain 0830 - 1230 i
o Tour r
-1230 - 1300 g3 1300 - 1330 e
Tour Report 1330 - 1400 Status - Crawford 1400 - 1500 l
Follow-Up A
Final Survey Plan Update - Borst
+
1400 '1500 Committee Reports NRC Meeting on Final Survey Plan - Crimi I
1500 - 1600 e
Closed Session i
Meeting with McCarter on December 2,1993 at FSV from 1600 to 1700.
P OVERSIGHT COMMITTEE MEETING December 2,1993 FORT ST. VRAIN STATUS AND SPECIFIC ISSUES
3 STATUS OF SPENT FUEL SHIPPING LEGAL PROCEEDINGS NEPA COUNTER SUIT - DISCOVERY PHASE DOE IS PREPARING AN EIS FOR THE INEL SITE WHICH WILL INCLUDE THE ENVIRONMENTAL ASPECTS OF TRANSFERRING FSV FUEL TO IDAHO. DOE HAS ANNOUNCED THAT THE EIS WILL BE COMPLETED BY JUNE 15, 1995. THE IMPLEMENTATION PLAN WAS RECENTLY RECEIVED AND IS BEING REVIEWED BY PSC.
es o 'tQ COMFr nr PSC APPEAL OF SHOSHONFeBANNOCK CASE DECISION PSC IS AWAITING A COURT DECISION BUT IT IS NOT EXPECTED l
ANY TIME SOON.
l l
l
1
_j i
STATUS OF DECOMMISSIONING CURRENT MAJOR DECOMMISSIONING WORK ACTIVITIES:
e THE SHIELD WATER SYSTEM IS FUNCTIONING WELL - NO FURTHER EQUIPMENT FAILURES. THE TURBIDITY HAS GENERALLY BEEN LESS THAN I NTU FOR SOMETIME NOW.
i BLOCK REMOVAL IS PROCEEDING WELL. AS OF DECEMBER I, 1993 A TOTAL OF 656 BLOCKS HAVE BEEN REMOVED AND 25 I
SHIPMENTS HAVE BEEN MADE TO RICHLAND.
LOW LEVEL RADIOACTIVE WASTE DISPOSAL RATES AT THE e
NORTHWEST COMPACT DISPOSAL SITE THE WASHINGTON UTILITIES AND TRANSPORTATION COMMISSION (WUTC) HAS RULED AND THE ESTABLISHED RATE IS $28.30/ CUBIC i
FOOT. THE SEMI-ANNUAL RATE ADJUSTMENT RESULTED IN A 1
RATE OF S19.61/FT8 WHICH IS BEING APPEALED BY U.S. ECOLOGY.
U.S. ECOLOGY HAS ALSO APPROACHED THE LARGE GENERATORS GROUP ABOUT A POSSIBLE SETTLEMENT OUT OF COURT.
U.S. ECOLOGY HAS APPEALED THE 1992 WUTC RULING AND WAS GRANTED AN INJUNCTION. THE JUDGE IN THE CASE HAS SET THE INTERIM RATE AT $36.10/ CUBIC FOOT (DOWN FROM LAST QUARTER OF $37.70). THE DIFFERENTIAL BETWEEN THE WUTC RATE AND THE COURT IS BEING PUT INTO AN ESCROW ACCOUNT.
THE RATES ARE SUBJECT TO REVIEW AND MAY BE ADJUSTED ON A QUARTERLY BASIS.
NOTE THAT THE ABOVE RATES DO NOT INCLUDE THE ASSOCIATED TAXES AND SURCHARGES.
~
STATUS OF DECOMMISSIONING (continued) r SAFETY PSC HAS GONE 496 DAYS WITHOUT A LOST TIME ACCIDENT.
THE FOLLOWING TABLE PROVIDES A
SUMMARY
OF THE SAFETY i
RECORD ON THE DECOMMISSIONING PROJECT FROM AUGUST 1, 1992 TO NOVEMBER 30,1993:
\\
SAFETY RECORD PROJECT AND ANNUAL THROUGH NOVEMBER 30,1993 Company Near Hits First Aid Recordable /
IAst Time Medical.
PSC 5
5 0
0 (1993)
PSC 5
7 3
0 Project WT N/A N/A 9
1 (1993)
WT N/A N/A 15 1
Project
STATUS OF DECOMMISSIONING (continued)
RADIATION PROTECTION PERFORMANCE t
TOTAL ANNUAL AND PROJECT EXPOSURES THROUGH NOVEMBER 30,1993 ARE SUMMARIZED BELOW:
RADIATION PERFORMANCE PERSON-REM DURATION ESTIMATE GOAL ACTUAL 1992 30.21 25.16 14.23 JAN. - NOV.1993 78.7 49.5 71.046 PROJECT TOTAL THROUGH 108.91 74.66 85.276 l
NOVEMBER 30,1993
)
STATUS OF DECOMMISSIONING.(continued) l LOW LEVEL WASTE SHIPPING STATUS e
PROJECT LOW LEVEL WASTE SHIPPING STATUS THROUGH NOVEMBER 30,1993 IS SUMMARIZED BELOW:
RADIOACTIVE WASTE DISPOSAL STATUS
)j Duration Number of Total Volume
- Total Activity i
Shipments (ft')
(Curies) 1991 Total 26 1,028 8,084 1992 Total 70 12,131 32,686 1993 Total 97 34,112 19,111 Project Total 193 47,271 59,881 i
l i
FSV PCRV UNDERWATER BLOCK REMOVALS 12/2/93 100 -
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Core Support Floor Blocks Total AllGraphiteitems*
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j Large Side Reflector Blocks Side Spacer Blocks Cask Shipments-Blocks / Posts Core Support Floor Poets gpgj
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Shaded area equals f
- Total also includes 60 Hex Reflectors total shipped.
and 2 Defueling Blocks.
i 9
i L
t i
FINAL SURVEY PLAN UPDATE TED BORST, CHP DECEMBER 2,1993
).-
FINAL SURVEY PLAN UPDATE FINAL SURVEY PLAN CONTAINED IN SECTION 4 OF DECOMMISSIONING BASED ON NUREG 2082 AND NUREG 5512 DESIGNED TO BE AN OVERVIEW - DETAIL TO BE INCLUDED IN IMPLEMENTING PROCEDURES PSC ELECTED TO PROCEED WITH DRAFT NUREG 5849-BASED FINAL SURVE PLAN REFLECTS CURRENT NRC PHILOSOPHY FEEDBACK INDICATED NO SIGNIFICANT COST DIFFERENCE PSC ELECTED TO RETURN TO NUREG 2082-BASED FINAL SURVEY PLAN FEEDBACK FROM NUMARC ON INDUSTRY PRECEDENT INDICATION FROM CONTRACTOR OF SIGNIFICANT COST DIFFERENCE NRC MEETING ON NOVEMBER 15-16 PROVIDED USEFUL INFORMATION USE OF NUREG 2082 COULD RESULT IN SIGNIFICANT REVIEW TIME USE OF DRAFT NUREG 5849 WILL EXPEDITE FINAL SURVEY PROCESS LARGEST NRC CONCERN IS IN AREA OF HARD TO DETECT NUCLIDES NRC EXPECTS " IMPLEMENTATION PLAN" SUBMITTAL PSC HAS DECIDED TO RETURN TO DRAFT NUREG 5849-BASED FINAL SURVE PLAN DRAFT NUREG 5849 APPROACH IS MORE TECHNICALLY CORRECT HARD TO DETECT ISSUE IS INDEPENDENT OF NUREG 2082 OR 5849 CHOIC COST DIFFERENCES APPEAR TO BE MANAGEABLE
m FINAL SURVEY PLAN UPDATE.
JOINT PSC/WT TECPEICAL TEAM ASSEMBLED TO ADDRESS FINAL SURVEY ISSUES l
MEETING WEEKLY TO DISCUSS ALTERNATIVES INDUSTRY EXPERTISE BEING UTILIZED NUMARC REPRESENTATION OVERSIGHT COMM11 Att GPU - 6ev Goo O SHOREHAM HARD TO DETECT NUCLIDES (HTDN)
ISSUE IS INDEPENDENT OF NUREG 2082 OR DRAFT NUREG 5849 LARGEST SINGLE COST IMPACT ON FINAL SURVEY PLAN HTDN POSITION READY FOR NRC SUBMITTAL BY JANUARY I,1994 APPROACH WILL BE DOSE-BASED - COMPARISON TO Co-60 DOSE "BIPLEMENTATION PLAN" BEING DEVELOPED LEVEL OF DETAIL TO BE INCLUDED WILL BE BETWEEN THAT CURRENTLY IN SECTION 4 OF DP AND THAT TO BE IN IMPLEMENTING PROCEDURES WILL BE BASED ON DRAFT NUREG 5849 RESOLUTION OF HTDN ISSUE SHOULD EXPEDITE NRC REVIEW OF PLAN ANTICIPATE SUBMITTAL TO OVERSIGHT COMMITTEE /NRC IN FIRST QUARTER OF 1994 i
?
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I
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,i i-December 21,1993 i
l p
Mr. Clegg Crawford Public Service Company.of Colorado P.O. Box 480
.l Denver, CO 80201 l
Subject:
Fort Saint Vrain Defueling/ Decommissioning Oversight Committee i
Dear Clegg:
With our current frequency of meetings, we find ourselves sometimes losing track of.
)
issues from meeting to meeting. To help us in this regard, we would like to initiate an "open items" list. The list will contain all the various topics and areas of concern we have discussed in previous meetings that we feel warrant followup or discussions in the future.
The enclosure shows the list we would like to use at our next meeting. We propose that we review the list at each meeting, following your status update. Some items on the list can be covered quickly, while others will probably be the topic of full presentations. We j
will update the list following each meeting based on our discussions.
t Please call me if you have any questions.
j Sincerely, l
4J F
William H. Hamilton j
Oversight Committee Chairman l
l Enclosure cc:
D. Warembourg M. Fisher l
I 1
. ~.
Eh
^
-.9 OPEN ITEM LIST (UPDATED DECEMBER 2,1993) l l
3 f
Topic Issue / Concern 1.
Final Site Survey Plan Implementation plan / methodology and completion schedule (February 28-510)-
2.
Heavy Load Handling Industrial safety of dropping shield bell on floor or work platform.
l 3.
- What is feed and bleed rate to keep up with tritium build up
- How will tritium diffusion into concrete be measured and integrated into final survey plan -
4.
Hard to-Detect Review tracking of radionuclides (Chlorine 36, Radionuclides Carbon 14, Tritium, Iron 55, Nickel 53, Nickel 59, 3
Others?)
5.
Demin Beds Performance at low activity concentrations I
6.
BOP Work Status update 7.
Divers Status update 8.
Release of Clean Material Review materials released to-date (amount, activity, where released to)'
9 Core Barrel Cutting Status of equipment development / testing
- 10. Core Floor Removal Status of development
- 11. Survey Result.s Update on any new survey results in PCRV including j
steam generator surveys
- 12. Initial Site Survey Report Copies to Oversight members
- 13. PCRV Water Clarity Clarity degradation during aggressive in vessel work t
T ~
THE WALL STREET JOURNAL MONDAY JANU 6
.)
,1993
-\\
=
ClOSina COSfS the other 14 that have closed earlier than pdhmu d C@ne pmnts an average of only 12.7 years; F' ort St.
e5 out that the last straw that caused it to Vram nn for & And mth tM avan Nuclear Utilities Face ma in =v *=r c=as a problem com-per-kilowatt cost of runmng a nucitar close Fort St. Vram w
- a vaats: eacks plant now edgmg tugher than the cost of a in the reactor's steam tubes-coannred pam. oenmment of enere IrnmenSe E>Xpenses The relatively small. 230-merawatt officials say pnvately that 25% of the Fon s1. vrain,mn1ces1 mt mm,. te rema nex, mmg reacton may be closed in the In Dismanth,ng Plants build in the 1970s. Takin
n =$t :333==oa: =*g it apart safely decade,o, econom1c,ea,ons.1h,t r narreeme
means utttties such as Public Service of mth state regulators, the utility's cus-Carae.=counned a eeharm paa Customers and Shareholders tomers will still be helpmg to pay for the plant'$ demise in the year 2003.
Please Turn to Pope A5. Cblumn !
Face Years of Fighting g$mg,2$*3,g1nfl1,m$"[,ol3,;,sClosingCosts: Utilities Larger-than expected costs from early 1m th 2" Over Bearing the Burden threatening some utilities with huge bills for which they are utterly unprepared.
Respiratorsa2ubber Boots
' saving for netirement O
ay ear y O u
The Nuclear Regulatory Commission OldNuclear Plants By RostzT Jo1{NSON requires utilities gracually to put aside as And ANN DE RoctricsAc muCD as $135 million for each of their Staff Reponen of Tur Wau. srnrrT JOURNAL.
nuclearplants to cover tne costs of disman-Omtmued erom arst rage FORT ST. VRAIN. Colo. - Nuclear thng "decommissionmg."ingovermnent of na vmg decommissioning funds over 1our power has caused utilities so many head-parlance. But NRC officials acEnowledge decades, are being caught short.
aches over the years that some are ready that trus sum is far short of the real amount Wall Street analysts say the utilities in-to just walk away from it. But they can't needed: they say they will soon issue dustry should have confronted the eco-sharply higher esumates of how much nomic reality of decommissionmg long even do that, Retiring old plants is turning out to utilities should put away for the end of the ago. To be licensed to use nuclear power.
be such a challenge that the visitors' atomic road.
utilities had to file plans shomng how they center at a plant here, which once told A recent Stanford Uc:versity study sug-would return an obsolete power-plant site schoolchildren about the marvels of atomic gests that unlities snould alreacy have to pnstine condition. But they didn't have l power, now entertains engmeers who accumulated a total of $33 billion to have to calculate the expected cost of doing so. I come from as far away as Japan to study enough for evemual plant dismantling, but and fewer than a dosen utilities ha the hugely costly and complex process of the NRC estimates that only 54 billion has such estimates public.
dismantlement.
been stashed so far. When Portand Gen-The financial facts of nuclear de-Fort St. Vrain is the ficst fully opera-eral Dectne Co. in Oregon abrup'Jy an-commissiomng by utilities will usiser in an tional commercial nuclear plant to be nounced plans earlier this month to close era of lengthy regulatory battles over how taken apart piece by piece, its owner.
its 67.5%-owned Trojan nuclear plant, the much of the costs can be passed along to Public Service Co. of Colorado. is among Utility's coffers contamed only 8% of the customers, industry officials predict. But the growing ranks of utility compames now 54S8 m$ ion estunated to be its share of the utilities themselves will almost cer-facing a harsh reality: Not only are some dismantling costs. It will try to wring the tain!y have to shoulder big chunks of the t nuclear plants too expensive to run, but it rest of its share from consumers in a cost - thus eroding their profit margms.
I may cost more to take them apart. in regulatory battle that may take years.
raising their debt totals and making it today's dollars, than it cost to build them in Rising Estimates more expensive for them to borrow.
i the first place.
"IjusthopeI'm retired fromratin l
It is a painful lesson painful for the The worst news is yet to come. Some Ity bonds when most of this happens."g util-
, companies, for their shareholders and for utinties are already ra:3mg esumates Daniel Scotto, a utilltrbond analyst at stiys t their rate payers. Nuclear plant disman-of anucipated dismanttmg costs farfugherDonaldson, Lufkin & Jenrette. "Teanng (ling, says James Greene. a utilities con-
, than those forecast by the NRC For away the layers of decommissionmg prob-sultant at the accounting firm of Arthur eaample. Amentan Dec=c Power Co..
lems is like peeling an omon. Your eyes Andersen & Co.. is "the big bogy out there based in Columbus. Ohio. recently in-tear more and more."
watting."
creased the d:smanthng forecast for its In the six years since Public Service of Costly Repairs two nuclear umts. Whose combmed 2."100-Colorado's Fort St. Vrain operaton went The Fort St. Vrain plant has become a merswatt capaeny is seven umes that of officially sour with a $200 million charge symbol of the problem. There were no acct-
, Fort St. Vram. to a sum m the range of against earnings. the utility's net income dents here, no radiation leaks. no alarms
! $555milliontoSt.lbillion-comparedwith has plunged to 13.8% of capital from 19.3%
i a 19S9 estimate of $340 million-despite a lessened corporate tax rate, and about meltdowns. There was just a long hst SMarly. Nebraska P its debt rating has been reduced four times of temporary closings and costly repatts.
tnet based in Columous. m h D>-
W mm & Pw's CO Neb., more than tnpied the dismantung cost forecast for its For now a number of utilities caught in es oPer tion y about 5% !t 836-megawatt nuclear plam last year to similar squ'eezes are mothballing plants time. Mark Stutz, a spokesman for the
$1.15 bilhon.
until they can accumulate dismantling utility, says simply: Our nuclear plant Moreover, the day of recronmg is far funds or discover lower cost ways to dis-dida I wort.
} Ined when they buut their plants. Nu-closer for many unht:
pose of the facilities. The NRC allows Fort SL Vrain was the first and onh.
utilities to wait up to 60 years before helium-cooled commercial reactor in the clear facihties are licenseo by the NRC to they must dismantle a plant that has been U.S. The rest are water cooled includmg operate for a supposed 40-year life evete.
taken out of service. But maintaining, but the 15 p! ants closed so far were open for Inspectmg and secunng such a facility can still cost up to 515 million a year. Says Ron Bins, director of Colorado's Office of Con-P A
\\
sumer Counsel: "You'll need generations customers about this plant.s costs,.. says And all that water must be chemically 0
UI treated to remove radioactive resms.
01 m er (foreseenin 1965.
Descen ants f the plan n
Und adMonal amounts M water
" ub ic Service of Colorado hired 1830s pioneer Marcelun St. Vrain, asked must be used for washmg workers' protec-i ro based General Atomics to de-that the plant be called somethmg else to tive clothmg. A quarantmed laundry has the Fort SL Vrain plant at this former save the family from embarrassment.
- 5e p fur-trading post 35 miles north of Denve..
ash up m 0 t a
An early pamphlet about Fort St.
This was still the age of innocence for Vrain put mamtenance requirements at expected to take.
nuclear-fueled electncity. It had been only little more than a two week refuelmg stmt
,,You wash all those suits and clean ions 11 years smee Lewts Strauss. chairman annually, in reality. the piant sat useless Itum that water. Then you cut up the wash-l of the Atomic Energy Commission issued for months at a time. In 1986, the Public Ing machine and the dryerand pack them his famous prediction that electncity could become "too cheap to meter." and Service Commission of Colorado stopped inside steel boxes. You chop up the floor only eight years since President Eisen-the utility from chargmg for Fort St.
underneath where the washer and dryer hower had waved a makeshift " magic Vram's power until it got costs under were, says James Krause wand" to open the nation's first commer-control. Three years later, fated with a houseElectncCorp.engmeer. a Westmg-consultingat 1
cial reactor near Pittsburgh. Walt Disney
!)ve-year repair job on the plant's cooling Fort St. Vrain. 'The last thing in the box is pubbshed a nuclear pnmer called "Our system, the utility gave up and closed the the Geiger counter you used to test every-place.
thing, and you bury that, too."
Fnend the Atom." Utility industry bm-chures depicted nuclear power matmg the Company Mficials muHed their option Public Service of Colorado figures it f
Arcue balmy enough for a tourist to sun-for 60 years, but will take 39 months from start to finish bathe on an iceberg, srppmg a tropical ultimately decided to start takmg the plant to complete the dismantling of the Fort St.
drmk.
apart last August "We just couldn't see Vram nuclear plant. But it plans to I "We all really believed that the nucIear guardmg the place for half a century,"
leave the outside walls standing, and even-i i
era would be one of decimmg electricity says Mr. Warem gurg.
tually rebuild the innards to bunt natural costs." says Duane Chapman, an econo-Why is dismantling a plant so expen-mist at Cornell University and a former save? Engineers cite the extenstve safety
..uys CimMege I
y consultant to the Department of Energy.
traming reqmred, the need to rotate of de pMa W m't have a "Never m my wildest creams did I think it workers to limit radiation exposure and nuclearreactor, of course. But we'll still be could be this expensive."
the lengthy plannmg of every move in making electricity."
When construction began on Fort St.
contammated areas. Thomas LaGuardia, Vrain in 1968. many environmentahsts an engmeer who consults with uulities cere still proclaiming nuclear power the about dismantling plants, says. "You need answer to fossiMuel poHution. The nuclear up to four hours to get ready to do some industry got one of its biggest boosts from jobs that would be simple in a fossil fuel the federal Clean Air Act of 1970, which plant. Sometimes you'll have to build toughened standards for coal plants. But mock-up reacmrs to pracuce.,so you don't by the time the plant opened in 1979 waste time on me real ming /
concerns about safety and waste disposal Nuclear dismantling is made tougher had long since replaced the rosy stenanos.
by the plant designs, wiuch cram all the At Fort.it. Vrain, those concerns trans-smsidu matdalinm me smanest possb lated into mountmg costs.
ble spaces to Ilmit radioacuve contamma-Public Service of Colorade had ongt' tion. "This is hot, sweaty work by people I
nelly planned to operate the plant with 54 weanng protective smts, respirators and corkers, but the number swelled to 400 rubber boots." says Mr. IAGuardia. " Pro-evsn before it opened. Then the work force ductivity will go way down on these t
ballooned again, to S37. under the public jobs."
j i
scrutmy that developed after the Three At Fort St. Vrain, the workers are Mlle Island nuclear accident.
findmg that nothing is simple. Just getting Meanwhile the plant was givmg its access to some of the radioactive areas of critics plentyof ammunition. A spokesman the plant means sledgehammenng aside for General Atomics says that the plant tons of steel pipes and cement walls.
"was sale. Unfortunately, there were The huge amount of water used to cool some bugs." The plant didn't consistently other nuclear reactors wasn't supposed to produce electncity at a cost that would be a problem at Fort St. frain's 1.4004e-provide the utility a profit under con-gree core because this plant uses helium to contml temperatures. But engmeers have sumer cost ceilings set by state regulators.
discovered they will have to pour a mdlion "An economic disaster redo's consumer counse." concludes Colo-gallons of water into the reactor vessel
- l. Mr. Bmz.
dunng dismanthng as a radiation shield.
"We were always under the gun from i
__m
3 9.
.e i
FORT ST. VRAIN DECOMMISSIONING i
UPDATE ON RECENT EVENTS i
PUBLIC SERVICE COMPANY OF COLORADO t
PRESENTATION TO THE NUCLEAR REGULATORY COMMISSION November 8,9,1993 Y
$ W$
furona (
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4 t
AGENDA Decommissioning Status Low Level Waste Shipping Safety and Radiological Performance Response to Chairman's Challenge (Whistleblowers)
Final Release Survey Present Decommissioning Concerns l
Supplemental Information 1
l w -.
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Page 1
t DECOMMISSIONING STATUS Removal of 1320 Tons of top head concrete I
complete ~
36' diameter rotary work platform installed 6" thick top head insulated steel liner assembly removed t
Removal of 1770 activated graphite components in progress llo dy s<Wge (3
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LOW LEVEL WASTE SHIPPING 4
i LLRW Shipments for disposal and volume reduction:
Disposal facilities at Beatty, Nevada and Richland,
~
Washington ukreer Volume Activity Shipments (cu.ft.)-
(Cl) 1 Direct 42,300 55,000 168 Disposal Sent to 31,100 160 45 Processor TOTAL 73,400 55,160 213
( % uqc Page 4
o SAFETY AND RADIOLOGICAL PERFORMANCE RADIOLOGICAL PERFORMANCE Total Project Estimate 433 person-Rom From Start of Project to November 1,1993:
Estimate 101.3 wAii f
Goal 90.9 person-Rem Actual 79.9 person-Rem LOW EXPOSURES DUE TO EFFECTIVE ALARA CONTROLS e
Page 5
RADIOLOGICAL PERFORMANCE (Cont'd) 10 skin contaminations,30 clothing contaminations Industry experience in Radiologically _ Controlled Areas:
Plants Contaminations per 10.000 RCA hours FSV 2.4 BWR Outages 14.0 PWR Outages 6.5 O positive bioassay results pu e_
Received full exemption from new 10 CFR 20 e
Page 6
s INDUSTRIAL SAFETY PERFORMANCE i
From Start of Project to November 1,1993:
Only 1 lost time work accident Lost work day incidence rate (per 200,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />)
Contractor 0.45 PSC 0
Severity rate (per 200,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />)
Contractor 5.1 PSC 0
Comparison with typical construction projects (per 200,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />):
Lost work day incidence rate 6.9 Severity rate 132 t
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r Page 7
RESPONSE TO CHAIRMAN'S CHALLENGE PSC is currently involved in a Whistleblower Complaint Destaffing plan called for release of all 38 contract HP Technicians,4 of which were released in March 1991 Downsizing due to inability to ship spent reactor fuel to Idaho National Engineering Laboratory 1 Technician released in March 1991 alleged release was due to raising safety concerns
-l Department of Labor investigated in April 1991 i
i NRC investigation completed in February 1992 Enforcement conference held May 10,1993 Notice of Violation / Imposition of Civil Penalty received in October 1993 1
PSC is appealing the Notice of Violation / Civil Penalty I
P e
I Page 8
j a
WHISTLEBLOWER-PROCESS i
i
'PSC supports NRC's position on whistleblowing and.
protecting whistleblowers If the process'does not obtain all of the evidence and reach an impartial decision, the process is suspect and justice and i
safety are not served Process assumes licensee is guilty until proven innocent NRC whistleblower investigations only pursue one side of the story NRC Enforcement Conferences are the first opportunity for licensees to present the other side of the story The final decision, made by people in Washington who i
personally heard neither side's story, may be politically motivated h ([%id - 4 kue is:s co w w cm ~,e 1
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1 Ps00 9
k FINAL RELEASE SURVEY i
Final Release Survey is major challenge of Project Project based on Release Criteria in Decommissioning Plan
> ' 5 pR/hr above background at 1 meter
~
10 MREM / year CS, Reg. Guide 1.86 contamination limits Fifnal Radiation Survey Plan approved in Decommissioning Plan
\\
Based on NUREG/CR 2082 and NUREG/CR 5512 guidance Successful Project completion requires Final Release Survey
(
criteria no more restrictive than Decommissioning Plan criteria C n C. u w eek. aloo g g o (Qy y 3CL
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Page 10
- i FINAL RELEASE SURVEY (Continued) i Goals of Final Release Survey:
i Demonstrate that levels of residual contamination pose no significant health risk Complete Final Release Survey in thorough and cost-effective manner Complete Final Release Survey within Decommissioning Schedule Release Part 50 license, allow for unrestricted use of site L( 70m v/ do (%v c(y
, rh k s..g Communications with NRC are vital to effective planning and implementation of Final Release Survey i
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Page 11
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PRESENT DECOMMISSIONING CONCERNS i
LLRW disposal costs M @C Final Release Survey Plan Continuing regulatory interface Unknowns associated with a project of this magnitude
)
Spent fuel disposition /High Level Waste repository ~3 FST_
fxcf 72 o&~cks Lack of a mixed waste disposal facility
-aan tu r Confirmation that FSV release criteria will not be subject to change r
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S SUPPLEMENTAL INFORMATION 1
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- w 3
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NUCLEAR REGULATORY COMMISSION
.Dt E j
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REGloN Iv e
URANIUM RECOV FIELD OFFICE
./
DENVER, COLORAoO 8022S j
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t-FEB 2 61993 g3g Docket:
50-267
'?72-009 f~
License: DPR-34 I
SNM-2504 Public Service Company of Colorado ATTN:
A. Clegg Crawford, Vice President Electric Production P.O. Box 840 Denver, Colorado 80201-0840
SUBJECT:
NRC INSPECTION REPORT 50-267/92-12, 72-009/92-12 (Notice of Violation)
This refers to the inspection conducted by Messrs. P. W. Michaud, D. C. Ward, and P. J. Garcia during the period of November 29, 1992, through January 29, 1993. The inspection was a review of activities authorized for the Fort St.
Vrain Nuclear Generating Station. At the conclusion of the in.spection, the findings were discussed with those members of your staff identified in the enclosed report.
Areas examined during the inspection are identified in the report. Within these areas,dhe. inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observation of activities in progress.
Based on this inspection, one of your activities was in violation of NRC requirements.
The violation involved the failure to notify the NRC within four hours of a liquid waste discharge flow path blockage which could have resulted in an uncontrolled release. While recognizing the fact that the release in progress at the time was below the maximum permissible concentration (MPC), the event could have had more serious consequences.
In accordance with 10 CFR Part 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosure will be placed in the NRC Public Document Room.
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely,
/
amon E. Ha, Director Uranium Recovery Field Office gww
. ;y { 4F L s'
Public Service Company 2
of Colorado FSB 2 619D
Enclosures:
Appendix A - Notice of Violation Appendix B - NRC Inspection Report 50-267/92-12, 72-009/92-12 w/ attachment CC W/encloss.9:
Public Service Company of Colorado ATTN:
M. H. Holmes Project Assurance Manager 16805 Weld County Road 19-Platteville, Colorado 80651 GA International Services Corporation Fert St. Vrain Services ATTH: David Alberstein, Manager P.O. Box 85608 San Diego, California 92138 Public Service Company of Colorado ATTN:
D. Warembourg Decommissioning Program Director 16805 Weld County Road 19-Platteville, Colorado 80651 Public Service Company of Colorado ATTN:
D. D.-Hock, President and Chief Executive Officer P.O. Box 840 Denver, Colorado 80201-0840 Kelly, Stansfield & 0'Donnell ATTN: Mr. J. K. Tarpey 1225 17th Street, Suite 2600 Denver, Colorado 80202 Chairman Board of Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1
1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413 4
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Public Service Company-
-3 PE 2 0 g3 of Colorado Department of Health 4
ATTN: Robert M. Quillin, Director Radiation Control Division 4210 East lith Avenue p
Denver. Colorado. 80220-a l-Colorado Public Utilities Commission
[
ATTN: Ralph Teague,'P.E.
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- 1580 Logan Street OL1-
-Denver, Colorado 80203 l
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j APPENDIX A NOTICE OF VIOLATION Public Service Company of Colorado Docket No. 50-267 Fort St. Vrain Nuclear Generating Station License No. DPR-34 During an NRC inspection conducted on January 27, 1993, a violation of NRC requirements was identified. The violation involved the failure to notify the NRC within four hours of a situation which could have resulted in an uncontrolled liquid waste release.
In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, the violation is listed below:
A.
Section 5.5.3.a. of the Decommissioning Technical Specifications, which were incorporated in License No. DPR-34 through Amendment No. 85, states that "The NRC Operations Center shall be notified of emergency and non-emergency events in accordance with 10 CFR 50.72."
10 CFR 50.72(b)(2) states, in part, that "the licensee shall notify the NRC as soon as practical and in all cases, within four hours of the occurrence of any of the following:
(iii) Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to:
(C) Control the release of radioactive materi al. "
Contrary to the above, the licensee reported a situation which could have resulted in an uncontrolled liquid waste release at approximately 4:00 p.m. on January 27, 1993, which was over 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after the situation was initially discovered.
This is a Severity Level V Violation (Supplement I).
Pursuant to the provisions of 10 CFR 2.201, Public Service Company of Colorado is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, NRC Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011-8064, and a copy to the NRC Uranium Recovery Field Office, P.O. Box 25325, Denver, Colorado 80225, within 30 days of the date of the letter transmitting-this Notice of Violation. This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each violation:
(1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved.
If an adequate reply is not received within the time specified in this Notice of Violation, an order may be issued to show cause why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.
Dated at Denver, Colorado this 26th day of February,1993
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APPENDIX B
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U.S. NUCLEAR REGULATORY COMMISSION URANIUM RECOVERY FIELD OFFICE 1
REGION IV NRC Inspection Report: 50-267/92-12 Facility License: DPR-34 72-009/92-12 ISFSI License: SNM-2504 Docket Nos.:
S0-267; 72-009-Licensee: Public Service Company of Colorado (PSC)
P.O. Box 840 i
Denver, Colorado 80201-0840 Facility Name: Fort St. Vrain Nuclear Generating Station (FSV)
Inspection At: FSV, Platteville, Colorado Inspection Conducted: November 29, 1992 through January 29 1993 l
Inspectors:
P. W. Michaud, Project Manager, URF0 P. J. Garcia, Project Manager, URF0 D. C. Ward, Project Manager, URF0 Approved:
[h}uj 2445 m
EdwaYd F. Hawkins,~ Deputy Director Date
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Uranium Recovery Field Office Region IV t
Inspection Summary Areas Inspected: Routine, unannounced inspection of the licensee's Audit Program, Plant Procedures, Surveillance Procedures and Records, Maintenance Program, Operational Safety Verification, Security Plan, Emergency l
Preparedness, Radioactive Waste Management, ISFSI Activities, and Transportation of Radioactive Materials.
i Results.
Within the areas inspected, one violation of NRC requirements was found.
e The licensee failed to report a blockage of the liquid waste discharge line to the NRC within four hours, as required. Since the discharge
)
occurring at the time of this event was below MPC, there were no radiological consequences. The failure to report an event which could have had more serious implications is of concern since it may indicate a lack of sensitivity to the reporting requirements which remain in effect, though the plant is in a decommissioning status.
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The annual emergency preparedness exercise was held during this inspection period. The response of the licensee's on-site organization and facilities was demonstrated satisfactorily.
During operations to remove the last of the dummy defueling blocks, two of the dummy blocks were inadvertently tipped onto their sides, making them inaccessible to the fuel handling machine's grapple. The licensee and the decommissioning contractor management team assessed the events and found no safety implications and no need to retrieve the blocks immediately. The licensee plans to retrieve the blocks manually after the PCRV is flooded and the top head opened.
All other activities observed by the inspectors appeared to be appropriately conducted and controlled, and no other significant issues were identified.
Summary of insoection Findinos:
One violation was identified (Section 7).
Attachment:
Attachment - Persons Contacted and Exit Meeting i
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3 DETAILS 1 PLANT STATUS On November 23, 1992, the NRC issued an Order authorizing the decommissioning
)
of FSV. Amendment No. 85 to Facility License No. DPR-34 was issued concurrent with the Order. This amendment replaced the Appendix A Technical Specifications with new Decommissioning Technical Specifications. The licensee implemented the new Technical Specifications (TS) and procedures based on these TS during this inspection period. Decommissioning activities l
are contracted to Westinghouse, who has subcontracted with MK-Ferguson (MKF) and Scientific Ecology Group (SEG).
Plant activities during the period included the removal of the steam generator secondary assemblies and the removal and disposal of dummy defueling blocks, metal clad blocks, and reflector blocks from the Prestressed Concrete Reactor i
Vessel (PCRV). The removal of blocks from the PCRV with the fuel handling machine was nearly complete at the end of this inspection period, with'3588 of 3652 blocks removed. Covers were welded onto the 12 steam generator penetrations and bolted onto the four helium circulator penetrations in preparation for filling the PCRV with water. Detensioning and removal of vertical tendons from the PCRV, removal of piping and electrical interference from the top head area, and horizontal core boring activities were performed in preparation for cutting of the PCRV top head concrete. These activities were performed in compliance with the conditions of the license and the i
provisions of the decommissioning order.
2 OPERATIONAL SAFETY VERIFICATION (71707)
The inspectors made periodic tours of the plant to observe activities in i
progress, verify plant status and assess the overall conditions. During these tours, attention was paid to the licensee's fire protection program elements, including the status of fire extinguishers, fire fighting equipment, fire doors, fire watches when required, and control of flammable materials and other fire hazards. The inspectors noted that plant housekeeping was i
maintained at a good level with occasional deterioration identified and corrected by the licensee. No other problems were identified during these tours.
The fuel deck and control room shift logs were reviewed along with the TS compliance log, clearance requests, and operations order book. The operability of plant systems was verified to be in accordance with the conditions of the license and the decommissioning order. Activities were observed to be conducted in accordance with approved work packages and procedures. The inspectors also attended selected plan-of-the-day meetings to maintain awareness of the status of plant activities and to identify areas that may require inspectors attention or followup to verify conformance with 1
license and regulatory requirements.
The inspectors periodically observed work in radiologically controlled areas on the turbine deck, fuel deck, and in other areas of the reactor building and
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the former new fuel storage building, which is now used for ' temporary storage of radwaste packages. The inspectors observed the posting control, Land j
maintenance of radiologically contro11ed' areas during tours of the areas.. The j
inspectors'also' periodically observed health physics _ technicians performing routine 'and special contamination and radiation surveys, and providing L l
radiological. control and assistance to workers. The inspectors selectively.
- observed job briefings'that were being provided to workers entering radiologically. controlled areas.: In addition, the lnspectors reviewed.
selected radiation work permits.and observed that workers complied with the-instructions concerning job initiation, protective clothing and equipment, and
, personnel dosimetry. The inspectors also observed that workers'used proper procedures for contamination control, including the performance of personal l
= contamination surveys upon. exiting of radiologically controlled areas. - No 1
problems were identified in this area, and the inspectors noted that.
radiological controls were being implemented in an excellent manner.
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On January 6,1993, during operations to remove dumy fuel blocks from the:
PCRV with the fuel. handling machine (FHM), two events occurred which resulted 1
in two dummy fuel blocks being tipped out of position. The licensee 4
determined that there were no safety implications to these events, and i
I subsequently decided to leave the blocks in the PCRV. The licensee. intends to retrieve these blocks manually, along with the remaining reflector blocks, after the PCRV.is flooded and opened.
It should be noted that.the
. decommissioning plan had previously called for removal of all the'dumy blocks in this fashion..
The first event occurred while removing dumy fuel blocks from Region I,.
located at the center of the core, which had 15 dumy fuel blocks-remaining.
While attempYini to retrieve a center column block, the FHM experienced an address failure in the automatic mode (this is not' uncommon with the FHM).
The operator re-zeroed the machine and returned it to automatic. The FHM then apparently tripped again. When the operator subsequently attempted to retrieve the block manually, the FHM mast camera showed the block to be lying on its side. The FHM oper'ations were temporarily suspended while the situation was discussed among the operators, the PSC fuel deck' manager, the MKF operations manager, and the. Westinghouse project director.
It was then determined that 1) there were no health, safety or radiological implications to the event, 2) the'Fim grapple head could not be used to retrieve a block on-
.its side and use of the FHM manipulator arm was not as desirable as later manual retrieval. under water, and 3) removal of the remaining dummy blocks should continue. No evaluation was apparently made of how or why the event occurred before the decision was made to continue operations. The FHM was
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then used to push the tipped block out of the way, and block removal operations were resumed approximately three hours after the initial event.
1 The next 13 dummy blocks were subsequently removed without incident, though the next center column block was removed manually due to an interruption in the FHM automatic program. While attempting to retrieve the final dummy -
block, the FHM again experienced an address failure. The operator manually re-zeroed the machine and returned it to the automatic mode. The FHM then automatically set the block back down and released it at the coordinates where the FHM had previously tripped. The block was thus slightly out of alignment
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5 and, having no surrounding blocks, was allowed to tip. A subsequent evaluation of the second event also concluded that there were no safety or radiological implications, though a more thorough analysis of the cause of the event was undertaken. The corrective actions implemented as a result of these events included a caution statement in the FHM operating procedure which requires that if the machine goes out of automatic at any time during a center column pickup, the move must be completed manually before returning to the automatic mode.
The events caused some concern over whether too much emphasis was being placed on maintaining production or schedules at the expense of safety. The licensee investigated the events, and determined that the cause of the first tipped block could not be established with any certainty.
This demonstrates the need to evaluate an event of this nature fu!1y at the time of its occurrence, since it may be impossible to recreate later. The events also caused some question on the level of involvement and effectiveness of PSC oversight of the contractors. These areas will be monitored closely during future inspections.
3 SURVEILLANCE AND MAINTENANCE ACTIVITIES (61700/61726)
Surveillance schedules and surveillance packages were routinely checked by the inspectors to confirm that surveillance tests were conducted at the required intervals. The inspectors observed portions of surveillances in progress and reviewed documentation of completed surveillances. The licensee initiated station service requests for components that failed to meet surveillance requirements, and retests were appropriately identified in surveilhnce schedules. The licensee's implementation of surveillance ' activities was acceptable with, regard to scheduling, tracking, and completion of surveillances required under the decommissioning technical specifications.
Maintenance activities were performed in accordance with approved work procedures. The activities often involved radiological aspects, as with work on the fuel handling machine, which were appropriately controlled and executed. No major maintenance activities occurred during this inspection period, and no areas of concern were identified.
4 OPERATIONAL STATUS OF EMERGENCY PREPAREDNESS (82701)
The annual emergency response exercise was held on December 8, 1992, and was observed by the NRC inspectors. The exercise scenario was developed to test the licensee's Decommissioning Emergency Response Plan's implementation in response to a realistic accident. The exercise involved a simulated drop of irradiated graphite blocks, which is a credible accident scenario involving potential offsite consequences.
The Control Room personnel properly classified the event, made the appropriate notifications within the required times, and verified habitability. The Technical Support Center (TSC) was manned in approximately 30 minutes.
Habitability of the TSC and the adjacent briefing room was verified periodically. The TSC Director maintained control and provided routine status j
briefings. Accountability was determined in just under one hour, with one person missing per the scenaria. There was some confusion over the missing
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person, since his name was similar to an individual who had recently left employment at the site. The confusion resulted in an initially fragmented approach to finding the missing person, rather than use of an organized search. This was the only weakness observed by the inspectors during the exercise.
1 Four radiological monitoring teams were deployed during this exercise to survey selected areas. Each team was deployed from the TSC and consisted of two health physics technicians. The teams were briefed on anticinated radiological conditions, dosimetry, protective clothing, and inst.rumentation requirements.
The inspectors concluded that the licensee had demonstrated the operational readiness of their Decommissioning Emergency Response Plan. The required facilities and equipment were adequately maintained, and appropriate staffing by qualified individuals was verified.
5 SECURITY PLAN AND ISFSI ACTIVITIES (81018/86700)
The licensee's Decommissioning Access Control Plan provides an industrial security program to prevent sabotage of decommissioning systems or radioactive waste storage, and to prevent theft of company owned materials. The inspectors verified that the site fence was maintained as required and that access to the restricted area was controlled through positive identification of each individual. Security personnel were observed performing routine tours and were equipped for continuous communication. Procedures were implemented by the licensee for personnel and vehicle access control, ' communications equipment, physt. cal barrier inspections, and record keeping.
The Independent Spent Fuel Storage Installation (ISFSI) is normally locked and alarmed, but unmanned. Routine tours of the facility by the security force and periodic surveillance activities were performed as required. The inspectors periodically verified the integrity of the facility and its conformance with the license requirements.
6 AUDIT PROGRAM AND PLANT PROCEDURES (40702/42700)
The inspectors attended meetings of the licensee's Oversight Committee on December 9 and 10, 1992. The Oversight Committee is made up of five industry executives who provide independent reviews of and input to the decommissioning activities. The topics covered during these meetings included updated radiation protection training associated with decommissioning, plans for the disassembly of the PCRV top head and activation analysis, and the protocol criteria for releasing bulk clean waste from the site. The licensee and contractor personnel made effective presentations on these subjects, and the oversight committee asked significant and critical questions.
Several recommendations were made on ways to improve various aspects of the ongoing decommissioning tasks and on considerations to be taken into account for future activities. The Oversight Committee requested additional information on PCRV contamination patterns, the process to be used to remove the final layer of material on the top head to open the PCRV, integration of safety programs among PSC and the various contractors, planning and scheduling for J
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i reactor vessel internals cutting and removal, and a redraft of the protocol paper on releasing bulk clean waste. The presence of the Oversight Committee and their level of involvement appeared to be extremely beneficial to the safe and efficient decommissioning of Fort St. Vrain.
The licensee was required to have the Decommissioning Technical Specifications, and procedures to implement them, in place by December 7, 1992. The inspectors verified that an entirely new set of procedures were implemented. The inspectors reviewed selected procedures and found them to be in accordance with the requirements of the Decommissioning Technical Specifications, though somewhat general in nature. The licensee indicated that the working level procedures were issued by the contractor (SEG). The inspectors reviewed selected SEG procedures and verified their adequacy.
7 RADI0 ACTIVE WASTE MANAGEMENT AND TRANSPORTATION (84101/86750)
On the afternoon of January 26, 1993, the licensee informed an NRC inspector that a liquid waste release the previous night had overflowed from the top of an underground oil separator and onto the ground. The release contained some detectable Cesium and Tritium, but was below the maximum permissible concentration (MPC) for each of these elements with no dilution.
The event was discovered at approximately 1:30 a.m. on January 26, 1993, when an operator went outside to reposition valves following the completion of the
-release and the subsequent firewater flush. The licensee evaluated the event and determined that, because the contents of the release were below MPC, it was not a reportable event. The licensee also determined'that since there was no oil flow aff site, it was also not reportable to the State or the EPA. The licensee investigated the cause of the event and determined it to be ice blockage downstream of the oil separator. This was attributed to recent extremely cold weather and the fact that fewer liquid waste releases are occurring during decommissioning than during previous operations. The licensee placed heat traci.ng on all accessible portions of the liquid waste discharge line. The licensee also took soil samples from the area where the oil separator had overflowed and found some soil contamination, which was removed and barrelled.
When the NRC inspector was informed of the event at approximately 3:00 p.m.
I the same day, the licensee was asked if it was reported in accordance with 10 CFR 50.72. The inspector then informed the licensee, after consulting with NRC management, that the event was reportable as a four hour non-emergency report under the requirements of 10 CFR 50.72(b)(2)(iii)(c), since it "could have prevented the fulfillment of the safety function of structures or systems i
that are needed to control the release of radioactive material." The licensee subsequently made the appropriate notifications. Since the report was not made within four hours as required, the licensee was informed that this was a violation of NRC requirements (50-267/9212-01).
The inspectors reviewed the shipping records maintained by the licensee. At the time of the inspection, a total of sixteen shipments had been made to
)
Richland, Washington in 1993. Additionally, the inspectors reviewed a i
representative number of the approximately 70 shipments made to Beatty, Nevada
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i Nevada, during 1992. All paper work required for proper documentation of each shipment was present and appeared to be in order.
The Radwaste Supervisor had reviewed all documentation and had initialed each form after his review, certifying that the work had been conducted properly and in accordance with procedures.
Appropriate radiological and safety surveys were noted to have been conducted on all shipments prior to leaving the facility. Photographic records indicate that all shipments were appropriately placarded and secured prior to leaving the site.
The inspectors interviewed a truck driver during the course of the inspection.
The driver appeared to be well trained on his responsibilities during the I
transport of radioactive materials. A wide range of questions vere asked of f
the driver concerning radiological and vehicle safety during trnnsportation.
In all cases the driver responded quickly and accurately about his required job duties, demonstrating good comprehension in the subject area. The truck driver was also asked if he routinely notified PSC upon arrival at his j
destination. He stated that he did not, but that the shipping manifest would be returned to Fort St. Vrain by mail, indicating that the shipment had reached its destination. Although the regulations do not require immediate notification that a shipment has been completed, the inspectors were concerned
'3 that an accident, especially during a period of inclement weather, could incapacitate the driver and no one would be contacted for days. This concern was raised at the exit meeting and licensee stated that they would review the matter.
The inspectors observed the removal from the refueling floor of a liner containing radioactive components taken from the PCRV for disposal at Richland, Washington. The liner was lifted by crane from the dedicated shielded storage area located on the refueling floor and lowered through the removable hatch down to an exclusive use trailer waiting in the truck bay below. A crew of four workers guided the liner into place on the trailer and promptly winched the container into the side loading cask mounted there. The inspectors noted that the operation was handled quickly and efficiently to keep exposures ALARA.
Unn'cessary personnel were excluded from the truck bay e
1 and refueling deck area during the completion of the task.
Individual workers appeared t.o know their job duties well,1oving in unison to complete the task promptly and without incident. Workers were observed to position themselves behind shielding to prevent unnecessary exposure during times when not actively engaged in securing the liner. Once the liner was secured within the cask, workers conducted the appropriate radiation surveys in preparation for shipment.
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The transportation program at the site appeared to be functioning adequately.
No concerns other than the minor one discueed above were identified by the inspectors.
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t ATTACIMENT t
1 PERSONS CONTACTED 1.k Licensee Personnel D. Blain, Health Physicist
- F. Borst, Facility Support Manager S. Chesnutt, Nuclear Licensing Engineer
- M. Fisher, Deputy Project Director
- M. Holmes, Project Assurance Manager J. Johns, Nuclear Licensing Engineer
- M. McBride, Project Controls Manager
- J. McCauley, Project Engineer M. Niehoff, Project Engineering Manager H. O'Hagan, Planning and Scheduling Coordinator
- G. Reigel, Operations Manager D. Seymour, Quality Assurance Engineer
- D. Warembourg, Decommissioning Project Director 1.2 Contractor Personnel C. Calton, Project Director, Westinghouse C. Cummin, Radwaste Supervisor, SEG T. Dieter, General Superintendent, MKF W. Hug, Site Operations Manager, MKF E. Parsons, Project Radiation Protection Manager, SEG
- D. Sexton, Radiation Protection Technical Support Supervisor, SEG M. Zachary, ALARA Supervisor, SEG K. Zahrt, Radiation Protection Operations Supervisor, SEG The ins'pectors also contacted other licensee and contractor personnel during the inspection.
- Denotes those attending the exit meeting.
2.
Exit Meetina An exit meeting was conducted on February 1, 1993. During this meeting, the scope and findings of the inspection were reviewed with the licensee. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspectors.
IK e, e.
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UNITED STATES pa macu
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NUCLEAR REGULATORY COMMISSION o
E,a REGION IV k
URANIUM RECOVE RELD OFMCE DENVER,COLORADOSimE NOV 151993 Docket: 50-267 72-009 License: DPR-34 SNM-2504 l
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Public Service Company of Colorado ATTN:
A. Clegg Crawford, Vice President Electric Production P.O. Box 840 l
Denver, Colorado 80201-0840
SUBJECT:
NRC INSPECTION REPORT 50-267/93-03, 72-009/93-03 i
This refers to the inspection conducted by Messrs. Paul W. Michaud, Pete Garcia, Jr., and Dana Ward during the period of June 1 through July 31, l
1993. The inspection was a review of activities authorized for the Fort St.
l Vrain Nuclear Generating Station. At the conclusion of the inspection, the l
findings were discussed with those members of your staff identified in the enclosed report.
Areas examined during the. inspection are identified in th'e report. Within' e
these areas 7 tHb inspection consisted of selective examination of procedures
~
and representative records, interviews with personnel, and observation of i
1 activities in progress.
No violations or deviations were identified during this inspection.
In accordance with 10 CFR Part 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosure will be placed in the NRC Public Document Room. Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, I
amon E. Hall, Director Uranium Recovery Field Office
Enclosure:
Appendix - NRC Inspection Report 50-267/93-03,72-009/93-03 w/ attachment
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'Public. Service Company 2
NOV 151993 of Colorado cc w/ enclosure:
Public Service Company of Colorado ATTN:
.D. D. Hock, President and Chief Executive Officer P.O. Box 840 Denver, Colorado 80201-0840 Public Service Company of Colorado ATTN:
D. Warembourg Decommissioning Program Director 16805 Weld County Road 19-Platteville, Colorado 80651 Public Service Company of Colorado
+
ATTN:
M. H. Holmes Project Assurance Manager 16805 Weld County Road 19-Platteville, Colorado 80651 Public Service Company of Colorado ATTN: Commitment Control Coordinator 16805 Weld County Road 19-%
Platteville, Colorado 80651 GA Internatibnat Services Corporation a
Fort St. Vrain Services ATTN: David Alberstein, Manager' P.O. Box.85608 San Diego, California 92138 Kelly, Stansfield & 0'Donnell ATTN: Mr. J. K. Tarpey 1225 17th Street, Suite 2600 Denver, Colorado 80202 Chairman Board of Commissioners i
Greeley, Colorado 80631 i
Regional Representative Radiation Programs Environmental Protection Agency i
1 Denver Place i
999 18th Street, Suite 1300 1
Denver, Colorado 80202-2413 4
.9,
'1; Public Service Company 3
. Department of Health ATTN: Robert M. Quillin, Director Radiation Control Divisioa
-4210 East lith Avenue Denver, Colorado 80220 Colorado Public Utilities Commission ATTN: Ralph Teague, P.E.
1580 Logan Street OLI Denver, Colorado 80203 l
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l APPENDIX U.S. NUCLEAR REGULATORY COMMISSION URANIUM RECOVERY FIELD OFFICE REGION IV i
NRC-Inspection Report:
50-267/93-03 f
72-009/93-03 Facility License:' DPR-34 ISFSI License:
SNM-2504 Licensee: Public Service Company of Colorado (PSC)
P.O. Box 840 Denver, Colorado 80201-0840 Facility Name: Fort St. Vrain Nuclear Generating Station (FSV)
Inspection At: FSV, Platteville, Colorado Inspection Conducted: June 1 through July 31, 1993 6
Inspectors:
P. W. Michaud, Project Manager, URF0 P. J. Garcia, Project Manager, URF0 D. C. Ward, Project Manager, URF0 Approved:
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O Edward F. Hakkins Deputy )irector Date Uranium Recovery Field Office Region IV Inspection Summary Areas Inspected: Routine, announced inspection of Operational Safety Verification, Solid Radioactive Waste Managemen't, Transportation of Radioactive Materials, and Occupational Exposure During DECON.
Results:
i All activities observed by the inspectors appeared to be appropriately conducted and controlled, and no other significant issues were identified.
Summary of Inspection Findinos:
No violations or deviations were identified during this inspection.
Attachment:
Attachment - Persons Contactad and Exit Meeting 5 ll l ?~EN$
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2 DETAILS 1 PLANT STATUS On November 23, 1992, the NRC issued an Order authorizing the decommissioning of FSV. Amendment No. 85 to Facility License No. DPR-34 was issued concurrent with the Order. This amendment replaced the Appendix A Technical Specifications with new Decommissioning Technical Specifications. The licensee contracted the decommissioning activities to Westinghouse, who has subcontracted with MK-Ferguson (MKF) and Scientific Ecology Group (SEG).
Plant activities during the period included removal of all concrete segments from the Prestressed Concrete Reactor Vessel (PCRV) top head. The PCRV shield water system was operated intermittently, and experienced several problems.
Water chemistry results indicated the radionuclide concentrations in the shield water were well below expected levels, though turbidity remained high.
Activities were performed in compliance with the conditions of the license and the provisions of the decommissioning order.
2 OPERATIONAL SAFETY VERIFICATION (71707)
The inspectors made periodic tours of the plant to observe activities in progress, verify plant status and assess the overall conditions. During these tours, attention was paid to the licensee's fire protection program elements, including the status of fire extinguishers, fire fighting equipment, fire watches when required, and control of flammable materials and other fire hazards. Thv inspectors noted that plant housekeeping had deteriorated slightly, especially in areas where more than one activity was in progress.
The licensee was informed that additional attention was needed to avoid tripping hazards and to control debris.
The refueling floor log, control room shift log, TS compliance log, clearance requests and operations order book were reviewed periodically. The operability of plant systems was verified to be in accordance with the conditions of the license and the decommissioning order. The inspectors also attended selected plan-of-the-day meetings to maintain awareness of the status of plant activities and to identify areas that might need additional attention or followup to verify conformance with license and regulatory requirements.
The Independent Spent Fuel Storage Installation (ISFSI) is normally locked and alarmed, but unmanned. Routine tours of the facility by the security force and periodic surveillance activities were performed as required. The inspectors periodically verified the integrity of the facility and its conformance with the license requirements.
The inspectors periodically observed work in radiologically controlled areas on the refueling floor and in other areas of the reactor building and the former new fuel storage building, which is used for temporary storage of radwaste packages. The inspectors observed the posting, control, and maintenance of radiologically controlled areas during tours of the areas. The inspectors also periodically observed health physics technicians performing
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I routine and special contamination and radiation surveys, and providing radiological control and assistance to workers. The inspectors selectively observed job briefings that were being provided to workers entering radiologically controlled areas.
In addition, the inspectors reviewed selected radiation work permits and observed that workers complied with the 1
instructions concerning job initiation, protective clothing and equipment, and personnel dosimetry. The inspectors also observed that workers used proper procedures for contamination control, including the performance of personal contamination surveys upon exiting of radiologically controlled areas. No problems were identified in this area, and the inspectors noted that radiological controls were implemented in an appropriate manner.
On June 29, 1993, a worker was injured while working in the south concrete segmenting tent on the reactor building refueling floor. Personnel working in the segmenting tents are required to be in full protective clothing due to the presence of loose contamination. The worker was making a core drill in one of the concrete segments which had been removed from the PCRV top head in order I
to remove a neutron detector. While attempting to slow or stop the core drill bit with his hand, the worker's glove was caught in the core drill bit. This caused his hand to be pulled backwards around the drill bit, resulting in three compound fractures. The individual was surveyed clean and transported offsite by ambulance. The inspector reviewed the licensee's and their contractor's response to this event and determined it was prompt and well executed.
The licensee continued to have problems with the PCRV shield water system during this inspection period. Two impeller failures occurred in pump No. 1, apparently due ~to preexisting defects in the impeller castings. The licensee also had some problems with the performance of the filters, which were apparently being bypassed somehow and therefore not removing particulate matter from the water. Turbidity of the shield water continued to increase, due in part to some concrete slurry from the top head cutting operations having entered the system. Lowering of the turbidity levels is desired to improve visibility for in-core component removal activities. Efforts were ongoing to correct these problems and improve the performance of the system.
No violations or deviations were identified in this area.
3 OCCUPATIONAL EXPOSURE DURING DECON (83100)
The inspectors reviewed selected ALARA packages and noted that the packages were well prepared for each job. The packages indicated that a good ALARA review had been performed to implement appropriate engineering techniques for control of airborne radioactivity and dose reduction from external exposure hazards. The inspectors also reviewed the monthly ALARA committee meeting reports.
In these meetings, the ALARA committee reviewed the project exposure history, radiological occurrence reports, ALARA suggestions, radwaste generation and exposure evaluations. Through June, 1993, the total person-rem exposure for the project was 28.6 Rem, compared to an estimated exposure of 433 Rem and an ALARA goal of 347 Rem. The licensee's radiological controls have continued to result in exceptional performance. Personnel seemed to be
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aware of ALARA concerns and objectives, with over 100 ALARA suggestions having been provided.
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Selected radiation work permits were reviewed by the inspectors, each of which appeared to be sufficient to provide adequate control of the radiation hazards involved. The inspectors noted that on the radiation work permits which
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required cutting or drilling of radioactively contaminated concrete, i
appropriate engineering controls were included, such as the use of HEPA filtering units. The inspectors reviewed records of airborne radioactivity samples and determined that the licensee performed airborne sampling as called for in various radiation work permits. Records of airborne samples indicated that airborne radioactivity levels were routinely below the licensee's action level for respirator usage.
The inspectors noted that respirators were used by the licensee for both industrial hazards and airborne radioactivity hazards. A review of selected records for individuals that used respirators indicated that they were qualified for the specific respiratory equipment used.
The inspectors reviewed radiological occurrence reports for 1993. The reports-
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identified suitable concerns for followup and the corrective actions which were taken. The inspectors noted that multiple occurrences were documented against two individuals in these reports. The inspectors also noted that i
after the last infraction by these individuals, they were denied access to the radiological controlled area; one for two days the other for thirty days. No further infractions by these individuals were documented after they were disciplined by being denied access. One report reviewed by the inspectors dealt with imdividuals that were sprayed by core drilling water that contained tritium. The licensee took appropriate corrective action in this case. The inspectors reviewed sample analyses from cutting and drilling operations and determined that tritium levels were very low and that the licensee's program for tritium bioassay was adequate.
No violations or deviations were identified in this area.
4 RADI0ACTIVEWASTEMANAGEMENTANDTRANSPORTATION(86750)
The inspectors observed the methods for removing the concrete slurry produced during the diamond-wire cutting activities from the PCRV top head liner.
Full barrels of decanted slurry are then solidified by evaporating the remaining liquid. The evaporation process is performed with electric heaters on the barrels, which are located under an exhaust hood connected to the reactor building ventilation system.
The inspector reviewed the licensee's procedure for determining the alarm setpoints for the liquid waste effluent monitors. Step 6.7.3 of procedure DPP 5.4.2, was recently revised in Iswa 3 to include an adjustment factor to account for tritium and other hard to detect isotopes. The value of this adjustment factor is provided by the PSC Radiation Protection Manager and based on maintaining the total concentration of all nuclides at or below MPC.
The inspector noted a potential problem in that the procedure did not specifically restrict this new adjustment factor to less than or equal to 1.
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i The revised procedure was in fact.used to increase the alarm setpoints on one occasion by using an adjustment factor greater than 1.
Though the specific instance did not present a problem, the potential existed to exceed MPC for the more. restrictive isotopes by using an adjustment factor greater-than 1.
This was discussed with the PSC Radiation Protection Manager who agreed to revise the procedure to include a restriction on the adjustment factor of less than or equal to 1.
t No violations or deviations were identified in this area.
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ATTACHMENT l
1 PERSONS CONTACTED j
1.1 Licensee Personnel
- D.- Blain, Facility Support Engineer
>l F. Borst, Facility Support Manager S. Chesnutt, Project Assurance Engineer
- T. Dice, Shift Supervisor
- M. Fisher, Deputy Program Director
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- M. Holmes, Project Assurance Manager M. McBride, Project Controls' Manager
- M. Niehoff, Project Engineering Manager H. O'Hagan, Planning and Scheduling Coordinator
- D. Warembourg, Decommissioning Program Director i
1.2.
Contractor Personnel l
- C. Calton, Project Director, Westinghouse T. Dieter, General Superintendent, MKF
- W. Hug, Site Operations Manager, MKF
- E. Parsons, Project Radiation Protection Manager, SEG
- D. Sexton, Technical Support Supervisor, Westinghouse M. Zachary, ALARA Supervisor, SEG K. Zahrt, Radiation Protection Operations Supervisor, SEG i
The inspectors also contacted other licensee and contractor personnel during the inspectton.~
- Denotes those attending the exit meeting.
2 EXIT MEETING t
An exit meeting was conducted on July 29, 1993. During this meeting, the scope and findings of the inspection were reviewed with the licensee. The licensee did not identify as proprietary any information provided to, or
-l reviewed by, the inspectors.
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UNITED STATES p stop\\
NUCLEAR REGULATORY COMMISSION f
REGION IV 8
611 RYAN PLAZA DRIVE, SUITE 400 o
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' AR LINGTON, TEXAS 760114064 JAN 2 41994 Docket:
50-267 72-009 License: DPR-34 SNM-2504 Public Service Company of Colorado ATTN:
A. Clegg Crawford, Vice President Electric Production P.O. Box 840 Denver, Colorado 80201-0840
SUBJECT:
NRC INSPECTION REPORT 50-267/93-04; 72-009/93-04 This refers to the routine, announced inspection conducted by Messrs. R. J. Evans and P. W. Michaud of this office on October 13 through December 9, 1993.
The inspection was a review of activities authorized for
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the Fort St. Vrain Nuclear Generating Station. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the enclosed report.
Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selected examination of procedures and representative records, interviews with personnel, and observation of activities in progress.
No violations or deviations were identified during the inspection.
In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter, the enclosure, and any response will be placed in the NRC Public j
Document Room.
Should you have any questions concerning this inspection, please contact Mr.
Robert Evans at (817) 860-8234.
I Sincerely, I
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, a, Dwig D. Chamberlain, Acting Director Division of Radiation Safety and Safeguards i
Enclosure:
Appendix - NRC Inspection Report 50-267/93-04, 72-009/93-04 f
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9 Public Service Company of Colorado cc: w/ enclosure:
Public Service Company of Colorado ATTN:
D. D. Hock, President and Chief Executive Officer P.O. Box 840 Denver, Colorado 80201-0840 Public Service Company of Colorado ATTN:
D. Warembourg Nuclear Operations Manager 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Public Service Company of Colorado ATTN:
M. H. Holmes Nuclear Licensing Manager 16805 Weld County Road 19-1/2 i
Platteville, Colorado 80651 Public Service Company of Colorado ATTN: Commitment Control Coordinator 16805 Weld County Road 19-1/2 Plattaville, Colorado 80651 GA International Services Corporation Fort St. Vrain Services ATTN: David Alberstein, Manager P.O. Box 85608 San Diego, California 92138 l
0'Malley and Associates, P.C.
ATTN:
- 5. B. O'Malley 950 South Cherry Street, Suite 520 Denver, Colorado 80222 i
Chairman Board of Commissioners Weld County, Colorado a
i Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413
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Public Service Company-
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- of Colorado Department of Health-ATTN:- Robert M. Quillin, Director Radiation Control Division 4210 East lith Avenue Denver, Colorado' 80220 Colorado Public Utilities Commission ATTN:~ Ralph Teague, P.E.
.1580 Logan Street OL1 Denver, Colorado 80203 t
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c-t APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV e
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? Inspection Report:
50-267/93-04 72-009/93-04 License: DPR-34 SNM-2504 Licensee:
Public Service Company of Colorado P.O. Box 840 Denver, Colorado 80201-0840 a
Facility Name: Fort St. Vrain Nuclear Generating Station Inspection At:
Platteville, Colorado Inspection Conducted: October 13-14 and December 6-9, 1993 Inspectors:
R. J. Evans, Radiation Specialist P. W. Michaud, Project Manager Approved:
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W. L. Fisher, Chief, Nuclear Materials Date Licensing Section n
Inspection Summary.
Areas Inspected:
Routine, announced inspection of plant status, operational safety inspection, organizational structure, surveillance procedures and records, corrective action process, onsite review of one licensee event report, and follow up for the corrective actions taken for a previous i
violation. Also documented in this inspection report are details of a decommissioning status meeting that was held between the licensee and the NRC during the inspection period.
J Results:
The licensee was observed to have maintained good control of the Fort St. Vrain station refueling floor housekeeping, site radiological boundaries and postings, and security oversight of the Independent Spent J
Fuel Storage Installation (Section 2.1).
A weakness in the control of the decommissioning procedure system was identified when several out-of-date procedures were discovered in the plant (Section 2.2).
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. A commitment that was documented in a previous NRC inspection report was not implemented as stated in the report. The licensee's explanation of their position on the subject was determined to be acceptable (Section 2.3).
The failure to maintain control over the required posting boards was determined to be a weakness.
This problem may have been contributed to by a' procedure which incompletely quoted 10 CFR 19.11 (Section 2.4).
The licensee's organizational structure was determined to be the same structure that was comitted to in the Decomissioning Plan and Independent Spent Fuel Storage Installation Safety Analysis Report.
Changes to the structure will require that a change be formally made to these documents (Section 3.1).
The licensee had developed a surveillance program that met the intent of.
the license requirements. The record keeping of the surveillance tests was determined to be a licensee strength (Section 4.1).
The corrective action program was determined to be accep' table for the level of work in progress at the facility.
Several procedural weaknesses and procedural enhancements were discussed with the licensee for improvements to the corrective action process (Section 5.1).
A meeting to discuss the decommissioning status of Fort St. Vrain was-held in the Region IV office. A copy of the meeting handout is attached to this inspection report (Section 6).
Summary of InsDection Findinos:
Licensee Event Report 50-267/93-001 was closed (Section 7.1).
Violation 50-267/92-012-01 was closed (Section 8.1).
Attachments: - Persons Contacted, Exit Meeting, and Decommissioning Status Meeting Attendance
. - Decommissioning Status Meeting Handout
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DETAILS 1 PLANT STATUS The major decommissioning task in progress at the facility was the dismantlement and decontamination of the radioactive portions of.the prestressed concrete reactor vessel (PCRV). The top head concrete has been removed, sectioned, packaged, and shipped offsite. The top head liner and the PCRV upper side reflector keys were removed by October 1993. The licensee I
then began to remove the remaining graphite core components from the PCRV.
The critical path in the graphite block removal process was the turnaround rate for the low level waste trucks and casks that were used to ship the blocks to and from the burial site in the state of Washington. At the end of the inspection period, 809 of 1770 graphite components had been removed and 31 of 82 cask shipments had been made. Following the completion of the core graphite block removal, the next critical path items are the removal of the core barrel and outer keys, followed by the removal of the care support floor.
Other activities in progress during the inspection period included helium purification piping removal and liquid and gaseous waste piping removal. The licensee was experiencing problems in the removal of the waste system piping that traversed the walls and floors. The licensee was making preparations to remove the waste system tanks by performing radiation surveys and by decontaminating the tanks. Other activities planned in the near future included removal of the radioactive core outlet thermocouples.
The licensee recently experienced problems with the shield water system filters and pumps. The filters were not working properly because of a manufacturing tolerance flaw which allowed flow to bypass the filter elements.
The licensee installed filter inserts to correct the tolerance error and the filters have operated satisfactorily since late August 1993. Also, the licensee experienced multiple pump impeller failures. The impeller failures were determined to be the result of casting defects during fabrication. The impellers have since been replaced. Water clarity continued to improve throughout the inspection period. Turbidity, which had once been at a level of 14 Nephlometric Turbidity Units (NTU), was below 1 NTU during the December 1993 inspection. The system's occasional mechanical problems have not affected core component removal activities.
Project completion was tentatively scheduled for late 1995.
2 OPERATIONAL SAFETY VERIFICATION (71707)
The purpose of this inspection was to ensure that decommissioning activities were being conducted safely and in conformance with license and regulatory requirements.
The following paragraphs provide details of specific inspector observations during this inspection period.
_4 2.1 Plant Tours Brief tours were taken in the Fort St. Vrain facility and the Independent Spent Fuel Storage Installation (ISFSI). The plant's refueling floor activities appeared to be well controlled from a housekeeping and radiological standpoint (this was an area of concern in the last NRC inspection report).
No inappropriate radiological postings were observed. The work areas were generally free of loose debris and tools. The ISFSI was also determined to be well controlled from a housekeeping and radiological standpoint.
Addit,ionally, the control of the ISFSI security was determined to be adequate.
At the beginning of the December 1993 inspection, a thin film was observed on the top of the water in the prestressed concrete reactor vessel cavity. The pool's skimmer pump was not in service at that time'. The next day, the pump was placed in service at a flow rate of 200 gallons per minute. Water ripples were evident during pump operation; however, the film on the top of the water was no longer visible. The water ripples did not stop core decommissioning activities.
2.2 Procedure Control As part of the operational safety verification inspection, the control of procedures associated with the site decommissioning activities was reviewed.
System operating procedures are provided for the individual systems which must be kept operable during decommissioning. The operations procedure set provides the step-by-step instructions necessary to ensure that required systems are operated in accordance with the decommissioning Technical Specifications and the Decommissioning Plan. The DPP 3.1.2 series of procedures provide the specific system operating instructions. A contrciled set of system operating procedures are provided in the main control room.
During the plant tour of the facility, the NRC inspector toured the service water pump house and the fire pump house. Located in each of the buildings is a metal box which was previously used to house system operating procedures.
The procedures were available for equipment operator use during local operations. The inspector noted that superseded or expired procedures were located in the boxes in the two exterior structures. The out-of-date procedures were indicative of a weakness in the control and distribution of,
system operating procedures.
If the licensee wants to distribute procedures.
for use by plant operators in the plant, then the procedures would have to be controlled to ensure the procedures are current and up-to-date. As an interim measure, the shift supervisor removed the out-of-date procedures from the i
storage boxes and discarded them. The inspector suggested to the shift i
supervisor that a sweep of the plant might be necessary to ensure that all out-of-date procedures were removed from all in-plant storage boxes.
Later in the inspection, the inspector was informed that this activity had been completed.
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2.3 Commitment Chance as Documented in an NRC Inspection Report The licensee informed the NRC that a commitment, as documented in a previous I
inspection report,-was not implemented as stated in the report. During this inspection period, the NRC inspector discussed the commitment with the i
licensee'and acknowledged their position on the subject.
NRC Inspection Report 50-267/93-03; 72-009/93-03, in part, documented an inspection of the radioactive waste management and transportation activities of the licensee. As documented in that report, the inspector reviewed the licensee's procedure for determining the alarm setpoints for the liquid waste effluent monitors. A step of this procedure was previously revised to include an adjustment factor to account for tritium and other hard to detect isotopes.
i The adjustment factor was calculated to ensure the licensee did not exceed the maximum permissible concentration (MPC) unity equation for liquid effluent releases. The inspector noted a potential problem in that the procedure did t
not specifically restrict this new adjustment factor to be less than or equal to one.
The inspector concluded that the MPC for the more restrictive i
isotopes could be exceeded by using an adjustment factor of greater than one.
The inspection report stated that this concern was discussed with the licensee's radiation protection manager, who agreed to revise the procedure to include a restriction on the adjustment factor to ensure the factor remained less than or equal to one.
During the December 1993 inspection, the licensee voiced concerns with the i
commitment that was documented in the inspection report. The radiation
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protection manager stated that he committed to revise the procedure to clarify how the adjustment factor would be used. The radiation protection manager did not commit to revising the procedure to ensure the adjustment factor always remained below one because the adjustment factor could be greater than one.
The licensee has since revised the procedure to satisfy the verbal commitment made to the NRC inspector; however, the actions taken did not strictly match the words contained in the previous inspection report. The concerns voiced by the licensee appeared to be reasonable; therefore, the inspector concluded the commitment documented in the previous inspection report may have been taken out of context.
Licensee actions taken appeared to satisfy the previous commitment.
2.4 Inadeauate Control of Reauired Postina Boards i
During tours of the facility, the required posting boards were inspected. The NRC inspector noted that the licensee had not maintained positive control over the documents that were posted pursuant to the requirements established in 10 CFR 19.11.
The control of the required postings was determined to be a weakness.
The NRC inspector noted that the licensee had six locations where NRC information was posted.
The inspector noted that the boards were inconsistent in postings.
One board had an out-of-date Form NRC-3 posted, two boards still had a 1992 inspection report posted, and a recent NRC Notice of Violation, 1
i with a proposed imposition of civil penalty, was on about half of the boards.
None of the boards had the licensee's response letters to a recent violation posted. The incomplete and inconsistent postings was indicative of a weakness in the control of the boards. The licensee should clearly delineate which locations in the plant needed to have posting boards and maintain positive control over the boards.
The inconsistent control of the required posting boa'rds may have been l
contributed to by an inadequate procedure. The procedure, " Government Correspondence," DPP 2.2.5, Issue 1, described the process for controlling and reviewing correspondence relating to Fort St. Vrain from government agencies.
Section 6.3.5, Correspondence Posting, lists the licensee's responsibilities in accordance with the 10 CFR 19.11 requirements. The procedure steps in Section 6.3.5 incompletely quoted the wording of 10 CFR 19.11.
For example, the requirement to post any violation response from the license was left out of the procedure. This deficient procedure was pointed out to the licensee.
2.5 Conclusions During the brief tours of the facility, the licensee was observed to have maintained good control of the refueling floor housekeeping, radiological l
boundaries and postings, and security oversight of the ISFSI.
High quality, accurate, and up-to-date procedures are needed for decommissioning and remediation of the site to ensure the processes are strictly controlled by the licensee. The discovery of out-of-date procedures in the plant was determined to be a weakness in the control of the decommissioning procedure distribution process.
A commitment that was documented in a previous NRC inspection report was not implemented as stated in the report.
The licensee's explanation of their position on the subject was determined to be acceptable.
The inconsistent control over the required posting boards was determined to be a weakness. This problem may have been contributed to by a procedure which incompletely quoted 10 CFR 19.11. The licensee should determine which boards are the official posting locations and maintain these locations appropriately.
3 ORGANIZATION (36800)
The licensee's organizational structure was inspected to ensure compliance with the Decommissioning Plan and Independent Spent Fuel Storage Installation (ISFSI) Safety Analysis Report (SAR) commitments. Also inspected was the licensee's method available for making changes to the organizational structure.
3.1 Details The Decommissioning Plan, Section 2.4, provides a description of the decommissioning organization and responsibilities for the Public Service
i i Company of Colorado. Section 2.5 of the Plan provides a description of the contractor organization and functions. The Plan provides a description of the key management positions and responsibilities for both the licensee and contractor organization. Also provided in the plan are figures that show the functional diagrams for the management chain of command. Additional details of the licensee's organization, including organization charts, are provided in the licensee's " Project Administration Manual," DPM 1.0, Issue 1.
The licensee organization currently consists of about 68 employees, excluding i
an additional 21 security personnel and 27 employees on loan to other organizations.
The Westinghouse Team Organization consists of 150-170 employees, depending on the activity in progress. The Westinghouse Team Organization consists of approximately 15 Westinghouse employees, 40-45 Scientific Ecology Group employees (used primarily for health physics and radiological protection activities), and 90-110 MK-Ferguson employees (used for site labor and labor management).
TheISFSISARdescribesthestaffneededtosupportISFSIoperation. Key members of the licensee's staff also perform functions required by the SAR.
The ISFSI SAR also includes a figure that shows the functional diagram of the j
ISFSI staff chain of command. The actual organization charts for both the Decommis;ioning staff and the ISFSI staff were compared to the charts provided in the Plan and SAR. Other than a few minor variations in the titles of key members in the chain of command, the actual organization charts reflect the same organizational layout as shown on the charts in the Plan and SAR.
The Proposed Decommissioning Plan and ISFSI SAR were previously submitted to the NRC for review. Since the documents are essentially commitments to the NRC, any changes made to the organization structure, as described in the Plan and SAR, will require formal changes be made to the documents. This concept was presented to the licensee because the licensee was considering changing the structure during the middle of 1994.
3.2 Conclusions The NRC inspector concluded that the actual organizational structure reflected the Decommissioning Plan and ISFSI SAR commitments, and that changes to the structure, as described in the Plan and SAR, will require that a change be,
formally made to the documents.
4 SURVEILLANCE PROCEDURES AND RECORDS (61700)
An inspection of the surveillance program was performed to ascertain whether the surveillance of systems and components was being conducted in accordance with approved procedures as required by Technical Specifications (TS). The inspection consisted of a review of the surveillance program, two surveillance 1
procedures, and the surveillance procedure scheduling process. The inspector concluded that the program was well controlled by the licensee.
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' 4.1 Details The procedure, " Surveillance Program," DPP 3.1.3, Issue 1, defined the responsibilities and controls for implementing the surveillance program. The procedure, in part, listed the requirements for the test procedures and reporting test results. The procedure also provided the systematic means for l
scheduling the surveillance program.
This procedure was reviewed in detail during the inspection. Additionally, the surveillance program was reviewed with the senior documents clerk, who has control over the record keeping of the surveillance program. At the time of the inspection, the licensee had 74 surveillance procedures, which varied in required time intervals of I week to 5 years. The inspector noted that the scheduling process apparently works well, as evidenced by the fact that there had not been any missed e
surveillances in 1993.
The NRC inspector witnessed the performance of two procedures, the monthly
" Fire Pump and Instrumentation Functional Test," and the weekly " Visual Inspection of ISFSI Cooling Inlet and Outlets." Both procedures were reviewed and were determined to be technically adequate, including the_ptocedure acceptance criteria. During the performance of the Fire Pump and Instrumentation Functional Test surveillance, the electric fire pump pressure was noted by the test performer to be unacceptably low. The value measured, 90 psig, was below the acceptance criteria limit of 142 psig.
The unacceptably low value resulted in a test failure. A station service request was issued to troubleshoot and repair the local pressure gauge, which was suspected to be sticking in place. Also during the diesel fire pump run, the pump packing was leaking excessively. This condition was' pointed out to the shift supervisor. The inspector had no concerns with the performance of the ISFSI surveillance procedure.
l 4.2 Conclusions i
The licensee has developed a surveillance program that meets the intent of the license and the program appears to be working well.
Of the procedures reviewed, the TS requirements were properly incorporated into the procedures and the surveillance intervals were being adhered to in a timely manner. The record keeping of the surveillance tests was determined to be a licensee strength.
j 5 CORRECTIVE ACTION PROCESS (92720)
The licensee's corrective action program was inspected to determine whether the licensee was effectively implementing a corrective action process that i
identified, followed, and corrected conditions adverse to quality, especially with regard to determination of the cause of the problem and elimination of both problem recurrence and occurrence of similar problems.
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5.1 Details 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, states that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant conditions adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.
10 CFR 72.172, Corrective Action, has similar requirements. The licensee committed to implement these sections of the regulations in both the Decommissioning Plan and the Independent Spent Fuel Storage Installation (ISFSI) Safety Analysis Report (SAR).
The licensee's corrective action program consists, in part, of problem reports, licensee event reports, station service requests, and 10 CFR Part 21 reports. The liseuee also subscribed to the Licensing Information Service, an industry sponsored newsletter. The licensee was not a member of the Institute of Nuclear Power Operations (INP0) program. Additionally, the licensee did not have an employee concerns program.
During the inspection, a selected sample of corrective action documents was reviewed. The NRC inspector concluded that the corrective action system currently in place works for the licensee, in part, because of the low number of documents in the system.
For example, licensee personnel had written 45 problem reports (prs) in 1993 at the time of the inspection.
However, several procedural weaknesses and procedural enhancements were discussed with the licensee for program improvements.
For example, the ISFSI SAR, Section 11.3.1.4, stated, in part, that all corrective action documents are analyzed to identify trends in quality performance, and that the results of these analyses are reported to the Corporate Vice President. The procedure "ISFSI Quality Assurance Manual,"
IPM 2.4, Issue 2, listed the corrective action documents. One of the corrective action documents was the station service request, which is used to document, authorize, and control maintenance on ISFSI equipment. At the time of the inspection, the licensee did net have a program in place to trend ISFSI station service requests. This was not of significant concern to the inspector at the time of the inspection because of the low number of station service requests written for the ISFSI.
The same corrective action paragraph in the SAR stated that the results of these analyses are reported to the corporate vice president. The mechanism on how this is accomplished could not be identified. The Performance Indicator report, issued monthly, apparently satisfied this requirement in the past, but this document is no longer used to track these or similar adverse trends. The licensee should develop a program to trend station service requests and submit the results of adverse trend analyses to the corporate vice president, or the licensee should remove this requirement from the SAR.
. The procedure " Problem Reporting System," DPP 1.0.4, Issue 2, provided the mechanism for bringing problems identified to the attention of the appropriate levels of management for resolution and/or corrective action. The process for handling problem reports (PR) was described in this procedure. The NRC inspector identified several weaknesses in the handling process. According to the procedure, if a PR is identified that involves a procedural problem, the problem is simply corrected (deficient procedure is revised).
The PR procedure then instructs personnel to " forward a copy of the resolved PR to Project Assurance and file the original of the PR for future use." Also, the PR procedure states that "if a contract issue is involved, the decommissioning engineering manager shall sign the PR, file the original for future use, and send a copy of the PR to Project Assurance." This method of operation has two basic problems.
First of all, this method short cycles prs involving procedure problems.
Under the current system, programmatic procedure problems may not be identified because they are not entered into the PR process.
Programmatic procedure problems are areas that could affect quality, I
therefore, the problem reports should be entered into the system for trending purposes. Second, prs involving procedure problems and contract issues do not l
follow the same flowpath as other prs.
The procedure provided irrstructions to
" file the original of the PR for future use."
This statement was ambiguom and should be clarified to prevent the accidental loss of original prs.
For example, the NRC inspector asked the current Decommissioning Engineering Manager where a recent PR involving a contract issue was located, but the manager could not immediately recall where the original was located. The control of these two types of prs should be the same as the other prs or the originals could become lost.
The " Problem Reporting System" procedure had additional weaknesses involving operability and reportability reviews. Once a PR has been entered into the l
system, Step 6.3.2 of the PR procedure provided instructions to forward a copy of the original PR to Project Assurance (licensing) for determination of reportability.
This PR procedure process has a weakness in that the reportability review is not documented as having been completed. Also, the PR procedure did not specifically reauire operations department review of the PR at any point in the PR process. This is of concern because the shift supervisor must ensure equipment and plant compliance with Technical Specification requirements.
Under the current program, a component could be inoperable and the plant could be under a Technical Specification limiting Condition of Operation but the operations department would not be aware of it because the PR process did not require operations notification. The reportability and operability determinations should be documented on the PR 1
with line item signatures.
This method would ensure that these reviews have been completed.
The control of quality assurance hold tags is another weakness of the " Problem Reporting System" procedure.
Step 6.3.4 stated that if a material nonconformance is identified, then hold tags are to be affixed to the nonconforming item.
The hold tag program was not defined or described in the current decommissioning procedures.
The concept of Hold Tags was a carryover from the period before decommissioning; however, the licensee failed to i
4 include this tag or a description for the control of the tag into the current program.
The procedure " Facility License Compliance and Licensee Event Report Program,"
DPP 2.2.2, Issue 1, described the Project Assurance department's activities associated with verifying compliance with the facility licenses and other documents governing facility operation.
Several procedural weaknesses were identified with this procedure.
Step 6.1.1 stated that Project Assurance reviews miscellaneous documents to determine reportability, including surveillance procedures. As documented in Section 4.1 of this inspection report, the surveillance procedure " Fire Pump and Instrumentation Functional Test" was a failure because of an incorrect pump pressure reading. This surveillance would have been a good candidate for a reportability review; however, the reportability review procedure did not describe how the surveillance, and other miscellaneous documents, end up in licensing for a review. The NRC inspector concluded that Step 6.1.1 of procedure DPP 2.2.2 was an ambiguous statement that should be reconsidered by the-litensee.
i Step 6.1.2 of the " Facility License Compliance and Licensee Event Report Program" procedure stated that written reportability evaluations will be completed and reviewed by the Project Assurance Manager or designee, using the Compliance /Reportability Evaluation Form (attachment A to the procedure).
During the inspection, three reportability evaluations that were attached to prs were reviewed. These three evaluations were not performed using the Compliance /Reportability Evaluation Form, or the format described on the form.
Additionally, the second review of the evaluation was not being performed in accordance with the procedure requirements. However, the inspector concluded j
that the failure to perform the second review on the three prs did not result in deficient reportability reviews by the licensee.
The procedure " Government Correspondence," DPP 2.2.5, Issue 1, described the.
process for controlling and reviewing correspondence related to the facility i
from government agencies.
Step 6.3.2 of this procedure stated that governmental letters relating to nuclear industry operating experience, applicable to the facility, will be routed to appropriate personnel. This step is performed using an uncontrolled " Document Review Sheet." The use of the sheet to assist in procedure compliance was proactive on the part of the licensee; however, this sheet should be controlled and should be included in the procedure as an attachment.
The procedure " Commitment Control," DPP 2.2.1, Issue 1, described the process for identifying, controlling, and administering NRC and license related commitments.
Step 6.6.4 of this procedure stated that commitment status reports will be issued periodically to project personnel by Project Assurance.
The inspector concluded the licensee's commitment status reports were not being distributed as required by the procedure. However, the licensee was meeting all commitments within the required time intervals even.though this subject area was not being tracked by a computer generated data base at the time of the inspection.
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t e 5.2 Conclusions Overall, the NRC inspector concluded the corrective action program was adequate for the licensee. NRC commitments were being met by the required due date even though this program was currently not computerized.
The inspector concluded that the current system works for the current level of work in progress at the facility with 45 prs written in the last year.
Several procedural weaknesses and procedural enhancements were discussed with the licensee for improvements to the corrective action process.
6 DECOMMISSIONING STATUS MEETING (30703)
On Tuesday, November 16, 1993, a routine, semiannual meeting to discuss the decommissioning status of Fort St. Vrain was held in the Region IV office.
The attendees of the meeting are listed in the Attachment I to this inspection report. The meeting agenda included the decommissioning status, low level waste shipping, safety and radiological performance, response to the Chairman's challenge involving whistleblowers, final release survey, present decommissioning concerns, and staffing levels. The briefing was, essentially 1
the same as the briefings that were given to headquarters personnel on November 8-9, 1993. A copy of the meeting handout is attached to this inspection report (Attachment 2).
7 ONSITE REVIEW OF LICENSEE EVENT REPORTS (92700) 7.1 (Closed) Licensee Event Report 267/93-001: Overflow of Radioactive Liouid Effluent Line On January 26-27, 1993, the licensee experienced an overflow of radioactive i
liquid waste from an underground oil separator in the release line. There was i
negligible radioactive contamination of the affected soil. The overflow was suspected to be caused by an ice plug in the line downstream of the oil
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separator. Corrective actions taken included the installation of heat tracing i
on 3 inch piping and a valve in the Valve Vault Area. The NRC inspector concluded that this event was an isolated incident and licensee corrective actions were appropriate.
8 FOLLOW UP ON CORRECTIVE ACTIONS FOR VIOLATIONS (92702) 8.1 (Closed) Violation 267/92-012-01:
Failure to Notify the NRC Within Four Hours Following the radioactive liquid eff'luent line overflow event of January 26-27, 1993, the licensee failed to notify the NRC within the required 4-hour time interval. Corrective actions taken included issuance of Operations Orders which discussed managements' reportability philosophy. The orders stressed the philosophy of "when in doubt, report it."
The NRC inspector concluded the corrective actions taken were appropriate and that this incident appeared to be isolated.
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6 ATTACHMENT 1 1 PERSONS CONTACTED 1.1 Licensee Personnel F. Borst, Facility Support Manager S. Chesnutt, Senior Nuclear Licensing Engineer R. Heggen, Contract-and Budget Coordinator M.. Holmes, Project Assurance Manager M. Niehoff, Decommissioning Engineering Manager G. Redmond, Senior Quality Assurance Engineer G. Reigel, Operations Manager D. Seymour, Senior Quality Assurance Engineer D. Warembourg, Director, Decommissioning Program
'1.2 Contractor Personnel G. Bickel, Project Control Manager, MK-Ferguson C. Calton, Project Director, Westinghouse B. Dyck, Licensing Engineer, Westinghouse Team Organization G. Howard, Project Engineering Manager, Westinghouse W. Hug, Site Operations Manager, MK-Ferguson M. Kachun, Lead Site Quality Assurance Engineer, Westinghouse V. Likar, Technical Services Manager, Westinghouse
'J. Parsons, Radiation Protection Manager, Scientific Ecology Group D. Sexton, Technical Support Supervisor, Scientific Ecology. Group P. Schick, Project Control Manager, Westinghouse The personnel listed above attended the exit meeting.
In addition to the personnel listed above, the inspector contacted other personnel during this inspection period.
2 EXIT MEETING An exit meeting was conducted on December 9, 1993.
During this meeting, the inspector reviewed the scope and findings of the visit with the participants.
The participants did not identify as proprietary any information provided to, or reviewed by, the inspector.
3 DECOMMISSIONING STATUS MEETING The personnel listed below attended the NRC-Fort St. Vrain Decommissioning Status Meeting on November 16, 1993, in the Region IV office.
3.1 Licensee Personnel A. Crawford, Vice President, Electric Production M. Fisher, Deputy Program Director, Decommissioning
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- e. 3.2-NRC Personnel D. Chamberlain, Deputy Director, Division of Radiation Safety and Safeguards R. Evans, Radiation Specialist, Nuclear Materials Licensing Section, Division of Radiation Safety and Safeguards W. Fisher, Chief, Nuclear Materials Licensing Section, Division of Radiation Safety and Safeguards J. Milhoan, Regional Administrator J. Montgomery, Deputy Regional Administrator C. Thomas, Acting Director, Division of Radiation Safety and Safeguards b
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