ML20033H047
| ML20033H047 | |
| Person / Time | |
|---|---|
| Site: | 07000025 |
| Issue date: | 04/15/1990 |
| From: | ROCKWELL INTERNATIONAL CORP. |
| To: | |
| Shared Package | |
| ML20033H046 | List: |
| References | |
| AI-78-10, NUDOCS 9004170198 | |
| Download: ML20033H047 (37) | |
Text
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AI-78-10 i.
DECOMMISSIONING PLAN FOR ROCKWELL INTERNATIONAL HOT LABORATORY LICENSED UNDER
-I SPECIAL NUCLEAR MATERIAL LICENSE SNM-21 I
l Rockwell International-Rocketdyne Division -
6633 Canoga Avenue Canoga Park, CA 91303 April 15,1990 I
PDR A DOC.k 07000025 C.
I Al-78-10 8
I DECOMMISSIONING PLAN FOR ROCKWELL INTERNATIONAL HOT LABORATORY I
LICENSED UNDER SPECIAL NUCLEAR MATERIAL LICENSE SNM-21 I
y Prepared by the Staff ofthe Radiation and Nuclear Safety
'_ g and Nuclear Operations B>
Departments I
i Rockwell International Rocketdyne Division 6633 Canoga Avenue Canoga Park, CA 91303 April 15,1990
Rockotdyne Divisi2n s
Rev
SUMMARY
OF CHANGE APPROVALS AND DATE f
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>.S 4/15/90 Entire document revised.
IAe' Paul M. Sewell Approvals & Date
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W. R. McCurnin
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Division Director 2
FOAM 734-C 1 REV,12 M lil
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CONTENTS
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1.0 G ENERAL INFORM ATION............................................
1-1 2.0 DECOM MISSIONING ACTTVITIES.....................................
2-1 2.1 DECOMMISSIONING OBJECTIVE AND SCOPE......................
2-1 2.2 DECOMMISSIONING DESCRIPTION...............................
2-3 2.2.1 Hot Cell and Decontamination Rooms...........................
2-4
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2.2.2 Other Posted Radiological Support Areas.........................
2-5 2.2.3 Ug uid Waste Syst e m..........................................
2-5 2.2.4 Radioactive Exhaust 2-6 2.2.5 Unpost ed Support Areas.......................................
2-6 2.2.6 Fission G as Thnks............................................
2-7 l
2.3 P R OCE D U R ES...................................................
2-7 2.3.1 Scope......................................................
2-8 2.3.2 Applicable Documents and References...........................
2-8 2.3.3 Equipment and M aterials......................................
2-8 2.3.4 Special Safety Precautions......................................
2-8 2.3.5 Work Instru ctions.............................................
2-8 2.3.6 Waste Volume Estimat es.......................................
2-8 2.3.7 M a nho u r Estima t es...........................................
2-8 2.3.8 Personnel Dose Estimates......................................
2-8 2.4 S C H E D U LE..................................................... 2-9 2.5 OllG ANIZATIONAL RESPONSIBILITIES AND AUTHORITY...........
2-11 2.'i.1 K ey Po si t i o ns................................................
2-11 2.S.2 Educational and Experience Requirements........................
2-13 2.6 T RAI N I N G.......................................................2-13 2.7 CONTRACTOR ASSISTANCE.......................................
2-14 3.0 RADI ATI O N PR OTECTI O N............................................ 3-1 3.1 HISTORICAL INFORM ATION......................................
3-1 3.2 ALARA P O LI C Y.................................................. 3-1 3.3 H E ALTH PH YSICS P ROG RAM.....................................
3-3 3.3.1 Radiation Protection Equipment................................
3-3 3.3.2 Radioactive Exhaust Systems...................................
3-4 3.3.3 Ai r Sa mpli n g................................................
3-5 3.4 CO NTRACTO R P E R S ON N E L....................................... 3-6 3.5 RADIOACTIVE WASTE M AN AGEM ENT............................
3-6 4.0 FIN AL RADI ATION SURVEY PLAN.....................................
4-1 4.1 S U RVEY AR E AS..................................................
4-1 iv
3 I
4.2 S AM PLING INSPECTION..........................................
4-1 4.3 B ACKG RO UND RADI ATIO N.......................................
4-2 l
4.4 S U RVEY IN STRU M ENTS..........................................
4-2 1
DATA AN AIXS ES.................................................
4-4 4.5 4.5.1 Sampling Inspection by Variables................................
4-4 4.5.2 Accepta n ce Crit e ria...........................................
4-5 I
4.5.3 Summary Presenta tion.........................................
4-6 5.0 FU N D I N G...........................................................
5-1 6.0 PHYSICAL SECURITY AND M ATERI AL CONTROL PLANS...............
6-1 TABLES 3.1.
Waste Volume Estimate - Unrestricted Use................................
3-8 FIGURES 2.1.
Layou t o f RI H L Fa cili ty................................................
2-2 2.2.
RIHL Decommissioning Schedule........................................
2-10 l
2.3.
Organiza tion Ch a rt....................................................
2-12 i
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1,0 GENERAL INFORMATION F
A special nuclear materials license was issued to Rockwell International, Rocket-dyne Disision, by the U.S. Nuclear Regulatory Commission (NRC) for operation of the Rockwell International Hot Laboratory (RIHL) and other facilities. In support of that l
license, a decontamination plan, AI-78-10, was submitted to the NRC. This decommis-sioning plan is a revision to that document, which reflects current planning for decommis-sioning the facility and incorporates the format recommended by NRC Regulatory Guide 3.65 for such plans.
Special Nuclear Materials Ucense No. SNM-21 was issued to Rockwell Internation-al, Rocketdyne Division,6633 Canoga Avenue, Canoga Park, CA 91303. It was renewed in June 1984 I
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l 2.1 DECOMMISSIONIN OB E
\\ AhD COP I
The objective of this program is to decontaminate the RIHL facility to levels which I
allow release for unrestricted use, as defined by Special Nuclear Materials Ucense No.
SNM-21, Annex B. The facility consists of four rectangular hot cells, abutted by decon-g tamination rooms. These areas are surrounded by a building structure which provides an 3
operating gallery in front, a senice gallery in the rear, and contiguous operation support, mockup, and administrative offices. The hot cells and decontamination rooms are con-l structed of reinforced and dense concrete. All surfaces of the hot cells and the floors of the decontamination rooms are clad with steel. The layout of the facility is shown in Fig-ure 2.1. In-cell equipment is remotely operated from the operating gallery using manipu.
lators, analytical equipment, and controls. Access to inside the cells is from the senice I
gallery, located at the rear of the hot cells. The decontamination rooms are located be-tween the cells and senice gallery. These rooms are where equipment is decontaminated and packaged. The decontamination rooms also serve as contamination control areas be-tween the cells and the senice gallery. Connected with the senice gallery is a hot manip-ulator repair room for senicing low-level, radioactively contaminated equipment. The facility ventilation system causes air to flow from the radioactively uncontaminated areas j
toward the areas of highest contamination potential, through roughing filters and high-efficiency particulate filters, and finally out a 54-in, exit-dia,73-ft-high stack (above grade). The ventilation ducts are installed in the basement directly below the hot cells.
The following areas of the RIHL facility will require decontamination and/or verifi-cation that radiation levels are below limits for release for unrestricted use and constitute the primary activities of the decommissioning program.
Fission Gas 'Panks Hot Cells and Decontamination Rooms (1,2,3, and 4) e Glove Box Lab (Room 139), Laboratory (Room 141), and Manipulator Main-tenance (Room 128)
Senice Gallery and North Hall (Room 147) e Attic (Above Senice Gallery and Decontamination Rooms) e Support Area Drain System (Rooms 149,147,141,139, and 128) and Build-e ing No. 468 Operating Gallery (Room 117), Mockup Assembly (Rooms 125 and 131) e
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Basement Radioactive Exhaust System e
o Radioactive Exhaust Stack Radioactive Drain System (Hot Cells and Decontamination Rooms) e Offices and Change Rooms (Rooms 100,101,102,103,105,107,108,109, 110,111,112,113, and 149)
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Battery (Room 137), Emergency Generator (Room 135), Heating and Venti-e lating (Room 133), and Air Conditioning (Room 157)
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Hot Storage (Room 153) and Air Lock (Room 155) e Roof
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Asphalt and Concrete (Surrounding Facility) and Loading Dock Health Physics and Engineering 'Itailers (Offices).
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2.2 DECOMMISSIONING DESCRIPTION
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Decontamination will be the primary focus of the RIHL decommissioning. This will be initiated after all special nuclear materials in inventory have been removed from the
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facility. The basic approach will be to decontaminate the areas of highest radiation levels first to minimize personnel exposure. Conventional methods will be used which have proven successful for other facilities.
The following decontamination methods will be utilized. Foam cleaning will be used in conjunction with localized brushing with detergent to remove surface contamination.
Scabbling or high-pressure water will be used to remove contaminated epoxy paint. The steel liners will be grit blasted to remove residual surface contamination. The steel liners will be cut mechanically or with oxy-acetylene or plasma arc torch to remove sections for sampling of the underlying concrete. Any areas of concrete that have contamination will have the surface layer removed by chipping hammers, scabblers, and/or grinders as deemed best. Any deep penetration of contamination may require scarfing, core drilling, or jack hammering to reduce the contamination to allowable limits. In general, methods will be selected to minimize the quantity of waste generated and personnel radiation exposure.
Some degree of radioactive contamination exists in much of the functional areas of Building 020. The general extent of removable beta and gamma contamination is as fol-
. lows:
2-3
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1.
Cells 1,2,3, and 4 and connected Decontamination Rooms-high to moderately high (.;>_5,000 dpm/100 cm contamination known to exist).
2 2.
Most of the cell senice gallery has some degree of low-level contamination 2
(< 3,000 dpm/100 cm ). Also, the hoc shop, hot laboratory, the hot storage area, air lock, and the loading dock.
3.
The remainder of the cell senice area, the operating gallery, the slave shop, and r
the passageway between the operating gallery and the hot shop, are probably L
contaminated, but at a lower level (< 1,000 dpm/100 cm ),
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The radiation levels in the cells average 50 mR/h, with drains reading 1 R/h. Most equipment in the cells has contamination of > 100,000 dpm/100 cm and dose rates up to 2
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500 mR/h (near contact). The decontamination area ranges between 0.01 and 0.10 rad /h.
Senice gallery, hot slave shop, and hot storage areas are in the 50 to 5,000 dpm/100 cm2 range with up to 20 mR/h in cracks, etc. The basement area, in general, is about I
2 1,000 dpm/100 cm, tank alcove,100 mR/h, pipes, pumps, and filter banks are generally contaminated. The radioactive contamination consists primarily of old mixed fission prod-(
ucts (Cs-137, Sr-90, Pm-147) with Co-60 and small amounts (< 0.01%) of uranium and trace amounts of plutonium. About 2.2 curies of contamination is estimated within the
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building and surroundings.
The primary operating areas of the RIHL have been cleaned periodically to mini-mize contamination during the years of licensed operation. The facility was designed to control and minimize contamination. No historical occurrences are identified which might adversely affect decommissioning safety.
2.2.1 Ilot Cell and Decontamination Rooms Initial decontamination of the cells may be performed remotely if radiation levels are deemed too high. The storage and transfer drawers will be emptied, decontaminated, and surveyed remotely to ensure that all high-radiation sources are removed. Contami-nated equipment will be cleaned in place (if practical) and then moved from the cells to the decontamination rooms. The false floors will be cleaned and removed. Remote handl-ing equipment will be removed from the cells. Final decontamination and disposition of the items will be performed in the decontamination room, senice gallery, or Radioactive
' Material Disposal Facility (RMDF), as conditions permit. Master-slave manipulators, periscopes, and other items penetrating the front wall will be removed and dispositioned.
The wall penetrations will be decontaminated and sealed to prevent recontamination dur-ing cell cleaning.
2-4
I When all of the materials and equipment have been removed, the paint will be com-1 pletely stripped from the cells and decontamination rooms. Radiation surveys will be per-I formed to determine if any contamination remains. Selected areas of the steel liners in the cell and decontamination rooms will be removed to permit radiation surveys of the underlying concrete to assure no contamination exists. Further work in these areas will be based on the results of these surveys.
4 There are approximately 270 through-tubes and 26 manipulator ports located in the hot cells and decontamination rooms. The plugs will be removed from the through tubes and manipulators will be removed to allow cleaning of the internal, metal lined penetra-tion surfaces. Chemical detergents and grit blasting methods of decontamination will be used to remove contaminated material. The steel through-tube liners may be removed with concrete core saw or jack hammer if fixed contamination or concrete contamination i
is found.
i 2.2.2 Other Posted Radiological Support Areas I
The hot cells and decontamination rooms are surrounded by other radiologically controlled areas used to support the in-cell activities. These include the service gallery, hot change room, hot storage room, air lock, hot laboratory, glove box laboratory, hot manipulator maintenance, basement, and cask storage yard. These radiologically posted areas, outside the hot cells and decontamination rooms, are controlled due to the pres-ence (or potential presence) of either surface contamination, airborne contamination, ionizing radiation, or a combination thereof. The approach to decontaminating these areas will be to remove the surface contamination and any other sources of contamma-l tion to the extent required to release these areas for unrestricted use.
4 2.2.3 Liquid Waste System The liquid waste system consists of a 3,000-gal collection / holding tank, located in Building 468 and associated floor drains, sumps, and sinks located in radiologically con-trolled areas. The radioactive drain piping system, associated with the hot cells and de-contamination rooms, is embedded in heavily steel-reinforced concrete, which creates a significant barrier to its removal. Rade studies and perhaps mockup testing will be neces-sary to establish the most efficient way to decontaminate these drains / areas for unre-stricted use. If these are to be abandoned in place, they will be decontaminated, and proven releasable to SNM-21, Annex B, limits.
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I When all of the materials and equipment have been removed, the paint will be com-pletely stripped from the cells and decontamination rooms. Radiation surveys will be per-formed to determine if any contamination remains. Selected areas of the steel liners in I
the cell and decontamination rooms will be removed to permit radiation surveys of the underlying concrete to assure no contamination exists. Further. work in these areas will be based on the results of these surveys.
I There are approximately 270 through-tubes and 26 manipulator ports located in the hot cells and decontamination rooms. The plugs will be removed from the through tubes and manipulators will be removed to allow cleaning of the internal, metal lined penetra-
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tion surfaces. Chemical detergents and grit blasting methods of decontamination will be used to remove contaminated material. The steel through-tube liners may be removed with concrete core saw or jack hammer if fixed contamination or concrete contamination is found.
2.2.2 Other Posted Radiological Support Areas I
The hot cells and decontamination rooms are surrounded by other radiologically controlled areas used to support the in-cell activities. These include the service gallery, hot change room, hot storage room, air lock, hot laboratory, glove box laboratory, hot manipulator maintenance, basement, and cask storage yard. These radiologically posted areas, outside the hot cells and decontamination rooms, are controlled due to the pres-l.
ence (or potential presence) of either surface contamination, airborne contamination, j
somzmg radiation, or a combination thereof. The approach to decontaminating these areas will be to remove the surface contamination and any other sources of contamina-tion to the extent required to release these areas for unrestricted use.
2.2.3 Liquid Waste System i
g The liquid waste system consists of a 3,000-gal collection / holding tank, located in B
Building 468 and associated floor drains, sumps, and sinks located in radiologically con-I trolled areas. The radioactive drain piping system, associated with the hot cells and de-contamination rooms, is embedded in heavily steel-reinforced concrete, which creates a l.
significant barrier to its removal. Trade studies and perhaps mockup testing will be neces-sary to establish the most efficient way to decontaminate these drains / areas for unre-L stricted use. If these are to be abandoned in place, they will be decontaminated, and E
proven releasable to SNM-21, Annex B, limits.
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The radioactive drain piping located under standard flooring in support areas
(-6-inch concrete) may be excavated and sectioned as necessary for disposal or electro-I polished in situ, depending on results on cell drains. This includes the drains in posted support areas. Ali Gaher liquid waste system components such as sinks or sumps shall be decontaminated, surveyed, and, based on the results, be released for unrestricted use or Abposed of as waste. The 3,000-gal collection / holding tank will be desludged, the con-tents solidified. md the tank, pumps, valves, gauges, etc., disposed of as radioactive.
j waste.
3 Building No. 468 which houses the 3,000-gal liquid radioactive waste tank will be decontaminated using previously described conventional techniques and released for un-i restricted use.
2.2.4 Radioactive Exhaust It is planned to use the primary exhaust system to support decontamination efforts i
in each radiologically controlled area. Effective use of the system will minimize personnel j
exposure and expedite the overall decommissioning program. Other high efficiency filtra-A tion exhaust systems will be used as necessary to control airborne contamination. The ex-I haust system shall be the last radioactive system removed from each radiologically con-trolled area. Generally, the ducting removal will proceed from the farthest point of de-parture from the blower or from each main duct trunk line. The ducts, filter housing, con-l trol system, etc., may be completely removed or decontaminated and dispositioned.
Where ducting is embedded in heavy concrete, every attempt shall be made to decontam-inate that portion of the duct so that it may be abandoned in place. Any ducting that can-not be decontaminated will be removed for disposal as waste. When all of the ducting system and blowers have been removed (or determined suitable for release for unre-stricted use), the stack shall be surveyed and, if possible, abandoned in place or demol-ished. If the stack exceeds the criteria for unrestricted use, it will be size reduced and dis-posed of as radioactive waste.
2.2.5 Unposted Support Areas Unposted support areas at the RlHL are listed below:
1.
Office areas (Rooms 100,101,102,109,112 and 113) 1 2.
Operating gallery (Room 117) 3.
Restrooms (Rooms 103,105 and 110) 2-6 l
4.
Mockup room (Room 125)
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- 5. _ Heating and ventilation room (Room 133) 6.
Generator room (Room 135)
I 7.
Battery room (Room 137) 8.
Manipulator repair room (cold side) (Room 128) 9.
Storage areas (Rooms 107 and 111) 10.
Cold Change area (Room 108) 11.
Grounds within the security perimeter 12.
RIHL Roof 13.
Health Physics and Engineering Office trailers.
These areas are expected to have minimal, if any, contamination and should not -
l contain contaminated items. They will be surveyed along with all the contents to verify that the requirements for release for unrestricted use are met. Areas which are found to exceed the acceptance limits will be decontaminated.
2.2.6 Fission Gas 'thnks a
Three fission gas tanks were installed underground at the north end of the RIHL facility, c.f. Figure 2.1. The installation was never used. These were excavated, surveyed to verify that no contamination existed and released for unrestricted use.
2.3 PROCEDURES All decommissioning activities will be conducted in accordance with approved, writ-l ten procedures. The approvals necessary for release of procedures will be tabulated and issued in a Release Plan of Action (RPA). Release of the RPA will require the approval of the directors of all departments participating in the decommissioning program. Engineer-ing Document Control will verify that required approvals are obtained prior to releasing g
each procedure for use. Changes to procedures will require the same approvals as the p
original issue. Detailed Working Procedures (DWPs) will be prepared for each method and operation used in decontamination activities. A DWP will also be generated for each area of the RIHL to be decontaminated. These procedures will use the following format as a guideline.
2-7
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23.1 Scope l
The physical activities and physical areas, to which the DWP applies will be l
l described.
23.2 - Applicable Documents and References Other key DWPs that have direct applicability to the scope will be listed along with any other documents that directly apply to the activity, 233 Equipment and Materials Specific items of equipment, tooling, and materials to be used will be listed in this section, including those routinely used and any special items.-
2.3.4 Special Safety Precautions Any special safety considerations will be listed in this section. This is to emphasize j
l safety considerations associated with individual procedures.
23.5 Work Instructions i
l This section will define the step-by-step instructions for performance of the activity.
Cautionary notes on critical tasks will be included where safety or success of subsequent a
tasks may be affected.
l 23.6 Waste Volume Estimates A description of the radioactive contaminated material that is expected to be found will be tabulated, i.e., volume, classification, type, and burial volume of packaging and f
number of containers.
23.7 Manhour Estimates i
6 Estimates of the types of manpower, the number of manhours of each type, and.
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total hours to complete the activity will be included.
23.8 Personnel Dose Estimates i
i Estimates of the expected personnel doses will be made based on the manhour esti-mates and available radiation survey data. These dose estimates are required to allow 4
1 2-8
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work crew assignment, considering individual dose and allowable limits, an to imple-ment the "as low as reasonably achievable" (ALARA) program.
2.4 SCHEDULE i
The schedule for decontamination activities on the RIHL is shown in Figure 2.2.
I-
. This schedule may be changed to use personnel effectively and to adjust for unforeseen problems or programmatic perturbations. Such changes will require management concur-a rence and approval. The general schedule reflects the philosophy of decontaminating the areas of highest radiation levels first and proceeding to those where little, if any, contami-l:
nation is expected to be found. Two exceptions to this are the radioactive exhaust and ra-dioactive liquid waste systems. These systems are necessary to efficiently conduct the de-contamination of the facility in a safe manner. The schedule shows the initial demonstra-l tion of decontamination methods and techniques on Hot Cell No.1. This has been
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successful'and decontamination of Hot Cell No.1 and Decontamination Room No.1 is in progress. Decontamination of the Glove Box Lab (Room 139), the Laboratory (Room 141), and the hot Manipulator Repair (Room 128) is proceeding in parallel to allow effi--
cient use of available personnel. The Service Gallery (Room 147) and adjacent north t
hallway, the basement, and attic are to be initiated near the completion of Hot Cell and Decontamination Room decontamination to maintain efficient use of personnel. The sup-port drains for Rooms 149, ~147,133,141,139,128, and 125 are scheduled for removal concurrent with the decontamination of these rooms. These drain lines are located below nominally 6 inches of concrete and may be decontaminated in situ or removed by conven-l tional methods.
I The Radioactive Drain System from the hot cells and decontamination rooms is scheduled late in the program. It is embedded deep below the cell floor and walls in-heavi!y reinforced, dense concrete. Removal will require a major excavation effort. Devel-f opment of electropolishing as a remote decontamination method may allow the drain 5
lines to be cleaned, surveyed, and abandoned in place. The decision on which approach to take will be based on in situ tests now in progress.
The Hot Storage (Room 133), Air Lock (Room 155), and the Radioactive Exhaust t
System are scheduled to phase into the effort when decontamination of Rooms 117 and 125 is complete. Again, this is to maintain efficient use of personnel. The exhaust system 1
decontamination and removal is staged to maintain service until work is completed in all radiologically controlled areas. Decontamination of peripheral areas, where little, if any, 1
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gg Schedule APERTURE 2-10 CARg Mao XvallaMeh h
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8 contamination is expected, is scheduled as the final work phase. These areas include j
administrative offices, cold change room, roof, yard, and surrounding asphalt covered and bare grounds.
2.5 ORGANIZATIONAL RESPONSIBILITIES AND AUTHORITY The operations of Rocketdyne Division of Rockwell International Corporation are under the direct management of the Division President. The president delegates to sup-port organizations the responsibility for assuring that all operations are conducted in a safe manner and in accordance with the provisions of applicable licenses and regulations.
The organizational structure is shown in Figure 2.3. Although the titles and structure are subject to change, the principle lines of authority will remain unchanged.
The Health, Safety and Emironment (HSE) Department is delegated responsibility for radiation safety, industrial hygiene, and industrial safety. In this capacity, the depart-1 ment is responsible for development of an overall safety program and serves as advisor to functional organizations to assure that programs are implemented satisfactorily.
While the responsibility for safety is delegated to all operational management, members of HSE have the authority to halt any unsafe operation. Approval to restart a l
halted operation requires the concurrence of HSE.
2.5.1 Key Positions I
- The key safety positions by title and safety related responsibilities are:
l Manager - Radiation & Nuclear Safety Director - Health, Safety & Environment Vice President - Human Resources & Communications President - Rocketdyne Division The key operational positions with safety responsibilities are:
Manager, Nuclear Operations Chief Engineer, Atomics International Division Director, Atomics International President, Rocketdyne Division
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l 2.5.2 Educational and Experience Requirements The minimum qualification requirements for operational and safety personnel with safety-related management and staff positions are:
2.5.2.1 Operational Personnel L
1.
Department Director and Higher Management Positions - Bachelor's degree 1
l3 or equivalent in one of the physical sciences or engineering disciplines. Eight m
years of experience in a technical discipline related to the technology asso-i ciated with department operations, including 1 to 3 years of management ex-perience. Demonstration of the management and technical capabilities to or-I gamze and direct technical groups responsible for conoucting the operations for which the department is responsible.
i 2.
Manager - Bachelor's degree or equivalent in one of the physical sciences or engineering disciplines. Five years of experience in a technical' discipline re-lated to the technology associated with operations, including 1 to 3 years of technical management experience. Demonstration of the management and technical capabilities to organize and direct the operations.
2.5.2.2 Safety Personnel 1.
I Dspartment Director - Health. Safety. & Environment - Bachelor's degree in j
one of the physical sciences or in engineering. Eight years of experience in a i
technical discipline, plus 1 to 3 years of management experience. Demonstra -
tion of the managerial and technical capabilities required to organize and.
I conduct the company safety program.
I 2.
Management Personnel - Radiation and Nuclear Safety - Bachelor's degree B
in one of the physical sciences. Two years of experience in radiation protec-tion operations. Demonstration of sufficient judgment and analytical capabili-ty to establish and maintain a technically sound and effective program for the i
function to be supervised.
The bachelor degree requirement for the above positions may be waived with ex-plicit review of the applicant's background by supervision one level above the position under consideration and with concurrence of the next higher level of management.
4 2.6 TRAINING
. Appropriate training is provided to all employees who are assigned to work in areas where their personal safety or the safety of operation require it. New employees and other personnel whose regular work assignments include exposure to radiation or radio-active materials must complete a formal training course covering the general aspects of 2-13
__o
I I
working with these hazards. Periodic refresher training is a part of this program. This training includes all aspects of radiation safety that are appropriate to the assignments.
Management establishes the necessary training requirements for assigned personnel. The Training Department assures that the training occurs and retains the training records.
Implementation of the program is conducted by the Manager, Nuclear Operations.
I-2.7 CONTRACTOR ASSISTANCE Minimal, if any, contractor assistance is anticipated. However, any contracted activ--
ity invohing contaminated areas or materials must be covered by detailed work proce-g dures and sufficient training to ensure compliance with safety requirements by all person-j nel involved.
I E
9 I
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2-14
H 15 l
l 3,0 RADIATION PROTECTION 3.1-HISTORICAL INFORMATION I
The RIHL was designed and constructed to provide hot cells and auxiliary support for the examination of irradiated nuclear fuels and reactor components. Examinations lI have been conducted with Sodium Reactor Equipment (SRE) fuel assemblies and moder-ator cans; fuel elements from the Organic Moderated Reactor Experiment (OMRE) and g
W Piqua reactors; and fuel test capsules irradiated at the Materials Testing Reactor (MTR),
j Engineering 1bst Reactor (ETR), General Electric Testing Reactor (GETR), Whiteshell-i Reactor No.1 (WR-1), and Hanford reactors. Three intact compact reactor cores (under license-exempt operations), for the System for Nuclear Auxiliary Power (SNAP) 8 Exper-g imental Reactor, SNAP 10FS-3, and the SNAP 8 Demonstration Reactors, have been dis-assembled and examined. Irradiated fuel rods from the SRE, Hallam, EBR-I, SEFOR, EBR-II, and Fermi were declad at the RIHL from 1974 through 1988.-
The cells and decontamination rooms used in performing these examinations were decontaminated periodically or after completion of a program to maintain the radiolog-ical activity at "as low as reasonably achievable" (ALARA) levels. No modifications have 5
been made to the structural design of the facility during the operational lifetime. The esti-mated contamination in the functional areas of the RIHL is provided in section 2.2.
1 3.2 ALARA POLICY Along with its other responsibilities in maintaining an effective program of indus-trial and environmental safety, Rockwell is concerned with minimizing any adverse effects to the health and safety of workers, the public, and the environment due to operations with radioactive materials. This concern is consistent with NRC requirements for main-4 taining radiation exposures ALARA.
The direct goal of radiological safety procedures, including design, review, opera-i tions, training, and monitoring, is to minimize personnel exposures to ionizing radiation.
This goal is achieved by effective implementation of Health Safety and Emironment Pro-cedure G-01, " Radioactive Materials and Ionizing Radiation."
The general objective is to minimize radiation doses received by both individuals and groups by eliminating unnecessary exposure and limiting exposure to that necessary to the proper performance of work. The primary responsibility for identifying the need i
3-1
_o
I i
g and the method for achieving this objective is assigned to the Radiation and Nuclear Safety Unit.
l Individual doses are monitored in terms of millirems per week, month, quarter, and year. Group doses are monitored in terms of person-rems in similar time intervals. Indi-vidual doses per quarter and year are directly compared with appropriate dose standards
[
as established in 10 CFR 20 and with prorated standards for weekly and monthly doses, j
I Engineering designs and operating procedures are structured to assure effective opera-tions and control of radiological hazards. As a result, the exposure of the general public is so low as to be undeterminable.
I Additional guidelines for ALARA implementation are taken from U.S. NRC Regu '
latory Guides 8.8 and 8.10 and DOE Order 5480.11.
While the maximum personnel dose limits are those stated in the regulations (10 CFR 20), a planning limit of 1.0 rem per calendar quarter whole-body dose is applied in practice. Pocket dosimeters and special monitors that are processed following the work provide special control during operation in planned high-exposure situations.
Radiation and contamination surveys and determination of airborne radioactisity l
concentrations are made by trained and experienced Radiation Safety technicians and engineers. Allowable working times and the need for special precautions, such as protec-I tive clothing and respiratory protective devices, are determined by Radiation Safety per-sonnel. Radiation and Nuclear Safety is independent of the operations groups and has the responsibility and authority to halt unsafe operations.
Entry into posted areas is controlled where there is likely to be external radiation or airborne radioactivity in excess of acceptable levels for continuous work. Varying degrees of control, consistent with the hazard involved, are exercised over posted areas by Radi-g ation and Nuclear Safety. The Controlled Work Permit (Form 719-L Rev. 7-87) is a p
means of restricting access to posted areas on the basis of types of personnel and poten-tial hazards. The highest degree of hazard is associated with an area in which the active contamination and radiation levels are of such significance that special rigid entry con-trols and safety precautions are necessary. The permit requires approval of the managers controlling areas where work is to be performed, the personnel who are responsible for performing the work, and the Radiation and Nuclear Safety representative. The work per-mit defines the work location, description of the activity, applicable detailed work proce-dures, and validity period. In addition, it details the specific safety equipment required, 3-2 l
}
the surveyed radiation levels, and any special safety instructions. The individuals perform.
ing the work are required to read and understand the requirements.
33 HEALTH PHYSICS PROGRAM 3.3.1-Radiation Protection Equipment 3.3.1.1 Instruments (Survey, Counting)
The kinds of equipment used in radiation detection by the HS&E department, for portable survey, air sampling, and sample counting, include the following:
1 DC Powered (portable instrumentation) i l
Category 1 p-y ionization-type survey meter; I
sensitivity - 0.1 to < 5,000 mR/h -
g Category 2 Thin-window pancake Geiger-type survey meter;
)
sensitivity - background to 5 x 10 counts / min 5
Category 3 a scintillation-type survey meter:
j sensitivity - background to 10 counts / min 5
Category 4 p-y Geiger-type survey meter; sensitivity - 0.1 to 200 mR/h 1
AC Powered I
Category 6 Counting scalers and sample changing system; oc, p sensitivity to a and p radiation -
0.1 to 106 counts / min i
~
Category 7 Count rate meter for oc and p radiation 1
1 Category 8 Air monitor p-y Category 9 Air monitor oc Category 10 Multichannel analyzer Nal (TI) or Ge detectors.
i Calibration and maintenance of this equipment are controlled by Radiation Instru-ment Services Calibration Laboratory and performed to manufacturer and/or user recom-mendations. Instruments are calibrated after repair and on a regular schedule. Battery-1 powered instruments are calibrated every 13 weeks; AC-powered instruments are cali-brated every 6 months. If the manufacturer recommends more frequent servicing, that in-terval is used for routine calibration. Calibration sources are traceable to the National Institute of Standards and Technology.
t 3-3
0 3.3.1.2 Respiratory Protection Respiratory protective devices are used for decontamination operations (e.g., cutting.
an exhaust duct containing residual radioactive material) or for emergencies that might produce airborne radioactive material in excess of the limits specified in Appendix B, l
'Ihble 1, Column 1,10 CFR 20, and subject to the conditions specified in 10 CFR 20.103c The issuance and use of respirators are under the direct control of the HS&E department. A minimum of a successfully completed medical review plus 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of general training in the use of the respirator is required for any person required to use jg respiratory protective devices. Additional training is required for special devices such as W
airline respirators or SCBA. The equipment itself may be issued only by HS&E person-nel. The proposed use is reviewed by an HS&E representative who determines the appro-priate device to be used. Each time a respirator is returned, it is serviced, cleaned, and sanitized before being reissued.
3.3.1.3 Protective Clothing Protective clothing requirements are established by R&NS to prevent contamination of personnel. Required protective clothing generally consists of laboratory coats and shoe g
covers worn over personal clothing. However, additional specialized clothing is used when
- W-warranted by the nature of the operation, quantity and quality of material involved, and/
or chemical and physical form of the material involved.
l 3.3.2 Radioactive Exhaust Systems t
The RIHL exhaust systems direct the flow of air from the outside of the building l
into the controlled areas. The air flow is from an area of lower contamination to an area of potentially higher contamination within the building. The systems are designed specifi-l cally to control airborne contamination. As such, these systems will be maintained and utilized in the decontamination activities. Portable pump / filter units may be used for small local situations, but primary reliance will be on the RIHL systems.
3.3.2.1 High-Volume Cell Ventilation Ventilation of the four hot cells and associated decontamination rooms is provided by a 12,540-cfm constant-volume blower system. A second identical blower is located in parallel and is automatically actuated in the event of a failure of the primary blower. The f
cell exhaust passes through in-cell prefilters and then through high-efficiency particulate air (HEPA) filters in the basement. When necessary to provide additional filtration, I.
3-4 I
1 19 l
'g HEPA filters will be installed in-cell, downstream of the prefilters. Cell ventilation is con-trolled by pressure instruments located in the operating gallery. Sufficient ventilation sys-I-
tem capacity is provided to create large flow rates into the cells when any cell door or other opening is made. When a cell door is opened, about 4,000 cfm are exhausted from the cell. This corresponds to a velocity of > 100 linear ft/ min through the opening into I
the cell. This flow rate is adequate to prevent the release of contamination from the cell into the adjacent decontamination room.
I lb allow more cell doors to be opened at one time, the exhaust system will be shut down for cells as decontamination is completed. As the Hot Cells and Decontamination I
Rooms are decontaminated to sufficiently low levels, it will no longer be necessary to maintain the engineered safeguards needed during the operational period. The require-l ments for these safeguards are specified in ESG-82-33, Sections 3.2.2.1 and 3.2.2.2 and relate to the use of hot cells and of other enclosures. The engineered safeguards provide for minimum airflow velocities and differential pressures as required to prevent exceeding airborne radioactivity concentrations that are greater than the occupational maximum B
permissible concentration, and for shielding in operations that could not be performed without exceeding regulatory radiation protection standards. The conditions requiring safeguards will no longer exist in several of the hot cells and decontamination rooms as decontamination is completed. Ventilation may be reduced in these areas as needed to support decontamination efforts in other areas.
3.3.2.2 Posted Area Ventilation g
The general exhaust blower provides local exhaust through HEPA filters for the hot change room, hot side of the manipulator repair room, hot laboratory, glove box lab, air-I lock, hot equipment storage, service gallery, Room 112, and operating gallery. Two iden-3 tical 22,890-ft / min constant-volume blowers are located in parallel to provide this ex-haust, and discharge to the stack. One blower is normally in operation and the second, or I
standby blower, is automatically actuated if the first blower fails. Prefilters are used on all ventilation systems and are changed frequently to extend the life of the high-efficiency filters.
3.3.3 Air Sampling Low-volume particulate air samplers (about 1 ft / min) or lapel air samplers (about 3
g 5 ft /h) are required in all locations where there is a potential for airborne radioactive 3
material in concentrations in excess of 25% of the occupational exposure standards, 8
3-5 1
I l-based on knowledge of the material, facility, and operations. To the extent practical, sam-pler inlets are located at positions representative of the air to which workers are exposed.
Samples are removed for evaluation in counting systems at the'end of each shift or more frequently, as necessary, to prevent inhalation exposures from exceeding regulatory stan-dards. Air must be drawn through filters by use of vacuum pumps (Gast Manufacturing Model 0211-V103-G8 or equivalent) equipped with filter holders (Gelman Instruments Model 4170 or equivalent), or by use of house vacuum lines equipped with filter holders l
(Gelman Instruments Model 4170 or equivalent). A typical lapel sampler would be Mine -
Safety Appliance Monitair or equivalent. Filter paper efficiency is at least 99% for par-g ticles of 0.3-diameter or greater. Provisions are made for the determination, with 20%
accuracy, of the volume of air drawn through the filter. Sampler flow rate calibration and adjustment are performed every 6 months.
3.4 CONTRACTOR PERSONNEL The same safety requirements apply to all personnel in a radiologically controlled area. Contractor personnel must utilize the same protective equipment training and profi-ciency as company personnel and operate with approved procedures.
3.5 RADIOACTIVE WASTE MANAGEMENT j
Disposal site and transportation requirements must be complied with when packag-ing hazardous waste. Material, form, concentration level, toxicity, quantity, etc., are some of the considerations that determine the type of packaging as well as the type of disposal site. For decontamination planning, only radioactive wastes will be addressed. Any other hazardous waste, such as chemicals and asbestos, will be disposed through normal inter-plant channels.
I Radioactive waste packaging has been an ongoing, routine function of the hot cell and Radioactive Materials Disposal Facility (RMDF). Procedures and guidelines are es-tablished and in place; these shall be used for the final packaging and disposition of ra-dioactive waste during the D&D program. If unique conditions arise that are not covered by existing procedures, they shall be addressed by exception in the individual detailed L
work procedure under which the work is being performed. All radioactive materials re-moved from the RIHL will be shipped to the RMDF for subsequent shipment to an L
authorized disposal site.
l 3-6
I I
I l
g Asbestos waste must be handled under special procedures and by specially trained personnel even if not radiologically contaminated. Existing asbestos handling procedures I
will be used where applicable. If unique conditions exist that are not covered by estab-lished procedures, these shall be addressed, by exception,in the detailed working proce-dure under which the work is being performed.
lI Every effort will be made to avoid generating mixed waste. (A mixed waste is de-I fined as a hazardous waste material on the Environmental Protection Agency list, that is also radioactively contaminated.) If mixed waste material is generated, it will be necessary
{
to work directly with the disposal site on a case-by-case basis to resolve the disposition.
I There currently are about 30 gallons of radioactive contaminated acid at the RIHL This material is in use for electropolishing operations. No additional solution will be gener-ated. This material will eventually be a waste product. The solution will fall into the mixed waste category under Emironmental Protection Agency rules. Also, there are
< 40 gallons of radioactively contaminated lubricants and hydraulic fluids in facility equipment and hydraulic systems, which are listed as hazardous waste in California.
l The estimated volume of LSA waste that will be generated during the decontamina-tion activities is 26,500 cubic feet. No TRU waste is expected. This estimate is based on radiological surveys, history of activities in the facility, and the decontamination level to be achieved. A breakdown of the assessed volume of waste for different areas of the facil-ity is presented in Table 3.1. The total activity estimated for waste is 2.2 Ci. This estimate is based on an extrapolation of data from 2860 ft of material g'enerated during RIHL de-3 contamination in 1989. The activity in these shipments i due primarily to Cs-137 and Sr/Y-90, with significantly lesser amounts of CO-60, Cs-134, Sb-125, and U/Pu.
3-7 I
l
I Table 3.1.
Waste Volume Estimate - Unrestricted Use Boxes Truck (I-117 Ft 3/ Box 10 Box / Truckloads Decon cells 25 2.5 Decon decons 6.5 0.6 Decon slaves and thru tubes l
Service gallery and storage 12.5 1.3 Room 141 (complete) 5.5 0.6 l
Room 139 (complete, includes 3.5 0.4 glovebox) l 1
Air lock 3.5 0.4 Cell and decon paint stripping (384 drums) 6.4 Cell and decon room liner removal 48 4.8 i
Windows (remove, decon and package) 4 0.4 Thru tube liner flanges 9
0.9 Drain removal 6
0.6 Cell and decon room scabbling 3
0.3 Service gallery, final 3
0.3 Operating gallery - face 4
0.4 I
Operating gallery - floor and trench 3
0.3 Basement 8
0.8 Cell exhaust system 16 1.6 Building exhaust 16 1.6 Basement 10 1.0 Attic 10 1.0 Waste tanks 10 1.0 Gas hold up tank 2
0.2 Excavate hold up yard and spots 16 3.2 (Heavy)
Final survey
.2 0.2 TOTAL 226.5 30.8 D635-0084 3-8
-_ - - -- J
I l
4.0 FINAL RADIATION SURVEY PLAN The RIHL facility and the adjacent ground within the security perimeter will be sur-veyed to demonstrate that all radiation levels are below those specified by SNM-21, An-nex B. A detailed work procedure will be prepared that covers each aspect of the survey.
~l This will assure that the survey is conducted in a reliable, reproducible manner and is_
documented uniformly.
+
4.1 ' SURVEY AREAS The RIHL facility will be divided into 10 areas for purposes of the final survey:
Hot cells and decontamination rooms Liquid waste drain system, including Building No. 468 Fuel storage tubes Cell penetration tubes Basement Service gallery and hallway (Room 147), hot manipulator repair (Room 128),
glove box laboratory (Room 139) air lock (Room 155), laboratory (Room l
141), and hot storage (Room 153)
Remainder of offices and support areas, including the HP and the engineering trailers Cask storage yard Remaining outside surfaces (roof, pads, trailers)
Soil samples 4.2 SAMPLING INSPECTION Throughout the decontamination process, surveys are performed to identify areas requiring further cleanup. As these areas are cleaned, surveys are performed to verify that
' l adequate decontamination was performed. After a portion of the facility has been decon-taminated, and there is little risk of recontamination, a final survey is performed to dem-
'ig onstrate that the limits for remaining radioactivity have been satisfied. This final survey is performed using sampling inspection methods with a uniform-biased selection of sam-ples. For small areas, a minimum sample set of 30 is chosen; for larger areas, the number of samples automatically adjusts the parameters of the acceptance test.
E 4-1 I
I The final survey data will be generated by measurements taken at specific locations in the facility. The entire facility will be mapped in a 3 x 3 m grid to assure uniform cov-I erage by the survey of each area. Sampling will be done in a 1 m square area within each grid block. The sample locations will be biased toward the areas expected to have the highest radiation levels. This philosophy shall be applied throughout the facility survey, including the surrounding grounds and support trailers (health physics and engineering offices). There are structural surfaces that are not amenable to large surface measure-l ments, e.g., pipes, beams, and conduits. For these type surfaces in the radiologically con-trolled areas of the RIHL, a minimum of 20% of the surfaces will be surveyed. For the g
uncontrolled office and peripheral support areas, a minimum of 5% of the floors, walls, and ceilings will be included in this final survey.
!l All of the inner surfaces of the radioactive waste drain system, fuel storage tubes, and cell penetration tubes will be surveyed 100%.
The final survey will include 5% of the ground surface within the facility perimeter fenceline will be sampled. In addition to the radiation instrument survey, a 1 liter sample
.l of soil will be taken for gamma spectroscopy and gross alpha / beta activity measurements, and if deemed necessary, chemical analyses.
Sampling of roof areas will be done at an average rate of 2%, with the sample loca-tions biased toward the areas expected to have the highest radiation levels. This will be
. l limited to instrument alpha and beta radiation measurements. The roof surfaces are not suitable for taking smears of removable contamination.
4.3 BACKGROUND
RADIATION g
Background radiation measurements will be made at the beginning of each work l
day, at mid-day and at the end of the work day to coincide with the daily calibration lg checks on survey instruments. The average of the backgrounds and efficiency factors de-e termined for each half of the work day will be used for correction of survey data taken in the time period.
l 4.4 SURVEY INSTRUMENTS Radiation measurements will be made using the following instruments or the l
equivalent.
Ludlum Model 2220-ESG Scaler /Ratemeter
.E 4-2 I
Ludlum Model 43-1 Alpha Scintillation Probe Ludlum Model 44-9 Thin-Window Pancake GM Probe Ludlum Model 44-2 High-Energy Gamma Probe 7bnnelec Alpha / Beta Counting System Canberra Series 10-MCA System with High-Purity Germanium Detector Measurements of the average and maximum alpha surface activities for the survey will be made with alpha scintillation detectors, sensitive only to alpha particles with ener-gies exceeding about 1.5 MeV. The detectors shall be calibrated with a Pu-239 alpha
.I source standard. The energy of the alpha particles from this source is similar to that of the alpha emitters handled at the RIHL -
- I.
Measurements of the average and maximum beta surface activities will be made with a thin-window pancake Geiger-Mueller tube. The detectors will be calibrated with a Sr-90 beta source standard. This calibration source is appropriate for the mixture of beta emitters present at the RIHL
- I Measurements of removable surface activity (alpha and beta) shall be made by wip-2 ing approximately 100 cm of surface area using standard smear disks. Measuiements in the liquid waste drain lines, the fuel storage tubes, and the cell penetration tubes will re-quire special remote smear handling equipment. The activity on the disks will be mea-l sured using a gas-flow proportional counter. The counter will be calibrated using Pu-239 lI and Sr-90 standard sources.
l Surface activities (total) on the inner surface of the 3-in.-dia liquid waste drain lines g
and the 3-in.-dia cell penetration tubes will be made with a special probe. One design
. hich is being built and tested is a gas-flow proportional counter which uses the pipe wall W
w h
as the cathode. This type detector will be able to differentiate between alpha and beta activity. Whatever detector is used will be calibrated for a pipe geometry using appropri-l ate standard sources. The fuel storage tubes and the 10-in.-diameter periscope and mas-ter/ slave manipulator sleeve through-tubes are large enough to accommodate a rectangu-l' lar geometry alpha probe and the standard pancake beta probe, but special remote posi-tioning equipment and longer signal cables will be needed.
l Gamma spectroscopy measurements will be made on soil samples to qualify and l
quantify the specific activities. The high-purity germanium detectors used for these mea-l surements will be calibrated using standards that approximate the geometry and density I
4-3
j
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l of the soil samples. Gross detectable beta activity of soil samples will be measured using gas-flow proportional counters. Calibration of these detectors is done by use of kcl to represent the sample self absorption.
All portable survey instruments shall be serviced and calibrated on a quarterly basis.
j l
In addition, daily checks and calibrations shall be performed on all instrumentation to verify acceptable performance. All daily calibrations and checks will be made at the be-ginning of the work day, at mid-day, and at the end of the work day.
4.5 DATA ANALYSES l
Data sets will be created for each of the ten survey areas defined for the RIHL decontamination program Generally, the following data will be defined for each area.
I 1.,
Alpha total activity averaged over 1 square meter and standard deviation 2
(dpm/100 cm )
2 2.
Alpha maximum activity and standard deviation (dpm/100 cm )
2 3.
Alpha removable activity and standard deviation (dpm/100 cm )
l 4.
Beta total activity averaged over 1 square meter and standard deviation 2
(dpm/100 cm )
2 5.
Beta maximum activity and standard deviation (dpm/100 cm )
2 6.
Beta removable activity and standard deviation (dpm/100 cm )
4.5.1 Sampling Inspection by Variables I
Acceptance inspection by variables is a method of judging whether a lot of items is of acceptable quality by examining a sample from the lot, or population. In the case of determining residual contamination in the RIHL, by applying sampling inspection by vari-g ables methods will provide acceptable confidence in the conclusion made about the level is of contamination, while reducing the overall sample size.
The test statistic, x + ks,is compared to the acceptance limit U.
l-where x = average (arithmetic mean of measured values) observed sample standard deviation s
=
k tolerance factor calculated from the number of samples to achieve the desired I-sensitivity for the test
=
I l
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I
-U = acceptance limit The sample mean, standard deviation, and acceptance limit are easily calculable I
quantities; the value of k, the tolerance factor, bears further discussion. Of the various criteria for selecting plans for acceptance sampling by variables, the most appropriate is the method of Lot 'Iblerance Percent Defective (LTPD). The LTPD is defined as the poorest quality in an individual lot that should be accepted. Associated with the LTPD is a parameter referred as a consumer's risk (p), the risk of accepting a lot with quality equal to the LTPD. NRC Regulatory Guide 6.6 (" Acceptance Sampling Procedures for Exempted and Generally Licensed Items Containing By-Product Material") states that i
the value for the consumer's risk should be 0.10. Conventionally, the value assigned to
- g the LTPD has been 10%. These a-priori determinations are consistent with the literature i3 and regulatory position and are the same values used by the State of California. Thus, based on sampling inspection, the acceptance of the hypothesis that the probability of'
.l accepting a lot as not being contaminated which is, in fact,10% defective is 0.10. The val-l l
ue of k, which is a function of a-priori determinations made for p and LTPD, is derived based on the sample size, n, i.e.
l k = K + /Kj-ab Kj Kj
- a = 1 2 (n - 1) ; b = Kj n
2 a
where k = tolerance factor K2 the normal deviate exceeded with probability of p, 0.10
=
(from tables, K = 1.282) l Ks = the normal deviate exceeded with probability equal to the ITPD,10%
lg (from tables, K = 1.282) l l
number of samples n
=
4.5.2 Acceptance Criteria The criteria for acceptance are presented below as applied to the radiation data.
1.
Acceptance: If the test statistic x + ks is less than or equal to the acceptance I
limit (U), accept the region as clean. (If any single measured value exceeds 90% of the limit, decontaminate that location to as near background as is possible, but do not change the value in the analysis.)
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- 2. - Collect additional measurements: If the test statistic i + ks is greater than the limit (U), but iitself is less than U, independently resample and combine all measured values to determine if i + ks < = U for the combined set;if so, accept the region as clean. If not, reject the region.
~3.
Rejection: If the test statistici + ks is greater than the limit (U) and i > =
U, reject the region.
4.5.3 - Summary Presentation The results of the surveys in the test areas of the RIHL will be presented in a tabu-lation which is easily interpreted in reference to regulatory limits. This summary will be g,
comprehensive and include all of the fimal survey results. It will clearly illustrate' that the facility meets the limits imposed for release for unrestricted use.
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D635-0084/ tab I
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5.0 FUNDING The estimated cost to complete decommissioning the RIHL is $11.1 million. DOE is liable for decommissioning of the facility as stated in Contract DE-AC 86-SF16021, Modification 004, dated 9/30/86, titled " Fermi Fuel Declad Program." A cost estimate
{
has been prepared for the DOE The funding requested by fiscal year is: $2.8 million (1990), $3.1 million (1991), $3.5 million (1992), and $1.7 million (1993). This funding is based on current planning and is reflected in this decommissioning plan.
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6,0 PHYSICAL SECURITY AND MATERIAL CONTROL PLANS The physical security and material control and accounting plans submitted in sup-port of license SNM-21 will continue to be implemented. The latest revision to these documents were submitted as Attachments 3,4, and 5 to the application for license re-l newal.* The document numbers and titles are:
ESG-80-17 Physical Security Plan for the Protection of Special Nuclear Mate-rials at the Hot Laboratories of the Rocketdyne Division of Rock-well International - Revision 6, May 19,1989 I'
RI/RD86-190 Fundamental Materials Control for Special Nuclear Materials, April 15,1989 RI/RD86-190 Special Nuclear Material Control Program for Actinide Burner I
I Supplement 1 Program (TRUMP-S), May 10,1989 i
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- Rockwell Letter 89RC06668, " Application for Renewal of License No. SNM-21, Docket No. 70-25, Issued to Rocketdyne Division of Rockwell International," R. T. Lancet to L C. Rouse, dated May 25,1989 I
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