ML20023D734
| ML20023D734 | |
| Person / Time | |
|---|---|
| Site: | 07002985 |
| Issue date: | 05/17/1983 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20023D730 | List: |
| References | |
| 22320, NUDOCS 8306020646 | |
| Download: ML20023D734 (24) | |
Text
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i DUKE P0k'ER COMPANY 1
Catawba Nuclear Station - Unit 1 Application for Special Nuclear Material License General Information Pursuant to 10CFR 70.21 and 70.22 Duke Power Company, et al*, hereby applies for a special nuclear material license for Catawba Nuclear Station - Unit 1, Docket No. 50-413.
Duke is the holder of Construction Permit No. CPPR-116 issued by the Atomic Energy Commission on August 7, 1975. CPPR-116 permits construction of Catawba Nuclear Station - Unit 1 on the shore of Lake Wylie in York County, South Carolina. Duke is an investor owned electric utility incorporated in the state of North Carolina with its corporate headquarters located in Charlotte, North Carolina. The company address is as follows:
Duke Power Company, P.O. Box 33189, Charlotte, North Carolina 28242. Further information regarding Duke Power Company and Catawba Nuclear Station is contained in the application for operating licenses for Catawba Nuclear Station filed with the NRC (Docket Nos. 50-413 and 50-414).
Activities Sought to be Authorized A.
The receipt, possession, inspection, use and storage of: uranium enriched in the U-235 isotope contained in fuel assemblies and in fission chambers, Plutonium-238 contained a Pu-Be neutron source, and a Plutonium-239 alpha so*1rce.
B.
The packaging of fuel assemblies for delivering to a carrier in accordance with 10CFR, Part 71.
Reauested License Duration It is hereby requested that the issued special nuclear material license for Catawba Nuclear Station - Unit i expire on April 30, 1985 or upon conversion of Construction Permit No. CPPR-116 to an operating license, whichever is earlier.
It is currently anticipated that new fuel may be received onsite as early as December 1, 1983.
Sp r ial Nue' ear Material to.be Covered A.
Nuclear Fuel Special nuclear material will be contained in a number of fuel assemblies which shall not exceed one hundred and ninety-six (196) although only one hundred and ninety-three (193) fuel assemblies are required to load the reactor core of Catawba Unit 1.
A license authorizing 196 assemblies is requested to allow for contingencies.
- Duke Power Company is submitting this application for itself and as agent for the North Carolina Municipal Power Agency Number 1, the North Carolina Electric Membership Corporation and the Saluda River Electric Cooperative, Inc. 8306020646 830517 PDR ADOCK 070*****
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-> The Cata\\ba Unit 1 initial core iis composed of the following:
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V 65' fuel assemblies containing 1.6 wt. % U-235 which is 7.4 kilograms s
.of U-235 per fuel assembly.
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- 64 fuel assemblies-containing 3.1 wt. % U-235 which is 14.3
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q kiTograms of -U-235 per fuel assembly.
q 3 fu assemblies containing 3.1 wt. % (or less) U-235,which is 14.3
^ Ailograms-(or less) of U-235 per fuel assembly (conti2gincy).
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w-I Thelcotal h ight of normal uranium is-4,61.5 kilograms (U)sper fuel assembly. 'The total weight of a fuel assembly is 665.4 kilograms. The-
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fuel-assemb'iescontainnoU-233,Pu,(epleteduranium,orstborium.
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A fuel assembifs is described in the Final Safety Analysis Repo{t (FSAR) s for the Catawba Nuclear Station, Section 4.2.
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Special nuclear material will be contained in several fission chambers (movable incore neutron detectors). Each movable detector contains U 0s 3
enriched to more than 90 percent U-235.
The total quantity of U-235 in each detector is less than five (5) milligrams. There willQe a total of six (6) operational detectors and possibly as many as six (6) spare detectors on site at any one time.; 'Therefore, there could be as many as 60 milligrams of U-235 (in dettetors) on site at any one time.
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Pu-Be Neutros Source-s
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A Pu-Be neutron source will be feaM r~ed which will be,sted,fdr check and s
icalibrationofneutronsurv;yindrumentationprigeichiuelloadingand m
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' for rysponse checking ofsexcore n' clear instrumettadoIh'fter fuel
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Thetypeandamountofn9terialisanominal5cucks l 1.
s Plutonium 238-Berr 1,ium (Pu-Be) neutron source.
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b NOTE: Request is for one (1) Standard 5-Curie source containing
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approximately 0.33 grams of plutonium 238.u Manufacturer
\\?s Monsanto Research Corp. The source model number is 2724-BT.
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The 5 curies Pu-Be' Neutron Source shall pe dot.bly encapsulated to g
insure integrity. Certi,fication of satisfactory leakztesting of the sBurce by the manufacturer shall te reqtdyed ugn receipt of s
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( ' '\\ the sourcif'Upon receipt, monitoring of the-sottree, and container w i, w 3
S.'(Package s 11,be performe tsing appropriagype 4psts and dose rate
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The source shall be shipped and stored in an appropriate container,
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which comes, as part of the purchase agreement with the manufacturer.
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As ' stated above, this source shall be used for check and calibration r
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of health physics portable neutron survey instrumentation, at-the e
Catawba Nuclear Station prior to initial fuel loading and for response
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%Ny checking of excore nuclear instrumentation after fuel loading.
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M The source shall be utilized inside and outside the health physics
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" instrument calibration facility. To accomplish the calibration, the ph source:shall be removed from its storage container and placed in 7y position for-calibration by means of an attached handling and v
extension rod. The source shall be controlled by procedure when it 9m
'is utilized outside the health physics calibration facility.
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. The Radiation Protection Provisions are described in FSAR Section 12.
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Plutonium Aloha Source p
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- A small Pu-239 source set will be used for check and calibration of alpha j* gy
. emitting alpha radiation that spans the ranges of the portable instru-survey instrumentation. The source set contains four small sources
%~.a, ments. The sources are 1.25 inches in diameter with an active diameter f%
. qx 3 3pf 1 inch and mounted on nickel alloy with a total activity of 0.5 micro-g 11 curies. <The set is manufactured by Eberline Instrument Corporation.
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Facilities and Enuinment A.
Storage Facilities Constructikof the Catawba Unit 1 Fuel Storage Facilities has been completed,.-and preoperational testing of the handling equipment shall be t
complete prior to receipt of new fuel. Detailed information regarding
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the fuel storage facility structures, systems, components, and design x
bases is provided in the Catawba FSAR Sections 9.1 and 3.8.4.
A general arrangement ~ drawing of the Catawba Unit 1 fuel building is shown on l m,. x
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ly prio{r to being_ removed from the shipping container.
New uel is received in the fuel receiving area and' stored temporarfly V
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the jhipping container, the assemblies are handled one at a time by the suxi,11ary hoist'~~of the fuel handling bridge crane. Upon removal from the d%
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hiippingcontainerthe-assembly.isplacedinthenewfuelinspection-Q" Q
y, hatch *or is transported by the auxiliary hoist to the New Fuel Storage C,
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-l' Building vr,the' spent fuel pool for storage. Fuel is transferred from b "4 the Stogage Building'tg the spent fuel pool in a similar manner.
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New Fuel Storage Facility 4
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- '(he design of the new fuel facility is based on the following
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. Gene a'l Design Criterion 2 - Design for. protection against
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General Design Criterion 3 - Fire Protection;
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General Design Criterion 4 - Environmental and missile design bases;
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General Design Criterion 5 - Sharing of structures, systems and components; General Design Criteri$n el ' Fuel storage, handling, and radioactivejcontrol; y
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Capab/,lityrof periodic inspection; a.
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Shielding for radiation protection;
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Provisions for containment and confinement;
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General DesigVCriterion 62
, Prevention of criticality in fuel storage and handling;.
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ANSI N18.2-1973,"Section 5.7.4.-1; J
"The design of spent tuel storage racks and transfer equipment shall be such,that the effective multiplication factor shall not exceed 0.95 with new fuel of the highest anticipated enrichment in place assuming flooding with pure water. The denign of normally dry new fuel storage racks shall be such that the effective mutliplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place assuming optimum moderation (e.g., a uniform density aqueous foam envelope as the result of fire fighting). Credit may be taken for the inherent neutron absorbing effect of materials of construction or, if the requirements of Criterion 5.7.5.10 are met, for added nuclear poisons."
Each unit at Catawba Nuclear Station will have new fuel storage racks located within a Neu Fuel Storage Building. Both the New Fuel Storage Building and the new fuel storage racks are designed as Category 1 structures. The new fuel storage racks are arranged to provide dry storage for 96 new fuel assemblies. The racks consist of vertical cells grouped in parallel rows, six rows wide and 16 j-
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cells'long, which provide-support for the' fuel assemblies and a
,rf minimum center-to-center distance of 21 inches between assemblies.
Y The new fuel storage. racks are constructed of structural steel shapes and plates with guides which provide for ea'sy entry of the 9h', i
. assemblies into:the racks. The racks minimize the horizontal dis-
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placement of the fuel assemblies.during storage.
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- Calculated valties of X,[f _ for the storage arrays, including thy / ' Z 7s
.f effects of cal'eulationa and geometrical uncertainties, are lens than those required by ANSI N18.2-1973, Section 5.7.4.1 when a full 47 loading of the' assemblies is considered.
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The computer codes and techniques utilized in the analysis have been validated against experimental data for water moderated UOz lattices I
with characteristics similar to the fuel analyzed.
,,7 The following assumptions are made in evaluating criticality safety:
Parameters are chosen to maximize K,ff.
Under postulated conditions of complete flooding by unborated water, the array is treated as an infinite array of infinitely long assemblies.
Under postulated conditions of complete envelopment by aqueous g
foam, the irray is assumed to contain an infinite number of assemblies and a range of foam densities is examined to ensure that the maximum reactivity is established.
No burnable poisons, control rods or supplemental neutron poisons are assumed to be present.
All assemblies are assumed to be 3.5 w/o enrichment U-235 and unirradiated.
Effects of reflectors other than water are included if their neglect would have been nonconservative.
t The interaction between array members is included.
The following accidents are considered in the criticality design of the new fuel storage area:
Flooding: complete immersion of the entire array in pure, unborated, room temperature water.
Envelopment of the entire array in a uniform density aqueous l
foam of optimum density (that density which maximizes the
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reactivity of the array), for example as a result of fire fighting.
Accidents resulting in an increase in K,ff because of geometrical changes of the racks or fuel handling accidents are not considered credible due to the following design bases:
The facility is designed in accordance with GDC 2 and 4.
The racks are designed to Seismic Category I requirements.
The racks and anchorages can withstand the maximum uplift force available without a signficiant change in geometry, i
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E The design of the Fuel Handling System and administrative procedures insure suberitical spacing of fuel assemblies. Only one fuel assembly wi."'
be handled at a time within each of the
. four areas specified on Figure 1.
The introduction and retention of moderators into the facility is prevented by the following:
The storage of the new fuel within the New Fuel Storage Building precludes flooding of the new fuel assemblies by the probable maximum flood (PMF) and the probable maximum precipitation (PMP).
There is no piping routed through this area, the rupture of which could introduce moderator into the storage building.
There exists a drainage system sized to handle all probable means of inadvertent flooding.
Administrative policy has been established to preclude the use of hydrogeneous fire fighting material into the storage building. All extinguishers are of the dry chemical or CO2 type.
Two gamma radiation monitors, which alarm both locally and in the control room area upon the detection of preset radiation levels, will be located in the New Fuel Storage Building. These gamma radiation monitors are self testing. Should either unit become inoperable, indication of the failed state is provided in the Control Room.
There are normally no combustible materials present in the storage buildin2 A manual fire fighting system of appropriate capacity and capability is present should combustible material be inadvertently introduced.
c Ventilation for this facility will be provided by the fuel handling area ventilation supply and the fuel handling area ventilation ex-haust subsystems.
2.
Spent Fuel Storage Facility The design bases of the spent fuel storage facility are the following:
The prevention of criticality during storage.
The prevention of damage of the fuel, f
Adequate radiation shielding.
l Protection against radioactivity release.
Adequate monitoring of the fuel storage.
-Ability to withstand design seismic loads.,
The fuel assemblies are held in a vertical position by the spent fuel pool storage racks. The fuel assemblies are supported within the fuel storage racks by a stainless steel plate located approxi-mately six inches above the fuel pool floor.
The individual storage cells are constructed using stainless steel plates. The four sides of the cell are 1/4 inch stainless steel plate. A 1/4 inch plate, with a 5 1/2 inch diameter hole in the center is welded to the sides of the cell for support of the fuel assembly. The support plate is located approximately 6 inches above the pool floor. A lead-in assembly is provided at the top of each rack for ease and assurance of insertion of the fuel assembly.
For each unit, there are provisions for 1418 fuel assembly storage spaces for spent fuel.
Administrative procedures require that new fuel assemblies stored dry in the spent fuel pool be stored in a checkerboard array such that, for a given assembly, the four diagonally adjacent locations contain an assembly, and the four immediately adjacent storage locations are vacant. This checkerboard array insures that no two fuel assemblies will be el,ser than the 21 inch center-to-center spacing of the New Fuel Storage Building. The exact positioning of each fuel assembly is speelfied by the Reactor Engineer prior to storage of the new fuel assemblies in the spent fuel pool. The supervisor in charge of new fuel storage assures that each assembly is properly positioned and records where each assembly is stored. The Reactor Engineer reviews the supervisor's records to verify that each assembly was properly positioned.
The use of hydrogenous fire fighting foam in this fuel storage area 3 excluded. However, two hose stations (two hoses) are provided for fighting fires which could occur in the fuel receiving area.
Dri chemical or CO fire extinguishers are also provided for this 2
purpose, and their use is encouraged while the hose stations are used as a backup. These hose stations are located on the operating deck level of the spent fuel pool and are capable of providing a minimum of 100 gpm of water at a pressure of 65 psig each. The maximum pressure available to these stations is 100 psig. Admini-strative controls preclude the fire fighting crews from using these hose stations to spray water into the spent fuel pool or from spray-ing the fuel receiving areas if new fuel is being transferred from the shipping containe to its storage location.
In addition, the spent fuel pool racks are designed so that water disperses through all the racks and cannot accummulate in any given single rack.
Under the above restriction, K values were calculated for storage ff array, including the effects oI calculational and geometric uncertainties. These values are less than those required by ANSI N18.2-1973, Section 5.7.4.1 when an infinite array of the assemblies is considered. The computer codes and techniques utilized in the analysis have been validated against experimental data for water moderated UO lattices with characteristics similar to the fuel 2
analyzed. --
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- The following assumptions are made in evaluating optimum moderation criticality safety:
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Parameters are chosen to maximize K,ff.
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The array is assumed to be infinite in all directions, and a range of foam densities is examined to ensure that the maximum reactivity is established.
No burnable poisons, control rods, or supplemental neutron poisons are assumed to be present.
All assemblies are assumed to be 3.5 w/o U-235 enrichment and unirradiated..
i Credit is taken for the inherent neutron absorbing effect of i
some of the rack structure in accordance with Section 5.7.4.1 of ANSI N18.2-1973.
Geometric uncertainties and misalignment of the assemblies in the rack are assumed to occur in the worst possible combinations.
A 16 group cross-section data set was obtained from the standard KENO library tape. Most of the microscopic cross-section' data is of the j
Hanson and Roach 16 x 16 cross-section data for several elements.
s The cross sections for U-235 and U-238 were corrected for self shielding. Utilizing this 16 group cross-section set the Monte Carlo code KENO is then used to perform the series of criticality calculations on detailed representations of assembly arrays.
Results of the analysis indicate that an infinite checkerboard array of assemblies,.under the criteria stated above, would have a value of K less than 0.98.
Separate analyses were performed to compare i
resuIksofthisoptimummoderationanalysismethodwiththose f
obtained by reactor vendors.
The heaviest objects which could possibly be moved over the spent -
fuel pool racks are the two spent fuel pool weir gates.. However, j_
these weir gates will not be moved from their storage locations while the spent fuel pool is being used to store new fuel.
Therefore, the heaviest object that may be moved'over the new fuel assemblies stored in the spent fuel pool is a fuel assembly. The spent fuel pool racks are designed to protect stored fuel assemblies from damage resulting from a dropped fuel assembly.
3.
Temporary Storage in Shippina containers The new fuel will be temporarily ~ stored in shipping containers in the fuel receiving area before being removed rud-placed in either the New Fuel Storage BuildingLor spent fuel storage pool. 'The Westinghouse shipping container is a reusable metal container designed for shock and vibration isolation, humidity control and i
leak tightness to -protect fuel assemblies from damage ;during normal i.
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handling and shipping at temperatures from -40*F to +150*F.
Each container may contain one or two fuel assemblies with or without core components. No more than twelve shipping containers are expected to be on site at any one time.
The container lateral clearance dimensions are 47" high by 45" wide.
A 191" long shipping container is used to ship the 12' fuel assemblies and weighs 6100 pounds fully loaded.
4.
Removal From Storate Fuel assemblies may be removed from storage for either of these two reasons:
Inspection - This involves a visual or otherwise nondestructive examination of the fuel assembly to determine its acceptability before exposure in the reactor core.
i Precharacterization - This involves measurements and examination to determine the pre-exposure characteristics of a given fuel assembly.
Fuel assembly movements while being removed from or replaced in approved storage, shall be controlled by approved operating procedures.
The ability to remove one fuel assembly from storage is necessary in order to perform the tasks of inspection and/or precharacter-ization. The design of the fuel handling system and the proper control of these movements by approved operating procedures, shall preclude the possibility of positioning two or more fuel elements within the limits previously analyzed in Section 2.2.3 of Regulatory Guide 3.15.
One single element cannot achieve criticality under optimum conditions of spacing, moderation and reflection.
B.
Health Physics and Chemistry Facilities The qualifications of those personnel responsible for the control of Special Nuclear Material at Catawba are contained in Attachments 2, 3, and 4.
The Health Physics and Chemistry facilities are centrally located in the Auxiliary Building for efficiency of operation. Laboratory facilities consist of a conventional chemistry laboratory, a radio-chemistry laboratory, and a shielded radiation survey instrument calibration room.
These facilities-are equipped for conducting the health physics and-chemistry programs for the station, for detecting, a slyzing and measuring all types of raciation and for evaluating any radiological
. problem that may reasonably be expected. Equipment for preparing environmental radioactivity samples and for radio-bioassay is also in-cluded.
In addition, a' Health Physics operations office is provided in this location. Measurements for internal personnel dosimetry purposes are performed in-the Administration Building. _ _ _ _-___________
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Should the Health Physics facilities not be totally functional upon receipt of the materials covered under this license temporary provisions will be made to ensure that adequate capabilities are provided. Facilities would be organized within: (1) the Health Physics office areas in the Auxiliary Service Butiding, (2) the Temporary Administration Building (for the Health Physics Body Burden Counting Facility), and/or (3) in those station areas that have been completed, prior to receipt of materials.
C.
Additional Facilities and Access Provisions Change room facilities are provided where personnel obtain clean protective clothing and other equipment required for station work. The change rocas service the reactor buildings, the Auxiliary Building, the spent fuel pools, and the Hot Machine Shop. These change room facilities are divided into clean and contaminated sections. The contaminated section of the change rooms is used for the removal and handling of contaminated protective clothing after use.
Provisions for change and personnel decontamination are also available in the first aid room in the Radiation Control Area. Showers, sinks, and necessary radiation moni-toring equipment are provided in all of the change rooms to aid the decontamination of personnel.
Equipment decontamination facilities are also provided at the station for large and small items of station equipment, components and tools. There is one such facility for the station, and in addition, a cask decontamin-ation area is provided adjacent to each spent fuel pool. A decontamin-ation laundry is also provided.
Decontamination of work areas throughout the station is facilitated by the provision of janitor's sinks on each floor level in the Auxiliary Bailding and in the reactor containments.
Should the permanent facilities not be totally functional upon receipt of materials covered under this license, temporary provisions will be made which adequately provide for protective clothing, personnel decontamin-ation and equipment decontamination.
Drains from all these facilities go to appropriate radioactive liquid waste drain tanks. Written procedures govern the proper use of protective clothing, the change rooms, and the decontamination facilities.
In order to protect personnel from radiation and radioactive materials, the Restricted Area of the station is divided into areas of increasingly controlled access depending on radiation levels. Protection of personnel from access to radiation areas that exist temporarily or permanently ac a result of station operations and maintenance is by means of appropriate radiation warning signs, barricades, locked doors, audible and visual i
indicators and alarms, as required by 10CFR20. Administrative controls are also used in conjunction with the above and keys are issued to l
authorized station personnel for access to the Radiation Control' Area of the plant and to limited access areas within the Radiation Control Area ut. der certain conditions.
A contamination control checkpoint that is equipped with appropriate j
monitoring instrumentation is located at the main access point. Addi-tional exits with appropriate monitoring instrumentation are provided to support work in the Aexiliary Service Building. All other personnel-access points into the Radiation Control Area in the Auxiliary Building are protected by restricted-in/ free-out doors, and are for emergency exits only. Stairs located on the north, south, east, and west sides of the Auxiliary Building are provided for personnel access from one elev-ation to another. Contamination control checkpoints are appropriately-located at the stairwell locations. Additional checkpoints are strateg-ically placed throughout the Radiation Control Area, to prevent the spread of contamination within the area.
A Radiation Work Permit system is also utilized to control access to the Radiation Control Area, radiation areas and high radiation areas.
Before leaving the Radiation Control Area, personnel are required to monitor themselves with personnel friskers (thin window G-M detectors (ccunt rate meters,) positioned near each exit door), to make sure that they are free of significant contamination.
Authorized personnel enter the Radiation Control Area through key-operated doors, generally on the Turbine Building-Service Building-Control Room-side of the station, and leave through these doors (after monitoring themselves with the personnel friskers) prior to exiting the Radiation Control Area.
Personnel who are required to utilize protective clothing obtain these items in the Change Rooms. They first enter the Change Room on the
" clean" side, consult the Radiation Work Permit, remove street clothes, don the required protective clothing, and then proceed to the job location.
After completing work, they remove outer contaminated protective clothing, at the exit of the Radiation Control Zone previously established for the work area. They then proceed to the " contaminated" side of the Change Room, where they remove any remaining protectiva clothing items, wash and shower as necessary, monitor themselves; and then proceed to the " clean" side, where they put on their personal clothing and leave.
All persons entering the Restricted Area of the station shall wear the personnel monitoring equipment (TLD, pocket dosimeters, etc.) prescribed by the Station Health Physicist in accordance with NRC Regulations and shall comply with applicable Radiation Work Permits.
All work on systems or in locations where radioactive contamination or external radiation is present requires a specific Radiation Work Permit prepared under the direction of the Station Health Physicist before work can begin. The radiological hazards associated with the job are deter-mined and evaluated prior to issuing the permit. The Radiation Work Permit lists the precautions to be taken including radiation levels (for external and internal exposure) protective clothing to be worn and any radiation monitoring that may be required during the performance of the work. The permit is issued to the people who perform the work; a copy is available for the Shift Supervisor and a working copy is main-tained by personnel in the Health Physics Section.
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-All persons working under a permit are required to read the instructions on the permit and to fill out the information necessary on their Daily Exposure Time Record Card before and after entering the Radiation Control Zone.. The information from the permit and the card is entered into the
. Radiation Exposure Control and Job Exposure Control computer programs and serves, in part, as a personnel monitoring record for the individuals involved.
The counting room is shielded on all sides to facilitate low level counting work. The instrument calibration room also has shield walls.
In addition, extensive shielding of components has been utilized in the Auxiliary Building for the protection of personnel, both for routine operation and for maintenance.
D.
Protective Clothina and Respiratory Protective Equipment Special " protective" and " anti-contamination" clothing is furnished and worn as necessary to protect personnel against contact with radioactive contamination. This consists of coveralls, lab coats, surgeon caps, hoods, gloves, and shoe covers. The change room is conveniently located within the station for proper utilization of this protective clothing.
Approved respiratory protective equipment is also available to supplement process containment and ventilation controls, for the protectic a of per-sonnel against airborne radioactive contamination and the possibility of internal radiation exposure. This equipment consists of full-face air-purifying respirators and self-contained breathing apparatus. Also, a 4
breathing air system has been installed in the station, and esspiratory protective equipment consisting of air-line full-face respirators, hoods, and plastic suits are available, should their use become necessary or desirable.
Maintenance of the above equipment is in accordance with the manufac-turer's recommendations and rules of good practice, such as those published by the American Industrial Hygiene Association in its
" Respiratory Protective Devices Manual." The use and maintenance of protective clothing and respiratory protective equipment is under the direct control of Health Physics Section and personnel are trained in the use of this equipment before using it in the performance of their work.
-The use of this equipment will be in accordance with the Technical Specifications.
1 E.
Portable and Laboratory Equipment Different types of instruments are selected-to cover the entire spectrum i
of radiation measurement requirements expected. This includes instru-ments for detecting and measuring alpha, beta, gamma and neturon radiation. These consist of' Counting Room and portable radiation survey /
l monitoring instruments. These instruments are required to provide-protection against radiation for station personnel-(for surveys required by 10CFR20.201); to control the release of effluents for the protection of the health and safety of the public; and to provid. for all other radiological measurements necessary for personnel and public safety and
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for the protection of property. Sufficient. quantities are available to i
allow for use,-calibration, maintenance and repair.
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Counting Room instruments for radioactivity measurements include the following:
Computer-based multi-channel gamma analyzer with multiple Ge(Li) detectors, used for identification and measurement of gamma emitting radionuclides in samples of reactor primary coolant, liquid and gaseous waste, airborne contaminants and similar samples.
Automatic and manual beta-gamma counter-scales used primarily for gross beta measurements of surface contamination on swipes.
Alpha counter-scaler used for gross alpha measurements such as uranium or plutonium in reactor primary coolant samples or alpha contamination from surface or air samples.
Dual channel liquid scintillation counter used for measurement of tritium in reactor primary coolant, liquid ahd gaseous wastes, and for gross measurement of beta activity other than tritium.
Shielded body-burden and thyroid-burden analyzer used for measure-ment of possible internally deposited radio-isotope for determina-tion of internal dose of personnel.
Note: Should any of this equipment not be totally functional upon receipt of material covered under this license, provisions will be made to provide similar counting capabilities onsite or to transport samples to the McGuire Nuclear Station for processing on equipment of a similar nature.
Portable radiation survey and monitoring instruments for routine use are selected to cover the entire range from background to high levels for the radiation types of concern. These include (with nominal range character-istics as indicated):
Beta gamma survey meters (Geiger counters, 0-100 mR/hr) used for detection of radioactive contamination on surfaces and for low level dose rate measurements.
Low and high range beta gamma ionization chamber survey meters (0 mR/hr - 1000 R/hr) used to cover the general range of dose rate measurements necessary for radiation protection purposes.
Neutron rem dosimeter instruments (0 mrem /hr - 5 rem /hr) used to measure the sum of thermal, intermediate and fast neutron dose rates for radiation protection purposes.
Alpha scintillation counters (0 - 1,000,000 dpm/100 cm, 30 percent 2
efficiency) used for measurement of alpha contamination on various surfaces that may result from any uranium or plutonium in the reactor primary coolant, for example.
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The Process Monitoring System is relied-upon for continuous monitoring of airborne radioactivity. This is supplemented by grab air samples collected and analyzed by Health Physics during maintenance and routine and abnormal operations where airborne radioactivity may be evolved.
Airborne gaseous, particulate, and iodine samplers are also available for routine use as well as an assortment of special purpose and emergency type radiation survey instruments including bubblers for tritium, gas sample containers, low volume air samplers, particulate filters, and activated charcoal and silver zeolite cartridges. All of this equipment is kept in areas under Health Physics control. Necessary emergency instruments are also located onsite and at a remote assembly point.
In addition to the portable radiation monitoring instruments, fixed beta gamma count rate meters are located at exits from the Radiation Control Area. These instruments are intended to prevent any contamination on personnel, materials or equipment from being spread to the unrestricted secondary systems areas of the station. Appropriate monitoring instruments are also available at various locations within the Radiation Control Area for contamination control purposes.
Portal monitors, (GM thin side windows, ~30 percent efficiency, 150-7000 cpm) are also utilized, as appropriate, to monitor personnel leaving the station.
All of the above instruments are subjected to initial operational checks and calibration and to a continuing quality control program to assure the accuracy of all measurements of radioactivity and radiation levels.
These instruments are recalibrated with standards whenever their operation appears statistically to be out of the accepted limits.
In addition, routine calibrations are performed periodically on all of this equipment and after all repairs. These-periodic calibrations are per-formed at least quarterly with the exception of calibrations of the GeLi detectors which are performed at least semi-annually. A shielded calibration range capable of exposure rates fro. essentially background to hundreds of R/Hr is used for calibration of 3ation monitoring instruments. Also available is a small (mci lev;.) source for certain low level calibrations and a nominal 5 Ci Pu-Be neutron source for neutron instrument calibration. The gamma sources are calibrated with an exposure rate meter traceable to the National Bureau of Standards.
The body burden and thyroid analyzers are calibrated using phantoms and solution standards of the radionuclides of concern (C060, COss, Cs184, Cs137, and I 131).
These detectors are used in conjunction with a computer-based multi-channel gamma analyzer and associated readout to obtain a permanent record.
Record of all calibrations are kept and persornel dosimetry, survey and monitoring records, etc., are maintained as required by NRC regulations.
F.
Personnel Dosimetry Equipment Personnel monitoring equipment consisting of thermoluminescent dosimeters (TLD's) and self-reading pocket dosimeters are assigned by the Station Health Physics Section and worn by all personnel (employees and visitors) l ! -
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personnel dosimetry equipment such as high range self-reading pocket
. dosimeters and extremity TLD's may be assigned as needed depending on the. radiological conditions encountered.
t Personnel whose jobs require them to frequently _ enter the' Restricted Area of the station may ordinarily be assigned a permanent personnel monitoring
' badge and a dosimeter. Whereas personnel working under a specific Radia-tion Work Permit in a job situation where a sizeable fraction of the quarterly. allowable dose may be received in a relatively short period of time may additionally be assigned a high range self-reading. dosimeter and/
1-or extremity monitoring equipment, depending on job conditions. Extremity monitoring equipment is used for jobs or situtations where extremity i
dose is expected to be limiting or controlling or in excess of the whole body dose. The use of additional personnel monitoring equipment beyond that routinely used depends on the job and on existing radiological conditions as evaluated and determined by Station Health Physics personnel.
Records of Radiation exposure history and current occupational exposure 4
are maintained by the' Health Physics:Section for each individual for whom r
~ personnel monitoring is required. The external radiation dose to
. personnel is determined on a daily basis by means of self-reading pocket dosimeters. Personnel monitoring badges (TLD's) are processed at least monthly. If necessary, they may be processed more frequently.
A body. burden analyzer system for routine screening of personnel to 4
determine internal exposure is available on-site.
Outside services for radio-bioassay and whole body counting may be used as required for backup and support of the program. The station equipment is sufficiently sensitive to detect in the thyroid, lungs or whole body a small fraction of the permissible body burden for those gamma emitting radionuclides expected.
Body burden analyses are performed as soon as practicable on individuals who have been newly-assigned a visitor or permanent Health Physics badge i
or who are, terminating employment or assignment with Duke Power Company.
In addition all permanently badged individuals _are required to receive'at least one body-burden analysis per year.
1 Anyone onsite, whether badged or not, who was involved in a radiological-accident where internal exposure was likely, would be given a body-burden scan as soon as practicable. If radioactive material uptake had occurred, appropriate action would be taken.
. Personnel monitoring. badges (TLD) are supplied by a central in-house
. service which is responsible for the calibration and maintenance of all TLD and TLD resdout equipment. Self-reading pocket dosimeters are calibrated and leak tested at the station as part=of the station instrument maintenance Quality Assurance program.
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Health Physics Proeram A.
Program Objectives The three basic objectives of the Health Physics Program at Catawba are to:
Protect station personnel Protect the public Protect the station Protection of personnel means surveillance and control over the internal and external radiation exposure of personnel and maintaining the exposure of all personnel within permissible limits, and as low as reasonably achievable in compliance with applicable regulations and license conditions.
Protection of the public means surveillance and control over all station conditions and operations that may affect the health and safety of the public.
It includes such activities as radioactive gaseous, liquid and solid waste disposal and the shipment of radioactive materials. It also involves conducting an environmental radioactivity monitoring program and maintaining an effective emergency plan.
Protection of the station means the continuous determination and evalua-tion of the radiological status of the station for operational safety and radiation exposure control purposes. This work is done in order to warn of possible detrimental changes and exposure hazards, to oetermine changes or improvements needed, and to note trends for planning future maintenance work.
The program organization is as follows:
The Station Manager is responsible for the protection of all persons against radiation and for compliance with NRC regulations and license conditions. This responsibility is in turn shared by all supervisors.
Furthermore, all personnel are required to work safely and to follow the regulations, rules, and procedures that have been established for their protection.
The Duke Power Company System Health Physicist establishes the Health Physics Program for Catawba that is designed to assure compliance with applicable regulations, licenses and reguiatory guides. He also provides technical guidance for conducting this program, audits the effectiveness and the result of the program and modifies it as required. He also provides technical assistance to tDe Vice President, Nuclear Production Department, who has management aut*3rity to implement the "as low as reasonably achievable", (ALARA), occupational exposure policy, to which Duke Power Company is committed.
r-The Station Health Physicist is responsible for conducting the Health
-Physics Program that has been established for the station. The Station i
Health Physicist has the duty and the authority to measure and control 1
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the radiation exposure of personnel to a level that is as low as reason-ably achievable and within regulatory exposure limits; to continuously evaluate and review the radiological status of the station; to make recommendations for control or elimination of radiation hazards; to train personnel in radiation safety; to assist all personnel in carrying out their radiation protection responsibilities; and to protect the health and safety of the public both on-site and in the surrounding area.
In order to achieve the goals of the Health Physics Program and fulfill these responsibilities for radiation protection, radiation monitoring, survey and personnel exposure control work are performed on a continuous basis for.all station operations and maintenance. This requires a Health Physics Technician on each operating shift. The extent of this surveillance is outlined below.
The Health Physics section performs the major portion of the Health Physics work for the station. Personnel in the Health Physics section normally work on the day shift, seven days a week, during periods of routine operation; and deploy onto the other shifts for major maintenance, shutdown and refueling work. The Health Physics Section is organized into three or more major units, each headed by a Health Physics Coordinator. These units are: (1) Staff, (2) Surveillance and Control, and (3) Support Functions.
For the purpose of defining and assigning work to be performed by the operating shifts and the Health Physics Sections, the routine station radiation surveillance work can be described as consisting of radiation monitoring, radiation survey, radiation exposure control and radioactive waste disposal activities.
The Health Physics Technicians on each shift perform radiation monitoring and exposure control work for the routine shift operations, particuarly on the back shifts (other than day shifts). This work is performed under the direction of the appropriate Health Physics Supervisor. A " Shift Health Physics Guide," prepared by appropriate Health Physics Supervisor, will designate routine work to be performed.
The Health Physic = Section also performs essentially all of the work necessary to calibrate and maintain (other than repair) the Counting Room instruments and the portable radiation monitoring instruments.
Duties concerning radioactive liquid, gaseous and solid waste disposal are performed under Health Physics direction. The detailed analyses and records required to characterize the nature of these releases, both qual-itatively and quantitatively, are under the control of Health Physics.
In-addition, solid waste disposal and shipments of radioactive materials are under the control of Health Physics.
Training and qualification of personnel in Health Physics is the responsi-bility of the Station Health Physicist and is performed under his direction.
All administrative aspects of training, such as scheduling and documentation are handled by Administrative services. Administrative services also administers the general standardized Health Physics Training.,
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The Health Physics Section also conducts the Offsite Radiological Monitoring Program for the station.
B.
Inspection of New Fuel Although no contamination of the new fuel assemblies is expected, upon receipt, each assembly will be swiped with disc swipes and counted for alpha and beta-gamma activity prior to storage, for the detection of any residual contamination.
In the.unlikely event loose contamination greater than 20 dpe/100 cua alpha ~and or 1000 dpm/100 cua beta-gamma above background is detected the person (s) monitoring the shipments will contact the Station Health Physicist or his designee who will then supervise contamination control.
The Station Health. Physicist or his designee will take the appropriate action to assure that the contaminated item is controlled.
Transfer of Special Nuclear Material The Westinghouse Electric Corporation is responsible for the shipment of new fuel to the Catawba Nuclear Station.
Should the need arise for Catawba Nuclear Station to package and transport new fuel, the station would ship the new fuel in Westinghouse Electric Corporation owned new fuel shipping containers. This would be in accordance with the provisions of 10CFR Part 71 and DOT regulations.
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The Security Plan and Contingency Plan associated with storage of new
-fuel in the Unic I spent fuel pool has been submitted to the NRC for review under separate letter dated May 10, 1983.
Emernency Plans The emergency planning effort is governed by the provisions of the Station Emergency Plan, other off-site agency plans and station implement-ing procedures.
' Financial Protectiqn Duke Power company will be obtaining 560 million dollars of Nuclear Liability Insurance for Catawba Nuclear Station Units l 'and 2.
This insurance coverage will satisfy the requirements specified in the Price-Anderson Act for licensed nuclear power plants.
InEEEE112R&
-The applicant requests exemption from the monitoring and emergency procedures requirements of 10 CFR 70.24.
This exemption is requested
.because the nature of the special nuclear material storage arrangements and procedural controls which the applicant proposes to employ, precludes any possibility of accidental criticality during receipt, unloading,
. inspection, storage or packaging of the new fuel assemblies. Mm
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ATTACHMENT 2 RESUME OF TRAINING AND EXPERIENCE LIONEL LEWIS, Systeu Health Physicist, has 30 years of practical Health Physics experience.
Certified in Health Physics by the American Board of Health Physics.1961 EDUCATION 1.
Bachelor of Arts (BA) in Pre-Medical Sciences, University of Vermont, 1949.
2.
Atomic Energy Commission Fellowship in Radiological Physics, University of Rochester and Brookhaven National Laboratory, 1952-53.
3.
Master of Science (MS) in Biophysics (Radiological), University of Rochester, 1955. Degree awarded after completion of thesis project on gamma and thermal neutron dosimetry of Brookhaven Reactor Thermal Column.
Miscellaneous:
1.
Graduate work in Limnology, Colorado State University, 1951-52.
2.
Course in Nuclear Radiation Detection for Medical Health Scientists, University of Michigan, 1968.
EXPERIENCE 1.
System Health Physicist - Duke Power Company, June 1967 to date.
Responsibilities include establishing and directing the Health Physics and Environmental Radioactivity Monitoring Programs for all company nuclear power stations, and participating in all other company nuclear activities as Radiation Safety Officer, advisor, and coordinator. Also participate in design, construction, licensing, and operational aspects of nuclear power stations.
Member of Nuclear Safety Review Board 1
2.
Health Physicist and Safety Coordinator - January 1961 to June 1967 Plant Superintendent - November 1963 to December 1964, Carolinas-Virginia Nuclear Power Associates, Inc., Carolinas-Virginia Tube Reactor (CVTR) Facility, Parr, South Carolina.
As Health Physicist, established and directed the health and safety program at the CVTR (a heavy water power-demonstration reactor) which consisted of health physics, general industrial safety, industrial hygiene, and fire-prevention.
Participated in design, construction, checkout, and operation of the CVTR. Duties included preparing testimony and testifying at facility license hearing and t
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ATTACHMENT 2 - PAGE 2 of 4 EXPERIENCE (cont'd) intervention hearing, assisting in preparation of facility license applications, and preparing By-Product, Special Nuclear Material, a0d Reactor Operator and Senior Operator license-applications. Also pa.ticipated in preparation of Final Hazards Summary Report, Tecnnical Specifications, and amendments to the Technical Specifications.
Served as member and secretary of Reactor Safety Committee; Responsi-bility as Health Physicist also included conventional radiochemistry.
Prepared comprehensive Health Physics Manual. Assisted in prepar-ation of nuclear fuel accountability procedure.
Duties as Safety Coordinator were to integrate reactor-safety activities of all plant personnel and sections within the plant with those of the Reactor Safety Committee and the AEC Division of Compliance, to ensure compliance with all AEC regulations and conditions of licenses, and to see that reactor plant was operated within the limitations of the Technical Specifications.
Duties as Plant Superintendent included, in addition to adminis-trative duties, directing activities of reactor plant Operations, Maintenance, and Chemistry sections of the Plant Operations Group during the one-year period of initial operation at power levels up to full licensed power.
3.
Supervisor of Health Physics, Safety, and Industrial Hygiene - 1957 to December 1960, Nuclear Division and Naval Reactors Division, Combustion Engineering, Inc., Windsor, Connecticut.
Experience includes establishing and directing Health Physics and Safety Programs at power reactor (SIC Prototype Submarine Power Reactor) and two critical facilities: Engineering and Materials Development Laboratories and Reactor Fuel Element Manufacturing Plant. Participated in design, construction, and operation of SIC Prototype Reactor.
Served on Reactor Safety Committee (SIC Prototype). Prepared Radiation Safety Manuals for Nuclear and Naval Reactors Divisions.
Contributed to preparation of U. S. Navy Radiological Safety Manual.
4.
Assistant Supervisor of Health Physics, 1955-57, Nuclear Division, The Martin Company, Baltimore, Maryland.
Experience includes assisting in establishing and directing Health Physics programs at Fuel Element Fabrication Facility and Fuel Element Laboratory.
Participated in licensing of Critical Facility and establishment of Health Physics program. _. _ _ - -
a
o ATTACHMENT 2
,DAGE 3 of 4 EXPERIENCE (cont'd) 5.
Health Physicist (Jr.), 1953-55, Brookhaven National Laboratory, Upton, New York.
Main responsibility was Biology and Medical Departments; however, experience as Health Physicist also includes Research Reactor, Radio-chemistry Laboratory, Radioactive Waste Disposal, Instrumentation Calibration, Personnel and Environmental Monitoring, Nuclear Engineering Facility, Accelerator and Cosmotron, Hot Laboratory, and emergency health physics responsibilities for the entire Brookhaven National Laboratory..
TECHNICAL PAPERS 1.
" Neutron and Gamma-Ray Dosimetry of a Thermal Neutron Irradiation Facility". Proceedings of the First Geneva Conference on the Peaceful Uses of Atomic Energy.
1955.
2.
" Thermal Neutron Equivalence of Whole-Body X-Irradiation".
Radiation Research Journal, Volume 4, Number 1.
1956.
3.
"Preoperational Health Physics Training of Reactor Personnel".
Presented at the Health Physics Society Annual Meeting, Gatlinburg, Tennessee, 1959, and published in Health Physics, Volume 6, Number 3/4.
1961.
4.
" Radiation Design and the Health Physicist". Presented at the annual meeting of the Health Physics Society, Las Vegas, Nevada.
1961.
5.
"The Development and Use of Meteorological Map Overlays for Emergency Purposes at a Power Reactor Facility".
International Symposium on Fission Product Release and Transport under Accident Conditions. Oak Ridge, Tennessee.
1965.
6.
"The Administration of Radiation Safety at the CVTR".
Nuclear Safety. May, 1966.
7.
"Eavironmental Monitoring for Nuclear Power Plants - A Utility Health Physicist's Viewpoint".
Invited paper, panel discussion, Southeastern Electric Exchange. Atlanta, Georgia.
1968.
8.
" Gaseous Wastes in a Nuclear Power Plant: Control, Monitoring, and Release".
Invited paper presented at NC Nuclear Environmental Workshop, sponsored by N. C. State University. Pinehurst, North Carolina.
1970.
9.
"The Terrestrial Radiological Monitoring Program at Duke Power Company's Oconee and McGuire Nuclear Stations". An invited paper, Southern Conference on Environmental Radiation Protection from __
ATTACHMENT 2 - PAGE 4 of 4 TECHNICAL PAPERS (cont'd)
Nuclear Power Plants, Region IV, Environmental Protection Agency.
1971.
10.
" Nuclear Radiation - What about It?" Invited paper presented before the general meeting of the National Board of Boiler and Pressure Vessel Inspectors.
St. Paul, Minnesota. May, 1972.
11.
" Health Physics Computer Programs for a Nuclear Power Plant".
Presented at Health Physics Society Annual Meeting.
Las Vegas, Nevada.
1972.
12.
"The Reactor RS0".
An invited paper for the Workshop on the Changing Responsibilities of the Radiation Safety Officer, at the annual meeting of the Health Physics Society. Miami Beach, Florida.
June, 1973.
1 MISCELLANEOUS INFORMATION 1.
Participated in symposium on Offsite Public Health Aspects of Nuclear Power Plant Operations, sponsored by U. S. Public Health Servic.. Region IV, Atlanta, Georgia.
1967.
2.
Participated in State Environmental Radiation Laboratory Symposium sponsored by Environmental Protection Agency, Region IV and Florida Department of Health, Orlando, Florida.
1972.
3.
Participated as a member of Radiation Emergency Response Committee for proposed Southeast Mutual Radiological Assistance Plan, Region IV, Environmental Protection Agency.
MEMBERSHIP IN TECHNICAL ORGANIZATIONS 1.
Charter Member - Health Physics Society; served on Public Infor-mation Committee, Standards Committee, Program Committee, President-Elect North Carolina Chapter.
2.
Member - American Nuclear Society - Reactor Operations and Environ-mental Sciences Sections.
3.
Member - American Industrial Hygiene Association - served on Radiation Committee.
4.
Completed four-year term on Panel of Examiners, American Board of Health Physics in 1975.
5.
Member - Power Reactor Health Physicists Group - served as Chairman for three successive terms.
6.
Member - National Council on Radiation Protection (NCRP) Scientific Committec 46 on Operational Radiation Safety.-. _ - -
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ATTACHMENT 3 RESUME OF TRAINING AND EXPERIENCE W. Phillip Deal, Station Health Physicist, has ten years of practical Fealth Physics experience in commercial nuclear power.
Education A.
General 1.
Graduated 1962, North Rowan High School, Spencer, N. C.
B.
Technical 1.
9/69 to 3/70, Duke Power Company, Nuclear Operator Training Program Source Use Experience 60Co..9 mci - 3.5 Ci Duke Power Company, Inst. Calib.
Pu-Be 5 Ci - Duke Power Company, Inst. Calib.
137Cs 8uCi - Duke Power Company, Inst. Calib.
Experience A.
Duke Power Company - October 17, 1966 1.
Station Health Physicist - Catawba Nuclear Station - June 1, 1978 to present. Responsible for the overall implementation and adminis-tration of the Station Health Physics Program.
2.
Assistant Health Physicist / Associate Health Physicist - Oconee Nuclear Station - July 1,1977 to May 31, 1978, responsible for Health Physics related training, and special projects development.
3.
Health Physics Supervisor / Assistant Health Physicist _- Oconee Nuclear Station - January 1, 1976 to June 30, 1977, responsible for support-functions section of the Health Physics Unit. Duties consisted of supervision of instrument calibration, dosimetry issue, waste re-lease accountability, maintenance of exposure record, radioactive l
waste shipments, and source inventory / control.
4.
Health Physics Lab Tech /Labman - Oconee Nuclear. Station - May 1973 to December 1976, assigned duties within Support-Functions section as described above, and in addition served as a technician on shift with Operations personnel.
t 5.
Operator - Oconee Nuclear Ocation - April 1969 to May, 1973, assigned duties nuclear pcwer plant operator.
6.
Operator - Buck Steam Station - October 1966 to April 1969, assigned i
duties as coal-fired steam plant operator.
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ATTACHMENT 4 RESUME OF TRAINING AND EXPERIENCE Ronda L. Clemmer, Station Health Physics Coordinator, has 10 years of practical Health Physics experience in commercial nuclear power.
Education:
1963 Holbrook High School Graduate 1964-1965 U. S. Naval Nuclear Power School, Mare Island, California (6 months) 1965 U. S. Naval Nuclear Prototype, Windsor, Connecticut (6 months) 1979 Industrial Radiation Safety - N. C. State (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 1981 Health Physics Retraining (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />)
Source Use Experience:
80Co
.9 mci - 3.5 Ci Duke Power Company, Inst. Calib.
Pu-Be 5 Ci Duke Power Company, Inst. Calib.
137Cs 8uct Duke Power Company, Inst. Calib.
Experience:
1971 - 1973 Duke Power Company, Oconee Nuclear Station - Utility Operator 1973 - 1975 Duke Power Company, Oconee Nuclear Station Shift Health Physics Representative May 1975 Qualified ANSI Technician two years Health Physics experience with Duke Power Company 1975 - 1977 Duke Power Company, Oconee Nuclear Station Health Physics Technician S&C Group May 1977 Qualified ANSI Supervisor four years Health Physics experience with Duke Power Company 1977 - 1979 Duke Power Company Oconee Nuclear Station Health Physics Supervisor 1979 - 1982 Duke Power Company Catawba Nuclear Station Health Physics Supe rviser 1982 - Present Duke Power Company Catawba Nuclear Station Health Physics Coordinator l
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