ML20247D590
| ML20247D590 | |
| Person / Time | |
|---|---|
| Site: | 07000754 |
| Issue date: | 03/27/1989 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20246N964 | List: |
| References | |
| NUDOCS 8905250490 | |
| Download: ML20247D590 (104) | |
Text
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SPECIAL NUCLEAR MATERIAL-LICENSE RENEWAL' APPLICATION-FOR THE
-VALLECITOS NUCLEAR CENTER i
MARCH'27, 1989 4
I LICENSE SNM-960 DOCKET 70-754 l'
GENERAL ELECTRIC COMPANY VALLECITOS NUCLEAR-CENTER P.O. BOX 460 PLEASANTON, CA 94566 8905250490 890421 di
{DR ADOCK 0700 4
SPECIAL NUCLEAR MATERIAL LICENSE RENEVAL APPLICATION FOR THE VALLECITOS NUCLEAR CENTER INTRODUCTION P
The General Electric Company has been engaged in nuclear energy work since 1955. One of the major installations at which the Company conducts this work is the Vallecit.os Nuclear Center located near Pleasanton, California.
OBJECTIVES OF THIS APPLICATION The basic objective of this application is to obtain a renewal of License SNM-960, Docket 70 754.
APPLICATION General Electric hereby submits the enclosed information describing all activities at the Vallecitos Nuclear Center in which special nuclear materials are utilized, 'except those activities subject to licensing pursuant to 10 CFR Part 50, in support of our application for renewal of License SNH460.
Several sections contained in the previous submittal have been eliminated due to reduced activities. '.The facilities previously described in the eliminated and/or replaced sections have been discontinued. On the basis of this information, General Electric requests that License SNM-960 be renewed.
By C. E. Cunningham l
Senior Licensing Engineer 1
Irradiation Processing f
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..i CONTENTS 1
r Section Eggg 1.0 CENERAL INFORMATION 1-1 1.1
-Corporate and Financial....................................
1-1 1.2 Location and General Description of Vallecitos Nuclear Center..............................................
1-1 1.3 Ceneral Plans and Uses of Special Nuclear Materials........
1-3 1.4 Special Nuclear Material Possession Limits.................
1-3 1.5 Principal Veilecitos Facilities............................
1-4 EiKEA 11 Bay Area Map...............................................
1-7 1.2 Vallecitos-Nuclear Center Building Locations...............
1-8 2.0 ORGANIZATION AND ADMINISTRATION 2-1 2.1 Organization...............................................
2-1 2.2 Delegation of Re sponsibility...............................
2 -1 2.3 Site Management Review Function............................
2-2 2.4 Responsibilities of the Nuclear Safety Component...........
2-2 2.5 other Components contributing To overall Safety............
2-3 2.6 Technical Personnel Capabilities..........................
2-4 2.7 Implementation of Criticality Control Program..............
2-4 i.
2.8 Implementation of Kadiation Safe ty Program.................
2-7 2.9 Training Programs........................................
2-9 2.10 Independent Review and Change Authorization Procedure......
2-11 Fi rure 2.1 VNC Administrative Organization Chart....................... 2-14 2.2 Nuclear Safety Organization................................
2-15 2.3 Radiation Safety Structure.................................
2-16 2.4 Criticality Safety Structure...............................
2-17 Addendum A, Personnel Resumes 2-1A l
3.0 RADIATION PROTECTION FACILITIES AND EQUIPMENT 3-1 4
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3.1 Portable Monitoring Instruments............................
3 1 3.2 Fixed Monitoring Equipment.................................. 3-2 3.3 Protective C1othing........................................
3-3 I
3.4 Film Badges and Pocket Dosimeters.........................
3-3
[
3.5 Radioactive Waste Facilities........<.......................
3-3 3-5 l
3.6 Emergency Equipment.........................................
3.7 Industrial Safety Equipment.................................
3-5 3.8 Criticality Alarm Systems..................................
3-5 3.9 Analytical laboratory Counting Equipment and Capabilities 3-7 l
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4~0' RADIATION FROTECTION FROCEDURES 4-1
, 4.1 Fersonnel Work Ru1es.......................................
4-1 4.2' Limits'of Radiation In Controlled Work Areas...............
4-1 4.3 ? '
Forsonnel Monitoring.......................................
4 3 4.4 Surveys..................................................... 4-4 4.5 Pos ting and labeling....................................... 4 6 3s 4.6-
.Vaste Disposa1.............................................
4-7 4.7
- Reports and Records........................................
4 7
- 4.8 Indus trial Safe ty Program.................................. 4 l4.9-
- Vallecitos Emergency Procedures............................
4 8
~4.10
. Radiation Safety. Review.....................................
4-10 f
. 5.0-CRITICALITY SAFETY FROGRAM TECHN01DCY 5-1 5.1 Introduction...............................................
5 1 5.2:
Minimum Critical Quantities At Isolated Locations..........
5-2 5.3 Determination of Critical Parameters of Accumulations and' Arrays of Fissile Materia 1...................
5-2 5.4 Adjustment Factors for Normal Operating Conditions.........
5 6.0' IABORATORY BUILDING 102 6-1 6.1 Incation and General Description...........................
6-1 6.2' Ventilation System........................,................
6-1 6.3 Decontamination Room.......................................
6-5 6.4 ~
-Air Samplers...............................................
6-6 6.5 Liquid Waste Handling......................................
6-6
.6.6 Emergency Power............................................. 6-6 IiEER
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6.1 100 Area and Surroundings..................................
6-7
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'6.2 Building 102 Main' Floor i. Radioactive Materials Areas......
6-8 6.3 Building 102 Exhaust System Schematic Diagram..............
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':6.4 Basic Stack Sampling System for Building 102...............
6-10 7.0 RADI0 ACTIVE MATERIAIA IABORATORY 7-1 f,I 7.1' Location and General Description...........................
7-1 7.2 General Plans and Uses of Materials........................
7-1
.7.3 RKL Fac il i ti e s............................................. 7 - 1 7.4 RML Criticality Control System.............................
7-6 ZiEEF 7.1 Radioactive Materials laboratory Hot Ce11s................. 7-10 l.
l-f Lieseee No. SNM-960 Decket No.70-754 Sect. No. contenen Pope i
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19.0 DELETED 10.0 CHEMISTRY, METALLURGY AND CERAMICS M50RATORY -
10-1 BUILDING'103 10.1';
Location and General Description...........................
10-1
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10.2' General Plans and Uses of Materia 1.........................
10-1 10-1 10.3:
Imboratory Facilities and Equipment.........................
'10.4 -
Building 103. Procedures....................................
10 5 10.5 Building 103 Criticality contro1...........................
10-7 1
Figure.
10.1 Building'.103 Floor Plan ' (Ground 14 vel)..................... 10 8
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- 10.2 Building 103 Floor Plan (Second'F1oor).....................
10-9 j
- 10. 3 ' Building 103 Vault layout..................................
10-10 11.0
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12-1
.12.0 BUILDING 105 12.1-Location and General Description............................'12-1 j
12.2 NTR Facilities.............................................
12-1 12.3-Other Laboratory Areas (Advanced Nuclear Applications).....
12-1 Firure 12.1 Building 105 Floor F1an....................................
12-2 13-1 13.0 WASTE BANDLING 13.1 Radioactive Liquid Waste Evaporator Flant (Building 349)...
13-1 13.2 Solid Weste Handling.......................................
13 4 ZiEELt 13.1 Waste Treatment F1 ant......................................
13-6 13.2 Hillside Waste Storage.....................................
13-7 i
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I 1.0 GENERAL INFORMATION j
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.1.1 CORPORATE AND FINANCIAL 1
This application is filed by General Electric Company, a New York corporation I
with a principal place of business 'at One River Road, Schenectady, New York.
A list of the names of the directors and officers appears in General Electric's latest Annual Report, a copy of which is forwarded to the j
Commission each year for inclusion in Docket 70-754. All correspondence may be addressed to 3135 Easton Turnpike, Fairfield, Connecticut 06431.
General Electric is not controlled by any. alien, foreign corporation, or foreign government; it is controlled by its Board of Directors and the Officers elected by the Board.
General Electric is a publicly held corporation whose stock is traded on the principal security exchanges. The applicant has no knowledge or any information indicating any appreciable ownership of General Electric stock by an alien, foreign corporation, or foreign government. No person owns of record or is known by General Electric to own beneficially one percent or more of the outstanding shares of its capital stock.
Financial information required by 10CFR70.22 appears in the aforementioned Annual Report.
1.2 IDCATION AND GENERAL DESCRIPTION OF VALLECITOS NUCLEAR CENTER The General Electric Vallecitos Nuclear Center (VNC) is operated in conjunction with the operation of a 100-kW nacier.r resekrch reactor, for the development and examination of reactor fuels, and to provide nuclear products and services.
It is located near the center of the Pleasanton quadrangle of Alt.neda County, Californie.. The laboratory is east of San Francisco Bay, approximately 35 air miles east-southeast of San Francisco and 20 air miles License No. SNM-960 Decket No.70-754 Sect. No.
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. north of San Jose. The site is indicated on the Bay Area map, Figure 1.1.
f The nearest sizeable towns are Pleasanton with a population of approximately
'40,000 located 4.1 air miles to the north-northwest, and Livermore with a l
population of approximately 50,000 located 6.2 air miles to the northeast.
A United States Veterans Administration Hospital with a population of approximately 500 is located about four miles to the east.
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The site is on the north side of Vallecitos Road (State Route 84), which is a two-lane paved highway. The Union Pacific railroad lies about two miles west l
of the site. The laboratory site consists of approximately 1,600 acres, about one-third of which is relatively flat. Approximately 1,400 acres of j
the site are leased for raising feed crops and cattle grazing.
j i
There is very little industrial activity within a 10-mile radius of the plant. A saml1 amount of light industry is located at Pleasanton, Fremont and Livermore, but these towns are not industrial centers. The city of
. San Jose to the south, 20 miles distant, and Oakland and San Francisco, 30 and 35 miles, respectively, to the northwest are the major industrial centers in the vicinity. In the southeast quadrant, there are no industries and very r, parse population for 20 miles and beyond.
i The property on which the laboratory buildings are located is drained by ditches leading to Vallecitos Creek. This creek discharges to Arroyo de la l
laguna near the north and of Sunol Valley, two to three miles southwest of l
the property, Water is supplied from the Hetch-Hetchy aquev,.:t by means of a 14-inch line capable of supplying over 3,000,000 gallons per day. A 500,000-gallon j
storage tank is provided on the laboratory site. One hundred thousand Ea11ons are reserved for fire protection.
Electrical power is supplied by Pacific Cas 6 Electric to the main laboratory substation from whence it is distributed to each building on the site.
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A sewage treatment system is provided in the southwest corner of the site.
1 Effluent from this system is disposed to site land.
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1.3 GENERAL PIANS AND USES OF SPECIAL NUCLEAR MATERIALS j
l This application requests authorization under Title 10, Code of Federal l
I Regulations, Part 70, to receive and possess the special nuclear material designated in Ssetion 1.4 herein; to receive and possess the special nuclear material and associated byproduct material produced by the irradiation thereof; and to use said special nuclear materials in research and i
development activities as defined in Section 70.4(j), ia chemical and physical analysis, and examination and investigation of nuclear fuels, associated materials and devices at the Vallecitos Nuclear Center.
1.4 SPECIAL NUCLEAR' MATERIAL POSSE 3SION LIMITS 1.4.1 Vallecitos Nuclear Center The special nuclear materials used in connection with activities authorized by Lic6nse SNM-960 at the Vallecitos Nuclear Center will no?. at any time exceed those limits listed in Section 2.1 of. Appendix A to L1. cense SNM-960.
1.4.2 Form and Enrichment Soecificationg l
The majority of.the Vallecitos Nuclear Center activities ara conducted in facilities and under procedural controls which accommodate any chemical or physical form and any U-235 isotopic content.
If specific limitations are placed on these parameters in connection with an irdividual activity in order I
to ttsure the radiation or nuclear safety of that work, the limit is described in the appropriate section of this application entitled, " General Plans and Uses of Materials".
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1.5 PRINCIPAL VALLECITOS FACILITIES Descriptions of the principal buildings and laboratories in which special L
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nuclear materials are used at the Vallecitos Nuclear Center site are set
-forth in this section with the primary objective of general orientation. The locations of these' facilities are shown in Figure-1.2. The specific activities conducted 'a W of these facilities and their safeguards j
equipment and procedures, ue discussed in later sections.
I 1.5.1 Radioactive Materials Imboratory The Radioactive Materials Laboratory (RML) is located in Building 102. This laboratory is a shielded facility equipped with remote manipulators to conduct experiments and analyses with irradiated reactor fuels'and other radioactive materials. The facility also includes a storage pool and dry pit l
storage.
1.5.2 Radiochemistry Laboratory Adjacent to RML, on the main floor of Au11 ding 102 and providing analytical support to it, is a radiochemistry laboratory equipped with standard chemical and' radiochemical apparatus. This laboratory primarily is used to analyze samples of materials prepared in the RML.
l 1.5.3 Meta 11urzv. Chemistry. and ceramics Buildinz l
A second major laboratu. - building in the 100 Area is the Metallurgy, Chemistry, and Ceramics Laboratory (Building 103). This two-story building I
consists of laboratories, variouuly equipped with laboratory apparatus designed to handle moderate quantities of radioactive materials, and offices.
The functions served by this facility are resesrch, development, and analytical chemistry services.
License No.' RNM-960 Decket No.70-754 sect. No.
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'1.5.4 Buildinn 105
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Just north of Building 102 is Building.105. The principal facilities located
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~in this building are an experimental reactor (the Nuclear Test Reactor) and 1 laboratories.
The' Nuclear Test Reactor. serves as a source of neutrons for neutron radiography., exponential experiments, irradiations, and as a ss;4itive device for reactivity measurements. The laboratories in Building 105 use only minute l quantities of special. nuclear materials.
1.5.5
'Enrineerine' Shoo Building 106 contains machine shops, instrument calibration facilities, maintenance support services, and the development shop.
1.S.6 Solid Radioactive Waste Storare Facility Solid radioactive wastes generated at the.various laboratory and facility
. locations.are stored in the waste storage facility located approximately midway.between the deactivated Vallecitos Boiling Water Reactor (VBWR) and General Electric Test Reactor (CETR) areas. This storage area includes shielded horizontal tubes for storing 5-inch and 7-1/2-inch-diameter waste liners ',
'1.5.7 Waste Evanorator Plant The Waste Evaporator Plant is located adjacent to the deactivated VBWR site.
This plant is used to concentrate and solidify liquid radioactive wastes I
generated at the Vallecitos Nuclear Center or other licensed facilities prior to transfer to authorized waste disposal firms or weste burial sites.
Such wastes may contain minute quantities of special nuclear material.
l License No. SNM 960 Decket No.70-754 Sect. No.
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.1.5.8' Reactors and Auxiliary Facilities The ESADA Vallecitos Experimental Superheat Reactor (EVESR), the VBWR, and p
k the CETR are deactivated.
1.5.9 400 Area The 400 Area consists of two buildings, 400 and 401-Building 401 is devoted chiefly to offices, while Building 400 currently is used for storage.of nonradioactive materials.
1.5.10 Buildine 104 This building is used for warehousing.
l 1.5.11 nuildine 300 This building contains a radiochemistry training laboratory and a reactor L
fuel cladding examination facility.
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VALI.ECITOS NUCLEAR CENTER BUIlh1NG IDCATIONS I
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i 2.0 ORGANIZATION AND ADMINISTRATION All Vallecitos Nuclear Center (VNC) activities'are conducted under the management of two organization components of the General Electric Company.
l This division is represented at VNC by Irradiation BUR Technolorv Processing (IP), which is responsible for radioactive materials handling and inspection, radioisotope encapsulation, nuclear safety, and landlord activities at Vallecitos; and by Chemical Technology Products & Services and by. Fuel & Plant Materials Technology, Products & Services, which are responsible for development and characterization of reactor materials and I
analytical functions.
This component provides consulting, l
Nuclear Plant Serviens Denartment retrofit, andupgradeservicestooperatingduelearpowerplantcustomersas required to operate and maintain BWR plants.
2.1 ORGANIZATION I
An organization chart for VNC is included at the end of this section as Figure 2.1.
An organization chart for the Nuclear Safety function is l
included as Figure 2.2.
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2.2 DELEGATION OF RESPONSIBILITY The General Manager of the components at VNC hac established a policy of l
protection of employees,'the public, and the environs from potential industrial, radiation, and nuclear hazards that could occur through
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activities conducted in each component's facilities. He has delegated the l
responsibility for implementing this basic policy through line managers to the manager and supervisor of each activity in which radioactive materials are handled, used or stored. Additionally, IP has experienced and competent l
staff personnel to provide expert advice and guidance to all cosaponents in matters of radiation and criticality safety. The Manager, IP, has been l
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' delegated responsibility to act as the chief executive safety officer for all j
VNC operations involving radioactive materials. Industrial safety is provided for VNC by the Industrial Safety & Hygiene function located in San Jose.
With respect to thc1 radiation and criticality safety programs, the basic line and staff relationships are illustrated in Figures 2.3 and 2.4, respectively.
2.3 SITE MANAGEMENT REVIEW FUNCTION The Vallecitos Technological Safety Council (VTSC) is composed of senior site management and/or technical personne1'and has been established as a site wide 3'
evaluation function with responsibility for management evaluations and
. reviews of VNC activities. Assigned functions include review of VNC activities to assure that major risks are identified, considered and resolved and that the application of policies and procedures is consistent between the site components. This function will review changes in product design or services that are first of a kind or that require deviation from established parameters and reflect critically on product or operational safety.
2.4 RESPONSIBILITIES OF THE NUCLEAR SAFETY C0KPONENT a.
Radiation Safety.
Establish and administer a radiation safety program to ensure the protection of employees and the genersi public. Provide monitoring support, a dosimetry program, an environmental monitoring program, and employee training in the methods for minimizing exposure
- through proper use of survey instruments and protective clothing and devices. Maintain all radiation exposure records required by regulatory
- agencies, b.
Nuclear Safety Como11ance and Review. Provide review of reportable incidents, new facilities or major changes to facilities and operations control standards, and professional advice and counsel on nuclear and radiation safety policy.70-754 Sed. S.
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IJsspsing. Represent "RC components in regulatory activities concerning radiation protection and licensing. Procure, administer and interpret NRC licenses and regulations, state and local government licenses and regulations, and provide advice and counsel to VNC components and their customers on regulatory matters.
I d.
Radiological Engineering. Assist in improved operation of VNC facilities and the overall radiation protection program through the analyses of existing systems, equipment and operation thereof and the recommendation of improved systems, equipment or methods. Generally, these studies will be safety and plant life oriented.
e.
Traininc.
Establish and administer a site-wide training program at VNC to ensure adequate knowledge of radiation control procedures.
f.
Criticality Safety.
Perform criticality control analyses to establish safe batches, geometries, concentrations, and spacing of special nuclear materials and equipment. Audit the criticality control environment and conduct educational programs in criticality matters, g.
Emerrency Preparedness and Resnonse. :4aintain the site-wide emergency procedures; coordinate and audit training.
25 OTHER COMPONENTS CONTRIBUTING TO OVERA1.L SAFETY In addition to the principal radiation and nuclear safety functions of the staff components previously described, the following components and their functions make significant contributions to the overall safety program, n.
Nuclear Materials Safecuards.
Establish and administer the basic system of special nuclear material control. Assure that SNM control policies and practices of all components are coordinated with the requirements of nuclear safety, licensing and shipping groups.
Perform reviews of compliance with internal control procedures and regulatory requirements.
License No. SNM-960 Docket No.70-754 Sect. No.
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Physical Security. Develop and operate security programs to safeguard l
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special nuclear materials and corporate property.
1 Irananortation and Materials Distribution. Assure all radioactive l
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materials shipments are in compliance with applicable Federal and State i
regulations.
d.
Industrial Safety and Hvriene Function. Develop programs to protect the employees from industrial hazards, including operation of medical and safety education programs.
2.6 TECHNICAL PERSONNEL CAPABILITIES The primary responsibility for operational radiation safety for the l
operations conducted in the various Vallecitos facilities involving special nuclear material rests with the supervisor or manager of each facility.
Equally important are the knowledge and experience of personnel in the Nuclear Safety function. Since the issuance of License SNM-960 in 1966, the Commission's Region V office has inspected VNC to assure that adequate levels of technical expertise are maintained in all positions. Resumis for key personnel are included as Addendum A to this section.
2.7 IMPLEMENTATION OF CRITICALITY CONTROL PROGRAM The program for protection against accidental conditions of criticality is implemented by means of functional responsibility assignments. Managers whose operations require the use of quantities of special nuclear materials approaching a theoretical minimum critical mass, or greater, are responsible for integrating and measuring the efforts of line and staff participants in this program. The principal participants and their responsibilities are outlined below.
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2.7.1 Responsibility of the Manarer 1
It is the responsibility of any manager of an activity involving special nuclear materials to establish and maintain a records system to accurately account for the quantities of such materials used and transferred, thereby providing a basis for controlling against excessive accumulation or loss.
It is the responsibility of any manager of an activity involving more than 500 grams of U-235, through the designated supervisor or engineer with l
specific responsibility for a given detailed activity area, to:
l Know and understand the specific limits of all nuclear parameters a.
applicable to his area of responsibility.
b.
Determine the need for a criticality analysis. This need arises whenever there is:
(1) a change in the parameters of special nuclear material such as enrichment, chemical and physical form, density, etc.;
l (2) a change in the parameters such as quantity or dimensions of unit accumulations; (3) a change in the spacing or dimensions of processing, handling or storage equipment; or (4) a change in environment such as the addition of fire sprinkler systems of possible consequence to criticality safety, Prepare in writing a request for criticality analysis stating the c.
activity or area involved, including a description of the special nuclear material, process description, and equipment description with appropriate charts, diagrams, and drawings. All special nuclear material shall be included whether license or license-exempt and whether as usable forms or as scrap and waste, d.
Upon receipt of results of a criticality analysis in writing from the Criticality Safety component, take all necessary steps to implement the appropriate controls such as:
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' (1)' ' Posting of limitations at strategic locations at the criticality
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areas where they apply.'
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Issue appropriate' procedures for establishing and maintaining equipment and materials in the arrangement and quantities
'specified by.the criticality analysis.
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(3)
Assure boundaries identifying each separate criticality area and I
position are marked conspicuously.
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' (4)
Assure all operating personnel are advised of any special limitations or controls other than as outlined above and are trained to follow the necessary precautionary procedures.
(5)
Establish procedures for handling off-standard processing conditions.
2.7.2 namoonsibility of criticality safety comoonent It is the responsibility of the Criticality Safety component to:
Review requests 1for criticality analysis to verify _ completeness of l
a.
submitted information and, as appropriate, personally inspect the equipment or location involved to obtain a first-hand determination of the possibility of interactions with other special nuclear material in
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l-b.
Make criticality control calculations to establish safe batches, I
i geometries, concentrations, and spacing of special nuclear materials and I
equipment in accordance with authorized computational methods.
l Furnish a written criticality analysis to the requesting operation for c.
each request received. The person in charge of the Criticality Safety l
component shall identify the analyses which involve systems in which the i
direct achievement of criticality may be the consequence of error in analysis and shall assure that such analyses are reviewed and verified.
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Perform a periodic inspection of the degree to which actual operations conform to physical situations on which the criticality calculations were based.
Establish a criticality control educational program to W hasize the e.
need for following precise instructions.
2.7.3~
Responsibility of Radiation Safety Comoonent i
It is the responsibility of the Radiation Safety component through its staff of qualified and experienced radiation monitors to:
In their normal rounds of duty be continuously alert to possible a.
criticality safety practices inconsistent with the appropriate criticality analyses and/or procedures.
b.
Se familiar with applicable parameters for criticality control and use these parameters as standards for inspection.
Verify that physical conditions and activities comply with the c.
conditions specified in the criticality analysis.
d.
Initiate immediate action to discontinue any SNM movement and shut down any operation which in its judgment does not fully comply with the criticality analysis.
2.8 IMPLEMENTATION OF RADIATION SAFETY PROGRAM The program for protection against radiation hazards is implemented by means of functions 1 responsibility assignments. Managers are responsible for integrating and measuring the efforts of line and staff participants in the program.
70 754 Sect. No.
2.0 Peg.
License No. SNM.960 Dedet No.
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/27/89 Amends Sect.(s)
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2.8.1 Resoonsibilitv of the Mananer It is the responsibility of the manager of an activity or area involving radioactive materials to:
Take all necessary steps to plan and organize the work of his component a.
in accordance with approved radiation safety standards and operational procedures.
b.
Identify needs for operational procedure revisions when there is a planned change in conditions such as types or quantities of radioactive materials or equipment modifications.
Integrate the results of reviews, inspections, engineering assessments c.
and investigations made by the Radiation Safety component to correct or improve operational procedures, controls and performance..
2.8.2 Responsibility of Radiation Safe 2v Comeonent It is the responsibility of the Radiation Safety component to:
Establish and administer through a system of instructions a radiation a.
control program and maintain all radiation exposure records required by regulatory agencies.
b.
Review and approve new operations in the design stage, identifying potential hazards and providing recommendations for their avoidance i
through education, training methods, and equipment application, Inspect each radiation area to assure compliance with radiation safety l
c.
procedures and limits.
Provide management with reports of inspection results showing trends in l
d.
activities involving radioactive materials.70-754 Sect. No. _
2.0 Pepe
- License No. 5 % 960 Dede No.
28 wwns
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,3u cTRAINING PROCRAMS
- 2. 9 -:
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s 2.9.1 Radiation Safety 4
I iEvery newl employee <at VNC normally' receives a radiation' safety' orientation, "New Employees Radiological: Safety Orientation",:within thirty days of e
reporting}tothesite.
If an employee's area manager determines that the-
-employee will be exposed regularly or frequently to radiation and/or radioactive materials,-lthe employee is instructed in radiation protection such Ehat hefis'able-to protect himself and is made aware of the degree'of h
hazard involved. A training program entitled, " Radiological Safety Ati'
~Vall'citos Nuclear. Center", is completed by each such employee normally e
within one' year of his starting date at VNC. This course is scheduled and conducted by the Radiation Safety component.
The.. training course includes the following elements:
I basic principles.of radiation safety, a.
b.
Company policies and operating procedures, c.
c.
radioactive materials handling methods and shielding requirements, d.
emergency procedures, ie.
requirements'of NRC regulations, and L
f.
NRC-license requirements.
Follow-up training' commensurate'with the work environment and the employee's
/
work perforn.ance is determined by employee supervision. The employee also receives on-going training in the form of on-the-job demonstrations, periodic safety meetings, etc.
Employees whose work assignments may include the need for the use of respiratory protection equipment also receive the " Respiratory Protection Training Course" (RPTC). The course provides both instruction and hands-on experience in the proper use and fitting of the respiratory protective-equipment.
l
- Liconee No. SNM-960 Decket No.70-754 Sect. He.
9o Pes
- 2-9 g
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- 2. 9. 2..
Criticality r;
-Every new employee who'will work regularly with special nuclear material in l
i areas in which quantities of special nuclear material. sufficient to form a critical mass are present is instructed in the principles of criticality:
safeguards and is made aware of,the degree of hazard involved. This instruction is completed within one year after the employee's starting date
.y at'that facility.
Area supervision _is' responsible for' informing all personnel at work or otherwise present in their area of specific procedures for criticality control and for~ appropriate administrative action to assure compliance with
~
these procedures.-
~
'2.9.3 lOther Prorrm==-
3
. Radiation monitors receive the " Radiation Monitoring Technicians Certification Course". This is an intensive training program designed to l
qualify participants as VNC-certified Radiation Monitoring Technicians.
Other courses are available from the Radiation Safety component on an as-needed basis.
l 2.9.4 Records h
Records of individuals receiving training courses are kept in a central computerized record file.
Ig L
i.
License No. SNM-960 Dodet No.70-754 3.,,. No.
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p 2.10 INDEPENDENT REVIEW AND. CHANGE AUTHORIZATION PROCEDURE 2.10.1 Definitions a.
. f"hanna Authorization. A Change Authorization is the VNC mechanism for L
independently reviewing and documenting changes. Any addition, alteration,' deletion, modification or substitution which results in a F
different position, course or direction not previously analyzed, adds a new capability, performs a different function, modifies performance characteristics, or introduces a hazard not previously analyzed will r
constitute a change. Descriptions of existing facilities and standard methods of operation normally are contained in documents supporting Federal and State licenses or previously approved operating instructions
.and may be used to assist in determining if the proposed activity constitutes a change.
b.
Facility. A facility consists of any permanent structure and that apparatus which, wholly or in part, is integrated with the structure, i.e., ~ shielding, piping, ventilation, reactor core, etc.
c.
Eauipment.. Equipment is defined as that apparatus which operates
. independently or.in conjunction with the facilities to perform the various VNC activities.
d.
hnerhant or Test.
An experiment or test consists of any device or combination of devices (not to include the facility or equipment wherein the experiment or test is housed) designed and operated for the purpose of obtaining a predetermined objective, or any nonroutine operation or manipulation of objects, systems or process streams for the purpose of gaining information.
s.
Procedure. Procedures are the written guidelines established by facility management to describe and define the methods and instructions for operations.70-754 Sea. W. _
2.0 Pope Lisonne W.
SNM-960 gg g, 2-11 3/27/89 w,gy g
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i 2.10.2 Dr225.1 Any addition, alteration, deletion or substitution which adds a new capability, performs a different function, modifies performance characteristics, or introduces a hazard not previously analyzed requires an independent review by use of the Change Authorization procedure. A Change Authorization is prepared whenever the work involves changes to:
Facilities, equipment, or procedures so that safety or regulatory L
a.
compliance considerations differ from those previously analyzed.
l b.
Preventive or corrective maintenance procedures for which no separate review procedures are established which could affect safety or violate a-license condition or technical specification, Radioactive material inventories and/or limits.
c.
d.
Hazardous or potentially hazardous industrial material inventories where such change is si nificant in terms of quantities or use.
5 The independent review of items is conducted so that the hazards (both direct l
and indirect) of the proposal are recognized and appropriate safeguards are provided to eliminate or reduce the probability and severity of potential accidents. While procurement, fabrication, selective installation or testing, etc., may proceed prior to the final CA review and approval, actual implementation of the proposed change may not proceed until this review and approval are received.
The Change Authorization is processed in accordance with a written prccedure f
and reviewed by the Nuclear Safety component and by the Industrial Safety component as appropriate.70-754 2.0 Pege License E. SNM-960 g g, g, g, 2-12 3/27/89 g
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2.10.3-Procedures I
Frenaration of channa Authorization. Knowledgeable individuals prepare
-a.
Change' Authorizations which describe in sufficient detail the nature of the changes and the effect on safety, including applicable drawings and specifications, acceptance test procedures (ATP's), quality control requirements (if applicable), means for assuring personnel-indoctrination for initiating method changes and specifying the.
-responsible supervision.
Certain minor changes and changes previously evaluated and documetated need only internal review and documentation. The Nuclear Safety component wil1~ assist the initiating component in making this determination.
b.
Review of Channe Authorization. All Change Authorizations (CA's) are reviewed by the initiating component, by the Nuclear Safety component and, as appropriate, by' Industrial Safety and are approved by the appropriate area or facility manager or his designated alternate.
Recommendations for. the chaage needed for safety may be added by the reviewers following discussion with facility personnel as appropriate.
The Nuclear Safety component has the responsibility for Determining whether a proposed change constitutes an unreviewed safety question or other license or technical specification violation. No CA should be l
implemented until review is completed and the CA is signed by the manager of the Nuclear Safety component or his designated alternate.
E c.
Distribution. A Change Authorization file will be maintained by each facility. A copy of the Change Authorization form must be filed with l
p the Nuclear Safety component, with other copies distributed as appropriate.
I 2.0 pop 70-754 bet. W. _
License S. SNM-960 Dedet W.
'a Date Amends lect.(s) 1
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, Chemical.
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_ Medical-A Radio-Eng.
_ Materials
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, logical Chemistry Services logical Products Technology Technology
& Special Processes Services l
Irradiation Processing I'" * #
Advanced Radioactive Remote Nuclear Operations Nuclear
, g,,,
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_ Pac 111 ties
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_ Shop
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" Waste Bandlins Instrument &
, Monitoring
-Eltstrical
& Dosimetry
-Transportation Maintenance
-Lic ensing
-Compliance
-Safesuards
-Quality Assurance
-Emergency Planning FIGURE 2.1.
VALLECITOS NUCLEAR CENTER ADMINISTRATIVE ORGANIZATION CHART (Components shown below the dotted line are located at VNC.)
License No. SNM.960 Decket No.
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FIGURE 2.4.
CRITICALITY SAFETY STRUCTURE i
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l Lloonee No.
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' ADDENDUM A.TO SECTION 2.0'-
4 RESUMES OF EEY MANAGERIAL AND SAFETY FERSONNEL 64
'.1.
R. W.'Daruitzel;. Manager, Irradiation Processing operation; B.S.,
A Chemical Engineering, University of.New Mexico,'1958.
7 Mr. Darmitzel' joined the General Electric Company, Vallecitos' Nuclear Center.
(VNC) in 1959 as an engineer responsible for the-planning and performance of the post-irradiation examination of nuclear fuel materials. In 1966, u
'Mr.' Darmitzel'became Manager. Isotope Production.and Development in which-capacity he-was responsible for the production of radioisotope products and
- the development of new radioisotope products, particularly for use 'in the~
i
. field of nuclear medicine. This position required interaction with such' regulatory agencies as the Food and Drug Administration, th. atomic Energy Commission, and the Department of Transportation. Mr. Darsitzel was active in the Atomic Industrial Forum and the American National Standards Institute.
From 1970 to 1975, Mr. Darmitzel became Manager, Radioactive Products and Services, and was responsible for both radioisotope production and the post-irradiation examination of reactor components in support of. the Nuclear Energy Division's reactor development programs.
In 1975,'Mr. Darmitzel assumed the position of Manager, IPO. He is responsible for the operstion of the radioisotope production and post-irradiation examination facilities, and the operation of the Nuclear Test Reactor and its support functions, including Nuclear Safety.
q Mr. Darmieni has been designated by the General Managers of the General Electric divisions represented at VNC as the manager responsible for nuclear Ie safety for the entire site.70-754 g.,,, g, 2.0, Adde,3;}gLA Pege SNM-960 y, g,,
' t.isense No.
2-1A
" 27/89 Amends lect.(s)
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' 3:
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A.2, J. M. therb; Manager, Nuclear Safety; B.S., Mechanical Engineering,
. Sacramento State College,1967; Licensed Professional Engineer, State
.of California (No. 15034).
Mr. Cherb joined the General Electric Company (CE)~ Vallacitos Nuclear Center l
S
. VNC) in 1965. After completing a training assignment in the radiation
(
1
.asfety organization he worked for most of his career in a variety of e
engineering and managerial positions'related to activities at VNC.
Engineering' positions held by Mr. Cherb include Plant Engineer, Test Engineer, Process Engineer, Project Engineer, and.Sonior Quality Assurance -
ls
- Engineer. -Re'sponsibilities' included maintaining and modifying equipment and systems,at;the Vallecitos Experimental Superheat Reactor and providing technical. direction.to operating units in'the conduct of experimental programs at the General Electric Test Reactor (CETR). He was. a maj or contributor to the establishment of the quality system at VNC which includes f
a Nuclear Regulatory Commission-spproved' Quality Assurance Program for-
' Shipping Packages for Radioactive Material in accordance with 10CFR71.
One year he' worked as a Field Engineer (Nuclear Systems) where he served as a resident QC representative during power reactor prassure vessel fabrications' and had similar assignments at several power reactor construction sites.
Mr. Cherb has'll years managerial experience in quality assurance at VNC, seven of'those years in the Nucisar-Safety organization.
In 1986 he was appointed Manager, Nuclear Safety. His primary responsibilities include developing the radiological.' safety criteria et VNC and monitoring the
~
radiological conditions to assure safety of workets and the public.
n 70-754 Seet. No. 2.0, Addendum A Page License No. -SNM-960 Dedet No. _
Dove _
Ar.pnds Sect.(s)
y.."
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't..
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A.3
- c. E. cunntneham, Senior Licensing Engineer; B.S., Physics, Iouisiana State University, 1958.
s 4
In 1958, Mr. Cunningham joined the General Electric Company in Richland,.
Washington,xunder the Technical Graduate Rotational Training Program. From 1960 to 1966Lwhile with General Electric in Richland, he was an engineer in the health physics group becoming responsible for all technical aspects of health physics for a. chemical reprocessing facility. His responsibilities included exposure records, development of administrative and technical procedures, developing bioassay schedules, liaison work with ot!.er site components and with the Atomic Energy Commission, dosimetry studies, and environmental surveillance.
From 1966 to 1967, Mr. Cunninghan was employed by Isochen, Incorporated, Richland,' Washington. His work chiefly involved an investigation into the merita of Cobalt-60'and Cesium-137 for commercial irradiation application.
In 1967,.Mr. Cunningham joined General Electric at VNC as a Critica11M Specialist, providing criticality work and consultation for other General Electric sites and backup capabilities for site health physics.
+
In 1971,.Mr. Cunningham became responsible for licensing and liaison activities between VNC and the various regulatory agencies.
From March, 1975,'to October, 1975 Mr. Cunningham assumed the additional responsibility for the establishment of the VNC Physical Security and SNM Accountability Program. Since Octcber, 1975, he ha.s reassumed his duties as the site licensing administrator.
Mr. Cunningham is a past member of the Board af Examiners of the American Board of Health Physics.
1
~
70-754 g, g, 2.0, Addendum A Pepe Lleense W. SNM-960 g g, 2-3A 3/27/89 g
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(
- A.4 L M. wurrav, Radiological Engineer; B.S., Mechanical Engineering,-
f
' Chico state College, 1963.
I Mr.,Murray has had 25 years of experience in nuclear safety and operations activities at the Vallecitos Euclear Center. He has expert capabilities in
- radiological engineering and risk evaluations, and his most recent assignment includes radiological engineering responsibilities for both the VNC and San Jose sites and responsibility for criticality safety.
O From 1963 to 1965, Mr. Murray's assignments on a training pro 5 ram included Nuclear Safety operations and technical analysis, BWR operations, and test reactor operations. He received a General Electric Test Reactor Operator's License in 1964 From 1965'to 1974f Mr. Murray held the positions of Facility Engineer, Product Manager, and Icotope Production Supervisor in the Remote Mandling Operation. His assignments ' included equipment design, project managing, sealed source fabrication, supervision of a processed isotope production crew,'and facility planning and modification.
Mr. Murray's assignments as Nuclear Saftty Engineer during the period 1974 to the present time include:
technical rick evaluations, computer-sided shielding. calculations, interns 1 and external dosimetry measurement and evaluation, radioactive affluent system evaluations and release limits,
' hypothetical accident evaluations, facility operating license application input preparation, stack monitoring system evaluations and recommendations, operating standards and procedures preparation and review, health physics is I
instruments evaluation and specification, and review of proposed site He became a California Registered Professional Mechanical activities.
Engineer in 1974 He attended and completed course work for the Health Physics Society School on Internal Radiation Dosimetry, Principles and Practices in June,1983, and the University of New Mexico-sponsored Nuclear Criticality Short Course in May, 1985.
1 -
Lieeneo W.
SNM-960 Dede No.70-754 Sed. b. 2.0. Addand"= A Pope 2-4A 3/27/89 g, g,g g
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A.5r E. J. Strain.' Rngineer; B.S., Physics, University of Utah,1952.
Upon graduation from college, Mr. Strain was employed at the California Research & Development Company in Livermore, California. Prime duty during this first year: of employment was performing linear accelerator target nuclear physics calculations. In 1953 he became a member of the operations The reactor facility
- Eroup for a water-boiler-type homogeneous reactor.
included a radioactivity counting room which, in conjunction with the
~
reactor,.provided services for the radiochemistry, instrumentation and health physics groups. The reactor facility was transferred from CR&D to the
~
LUniversity of, California's Livermore Radiation Laboratory in 1954.
Mr. Strain continued in the same position at the reactor with UCiRL.
- q Mr. Strain joined tho' General Electric compeny Vallecitos Nuclear Center in 1956 as a reactor supervisar and analyst for the Vallecitos Boiling Water Reactor Operations.. In this. position he was responsible for planning, (j
supervision, review and audit of operations and maintenance work in radiation
]
and contamination zones, and special nuclear and byproduct materials
, custodian. During this period he was one of the operations personnel trained and qualified to perform the-radiation monitor duties at VBWR. Following deactivation of the VBWR, Mr. Strain transferred to the Nuclear Test Reactor in 1962. Primary responsibility at NTR was preparing an updated hazards summary report and a license application amendment, including safety
-The new analysis, for a pressurized mixed-oxide subcritical experiment.
license, including authorization for performing the mixed-oxide suberitical experiment, was obtained in 1966. Mr. Strain then joined the Nuclear Safety group at Vallecitos as an engineer and is presently working in that capacity.
Prime. duties in this position have been to serve as a nuclear safety engineer for the EVESR, NTR, CETR, and the Plutonium Laboratory. Mr. Strain has also served as a member of the VNC Laboratory Safeguards' Croup and its Change Authorization Subgroup for several years during which time he participated in reviews involving all projects at VNC. Currently, he is the compliance angineer for the NTR and the deactivated reactor facilities and the Site Emergency Planning Coordinator.70-754 led. W. 2.0. Addendum A Pege Licensehde. SNM-960 gg g, 2-5A 3/27/89 g
g
k C
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(
3.0 RADIATION FROTECTION FACILITIES AND EQUIPMENT i :l -
.)
' Permanent laberatory buildings and other nuclear facilities make up the Vallecitos Nuclear Center complex. The buildings and facilities are designed, equipped'and maintained to perform work with radioactive materials in'a manner providing a high degree.of safety for persons amployed at the laboratory and for residents of surrounding areas.' Typical major equipment and facilities available for. the protection of health, life, and property are
- =
3 set forth in this section.
L
,y 3.1
. PORTABLE MONITORING INSTRUMENTS i
Monitoring instruments from the following list are available in adequate.
' supply to provide for essential monitoring and for scheduled calibration and maintenance.
sPORTABLE MONITORING INSTRUMENTATION Instrument Tvoe Rante 1.
GM Detector 0-500 k cym, beta-Samma
- 2. ~lonization Chamber 0-300 mrad /h, beta-gamma (low energy) 3.
Ionization Chamber (CP) 1-250 k mR/h, gamma 4-1,000 k arad/h, beta 4.
Ionization Chamber 1-1,000 k mR/h, gamma
-(gas multiplication)-
20-20,000 k arad/h, beta 5.
Geiger Tube (telescopic) 1-1,000 k mR/h, gamma 6.-
Scintillation Counter -
0-500 k cym, gamma Sodium Iodide (T1) 7.
Neutron Ren Meter (BF )
0.54,MO mRom/h, neutron 3
8.-
Alpha Survey Probes 200-1,000,000 dps; alpha l
(gas proportional and ZnS) 9.
Portable Air Samplers 0 8 cfm 70I/54 g, g, 3.0 p.,,
Lisease E. SNM-960 gg g, 3-1
.3/27/89
J f.
3.2 FIXED MONITORING EQUIPMENT
, ! Listed below are types of equipment installed for monitoring quantities or concentrations of' radioactivity.
, Air samplers and monitors utilizing CM, proportional, scintillation, and
.o semicondue?:or detection modes with moving and fixed filtering units which are cepable of alarming at air concentrations equivalent to MPC's in less than four hours for most of the commonly encountered radio.
'H Fixed filter units consist of 47-or 50-mm-diameter filters l
isotopes.
and constant flow control regulators. Stack sampling and monitoring units include isokinetic probes with GM, proportional, scintillation, semiconductor and/or flow-through ion chambers and appropriate filter media.
6
. Fixed gamma monitors with ranges from 0.1 mR/h to 10 R/h are' located in o
areas with potentially hazardous gamma fields, Hand-and-shoe counters and/or hand held probes are provided at principal l
o exit points for beta-gamma and alpha as required.
A whole body counter (shadow shield principle) utilizing a 5-inch by o
5-inch NaI crystal is capable of detecting 0.01-0.1% of the maximum permissible body burden of several common gamma emitters and 14 for most other gamras emitters.
Environmental surveillance is provided by a number of TLD dosimeters o
located on the VNC site and at its perimeter. Four permanent environmental air sample stations also are located on site.
1 l
l 70-754 3.0 License W. SNM-960 ggg gg p,p 32 3/27/89
t y
3
-3.3 FROTECTIVE CIDTHING r
Protective clothing is provided to assure the necessary protection of 1' ' personnel. The amount and. type of protective clothing required for a
~
' specific acaivity.or area are determined on the basis of contamination
.)
potential.'.Available protective clothing includes caps, laboratory coats, coveralls, boots, overshoes, shoe covers, gloves, and respiratory protection, either filtered or independent air-supplied types. Other special protective equipment is available for use from time to time.
1
..l FIIJi.RADGES AND POCKET DOSIMETERS 3.4 Film badges are worn where there is potential for radiation exposure.. Albedo.
[
. neutron dosimeters are used where appropriate. Pocket ionisation chamber dosimeters may be used.. If. pocket dosimeter results indicate an off-scale or I
unexpectedly high reading, the badge is processed; and if these results are confirmed, the circumstances are investigated and the individual is removed from radiation work if appropriate. TLD extremity dosimeters are worn when radiation exposure to the hands is expected to exceed total body exposure to a significant degree. Film and TLD dosimeters are serviced by a concaercial vendor.
i 3.5 RADIOACTIVE WASTE FACILITIES s
The wastes contcining the most significant quantities of special nuclear material are items such as irradiated fuel specimens which have been examined and analyzed at.the Building 102 hot cell complex.
Some of the wastes are
-delivered to a licensed waste contractor while others are stored at the Hillside Storage complex. Other radioactive wastes ordinarily contain only
.l gram or milligram quantities of special nuclear materials. Facilities used
~
for waste storage and handling at the site are specified in the sections which follow.
SNM 960 70-754 3.0 Liesene No.
ggg g,g p,p 3-3 3/27/89
3.5.1 Drv Wastes Dry contaminated wastes are placed in sealable drums, tubes, boxes, or casks
.available at each laboratory or facility where such wastes.may be generated.
Each laboratory or facility maintains a designated area for temporary waste
-storage. Dry wastes are transi,rred to Building 102 for final inspection and
]
[',
any necessary repackaging. Wasta packages then are transferred to the site radioactive material storage facility, Hillside storage.
)
f
.. The facility has a horizontal tube facility for storage of high-level
)
radioactive material contained'in sealed encapsulations called " waste f
liners. The horizontal tube facility is made of two rows of 40-foot-long a
concrete-lined steel pipes mounted horizontally and covered with earth.
Eleven of the tubes have a 6-incu inside diameter, and seven have a 10-inch inside diameter. The tubes in either row are spaced on 3-foot centers, and the rows are spaced 3 feet apart.with the tubes in the bottom row offset halfway between the tubes in the upper row. Shielding is provided on the top and sides of the facility by a minimum of 6 feet of compacted earth.
Shielding at the front and back ends consists of 3 feet of concrete in which the pipes have' been anchored, plus concrete-filled step plugs with a minimum of 3 feet of concrete shielding in the plug. Additional above-ground space for lower level waste or other materials is available within this fenced and posted facility.
3.5.2 Liauid vastes 1.iquid wastes are routed from laboratory sinks and gravity drains leading froe sources known to be or potentially contaminated through regulated pipe lines to retention tanks located ir, each building where such wastes are generated. Such wastes are transferred periodically to a waste treatment plant for concentrating and solidifying the liquid wastes which are described l
in Section 13 of this application.
l 70-754 3.0 Liessee W. SNM-960 g, g, g g,
- p.,
Dee, _
Amends Sect.(s)
Othar liquid waste (excluding sanitary waste) flows through a separate piping system into any three of four 60,000-gallon retention basins. After sampling j
and determining that radioactive contamination, if any, is within permissible I
discharge levels, the water in the basin is released.
Sanitary wastes are treated, and the waste waters are sprinklered on site.
3.6 EMERGENCY EQUIPMENT A vehicle is available to Radiation Safety and can be equipped quickly with a supply of protective clothing, first aid equipment, respiratory protection equipment, and portable instrumentation and sampling equipment for use during emergencies.
Emergency equipment also is stored in selected areas on site.
3.7 INDUSTRIAL SAFETY EQUIPMENT In conjunction with the radiation safety program at VNC, industrial health and safety of VNC personnel also is emphasized.
Some of the major protection facilities and equipment which are available include portable extinguishers, sprinkler systems, a dispensary attended part time by a registered nurse, and l
a wicte range of typical industrial safety equipment.
l 3.8 CRITICALITY ALARM SYSTEMS In any Vallecitos Nuclear Center area in which special nuclear material containing more than 500 grams of U-235 is used or stored and does not l
otherwise qualify as a "suberitical area" as defined in Section 3.14 of Appendix A, a monitoring system, including gamma or neutron-sensing devices which will energize an audible alarm in the event of criticality, is installed and maintained. The system in use on site is described in the following paragraphs.
License No. SNM-960 Docket No.70-754 sect. No.
3.0 Page 3-5 Date 3/27/89 Amends sect.(s)
/ i J
93-
,m 3. 8.1.
LII crvatal system
! t '.
This monitoring system consists of a lithium-iodide crystal coupled to a-
.photomultiplier tube. Failure of any detection circuit component which would
-Failure-prevent criticality detection activaties a warning light on the unit.
of any signal producing component-is detected during the monthly test.
i;
'The alarm level of the system is pre-set at not less than 5 mrem /h nor'more i
'than'500 mrem /h (this setting is an exemption from 10CFR70.24 contained in License SNM-960). The system is capable of energizing the alarm when the
' radiation' level.at a distance of 1 foot from the special nuclear material.is -
0 300 R/h or 2.1 x 10 neutrons,per square centimeter per second.
Sensing devices are pocicioned within 120 feet,. air equivalent, of every-required location where special nuclear material is handled, used or stored.
The system is tested by exposing the detectors to appropriate sources and sounding the alarm monthly. The alarm system is designed so that the alarm continues to sound until reset. The alarm is clearly audible in all locatic.ns where radiation exposure may result from an accidental criticality-If a facility does not have emergency backup power, all movements incident.
of SNM are suspended during a power failure.
3.8.2-Excented Areas Criticality sensors are not provided below the surface of the water in the i
I RML storage pool or within the RML cells, nor in the horizontal tube solid i
waste facility. Shielding surrounding the special nuclear material is as follows:
RML Storage Pool 16 feet of water RML Hot cells 1.5 to 3.0 feet of high-density concrete Horizontal Waste Storage 5.5 feet of compacted earth SNM-960 70-754 3.0 Lisease No.
y, g, g,g g, p,
Dee.
Amends Sect.(s) e
l 3.9 ANALYTICAL 1ABORATORY COUNTING EQUIPMENT AND CAPABILITIES The following is a suannary of the normal capabilities of the analytical laboratory counting room for radiation safety samples.
Samnle Tyne Instrument Detection Limit
- Air and Exhaust Alpha Proportional 3x10"f5pCi/cc Stack Samples Beta Proportional 7 x 10' pCi/cc l
2 x 10~13 pCi/cc-Charcoal Cartridges (I-131)
NaI(T1)
~
Smears Alpha Proportional 7 x 10 pCi Beta Proportional 3 x 10' pCi
~
Water (Retention Basin)
Alpha Proportional 3 x 10 pCi/cc Beta Proportional 5'x 10'8 pCi/cc l
- Based on standard sample size and counting times.70-754 3.0 p,p Llosaae No. SNM-960 gg g, g, g, 37 3/27/89 DWe Amends Sect.(s)
4.0 RADIATION PROTECTION PROCEDURES A system of Vallecitos Safety Standards establishes the site radiation and criticality protection and regulatory compliance programs.
The manager of the Nuclear Safety component issues the standards with review and comment of the managers of the major organizational components located on the site.
Currently, there are about 50 standards dealing with radiction protection matters.
The principal features of these are summarized belo-.
4.1 PERSONNEL WORK RULES Requirements are established to prevent or minimize the hazards of radioactivity and radioactive materials.
Eating, storing or preparing food, smoking, or etoring tobacco are not permitted in areas where there is a potential for contamination with radioactive materials.
Food containers may not be used for storing or handling radioactive materials.
General Electric furnishes protective clothing for service in areas Where s
contamination is likely to contact personal clothing. Protective clothing standards are set by the site Radiation Safety function to assure effective quality, positive identification, and to avoid use for other than its intended purpose. The amount and type of clothing for any specific activity are assessed on the basis of potential for personnel contamination.
4.2 LIMITS OF RADIATION IN CONTROLLED WORK AREAS All Vallecitos locations Where there is a potential for radiation exposure are classified (radiation area, high radiation area, etc.) in accordance with the definitions of 10CFR20, Sections 20.202 and 20.203.
General Electric's philosophy of protection is to keep radiation exposure at the lowest reasonably achievable level in all cases.
License No. SNM-960 Docket No.70-754 Sect. No.
an Pope 4-1 Dee 3/27/89 Amends Sect.(s)
'b With respect to operations which could produce airborne radioactive N
' contamination, managers of facilities are responsible for providing ventilation equipment to meet the concentration limits of 10CFR20.103 without the necessity, or credit, for personal respiratory devices during routine operations. In certain nonroutine situations where adequate ventilation equipment would be impractical' or could not ensure control of airborne insterial, respiratory protection of demonstrated integrity is utilized.
The respiratory equipment currently in use'at VNC is approved by the National Institute of Occupational Safety di Health (NIOSH) and, as such, achieves compliance with 10CFR20.
MSA Clearvue facepiece or Ultraview facepiece with a NIOSH-approved a.
cartridge, or the MSA Ultra-Twin facopiece with a NIOSH-approved cartridge, or equivalent apparatus may be worn where concentrations do not exceed 50 times the concentrations of. Appendix B, Table I, If radioactive iodine or other radioactive gases are present in 4
concentrations in excess of concentrations of Appendix B. Table I, a similar NIOSH approved respirator may be worn, or an air-supplied airline respirator may be substituted for the air-purifying type.
b.
NIOSH-approved plastic hoods with positive pressure, continuous flow air supply may be worn where concentrations may exceed 50 times the
^ concentrations of Appendix B, Table I.
U.S. Divers Company's Surviveir self-contained breathing apparatus c.
'(SCBA) or other NIOSH approved SCBA having a backpack air supply, hose, harness, pressure-demand regulator, and a facepiece may be worn where concentrations exceed the limits of protection provided by air-purifying and airline respirators and for situations in which airborne
-concentrations have not been evaluated.
VNC has adopted only equipment that is approved by NIOSH.
70 754 4.0 L
License W. SNM 960 gg g, g,g g, p,
Dope Amends Sect.(s) a_-_-_-_-__---_-___-_---__
l
>s
[
P No individual will be' permitted to work more than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> per week under I-conditions requiring _ masks.. No' individual.will be permitted to work in a.
i
. mask until he has received a medical ~ clearance for respirator use and has been thoroughly instructed in methods of proper use, fitting, and field tasting of respirators.
y' j'
^
14.3 PERSONNEL MONITORING Instructions for the use of film badges, finger TLD dosimeters, and pocket dosimeters include the proper part of the body on which the device is to be worn and procedures to prevent spurious readings.
In addition, personnel are
~
instructed to use monitoring instruments upon leaving a radioactive materials Handiand-shoe' counters are also provided at some locations.
area.
In order to determine the extent to which individuals may have' internally accumulated any radioactive materials, two bioassay procedures are established. Persons who work routinely with radioactive materials.
particularly alpha and beta emitters, are tested by the urinalysis method approximately quarterly but at-least annually by this means,-or more
~ frequently depending upon the potential for exposure to mechanisms which could cause ingestion, inhalation or absorption of such materials. Spscial' analyses for other radioactive materials'are performed whenever a need is indicated.
Persons who work routinely with radioactive materials which are readily
-)
' detectable gamma emitters may be tested additionally, or alternatively, by the whole body counting technique. A whole body counter is operated by the Nuclear Safety function. These tests are perforised at least annually. More frequent schedules are maintained for personnel who routinely handle radioactive materials. The whole body counter technique provides the advantage of producing results more accurately and more promptly than urinalysis and, therefore, is used wherever practicable.
4.0 p.g.70-754 g g, Lleense W. SNM-960 g g, 4-3 3/27/89
(
.The Nuclear Safety function is advised of all additions to the work force,'
changes in individual assignments, termination of employees, and any occurrence which may have resulted in internal deposition. This procedure assures that' appropriate schedules are maintained for biological assay.
4.4 SURVEYS Surveys to assure radiation safety are made routinely in order to datect'any
. unfavorable trends or conditions. Special surveys also are conducted as warranted by the suspected or potential presence of radiation or radioactive I
material. Routine survey. schedule.s are established whenever work programs
~
are. initiated and are changed when the work is discontinued or the scope of
. work is revised.. Survey methods are formulated to meet tho' needs of the
. particular type of radioactive materials used and in the light of the equipment capability. Original survey reports are retained by the Radiation Safety function with copies. to the supervisor of the surveyed area for information and action as appropriate.
Environmental surveys are made both within the VNC site and at off-site
. locations. These surveys involve samples of surface and ground water, soil, vegetation, and air and are used to verify compliance with the airborne l
concentration limits of 10CFR20.
The allowabis stack discharge dilution factor, i.e., the ratio of the affluent concentration in a stack to the concentration at the site boundary, for any given discharge stack on site is a function of the actual annual meteorological conditions, the stack flow rate, and the distances from the stack to the site boundaries. Technically, there could be an allowable dilution factor for each radioactive affluent stack on site; but for simplicity and conservatism, the building 102A stack, which has the largest potential source term and the highest flow rate, is used to determiae a single allowable dilution factor for all SNM 960 licensed stacks on site.
4.0 70-754 g, g, pop SNM-960 W h.
Liaones E.
44 l
3/27/89
To calculate the worst-case airborne affluent concentrations at the_ edge of
'the site in the event of a release, it is assumed that all releases are from i
ground level during moderately stable (Type F) atmospheric conditions with a one-meter per-second wind speed. No effective stack height is assumed because of limited stack heights and their proximity to buildings. The wind is assumed to blow toward the point of interest during the release. Using calculational methods and data outlined in Application Amendment No. 48 to Docket 50 18, HW SA 2809, and Regulatory Cuide 1.3, a concentration of 3.6 x will result from an average emission of 1 pCi/sec from a l.
10"?pci/cas 360 meter distance source.
Notwithstanding the above calculations, all effluent air from operations involving potential particulate activity is filtered; and, therefore, special l
nuclear material concentrations have been and are expected to continue to be routinely well below permissible levels for unrestricted locations at the points of discharge from the stacks.
Routine calibrations of survey instruments are performed as follows:
Alpha calibrations by use of sources which are traceable to plutonium a.
and t.tanium standards prepared by the National Bureau of Standards, b.
Beta calibrations by use of a slab of natural uranium in radioactive equilibrium with Th-234 and Pa-234.
I Neutron instriment calibrations are performed by measuring a response to c.
a pulse generator and to an Am-Be neutron source.
d.
Camma calibrations using Cobalt-60 sources standardized with a meter f
which, in turn, was calibrated with traceability to the National Bureau of Standards.
All radiation monitoring instruments are calibrated as frequently as deemed necessary to assure reliability during use. Portable radiation monitoring instruments are calibrated on an annual basis, before initial use, and after l
repair.
4.0 p,p 70-754 g, g, SNM-960 gg g, l
LleenseHe,
- Dee, Amends sect.(s)
l J
Stack particulate monitor systems are calibrated by placing a uniformly v'
distributed radioactive source in the same geometry as the, filter paper used for collecting particulate. Stack gas monitor systems are checked routinely by placing a reference gamma source on the side of the Kanne chamber and observing whether the response falls within prescribed limits. Calibration with a known radiogas standard has been performed to verify this procedure.
l Iodius monitors also are source checked routinely.
4.5 POSTING AND 1ABELING Instructions are established implementing the posting and labeling requirements of 10 CFR Part 20. Additional precautionary signs may be utilized to meet special requirements and detailed procedures. The area supervisor is responsible for maintaining the proper posting and labeling.
However, in view of the degree of control inherent in the security of the site and the required prerequisite training program for employees, there is an exemption from the provisions of 10CFR20.203 for high radiation area alarms which Caneral Electric deems 1.hould be continued.
An exemption from the alarm provisions of 20,203(c)(2) is requested for any area or location which is:
a.
(1)
Used for the temporary placement of strong, weatherproof containers such as shipping casks not exceeding the limits of 49CFR73.393, i.e., 200 mR/h at at.y point of readily accessible surface; or (2)
An established area in which there may be radioactive materials in containers, the readily accessible surfaces of which do not exceed dose rates of 1,000 mR/h at any time and which do not exceed 100 mR/h at the barrier for any continuous period greater than 30 days, such as radioactive waste pickup areas from which wastes are collected on a weekly or monthly basis; and is 1
1 SNM 960 70 754 4.0 Lleense No.
gg p,p 4-6 3/27/89 Owe Amends Sect.(s)=
L_--__---
s b.
Incated within che 94-acre inner fenced exclusion area of the site; and is 1.
Bounded by a distinctively colored rope or chain providing a barrier at c.
dose rate values not greater than 100 mR/h and posted as e High Radiation Area in accordance tach 10CFR20.203(c).
4.6 WASTE DISPOSAL Detailed procedures for the packaging, storage and removal of contaminated material which no longer is useful are established by operating components and reviewed by the Nuclear Safety function. The procedures define low, high, and intermediate levels of solid waste on the basis of a contact dose rate at the surface of outer packaging and provide specifications for container packaging to prevent loss of contents, repackaging of damaged units, labeling of contents and similar requirements. Procedures for handling or disposal of liquid wasi.es in the various waste treatment facilities described in Section 3 are detailed in similar instructions. Area supervision is responsible for adherence to proper handling procedures, for obtaining approved containers and for arranging for transportation of wastes to the appropriate site storage or treatment facility.
4.7 REPORTS AND RECORDS Records of surveys, personnel exposure records, and other records indicating the degree or nature of individuals' exposure are maintained by the Radiation Safety function. These records currently are retained for an indefinite period.
l l
i 70~
- 4.0 Lisease No. _ Sm-MO Dedet No. _
Sect. No.
p, 4-7 3/27/89 m,
m
0 4
- .4.8 INDUSTRIAL SAFETY PROGRAM E
. The effectiveness of the radiation and criticality safety program is in large measure influenced by a sound industrial safety system, and vice versa. Good housekeeping' practices, modern fire prevention equipment, and training in handling toxic or explosive chemicals are but a few benefits of industrial health and safety which carry into the nuclear areas.
For that reason,
' compatible program philosophies and policies are maintained. General Electric places direct responsibility for safety on the individual employee and for directing safe practices on his area manager.
2 0the'r important element.s include pre-employment medical examinations; medical
.]
. surveillance of personnel; control of flammables, explosives, and toxic
- materials;.and administration of fire prevention equipment requirements in
-accordance with applicable codes and regulations.
'4.9 VALLECITOS EMERGENC
Y. PROCEDURE
S
' Plans foi the prompt and rapid response to emergency situations are set forth-in the emergency procedures. The procedures are~ revised from time to time to reflect changes'in equipment and organization at the site. The procedures address emergencies which could arise from accidental criticality or release of radioactive material.
.4.9.1
-Nuclear Emergencies The accidental criticality alarm is a distinct sound which is activated
-automatically when criticality detection instrumentation in that local area l
reaches a pre-set limit. Personnel in the affected area immediately evacuate.
I i
License No. SNM-960 Decket No.70-754 Sect. No.
4.0 Pege 4-8 Doce 3/27/89 y, g,,, g
I' w
At the same time, the origin of the signal is designated at the main security building., The site-wide public address system is used to instruct all personnel at the site concerning further action. Upon the sounding of any alarm, an assessment is made of the indicated or reported situation and whatever action is deemed necessary to minimize personne1' injury and property damage'is taken. A dedicated telephone network can be used for conference-contact with VNC site managers representing most fields of technical competence'at the site,-with the person reporting the emergency, and with the
~
security building.
If the nature and severity of the emergency requires evacuation, either complete or partial, instructions are broadcast.
Personnel who have been trained in the use of survey instruments and protective apparel are available. Guidelines to assist management in formulating dose limits for emergency exposures for the protection of human life, recovery of victims, and protection of health and property are available. A system which has undergone a criticality accident will be left undisturbed until competent review has produced a plan to cope with the situation. Calibrated instruments are available at strategic locations.
Injured persons, if any, may be taken to the hospital in Livermore.
Approximate travel time to the hospital is 15 minutes. Arrangements have been made with the hospital to receive and care for injured persons who are contaminated, with supervision by competent Nuclear Safety personnel.
In the event of spurious alarms, the all clear will be announced on the public address system.
4.9.2 Non-nuclear Emergencies Other alarms (automatically and manually operated) are used for fire or other non-nuclear emergency.
The senior supervising person at a facility will decide when an emergency exists and will take appropriate action for the safety of people and equipment.
l I
License No. SNM-960 Decliet No.70-754 Sect.No.
4.0 Pege 3/27/89 Amends Sect.(s)
Dete _
I 4.10 RADIATION SAFETY REVIEW The design of-new VNC facilities, or major changes in existing facilities in which radioactive materials are to be handled, must be reviewed by the Radiation Safety function to insure that adequate health protection facilities are provided. Policy design and operating elements are considered in this review. Materials containment, exposure reduction (both operating and maintenance), design life, and assessments of potential hazards are used in the final design phase whereby such hazards are engineered out.
The Area Manager concerned, or his designated representative, is responsible for initiating requests for. radiation safety review of proposed facilities
.and radiation safeguards.
I License No. SNM-960 Decket No.70-754 Sect. No.
4.0 Pee
- 4-10 Dete 3/27/89 Ms Sect.(s) l
i9, f
5.0 CRITICALITY SAFETY FROGRAM TECHNOLOGY
5.1 INTRODUCTION
Administrative responsibilities for implementing the utiticality safety program at the Vallecitos Nuclear Center (VNC) are given in Section 2.0 of
.this application. The responsibility of qualified personnel to maintain activities at all times within specified limits is=amphasized.
the.
In making a criticality analysis, the two-contingency criterion is used:
occurrence of at least two unlikely, independent, and concurrent changes in one or more of the conditions originally specified as essential to nuclear safety is required before a nuclear accident is possible. All analyses receive an independent review.
An organizational component competent in neutron physics and the technical methods for criticality computations shall verify the use of any new computational methods for criticality control.
In the discussion below, information about computational tools useful for performing criticality analyses at VNC is presented. Sources of nuclear parameter data useful in performing hand calculations are discussed.
Critical experiments useful in evaluating the computational tools'for criticality analyses are identified. Any negative biases that must be applied are identified. Finally, the adjustment factors for normal operating conditions are presented.
I 5.0 p,
SNM-960 70-754 g, g, LleenseMs.
g, g, Dee.
Amende Sect.(s)
I L
5.2' MINIMUM CRITICAL QUANTITIES AT ISOLATED LOCATIONS 1
When only small quantities of special nuclear materials are required at any one time within an isolated facility, accidental criticality hazards are easily and definitely avoided by control of inventory to quancities substantially lower than the minimum quantities of such materials that have been experimentally found to be capable of criticality under optimum laboratory conditions.
As used in this section, the term " isolated" means physically separated from other areas where special nuclear materials are used. The procedures for accounting for inventories of SNM normally will be under specifically designated custodians. The minimum neutronic isolation barrier is taken as the equivalent of 12 inches of water or a distance of 12 feet, or the greatest distance across an orthographic projection of either accumulation or array on a plane perpendicular to a line joining their centers.
5.3 DETERMINATION OF CRITICAL PARAMETERS OF ACCUMULATIONS AND ARRAYS OF FISSILE MATERIAL 5.3.1 Eggerimental critical Data Experimental critical data are available in the literature from which critical parameters for fissile materials such as mass, sphere volume, cylinder diameter, slab thickness, and concentration may be determined.
Data are often presented so that values of critical parameters may be read directly from curves or tables. Some documents from which critical parameters may be obtained are:
a.
Karlsruhe Symposium papers, 1961, b.
A.H.S.B. Handbook 1, " Handbook of Criticality Data".
c.
TID-7028, " Critical Dimensions of Systems Containing U-235, Pu-239 and U-233".
License No. SNM-960 Docket No.70-754 Sect. No.
5.0 Pe9*
Dete 3/27/89 Amends Sect.(s)
~
i L_____._..__.___
I
7 fa
.d.
TID-7016, Rev. 1 and Rev. 2, " Nuclear Safety Guide".
e.
1AMS-3067, "Los' Alamos critical Mass Data".
.- f.
IAMS-2537, " correlations of Experimental and Theoretical Critical Data".
- g. -
Y-1272, "Y-12 Plant Nuclear Safety Handbook".
.h.
ARH 600, "C.:iticality Handbook".
i.
Oak Ridge Criticality Data Center Report, ORNL-CDC-5.
R j.
DP-1014, " Critical and Safe Masses and Dimensions of Lattices of U and UO Rods In Water".
2 When the geometry of the accumulation is other than a simple one, data obtained from this literature may be used to determine critical parameters by-E using simple buckling conversion methods or by using' approximate, but conservative, values for the accumulation dimensions.
q 5.4 ADJUSTMENT FACTORS FOR NORMAL OPERATING CONDITIONS Criticality analyses shall utilize calculative methods that have been evaluated by the calculation of models of critical experiments of similar geometrical configuration and material content; the maximum negative bias in the effective multiplication factor, defined as (-Akbias) - calculation
~
k,,p,,
g, determined in the evaluation shall be applied to the' analysis as given in the relationship below. The effective multiplication factor of the l
accumulation or array of accumulations shall satisfy the relatic. whip:
[1
+ I'A bias) s 0.95 k,,7.+ Akuncertainty 5.0 p.,.70-754 Sed. W.
Licenae W. SNM 960 Dede W.
5-3 I
3/27/89 Amends Seet.(s)
Dee i
where:
k,,7 is the calculated offective multiplication factor by the l
is the statistical uncertainty (2a) in the calculative method, Ak,,,,,,,g,gy mean value of the calculated effective multiplication factor, and (-AQ,)
l
' is the maximum negative bias determined from the validation of the calculative method. If the bias is positive, the (-Akbias) term shall n t be used, so that overstatements in the effective multiplication factor shall not l
be corrected.
I Normally, subcritical values of nuclear parameters for individual accumulations under normal operating conditions shall be as shown below.
However, in no case shall the effective multiplication factor exceed 0.95.
l For accumulations limited by mass, the normally suberitical mass shall l
a.
not exceed 74% of the critical mass.
b.
For accumulations limited by volume, the normally suber!tical volume shall not exceed 76% of the critical volume.
For accumulations limited by dimension, the normally suberitical c.
cylinder diameter or slab thickness shall not exceed 90% or 884, respectively, of the critical dimension where U-235 is the fissile constituent, nor 856 in all other cases.
d.
For accumulations limited by concentration, the normally suberitical concentration shall be less than that concentration which is equivalent to the following atomic ratios of hydrogen to fissile material:
H/U-235, 5200; H/U-233, 7600; and H/Pu (fissile), 7600, Not more than 45% of the critical mass shall be contained in any e.
accumulation in a criticality limit area in which double batching is credible.
Normally, suberitical values of nuclear parameters of interacting accumulations shall be:
Liesrae W. SNM-960 Desket W.,70-754 Sect. W.
sn Pop 5-4 Deee 3/27/89 4,,,g, g,,,,(,)
1 l
Not more than'69% of the critical number of accumulations whenever the l
a.
critical miraber has been determined by experiment; or q
b.
Not more than 50% of the critical number of accumulations in any array calculated by the density analog method (Ref: ARH-600); or l
The number of accumulations datermined by the solid angle method as i
c.
described in TID-7016, Rev. 2 Paragraphs 4.29 through 4.34.
The method may be applied only to accumulations that are well moderated (e.g.,
aqueous solutions) at the point of the maximum effective multiplication l
factor, and array reflection avst be no more effective than a thick l
l water reflector spaced at distances from the accumulations comparable to the spacirg between accumulations, a
1 I
Lieense No. SNM-960 Decket No.70-754 Sect. No.
sn Pope I
5-5 g
3/27/89 knds Sect.(s)
___,L, 6.0 1ABORATORY BUILDING 102 I
6.1 IDCATION AND GENERAL DESCR1.
..;N Building 102 is.a single story with basement, concrete and steel structure located as shown in Figure 6.1.
The predominant feature of the building is the multikilocurie facility known as the Radioactive Materials Laboratory (RML). General support laboratories for these facilities also are located in the building. The current layout of the building is shown in Figure 6.2.
The laboratory areas of the main floor are separated from general office areas by a fire wall. -Inherent in the building design are general service facilities such as the main ventilation system, decontamination rooms, and other areas which are described later in this section.
l t
6.2 VENT 11AT10N SYSTEM 6.2.1 Air Sunniv l
Inlet air to Laboratory Building 102 is provided by 30 air conditioning units furnishing modulated, filtered and tempered outside air to the operation and office areas. The capacity of the inlet air system is a nominal 65,000 cubic feet per minute.
6.2.2 Direction of Flow Airflows in the laboratory are from areas of low radioactivity toward and through areas of higher radioactivity. The arrows in Figure 6.2 indicate a
general direction of flow. The system of directional airflow minimizes the possibility of accidental contamination of nonradioactive areas. Airflows are all single pass and continuous except a portion of the RML operating gallery air which is recirculated through absolute filtsrs.
Linease No. SNM-960 Docket W.70-754 Sect. No.
6o Pepe 61 g
3/27/89 w, g,,,,(,)
n 4
The operating areas are supplied at rates of 6 to 40 changes per hour.
A minimum of six air changes per hour is being exhausted from the RML cell This amount of airflow to the cells is adequate for control of spaces.
contamination.
6.2.3 Pressure and Flow Rates small static pressures (-0.01 to -0.03 inches of water) are used in the isolation of thin walled general work areas, while static pressures from 0.02 tc 0.20 inches of water are used-to isolate the RML cells. The flow diagram as shown in Figure 6.2 indicates the airflow. Glove box contamination control is effected by maintaining from 0.5 to 1.0 inch of water negative static pressure. At this static pressure, air velocities of 125 linear feet i
per minute or greater are achieved through openings, an adequate rate to provide the necessary control.
6.2.4 Filtration and Exhaust The main exhaust equipment is located in' Building 102A (Figure 6.1).
During regular working hours and.any other times radioactive material is handled, two of six exhaust fans are'used to provide a normal exhaust flow of approximately 43,000 cfm with the other fans in n standby status. Exhaust air from Building 102 is routed to Building 102A via overhead ducts..At Building 102A this exhaust air is passed through 90 HEPA filters and then l
discharged to the atmosphere through a 66-inch-diameter, 75-font-high stack.
Exhaust air from the RML Hot Cells is prefiltered at the outlet of each cell and is routed to a filter bank of 10 HEFA filters in the Building 102 basement before it joins the main exhaust stream at Building 102A. Two activated charcoal filters are available as needed. Two booster,?ans are available as required to aid in the transfer of the RML Hot Cel effluent from Building 102 to Building 102A.
Lleense E. SNM-960 Ceeket W.70-754 3.es. m. _
r, 0 Pese 62 pee.
3/27/89 w, g,,,,(,)
4 The filter banks and the exhaust connections, at their point of origin, util'.se high-efficiency filters. The filter system is 99.954 efficient for.
0.3-pe-diameter homogeneous particles of dioctyl sebacate. The filters are constructed of fire resistant materials and are housed in noncombustible duct work.
The exhaust connections have acijustable dampers where necessary to provide the capability for belancing flows throughout the system.
i Figure 6.3 is a schematic diagram of the ventilation exhaust systsm.
l Fire protection'is provided to the exhaust system by: '(1) a CO 8uPpression 2
. system and a water fog system for the Het Cell basement exhaust filter bank; 3
f (2) a water spray system for the Hot Cell charcoal filters; and (3) a water fog spray for the main exhaust ducting from Duilding 'A02.
None of the water l
suppression systems can discharge into facilities (e.g., Hot Cells, glove boxes) containing SNM.
i 6.2.5 Emergency Exhaust I
in the event of the loss of utility power, emergency power is provided by a 335-kW diesel-driven electric generator. The generator is activated automatically by power loss and is designed to reach full capacity within one l
minute. #di critical equipment normally supplied with normal power through the main building switchboard will be supplied with emergency power. This equipment includes two exhaust fans, fans supplying air to areas where radioactive materials are handled, and the main stack monitoring equipment.
I l
L The generator is supplied with fuel from a 1,000-gallon storage tank.
At sn l
estimated fuel consumption rate of 20 to 25 gallons per hour, this supply should be sufficient for at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of continuous, full-load operation.
I License No. SW-960 Dssket No.70-754 Sect. No.
6.O Pepe j
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The building alarm panel and exhaust far.s are supplied with normal power through the Motor Control Center (MCC). The solid-state logic portion of the MCC is a solid state, battery powered computer that provides for automatic switching of equipment during transitions between normal and amergency conditions. During normal and emergency power operations, it performs the I
following functions automatically:
If one of the operating fans should fail, it will provide an audible and l
a.
visual alam; identify and shut down the failed fan; and start a standby fan. The control panel will indicate which fan has failed, b.
In case a fire is identified by a detector in the exhaust ducting, the computer operates the exhaust system on reduced flov until the " fire" l
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condition is cleared manually.
4 6.2.6 Etad Monitorine The Building 102 stack monitoring system was designed for the monitoring of the primary isotopes releared during normal operations in Building 102 aN 102A. The system has been designed to detect beta-gamma particulate and j
gaseous activities. Each detector output signal is transmitted to a processor /ricorder which processes the data into a desired format and The flow schematic of displays.information on a recorder printout or trace.
this system is shown in Figure 6.4.
Noble G g,,fonitor 6.2.6.1 A
The The flow-through ion chamber is a detector to measure activity in gases.
test gas flows between the outer pair of electrodes, called the " scrubber",
I to which a potential is applied. Here, ionization from previous decays is cleared from the gas. The uncharged test gas then passes into the 16.3-liter f
activs volume. The ions produced by any radioactive decay occurring in the active region drift to the collecting electrode. The resulting current is j
measured by the electrometer. The sample flow rate is 3 cfm.
~7 The response for Xe-133 is 1.45 x 10 sup/pci/cc.
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The beta-gamma particulate monitor is utilized to evaluate the total amount of beta particulate activity and the total amount of a pre selected gamma-emitting isotope in a sample airstream withdrawn from the stack. The sampled air.(1.5 cfm) passes through a 47-mm filter which will remove l
i approximately 99.9% of all particulate greater than 0.3 pm in diameter.
The beta-gamma activity collected on the particulate filter is monitored by a 2-inch-diameter pancake-type GM detector having a window thickness of 7 to 8 2
l mg/cm which is located approximately 1/2-inch from the face of the filter.
The CM dete-ctor portion is calibrated with a Cl-36 beta source for whicht the efficiency is approximately 35% (4w).
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6.2.6.3 Permanent Records j
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The 47-am filter and a charcoal cartridge are sent to the site counting room for analysis (alpha, beta-gamma and iodine). These analyses provide the basis for the permanent stack effluent emission records.
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6.3 DECONTAMINATION ROOM A decontamination room is available for cleaning tools and equipment.
Equipment in the decontamination room includes standard filtered airflow-type hoods connected to the building ventilation system, a large stainless steel I
sink draining to the building radioactive waste system, and miscellaneous other items of equipment typical of that used for decontamination work. The room is located in a high radiation area.
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' AIR SAMPLERS The air in varicus parts of laboratory rooms, offices and other occupied areas is sampled on a continuous basis. Approximately 80 sample stations in the RML and radiochemistry areas have been,provided. Air is drawn through each sample station filter paper.
Flow rates routinely are determined using standard equipment for such determinations.
Filter papers are removed from the stations on a routine schedule and counted for alpha and/or beta-gamma activity.
Continuously monitored, alarming air sample stations also are provided in the RML.
In addition, spot check' air sampling is done whenever a new operation is introduced to assure absence of significant contamination in the breathing air; and several continuous air monitors are in use for alpha and beta-gamma l
determination. Alarms are sounded if concentrations increase significantly.
6.5 LIQUID WASTE HANDLING Sinks and gravity drains in Building 102 radioactive materials areas are routed to a filter which removes particulate.
The liquid then is pumped to a 3,000-gallon tank.
The tank periodically is emptied, and the waste liquid is sent to the liquid' waste evaporator facility to be concentrated, solidified, and transferred to a commercially licensed waste disposal contractor. This liquid waste material contains only minute quantitles_of SNM.
High-level liquid wastes are solidified at the hot cell work stations for l
transfer to the waste contractor.
6.6 EMERGENCY POWER In addition to the exhaust system as described in Section 6.2.5, the Building q
102 emergency generator also provides emergency power to the criticality alarms, the breathing air alarm (loss of pressure alarm), the fire alarms, the telephone equipment, and an annunciator panel which transmits these l
alarms to the security building.
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- 7.0 RADIOACTIVE MATynmm IASORATORY 7.1 1DCATION AND GENERAL DESCRIPTION
- The' Radioactive Materials Laboratory (RML) is the term applied to a series of shielded facilities located in Building 102 used for the examination of irradiated reactor fuels and irradiated hardware and radioisotope production, including support and service areas. A plan of the RML area is shown in Figure. 6.2.
Although there is a number of hot cells in the RML complex, special nuclear material licensed under License SNN-960 is handled predominantly in Hot Cells 1, 2, 3, 4 and 5;-and, accordingly, only those cells will be addressed in detail in this section' A sketch of these RML cells is. included as Figure :7.1.
x 7.2 GENERAL PLANS AND' SES OF MATERIALS U
'RML provides a high-level shielded facility in which safe and efficient non nuclear testing and examination of irradiated reactor fuels, materials and components may be conducted. The facility also is equipped for the handling of large quantities of byproduct materials.
l 7.3-RML FACILITIES 7.3.1 ceneral Description Equipment and facilities of the Radioactive Materials. Laboratory are classified into several major areas:
the cell operating area, the general laboratory and cell service corridor, and the cell area. All areas are on n e main floor of Building 102 as shown in Figure 6.2.
These areas are
' described more fully in the sections immediately following.
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-7.3.2 HigkImval calls (Hot cells 1, 2, 3 and 4) y
. Tho' cell area is the heart of the Radioactive Materials. laboratory as it is
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-here that work on irradiated material is performed. A double or-twin cell'
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block for high level work is located on either side of the south and of the service corridor, constituting a total of four high-level cells. Each of 4
these ' cells can contain safely in excess of one million curies of 1-MeV E
gasunas. The high-level cell walls are 36 inches thick up to a height of 12 feet, where they are reduced to 24 inches to take advantage of reduced l~
shielding requirements and to provide a set-back for overhead manipulator The shielding material is a high-density concrete made with
. rails.
ferro-phosphorous aggr*5 ate with a poured weight of 300 pounds per cubic foot.. Shielding in the. vertical direction is provided by the 3-foot-thick
. concrete roof above the cells.
Each cell can be equipped with an overhead bridge-mounted manipulator and a l
- 3-ton bridge crane running on the same rails. The units are designed to g.g traverse the entire length of the cell and the radiation lock and can be moved'out of the cells on removable extension rails. This equipment provides
. handling capability for transfer of irradiated materials in heavy shielded b
. casks.
The. viewing window at each operating station is a 3-foot-thick, lead glass L
window with glass of sufficient density (at least 6.2 gs/ce) to give
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Each of the high-level cells as shown by Figure 7.1 has a 6-foot-long radiation lock for entry from the access corridor. These radiation locks The are formed by hydraulically operated bi-parting steel shielding doors.
outer door is 18 inches thick, and the inner door is 15 inches thick. A a
3.5-foot-diameter by 9-foot-deep pit is provided in the radiation lock floor
.to accommodate very large casks and permit long irradiation assemblies to be
. withdrawn. The access corridor width and floor loading specifications were based on the use of a 15-ton-capacity fork lift truck in this area. An intercell. transfer system is provided in the common wall between each pair of cells..
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l Each high-level cell has an operating area approximately 17 feet long by 6.5 feet wide and is 14 feet high inside. There are four operating stations in each cell. Three of the stations are located along the side wall of the cell and the fourth at the end wall.
H ere is a total of 16 operating stations available. Typically, the elements of each cell operating station are a pair of through-wall master-slave manipulators, a viewing window, and the "in-cell" work or experiment apparatus required for carrying out one or more functional tests or operations. Six to seven feet of width are available at each station internally and externally. This permits adequate space for one or two operators and the miscellaneous control and operating equipment which is required on the cold side, Water is piped to the hot cells by an independent supply system. A 25 gallon vented storage tank located on the top of Cell 1 is filled by a water line not directly connected to the tank. This tank feeds an adjacent 42-gallon pressure tank through a connecting check-valved pump line which, in turn, supplies pressurized water at a maximum of 60 psi to a process water header with branches to each hot cell. This independent system provides positive assurance against feedback of cell water to the potable water system as well as preventing large quantities of water (more than 42 gallons) to enter the cells in the event of piping failure. This system currently serves only l
Cells 1 and 5.
7.3.3 cell Door Interlock system The high-level cells are equipped with inner and outer steel shielding doors which form a radiation lock. The cell doors are hydraulically operated and controlled by a panel at the cell operating face. The operating controls are interlocked such that the outer door normally cannot be opened when the inner door is open. A key-locked override switch is provided for unusual circumstances such as cell decontamination activities.
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'7.3.4' cell 5 i
A smaller cell designed for conducting metallographic work occupies the south
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end of the main access corridor. This cell is used primarily for the preparation of samples for metallographic examination and micro-hardness tasting. Remotely operated equipment for this cell includes sample mounting, polishing, cleaning, and etching equipment as well as a remotely operated
'i micro hardness. tester. - Meta 11ographic samples after polishing and etching can be checked with the optics of the hardness tester before being trans-ferred remotely out of the cell to a modified research metallograph contained in a shielded enclosure. The working area of the metallography cell is approximately 5.5 feet deep, 8 feet wide, and 8 feet high. The front wall contains two 18-inch by 13-inch lead glass windows for direct observation of the working area. The call walls are 18-inch-thick magnetite concrete. The back of the cell', which opens onto the access corridor, contains a safe-type door for equipment and personnel access. Metallographic or micro hardness test samples are introduced to this sample entry port by means of a special
.j transfer cask.
7.3.5 Pool Facility A 16-foot by.8-foot by 16-foot-deep, stainless steel-lined water pool is l
available for underwater transfer, examination, repair, assembly or disassembly, and storage of irradiated materials. Overhead crane facilities are available for cask handling in the pool or on the floor of the room..
A filtered exhaust port is located adjacent to the pool. Dose rates of the surface of the pool are of the order of 20 mR/h at maximum storage capacity.
The pool is equipped with portable types of tools including tongs, hooks, tables, racks, and lights. The water is circulated at about 30 gpa through strainers and a tank containing 6 cubic feet of nuclear grade ion exchange resin, thereby maintaining excellent water clarity and decontam nation.
i Rosin is replenished as necessary. Spent resins are removed to waste drums for disposal.
License No. JNM-960 -
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,,s 7.3.6 In-Floor Dry Stormen Pit Located immediately adjacent to the RML storage pool, a dry storage pit is available for temporary storage of irradiated fuel assemblies, rods or other j
i rod-shaped materials. This storage device consists of 19 recessed pipes l
fabricated of 6-inch Schedule 40 steel pipe 46.5 inches long attached to two horizontal circular steel plates 91.25 inches in diameter and 6 inches thick.
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3 device rests on a ledge cast in the concrete floor so that the ton surface is flush with the floor with the tubes extending downward below the floor surface. The pipes are arranged to provide a minimum center-to-center spacing of 18. inches. Twelve-inch-thick stepped plugs provide shielding of sources up to 500 R/h at 1 MeV to dose rates of 2.5 mR/h.
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Fuel materials are placed in sealed tubes called waste liners prior to insertion into the' storage pipose Movement of the material is accomplished by means of shielded transfer casks which have top and bottom entry ports.
Handling tools which are attached remotely to the waste liner can permit 1
' insertion and withdrawal of stored objects without encountering direct radiation from the top of the open cask. Each tube has two 1-inch-diameter l
holes located 3 inches above the bottor; and one 1/4-inch hole in the bottom plate to prevent accumulation of water, 7.3.7 other cells Cells 6, 9, 10 and 11 (see Figure 6.2), are used in support of research and l
development activities authorized under California License No. 0017 60.
7.3.8 nadiochemiserv u beratory The radiochemistry laboratory, used principally for the analysis of samples of irradiated fuel materials from the RML, is located immediately adjacent to the RML area. Hoods or glove boxes are connected to the previously described 102 ventilation system and are designed to provide a minimum face velocity of l
125 linear feet per minute.
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7.0 Pepe 7-5 3/27/89 Ms bd.M g
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7.3.9 Eervicina Areas and Eauinment The Radioactive Materials Laboratory section of Building 102 also contains several shielded cells used in non-SNM work, a waste analyzing area, an l
equipment decontamination room, a machine shop, equipment storage areas, and a manipulator repair room.
I l
A hot shop facility now occupies the former Plutonium Analytical Laboratory (PAL) next to the radiochemistry laboratcry. The PAL was surveyed completely
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(except for the area behind a metal security wall) and decontaminated following the discontinuance of plutonium operations. The area now is used for the repair of contaminated equipment. Only traces of SNM are present in the contamination.
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7.3.10 Criticality Alarm Sensors The Radioactive Materials Laboratory is monitored for criticality accidents by a detector located at the storage pool area entrance (Figure 6.2).
1 l
7.4 RML CRITICALITY CONTROL SYSTEM Special nuclear materials used in connection with RML operations are principally in the form of oxides in irradiated fuel elements and experimental capsules. The spectrum of material and activity types may be quite broad.
"r is required flexibility has been taken into account in the establishment.sf criticality centrols and is reflected in considerably larger
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l safety margins than might be appropriate to more routine or repetitive situations. For example, each CIA in each fuel examination cell is limited to 45 percent of a critical number of units (fuel rods, assemblies, etc.);
and each fuel examination cell is limited further so that criticality is not possible if all of the fissile material in the cell comes together j
simultaneously under conditions of optimum water moderatica and full water reflection; but normal activities preclude moderation to any degree, and License No. SNM-960 Decket No.70-754 Sect. No.
7.0 Pope 7-6 Dete 3/27/89 Ms Sed.(s)
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.' sources of water to the cells are limited and virtually accident proof.
Criticality controls governing RML and supporting activities are described in
~ the following sections.
7.4.1 Receiving RML work is predominately on the post-irradiation fuel examination work but l
also involves work with various radioisotopes, including encapsulations, and the 'examit.ation of experimental fuel capsules.
Shielded casks used for SNM shipments or on-site transfers have received criticality analyses demonstrating their safety either to receive NRC certification or to meet the requirements of License SNM-960.
Quarantine zones in the radioactive materials storage area (Hillside Storage)
'and the north side of Building 102'are used for incoming shipments from off site awaiting evaluation of the contents. Containero are spaced according to l
DOT and NRC requirements.
Casks of irradiated material also may be received on the Building 102 dock.
l.
7.4.2-S.t.orage Irradiated special nuclear materials may be stored in shielded shipping or
' transfer casks (temporary), in the RML pool, or in the dry pit storage facility..When material is stored in casks, the spacing between casks and the number accumulated in a given area vill be consistent with the DOT
. requirements for safe shipments.
Pool storage is limited to locations which have been analyzed both for the critical parameters of the accumulation of material in each storage location and the interaction between storage of uranium or mixed oxide irradiated fuel rods or fueled experimental capsules. All storage locations are isolated
'from one another by a minimum of 12 inches of water, and the locations have been analyzed for all varying degrees of interspersed moderation. The fuel rods / capsules stered in each rack are limited to 45% of the smallest estimate Lleones No. SNM-960 Decke W.70-754 be. W.
7.0 Pope 7*7 g
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of the initial number of pins with consideration given to the fissile composition, pellet diameter, and enrichment. When two or more types of t
rods / capsules are stored in the same location, the limit for the most restrictive type is used. The pool is further limited so that criticality is not possible even if all the SNM in the pool came together simultaneously.
The dry pit limit is currently 300 grams fissile for each of the 19 positions.
Each position hr4 a minimum edge-to edge spacing of 12 inches.
Fuel rods or assemblies are scaled in aluminum waste liner cans similar in design to DOT Specification 2R containers which exclude water in the "as-stored" situation.
There are no credible means for rearranging the tube spacing.
If the pit were flooded accidentally, even with optimum moderation within the sealed liner units, the fully loaded facility would be safe since each unit would be isolated by 12 inches of water.
7.4.3 Doerations Criticality controls for the hot cells and for cold limit areas are based on the principle of limiting the quantity of fissile materials in terms of 45%
of the minimum critical mass.or number,of units for each limit area.
In determining the minimum critical mass or number of units (e.g., irradiated fuel rods), such factors as the form (e.g., fuel rod, powder, solution),
enrichment, and composition of the special nuclear material and the available moderation and reflection are considered.
The limit area control is implemented by an electrenic data processin5 system. An inventory log is kept for ecch limit area showing the quantity of l
fissile material in each unit transferred to and from the limit area.
Each cell is isolated by 36 inches of concrete and from its interlock by a 15-inch-thick steel door.
License No. SNM-960 Decket No.70-754 s,c,. No.
7.0 Fey
/9 Dete Amends Sect.(s)
The analytical services provided to RML by the Radiochemistry Laboratory necessitate only small samples of special nuclear material in the lab area.
Simple mass control therefore is established. Not more than 300 grams U-235 are permitted within the confines of the Radiochemistry Laboratory at any l
time. If mixtures of these fissile isotopes are present, the lowest limit applicable is the limit for the entire laboratory.
7.4.4 Waste Dianosal Upon completion of hot cell work, fuel-bearing materials and components are sealed.in prenumbered waste liners and removed to one of the storage locations described above and in accordance with the criticality control limits of the storage location or transferred to the Hillside Storage location in a shielded cask.
Other forms of waste materials are handled in previously described Building 102 liquid waste and ventilation systems.
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u 10.0 CBEMISTRY, METALLURGY AND CERAMICS IABORATORY - BUILDING 103 10.1 IDCATION AND GENERAL DESCRIPTION 1
The Chemistry, Metallurgy and Ceramics Laboratory, Building 103, is a two-story building with a partial basement located directly across the access l
l' road from Building 102 (Figure 6.1).
The building has a total floor area of approximately 22,000 square feet, including approximately 11,000 square feet of laboratory space. A corridor runs the length of the building on each floor. The laboratories are located on one side of the corridor and offices on the other as shown in Figures 10.1 and 10.2.
10.2 CENERAL PIANS AND USES OF MATERIAL Special nuclear material is used in research and development activities including analytical, test specimen fabrication, maintenance, and calibration work and for sources of radiation necessary to support research and development programs. At the conclusion of experiments, radioactive materials usually are re-worked, reused, stored, transferred off site to persons authorized to receive, or discarded as wastes.
l 10.3 IABORATORY FACILITIES AND EQUIPMENT 10.3.1 General Facilities located on the first floor include electron microscopes, x-ray equipment, microscopes, machine shop equips.ent of various kinds, ultrasonic equipment, electrochemical equipment, mounting presses, gas purification trains, hoods, and similar experimental equipment deemed necessary.
Lleonee No. SNM-960 Decket No.70-754 3.e,. No.
10.0 Pop 10-1 3/27/89 M s b et. M g,
The chemistry areas on the second floor consist of typical chemical laboratories, a counting room, and an instrument room.
Equipment in the.
chemistry laboratories includes the following: various types of spectrophotometers, fluorimeter, gas chromatography and a plasma emission spectrometer; other miscellaneous laboratory equipment; lead caves and glove l
boxes; vacuum systems, including necessary instrumentation; hoods designed for handling radioactive materials; counting instrumentation; and mass spectrometers of various types.
A concrete storage vault for special nuclear material is provided on the ground floor. The vault has walls and ceiling of 8-inch minimum thickness, without penetrations, and a single locked door.
A waste storage building of corrugated steel and aluminum on a concrete pad is provided for temporary storage of packaged wastes and scrap materials which result from the licensed chemistry, metallurgy, and ceramics activi-ties. They consist of paper, glassware, plates, rods,. wire, samples, and l
other waste residue materials which result from such research, development, and analytical activities. These wastes normally will contain small amo 4ts of byproduct, source, and special nuclear materials. They are transferred to the site waste handling component for final packaging and disposal.
10.3.2 Ventilation system l
Inlet air supply for Building 103 is provided by air conditioning units furnishing filtered and tempered outside air to the building. Air passes from the office areas through 2-inch fiberglass roughing filters in the laboratory door grills, thereby minimizing the passage of lint and dust into i
the laboratories. These filters also minimize backflow of potentially contaminated material in the unlikely event of complete exhaust system failure. Laboratory doors are equipped with automatic closers.
Air is withdrawn through the hoods and glove boxes passing first through individual high-efficiency filters at each hood or glove box. Only filters having a minimum efficiency of 99.97% for 0.3-micron-diameter homogeneous l
particles of dioctyl sebacate, DOS, are used in the laboratory effluent License No. SNM-960 Decket No.70-754 Sect. No. __. 10.0 Page 3/27/89 Amends Sect.(s)
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ventilation system. From the individual filters, the air is conducted through a second filtration in one of two parallel banks of high-efficiency (99.95% for the system) filters. Thus filtered, it is discharged through a 48-foot-high, 5-foot-diameter stack. The high-efficiency filters are fabricated of fiberglass to provide high resistance to fire.
Filter frames are metal or chemically impregnated for resistance to fire, and permanent duct work is metal or polyvinyl chloride.
Each laboratory room used to conduct activities with radioactive materials is equipped with air sampling devices. However, airflow rates are adequate to permit routine operations with nuclear materials without the use of personnel respiratory protection.
The main exhaust blower operates at approximately 36,000 cubic feet per minute.
If complete ventilation failure occurs, an evacuation alarm is sounded automatically.
From 9 to 12 air changes per hour are provided for most laboratory rooms.
However,-in some rooms the airflow rate may be as high as 15 air changes per hour. Hood exhausts are dampered individually to maintain minimum face velocities on the order of 125 linear feet per minute across the openings.
Glove boxes are operated at approximately -0.5 inches of water with respect to the room. Appropriate instrumentation indicating airflow and/or differential pressure is available.
The efficiency of the Building 103 ventilation filter system has been demonstrated by years of exhaust stack sampling data.
10.3.3 Radioactive Waste Facilities l
Dry contaminated waste materials generated in Building 103 are packaged as indicated in Section 3.5.1.
Waste containers are transferred directly to the site radioactive material storage facility or may be placed in a waste storage building approximately 11 feet by 12 feet constructed on a concrete pad adjacent to Building 103 (Figure 10.1).
This building is of corrugated steel and aluminum construction. The waste storage building is conspicuously posted in accordance with 10CFR20.203. Vaste packages stored in this area are removed periodically to the site waste storage facility for delivery to a licensed waste disposal contractor.
License No. SNM-960 Decket No.70-754 Sect. No.
10.0 Pope 10-3 gy,,
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Liquid contaminated wastes ori inating in Building 103 are routed from 5
laboratory sinks and gravity drains through regulated pipes to waste retention tanks. Two tanks, each of approximately 5,000-gallon capacity, are provided; :The tanks are' equipped for representative sampling and for draining to drums or tank trucks. All tank wastes are sent to the site waste
. evaporator for concentration and solidification.
10.3.4
. criticality Alarm l
L Building 103 currently is monitored for a criticality accident by three'
. detectors as described in Section 3.8.
Approximate locations of the sensors are'shown in Figures'10.1'and 10.2.
l 10.3.5'. Fire Protection Building'103 is provided with an automatic sprinkler system. Fire extinguishers also are located strategically throughout the laboratory areas.
Accidental fires in the hood enclosures are of low potential. Three types of
. fuel are available for such accidents: general paper and plastic combustibles, flammable solvents, and pyrophoric materials. VNC fire prevention procedures minimize the fire potential; however, extinguishing equipment and materials are provided at strategic locations in the building.
a
. Additionally, smothering agents such as Metl-1-X, asbestos cloth, etc., are available in each enclosure where sufficient fire probability exists.
Special precautions are taken when quantities of special nuclear materials l
are handled in hoods or glove boxes to minimize fire hazards in these enclosures. For example, metal containers will be used for pyrophoric materials.
License No. SNM-960 Decket No.70-754 Sect. No.
10.0 Pes
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10.4 BUILDING 103 PROCEDURES g,
k In addition to the VNC general procedures described elsewhere herein, other l
procedures specific to the health and safety of experimental and supporting t
operations conducted in the chemistry, metallurgy and ceramics laboratories, s
Building 103, are employed. The essence of these procedures is as follows.
10.4.1; Aggyev and cone==ination control Procedures Monitoring and step-off procedures are observed at points of transition from controlled areas to uncontrolled areas. In addition, hand-and shoe monitoring stations are provided at the building exits. Surveys at the beginning of irradiated experiments assure the adequacy of local shielding.
10.4.2 Clove Box and Hood Station Procedures Basic operations to be performed at glove box or hood stations normally involve gram or milligras quantities of uranium and TRU materials.- TRU powders in excess of 50 microcuries are handled in S ove boxes only.
l Blending, sample and specimen preparation, and analytical operations are the most frequent activities utilizing such material. Standard procedures for these activities are used.
10.4.3 ventilation system Maintenance The filtered and monitored ventilation system for Building 103 is described in Section 10.3.2.
The high degree of efficiency of this system is due in l
part to careful maintenance and operating procedures. The main filter bank in the' system is dual so that the system can be run on one bank while the other bank is being changed. The duct velocities are low enough to allow the main flow control and balancing to be done at each of the primary filter box connections. The main system basically runs as a large manifold of relatively constant suction so that primary connections can be made to the system as needed allowing the overall blower capacity of the main system to be maintained. Damper locations are restricted to dampers at the exit of Liessee W. SNM-960 Dode No.70-754 sec m.
10.0 Pope 10-5 3/27/89 Ms Sec.M b
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each primary absolute filter to control the flow of that primary branch and dampers that normally are open at the main filter bank to permit running the system on one bank while the filters in the other bank are changed.
All filter boxes are provided with differential pressure gages to measure pressure drop across the filters. The flows, correlated with the pressure changes, give a good indication of the state of the filters. Any filters found to be below standard are changed.
Filter boxes are surveyed by Radiation Safety for contamination prior to changeout. This survey on the i
l-second bank of filters also is a measure of the effectiveness of the first filtere.
Exhaust air from the Building 103 laboratories is sa:npled continuously. The l
filtered air that leaves the building is exhausted up a 48-foot stack which extends 20 feet above the building. The exhaut.t air is sampled continuously for iodine and particulate activity at a point 24 feet below the top. The particulate are collected on a high-efficiency, 47-mm membrane filter; and l
iodine is collected on a charcoal cartridge. The filter and cartridge are evaluated on a weekly basis.
10.4.4 F.aste Disnosai Normally, dry contaminated wastes occur in the form of paper towels, glassware, and similar nonsalvageable laboratory apparatus as well as small quantities of uranium as, for example, test specimens and analytical sample residues. These materials are packaged, sealed, recorded as to quantity and type of nuclear material, placed in the designated waste storage building (Figure 10.1), and transferred to the site radioactive materials storage facility until final disposal from the site.
Storage of dry wratas in Building 103 is limited to packages having surface dose rates or isJ mR per hour or isss.
The two 5,000-gallon-capacity radioactive liquid waste tanks are monitored routinely and, when approximately filled, are sampled and their contents routed to the site waste evaporator.
License No. SNM 960 Dochet No.70-754 Sect. No.
10.0 Pope 10-6 g
3/27/89 g, g,,g
10.5 BUILDING 103 CRITICALITY CONTROL r
The~ activities and work programs conducted in Building 103 are experimental 1' and analytical rather than industrial. Experimental operations rarely require large quantities of fissile special nuclear materials at a time but do require a wide variety of forms, compounds, and concentrations.
Critienlity control of these materials is relatively simple for it is based on the issuance of a safe quantity by a materials custodian for use in any-single laboratory room'at any one time for the majority of the activities conducted. Stock supplies of special nuclear materials normally are stored in a vault when not in laboratory use. Directions for storage arrays in the vault limit the amount and spacing of the material to safe systems assuming optimum moderation and reflection. Full or partial flooding of the vault is unlikely; however, since there is no water service in the vault and, except for the steel. vault door and electric lighting in the ceiling, there are no other penetrations.
The storage vault is isolated by conerste walls, ceiling and floor e least 8 inches thick. It is situated between a rest room and a lunch room. Areas immediately above the vault on the second floor do not contain special nuclear material. No special nuclear materials are permitted in any of the adjacent rooms. Within the stcrage vault, special nucIsar materials are stored in specified locations. Figure 10.3 shows the current configuration.
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12.0 BUILDING 105 i
Vu 12.1-LOCATION AND GENERAL DESCRIPTION i
i Building 105 is located immediately' north of Building 102 in the 100 Area 1
(Figure 6.1).
The building houses the Nuclear Test Reactor (NTR) and the l
Advanced Nuclear Applications laboratory (Figure 12.1).
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12.2 NTR FACILITIES The Nuclear Test Reactor (NTR) is licensed pursuant to 10CFR50 (License l
R-33).
SNM licensed pursuant to License SNM-960 is taken to the NTR North Room or the NTR South Cell (Figure 12.1) primarily for purposes of neutrographic examination. Such material is in the form of sealed units, and i
only one safe batch (<45% of the r.inimum critical accumulation) is permitted in the North Room or Eouth Cell at any one time. The materials are handled pursuant to the safety provisions and requirements of License R-33.
12.3 OTHER 1ABORATORY AREAS (ADVANCED NUCLEAR APPLICATIONS)
Other laboratory areas in Building 105 use SNM only as sealed sources, standards, foils, or as electronic components under general license.
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13.0 WASTE BANDLING 13.1 RALI0 ACTIVE LIQUID WASTE EVAPORATOR PLANT (BUILDING 349)
Activities described in this section involve the transfer of liquid radioactive wastes including those containing small quantities of sg,ecial nuclear material generated at various facility and laboratory installations at the Vallecitos Nuclear Center site to the Radioactive Liquid Waste Evaporator Plant, Building 349; the concentration of these wastes in Railding 349; the discharge of processed effluents; and the transfer of concentrated waste materials to a licensed waste disposal contractor. The estimated annual plant throughput is 100,000 gallor3 of liquid waste.
Each facility operated under License SNM 960 which generates liquid l
radioactive waste (i.e., Buildings 102 cr.d 103) are equipped with liquid radioactive vaste retention tanks. All lines in each building which potentially could carry radioactive materials are connected to the retention tanks. The tanks are emptied as necessary.
Prior to transferring liquid wastes to the Waste Evaporator Plant, samples are taken for gross alpha and beta, total uranium and plutonium, and U-235 analyses. During the first half of 1988, uranium levels ranged from <0.02 to 24.0 ppm with an aversge of 5.7 ppm. A log of the total quantity of U-235 in the plant at any one time is maintained. The entries to this log are based on the analyses of samples of the liquid waste, and the U-235 inventory has l
been typically between 50 to 70 grams.
The Waste Evaporator Plant is not a production or utilization facility within the meaning of 10 CFR Part 50, Section 50.2(a)(3) ["a facility designed or used for the processing of irradiated materials containing special nuclear materials...").
None of the components of the plant make possible the separation or purification of isotopes from each other.
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i 13.1.1 1 mention and ceneral neserintion h e Waste Evaporator Plant is housed in a metal building constructed on a poured concrete pad with integral concrete shielding walls around the processing vessels at high activity points. The shielding is designed to provide radiation levels in operating areas from normal wastes to less than 5 mR/h.'
The building entrance opens into a change room and operations control room.
General radiation levels in this area normally are less than 2 mR/h. Figure 13.1 shows the floor plan-for the building.
he Waste Evaporator Plant is located in an area to the northwest of the deactivated VBWR facility (Figure 1.2).
13.1.2 Liouid Waste Processine Liquid wastes are collected from the various site accumulation tanks and
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transported to the Waste Evaporator Plant by fork lift truck in a specially designed 1,500-gallon stainless steel tank. The waste transfer tank is equipped with internal vertical baffles to prevent cyclic shifting of the liquid material during transport. All external protuberances such as pipes, valves, gages, etc., are arranged or guarded in such a manner that they cannot come in contact with other vehicles or objects upon the roadway yet are readily accessible for manual operation for loading and unloading.
Experience in transporting liquid wastes in this way shows dose rates to the fork lift truck operator generally have been less than 5 mR/h.
In the direct evaporation process, the liquids are pumped continuously from the feed tank or chemical treatment tanks directly into the evaporator.
Here, the wastes are concentrated through a vertical-tube, natural convection ovaporator. The vapor is treated in associated equipment, including a Effluent high-efficiency desister entrainment separator and a condenser.
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l and disposal by evaporation. If further decontamination is necessary, the water can be re-routed to the feed storage tank for reprocessing or ion exchange treatmont.
13.1.3 ggnerete Mixinn station 1.iquid waste concentrates from the evaporator are collected in a receiver and i
discharged into DOT Specification 17-H 55-gallon drums or other approved containers. The concentrates then are mixed with a cement-diatomaceous earth i
mixture or equivalent for solidification. After solidification, the drums are sealed and prepared for disposal as dry solid waste. The drums as loaded will provide compliance with appropriate DOT regulations.
Solidified waste contained in drums is monitored; tagged with the radionuclides present, curie l
content, and radiation level; and stored awaiting removal from the Vallecitos site by a licensed waste disposal contractor or other approved methods.
Wac e drums are not buried at the VNC site.
Small amounts of liquid wastes incompatible with evaporation may be solidified directly in 55-gallon drums.
Storage of dry wastes at the Evaporator Plant is limited to packages having surface dose rates of 100 mrem per hour or less. Waste drums having radiation readings in axcess of 100 mrem per hour are transferred promptly to the high-level waste storage area. The building waste storage ares. is classified as a Radiation Area and so posted.
l 13.1.4 ventilation Effluent air from equipment and locations such as tank vents, hood vents, and sample points not requiring pre-filtering is exhausted directly to the main building filter exhaust system. Effluent air frca the vented side of the condenser along with other points of suspected higher activity is manifolded into pre-filters before being exhausted into the main building exhaust ducts..
License No. SNM-960 Decket No.70-754 Sect.Ne.
13.0 Pope 13 3 3/27/89 g, g,,g p.,,
4 The main exhaust then is directed through a HEPA filter system (99.954 efficient for 0.3-pa particulate). Filters appropriate for high relative humidity service are used.
The filtered air then is discharged to the atmosphere through a continuously sampled stack at a point approximately 7 feet. above the roof of the building.
Approximately 18 air changes per hour maintain proper contamination control.
Air discharged from the stack at the rate of 3,000 cfm is driven by an electrically powered blower mounted at the base of the stack.
13.1.5 Monitorine Procedures Monitoring and/or step-off procedures are observed at points where each potentially contaminated replated area exits into the clean or nonregulated areas.
In addition, survey instruments are provided at convenient locations for final surveying. The doors to the process equipment area are alarmed or padlocked in accordance with 10CFR20.203(c)(2).
13.1.6 criticality control The facility meets the requirements of a suberiticality area as defined in Section 3.14 of Appendix A to License SNM-960.
13.2 SOLID WASTE HANDLING A single organization has been designated by VNC management as responsible to assure that all solid vastes leaving the site meet the appropriate regulatory requirements. For the purposes of this application, it will be referred to as the Waste Hundling Function (WHF).
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13.0 Pope 13'4 g
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l' 13.2.1 Solid Waste Accumulation j'
Solid wastes are accumulated at each location where radioactive materials are l
handled. The majority of wastes fits the Low Specific Activity (LSA) category as defined in the Department of Transportation regulations.
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For each vaste accumulation, the generating component is responsible for maintaining a listing of all material in the accumulation. Each accumulation l
with its listing is forwarded to the VHF for final inspection and/or repackaging.
The WHF is responsible for implementing waste volume reduction. methods, as appropriate.
13.2.2 Solid Waste Storace Solid waste materials are stored in the site radioactive materials storage facility described in Section 3.5.1 and shown on Figure 13.2.
Vaste materials that include or are associated with significant quantities of special nuclear materials are placed in containers called weste liners.
Waste liners are stored in the horizontal tube facility bunker. Fifty-five-gallon drums and boxes containing lower level wastes are stored in a covered facility.
Limits on the maximum quantity of special nuclear material that may be loaded into any containers for purposes of waste storaEe have been established by Nuclear Safety. For 55-gallon drum storage, calculations were made assuming optimum water moderation and spherical geometry for the individual masses j
within each drum. No credit was taken for neutron absorption by the j
materials between the individual units. On this basis, limits were
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established which provide criticality safety for an essentially infinite array of these drums. This limit is set in this fashion since it is possible to store drums above grade in almost any geometry. However, normal storage arrangement would be a planar array one drum high.
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Attachment C Basis For Eliminating the VNC Radiological Contingency Plan q
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Basis For Eliminating the VNC Radiological Contingency Plan NUREG-1140 (ReSulatory Analysis of Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licenses) identifies a fire, a UF cylinder 6
rupture and a criticality accident as the only types of accidents for which the protective action guide doses could theoretically be exceeded. For VNC, each of these scenarios may be addressed as follows.
A.
IlE : There are no facilities at VNC for handling UF 6
p B.
Eira: The issue of release of radioactive material by fire was addressed in the " General Electric Vallecitos Nuclear Center Safety Evaluation Report, May, 1984" (SER), pp. 53-55.
This report found no potential for significant release of SNM in the event of a fire.
The SER was based in turn on the NRC Final Draft, " Accident Analysis for the General Electric Company Vallecitos Nuclear Center at Pleasanton, California, Related to License Renewal of Special Nuclear Materials License No. SNM-960", October, 1978. Of the areas identified as having a potential for serious accidents, except for criticality accidents, only thc Radioactive Materials Laboratory (RML) in Building 102 and Building 103 still utilize SNM.
For the RML the draft (p. 3) notes that the ventilation and filtration systems limit the release of SNM in case of a fire to insignificant quantities. The limiting scenario, dissolving of fission product target capsules, is no longer a licensed activity.
For Building 103 the limiting accident was an explosion rather than a fire. The maximum site boundary dose calculated was bo'ne dose of 0.36 Rem.
(Note: This was due to the inventory of Pu at Building 103 expected in conjunction with the operation of the mixed-oxide fuel laboratory in Building 102. The Building 102 laboratory is no longer a licensed activity.)
In view of the above, fire is not a credible scenario for requiring a contingency plan.
C.
Accidental Criticality: Although the license limits for SNM at VNC require criticality accident detectors, the actual potential for a criticality is nil. All past accidents have involved either liquid or metal systems. Neither exist at VNC.
SNM in SNM-960 facilities is almost entirely in the form of low-enriched UO2 p wder or irradiated power reactor fuel segments.
However, in order to provide a maximum boundary dose case, an accidental j
l criticality was assumed at Building 103. The combination of relatively thin shielding walls and close proximity to the site boundary will make it the worst case.
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The evaluation which follows was made using Regulatory Guide 3.34.
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results will be conservative.since the equations assume optimum.
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are physically impossible, i.e., there are no' containers, tanks, sumps, c
etc.. that.would allow such an accumulation of SNM.'
p' Evaluation of Accidental Criticality At Buildinn 103 c
Erohlam: Calculate the-radiation dose at the nearest site boundary from a.
criticality burst that would occur at the most probable location at VNC which-contains fissile material as' licensed under SNM-960.
Include both the prompt gamma and neutron dose and the released fission gas cloud dose.
Reference:
USNRC Regulatory Guide No. 3.34, " Assumptions used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Uranium. Fuel Fabrication Plant", Rev.1; July,1979.
Elyan:.
Prompt Gamma Dose, D(G) - 2.1E-20*N*d"-2*e'-(3.4d).
o Prompt Neutron Dose, D(N) - 7E-20*N*d*-2*e'-(5.2d) o Where:. D(G) - prompt grana dose', Ram D(N) - prompt neutron dose, Rem N
.- number of fissions d
- distance from source, km o.
Dose Reduction Factors for Concrete (Regulatory Guide 3.34):
'l Concrete Reduction Factor Thickness Gamma Neutron First 8 inches 2.5 2.3 First foot 5.0 4.6 Each added foot 5.5 20.0 o ~
Off-Site Person Breathing Rates (Regulatory Guide 3.34):
Time Period Rate. ec/see First 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.47E+02 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.75E+02 C-2 l
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o The internal and submersion dose conversion factors are taken.' rom ICRP 30.
The noble gas submersion dose to the lens of the eye is used for the whole body dose. The iodine doses are to the thyroid.
Where dose conversion values are not given in ICRP 30, the following cloud dose equations from the Regulatory Guide are used.
Radioactive Cloud Dose Rates:
Beta Dose Rate, DC(B) - 0.475*E(B)*C*[D(d)/D(B))
Gamma Dose Rate, DC(G) - 0.25*E(G)*C l'
Where: DC(B) - beta dose rate, center of cloud, rad /sec DC(G) - gamma dose rate, center of cloud, rad /sec E(B) - average beta energy, MeV/ dis E(G) - average gamma energy, MeV/ dis C
- concentration in cloud, Ci/cu m D(d)/D(B) - Energy Dependent Attenuation Factor o
The cloud concentration is determined from the diffusion factor which is taken from Figure 3A of the Regulatory Guide and is a function of distance. When the distance is 427 meters:
Diffusion Factor for 0-8 hours - 2.7E-03 sec/cu m Diffusion Factor for 8-24 hours - 6.4E-04 sec/cu m o
The fission product cloud consists of one-half the release fractions specified in Regulatory Guide 3.34 because the criticality accident must take place in a matrix of oxide-type fuel, both clad and unclad, not a dissolved solution of fissile material. The fission product cloud is released from ground level, not an elevated stack release. The assumed release fractions are:
Noble Gases 50.0%
Iodines 12.5%
Solid F.P.'s 0.0%
==
Description:==
The fuel vault in Building 103 is the criticality limit area (CLA) which has a fissile limit larger than most of the other CLA's on site and is nearest a site boundary. The Building 103 vault is assumed to have an accidental criticality involving twice the maximum limit of the vault, which is 1,589.6 grams of U-235 in various enrichments.
Some major event is postulated to bring the 3,179.2 grams of U-235 together in an optimally moderated and reflected configuration resulting in a single burst which displaces the U-235 so that no additional criticality events occur. The walls of the vault are assumed to remain standing, to contain the moderating / reflecting water, during the criticality burst.
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o Ingng:
I.3 4
. Total number of fissions 1.0E+17 fissions
\\
8 inches
' Concrete wall thicknessi t
Distance.to siteLhoundary - 427 meters
/
Results:
Promet Doses in D(G) 1.08E-03 Rem D(N) 1.81E 03 Rem I
L Cloud Contingrations and Doses Conc.
Dose Factors Committed-Qggtg_
, Curies-LIntegral Internal Submer. Organ,
- Skin, Nuclide Released uCi-s/ce' Ram Rem
-Kr-83m
'8.0E-01 2.2E-03 1.7E-04: 1.7E-04 3.7E-07 '3.7E-07
.Kr-85m.
~7.6E-01
-2.1E-03' 3.5E-02 8.5E-02 7.2E-05 1.7E-04 Kr-85
'8.0E-06
.2.2E-08 6.1E-04 4.8E-02 1.3E-11 1.0E-09 Kr-87 5.1E+00 1.4E-02 2.8E-01 5.1E-01,3.8E-03 7;0E-03 t:
Kr-88'
-3.3E+00 8.9E-03 4 6E-01 5.3E-01 4.1E-03' 4.7E-03
~
Kr-89 2.1E+02 5.8E-01
. 0E-01 5.2E-01.2.3E-01.3.0E l-Xe-131m. 4.0E 1.1E-06 3.1E-03 1.7E-02 3.3E-09 1.8E-08 Xe-133m 9.1E-03 2.4E-05 7.7E-03 3.8E-02 1.9E-07 9.3E Xe-133
=1.4E-01 3.7E-04 9.4E-03 2.0E-02 -3.5E-06 7.4 E-06 Xe-135m :1.1E+01 3.0E-02
.9.6E-02 1.1E-01 2.9E-03 :3.3E-03 Xe-135 1.8E+00 4.9E-03 5.6E-02 1.2E-01 -2.7E-04: 5.9E-04
~
Xe'-137
-2.5E+02 6.8E-01 4 '. 0E-02 7.7E-01' 2.7E-02 5.2E-01
--Xe-138; 6.5E+01.
1.8E-01 2.8E-01 4.1E 01 4.9E-02 7.2E-02 Noble Gas' Submersion Total 3.2E-01 1.2E+00 LI-131 4.3E-02
-1.2E-04 1.1E+00 1.4E-02 4.5E 1.6E-06
.I-132 5.4E+00 1.5E-02 6.3E-03 1.4E-01 3.2E-02 2.0E-03 I-133 8.0E-01 2.2E-03 1.8E-01 9.5E-02 1.3E-01 2.1E-04 I-134 2.3E+01 6.2E-02 1.1E-03 1.9E-01 2.4E-02.1.2E-02 Ie135
.2.4E+00 6.5E-03 3.1E-02 7.4E-02. 6.9E-02 4.8E-04 Thyroid Total 3.0E-01 Iodine Submersion Total 1.4E-02
'*-Units for the inhaled dose conversion are: Rem /pCi inhaled, e.xcept for the lens exposure from noble gas, which are the same units as submersion.
- Units of submersion dose conversion are: Rem-cc/pci-sec.
l' C-4
p;-
J Bummary:
?
j-The.'whole body dose received from an eight-hour exposure at the nearest site.
j
. boundary (to.the west) from a criticality accident in.the Building 103 vault i
L results from the sum of the-prompt gamma dose, the prompt neutron dose, and
[
the submersion dose to the lens of the eye. This sum is 0.32 Rem.
The skin dose received from the same eight-hour exposure at the site boundary p.
P is the submersion dose:to the skin from the noble gases and iodines. This et-
-dose is 1.24 Rem.
-The' committed thyroid dose received from the same eight-hour exposure at'the
~
-site. boundary is from the iodines. This dose is 0.30 Remi
==
Conclusion:==
The event assumed in this evaluation does not appear to be-1
! credible. Nonetheless, the calculated radiation dose at the nearest site
~
. boundary is 0.32 Rem.whole body and 0.30 Rem thyroid. These values are well below the EPA recommended PAG values of 1.0 Rem and 5.0 Rem, respectively.
p Therefore, a contingency plan for dose and releases from an accidental criticality involving tasterials authorized under License SNM-960 is not needed.
j.
r C-5
,,