ML20236E487
| ML20236E487 | |
| Person / Time | |
|---|---|
| Site: | 07000025 |
| Issue date: | 05/19/1989 |
| From: | ROCKWELL INTERNATIONAL CORP. |
| To: | |
| Shared Package | |
| ML19311A788 | List: |
| References | |
| ESG-82-33, ESG-82-33-02, ESG-82-33-2, NUDOCS 8906050283 | |
| Download: ML20236E487 (152) | |
Text
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'c ESG-82_-33 HEALTH AND SAFETY SECTIONS FOR RENEWAL APPLICATION OF THE SPECIAL NUCLEAR MATERIALS LICENSE SNM-21, DOCKET 70-25, ISSUED TO ROCKETDYNE DIVISION OF ROCKWELL INTERNATIONAL AUGUST 25,1982 REVISED FEBRUARY 29,1984
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DECEMBER 19,1988 MAY 19, 1989 RockwellInternational Rocketdyne Division 6633 Canoga Avenue Canoga Park, California 91304 O
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Page V
PART I LICENSE CONDITIONS l
1.0 ' Standard Conditions and Special Authorizations................
I.1 -1 1.1 Name....................................................
1.1 -1 1.2 Location................................................
1.1 -1 1.3 License Number..........................................
1.1 -1 1.4 Possession Limits.......................................
1.1 -1 1.5 Location Where Material Will Be Used....................
1.1-2 1.6 Definitions.............................................
1.1-2 1.7 Aut ho ri zed Ac t i vi t i e s...................................
1.1-4 1.8 Exemptions and Special Authorizations..................
1.1-4 2.0 General Organizational and Administrative Requirements.........
1.2-1 2.1 Licensee's Policy.......................................
I.2-1 2.2 Organizational Responsibilities and Authority...........
I.2-1 2.3 Safety Review Committee.................................
I.2-3 2.4 Approval Authority for Personnel Selection..............
I.2-4 2.5 Personnel Education and Experience Requirements.........
1.2-4
'2.6 Training..............................................
I.2-6 2.7 Op e ra t i n g P ro c e d u re s...................................
1.2-7 2.1 Audits and Inspections-.................................
I.2-8
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2.8.1 Reporting Requirements........................
1.2-9 2.8.2 Periodic Saf ety Review Provi sions.............
1.2-9 2.8.3 T ransmittal of Findings to Management.........
I.2-11 2.8.4 Enforcement of Safety Rules and Practices,....
1.2-12 2.9 Investigating and Reporting Off-Normal Occurrences......
1.2-12 2.10 Records.................................................
1.2-12 3.0 Radiation Protection...........................................
- 1. 3-1 3.1 Special Administrative Requirements.....................
I.3-1 3.1.1 Controlled Work Permit........................
I. 3-1 3.1.2 ALARA Policy..................................
I. 3-1 3.1.3 Radiation Safety Plans........................
1.3-2
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Page 3.2 Technical Requirements..................................
1.3-2 3.2.1 Access Control................................
I.3-2 3.2.2 Ventilation Requirements......................
I.3-2 3.2.3 Instrument (Survey, Counting, Criticality Monitors).....................................
I.3-8 3.2.4 Internal and External Exposure................
I.3-9 3.2.5 Sealed Source Leak Tests......................
1.3-14 4.0 Nuclear Criticality Safety.....................................
1.4-1 4.1 Special Administrative Requirements.....................
I. 4-1 4.1.1 Criteria for Establishing Plan of Action......
- 1. 4 -1 4.1.2 Verbal Approval..........................
1.4-1 4.1.3 Nuc lea r Sa f ety Analys i s........................
1.4-2 4.1.4 Approval......................................
1.4-2 4.1. 5 Inspection....................................
I,4-2 4.2 Technical Requirements..................................
1.4-3 4.2.1 Single-Unit Criteria..........................
1.4-3 4.2.2 Array Criteria................................
I.4-6 g
4.2.3 Limi ted Mode ra tion Cri teria...................
1.4-11 5.0 Environmental Protection.......................................
1.5-1 5.1 Effluent Control Systems................................
1.5-1 5.2 Environmental Monitoring................................
1.5-3 5.3 Liquid Radioactive Waste.....................
1.5-3 6.0 Special Process Commitments....................................
I.6-1 7.0 Decommissioning Plan...........................................
1.1-1 8.0 Radiological Contingency Plan.................................
1.8-1 PART II SAFETY DEMONSTRATION 9.0 Overview of Operation..........................................
11.9-1 9.1 Corporate Information..................................
11.9-1 9.2 Financial Qualification.................................
11.9-1 9.3 Summary of Opera ting Obj ectives and Process.............
11.9-1 9.4 SiteDescription........................................
11.9-1
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CONTENTS
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Page 9.5 Location of Buildings on the Site.....................
II.9-1 9.6
. Maps and Plot P1an......................................
11.9-?
9.7 License Ilistory.........................................
11.9-2 9.8 Changes of Procedures, Facilities, and Equipment........
11.9-2 10.0 Facility Descriptions..........................................
11.10-1 10.1 SSFL Facilities.........................................
11.10-1 10.1.1-Hot Laboratory................................
11.10-1 11.0 Organization and Personnel.....................................
11.11-1 11.1 Organi:ational Responsibilities.........................
11.11-1 11.2 Organizational Charts...................................
11.11-2 11.3 0 ganizational Procedures...............................
11.11-2 11.3.1 Responsibility Assignments....................
11.11-2 11.3.2 Internal Inspection and Review Program........
11.11-6 11.3.3 Initiation of Projects Involving Potential Hazards.......................................
11.11-9 11.3.4 Changes in Policies, Procedures, Facilities and Equipment..............................,...
11.11-12 11.3.5 Records of Authorization......................
11.11-13 11.3.6 Records Retention.............................
11.11-13 s
11.4 Functions of Key Personne1..............................
11.11-13 11.5 Education and Experience of Ke Training......................y Personnel...............
11.11-13 11.6 11.11-13 12.0 Radiation Protection Procedures and Equipment..................
11.12-1 12.1 Procedures..............................................
11.12-1 12.2 Posting and labeling....................................
11.12-1 12.3 Personnel Monitoring........................
II.12-2 12.4 Surveys.................................................
11.12-2 12.5 Reports and Records...................................
11.12-3 12.6 Instruments.............................................
11.12-3 12.7 Protective Clothing...................................
11.12-4 12.8 Entry and Exit Procedures..............................
11.12-5 l
12.9 Administrative Control Levels..........-................
11.12-5 12.10 R e s p i ra t o ry P r o t e c t i o n..................................
11.12-5 r
ESG-82-33 iii
m CONTENTS Page 13.0 ' Oc c upat ional Rad iation Expos ures.. -...........................
11.13-1 13.1 Occupational Exposure Analysis........................
11.13-1 13.2 Measures TEken to Implement ALARA.......................
11.13-1 1
13.2.1 Documentation...........................
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!I.13-2 13.2.2 Reassessment..................................
11.13-2 l-13.2.3 Internal Review.........................
1-11.13-2 13.3 B i o a s s a y P r o g ra m........................................ 11.13-3 13.4 Air Sampling Program.................
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13.5 Surface Contamination................
11.13-5 i
11.13-5 13.5.1 Surveys.......................................
11.13-6 13.5.2 L
Smear Test Limits and Action Guides...........
11.13-6 13.5.3 Survey Instrument limits......................
11.13-6 13.5.4 Removal nf Material and Equipment and Release of Facilities.........................
11.13-8 13.5.5 Contamination Limits for Radioactive Materials Shipments and Radioactive Waste.....
11.13-8 13.6 Shipping and Receiving..............................-....
11.13-8 13.6.1 Release of Exclusive Use of Vehicles.................... Transport 11.13-10 14.0 Environmental Saf ety--Radiological and Nonradiologica1..........
11.14-1 15.0 Nuclear Criticality Safr.y................................
11.15-1 15.1 Administrative and Technical Procedures.................
11.15-1 15.1.1 Documentation Requirements....................
11.15-1 15.1. ?
Nuclear Safety Analysis Requirements..........
11.15-1 15.1.3 Modifying Existing Procedures.................
11.15-3 15.2 Preferred Approach to Desi r
Basic Assumptions.........gn............................
11.15-4 15.3 11.15-4 15.4 Analytical Methods and Validation Ref erences............
II.15-4 15.5 Data Sources.....................
11.15-4 15.6 Fixed Poisons....................
11.15-4 15.7 Structural Integrity Policy and Revision Requirements...
11.15-4 15.8 S p e c i al C o n t rol s........................................
11.15-4
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CONTENTS 10 V
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16.0 Process Description and Safety Analysis.........................
11.16-1 1
16.1 Rockwell inh enational Hot laboratory....................
11.16-1 16.1.1 Operations Description........................
11.16-1 16.1.2 Safety Analysis of Each Step..................
11.16-8 16.1.3 Safety Features of Each Step..................
11.16-8 17.0 ' Accident Analysis.................................
11.17-1 Appendices A.
1981 Annual Report--Rockwell International'i..............
A-1 B.
Environmental Reports...................................
B-1 C.
Cri ticality Stud ies f or Fuel Handling................... C-1
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Annual Review of Radiological Ccntrols for 1979, 1980 and 1981................................................
D-1 References...........................................................
R-1 ESG-82-33 v
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Page 3-1. Enclosure Requirements for Highly Radiotoxic Nuclides..........
1.3-4 3-2 Modification Factors for Enclosure Requirements................
I.3-4 3-3 ' Acceptable Removable Surface Contamination.....................
I.3-12 4-1 Plan-of-Action Criteria........................................
- 1. 4-1 4-2 Reflection Factors for Fissile Material S Hot-Cell Internal Dimensions.............ystems................
I.4-10 10-1 11.10-1 10-2 Vi ewi n g Wi nd ow Pa rame te rs.....................................
11.10-3 12-1 Respirator Issuance Schedules..................................
11.12-6 13-1 Acceptable Removable Surface Contamination.....................
11.13-7 13-2 Survey Instrument Surface Contamination Limits.................
11.13-8 13-3 Acceptable Limits for Residual Radioactivity (Rockwell International /Rocketdyne Division)..................
11.13-9 13-4 Contamination Limits for Radioactive Materials Shi Waste.............................................pments and 11.13-10 FIGURES 10-1 Facility Layout of RIHL........................................
11.10-2 10-2 Ventilation Flow Diag ram f or RIHL.............................
11.10-5 10-3 Ventilation Schematic for the RIHL.............................
11.10-7 f~s 10-4 Location of Ventilation Filters at the RIHL....................
11.10-9
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15-1 Nuclear Safety Analysis Cover Page.............................
'11.15-2 16-1 RIHL Process Description.......................................
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L1 1.0 STANDARD CONDITIONS AND SPECIAL AUTHORIZATIONS i
7-Q 1.1.NAME Rockwell International Corporation, acting through its Rocketdyne Divi-
.c sion, is responsible for operations under Special Nuclear Materials License No. SNM-21, Docket 70-25.
Rockwell International is incorporated in the State I
of Dettware with corporate headquarter; in El Segundo, California.
1.2 LOCATION Operations under the license (SNM-21) are carried out at the following
' Santa Susana Field Laboratories The Santa Surina Field Laboratories (SSFL) are loct ted in the Simi Hills in Ventura County, California.
Operations under this license are carried out in the Rockwell International Hot Laboratory (RIHL)
(Building 020), and other buildings, as z:uthoriced by Licensee review for small quantities.
1.3 LICENSE NUMBER The license under which operations described in this report are conducted
'is SNM-21 (Docket 70-25).
The previous expiration date of this license was September 30, 1982. Renewal was on June 28, 1984.
The current expiration date is Ju.ne 30, 1989.
1.4 POSSESSION LIMITS Maximuni Amount that Licensee May Possess Byproduct Source, and/or Chemical and/or at Any One Time Under Special Nuclear Material Physical Form this License A. Uranium er.riched in the A. Any enrichment or A.Less than 5 kg U-235*
U-235 isotope form except UF6 2L B. Pu (principally Pu-239)
- 8. Any B.Less than 2.0 kg total Pu*
- Enriched uranium and plutonium authorized in combination less.than 5.0 effective kg calculated on kg U-235 + 2.5 x kg Pu C. Pu (principally Pu-239)
C. Daled tources C. 1.0 kg total Pu (as Pu-Ee sources)
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t 1.5 LOCATION WilERE Mi.TERIAL WILL' BE USED
- f Reference Item A, Section 1.4 Used at Hot Laboratory, Building 020,
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SSFL site Reference Item B,' Section 1.4 1.
Used at Hot Laboratory, Building 020, SSFL site; less than.2.0 kg Pu in irra-diated or unirradiated fuel; less than 1.0 kg Pu in process.
2.
Less thin 2g total Pu for gamma spectroscopy and radiometric counting analysis used in any authorized building at the SSFL site in actor-dance with NRC-approved radiation safety criteria.
3.
A maximum of 0.1 gram of Pu for
- hemical analysis in an authorized Milding at the SSFL site in accor-cance with NRC-approved radiation sdety criteria as described in Set-tien 3.2 below.
. Ref erence item C, Section 1.4 1.0 kilogram total Pu in any authorized building at the SSFL site in accordance with NRC approved radiation safety
. g criteria.
1.6.
DEFINITIONS Generally standard reference terms are used throughout this document.
The following brief definitions are supplied for terms that may be unique.
Approval authority The person responsible for approving oper-ations involving nuclear fuels following a f avorable recommendation f rom the fuels Committee of the Nuclear Safeguards Review Panel.
Criticality control area A building or subdivision within a. build-ing where fuel handling is performed and between which f uel movements are controlled.
Criticali+.y safeguards advisor The person responsible for the out-of-reactor Criticality Safeguards and Control Program.
ESG-82-33 1.1-2
h Criticality safeguards and control The function responsible for the tevelop-ment and administration of the procedures r~s iif and controls for the f abrication, storage, and shipment of both fresh and irradiated fuels.
This does not inc1:.de loading or unloading reactors, or critical or sub-critical experiments.
Criticality safeguards coordinator The person. responsible for the administra-tion of out-of-reactor fuel handling pro-cedures and practices.
I Custodian A person assigned responsibilities for accountability records, vouchers, and transfers of source and special nuclear materials, in and out of a material balance area (MBA).
Double contingency Two independent and unlikely events that are concurrent in time.
Feasibility report A formal report (submitted to 001 Opera-tions Of fices required for contractor work) describing a proposed fuel process, including appropriate criticality control, accountability, radiation safety, and security procedures, fU)
In-process storage station A storage area, located within a process-ing area, for fissionable material that is soon to be processed.
Material Balance Area (MBA)
A room or group of rooms which, for the convenience of accountability records, is considered as a unit.
Nuclear safety analysis The written evaluation of the nuclear safety involved in all processing steps in a given fuel handling project.
Reactor fuels committee Committee responsible for review of the out-of-reactor Criticality Safeguards Program.
SS materials label Card for attaching to a container giving the quantity and type of material contained.
Safe batch The maximum quantity of fissionable mate-rial allowed at a given work station.
The double-contingency philosophy is used in determining the maximum safe batch size.
ESG-82-33 1.1 -3
-Safe cart A fissionable-material-bearing vehicle that, when used in an array of similar d(3-vehicles, requires at least ten such vehicles for criticality under normal con-ditions of handling.
Rocketdyne operating policy (ROP)
Standard procedure that has been approved by Rocketdyne Division defining a specific method of operation.
Vault storage area An area, removed from the fuels processing areas, for the storage of fissionable materials.
In general, fuels are stored here for much longer periods of time than in the in-process storage stations.
Work station A specific area for performing a given operation.
It may contain a given piece of equipment (such as work table, lathe, sand blaster, plating bath, etc.) required for a given process operation' and one safe batch of fuel.
Work station limit card A card posted at a work station stating the maximum allowable limit or safe batch for each type of fuel to be processed at a given station.
O 1.7 AUTl10RIZED ACTIVITIES
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e Authorized activities are the conduct of broad research and development programs and fuel-element decladding utilizing special nuclear materials of various forms and enrichments. Also authorized are disposing to approved sites of scrap material and waste and transferring and shipping of finished material and specimens resulting f rom all activities.
1.8 EXEMPTIONS AND SPECIAL AUTHORIZATIONS There are no exemptions or special authorization required.
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2.0 GENERAL ORGANIZATIONAL AND ADMINISTRATIVE REQUIREMENTS 2.1 LICENSEE'S POLICY a
It is the policy of the corporation to comply with all the requirements of law, to operate the facilities in a safe and efficient manner and within the requirements of all license conditions under which the activities are authorized. Rocketdyne Division has as policy to carry out all the conditions of the license, SNM-21, in accordance with the corporate policy.
Additional-ly, Rocketdyne is committed tu safe operating practices and to keeping radia-tion exposures to employees and the general public as low as reasonably achievable (ALARA).
2.2 ORGANIZATIONAL RESPONSIBILITIES AND AUTHORITY As a separate operating division.of Rockwell International Corporation, Rocketdyne is under the direct management of the Division president.
Research, engineering, manufacturing, and other support activities are accom-plished within separate subdivisions organized into functional organizations.
The president delegates to these organizational heads the responsibility for ensuring that all operations are conducted in a safe manner and in conformance with the provisions of applicable licenses and regulations.
In addition to the functional organization described above, there is a program management organization through which program managers are assigned the responsibility for the planning, control, and direction of specific pro-grams. The actual performance of the various associated tasks is accomplished f
by the functional departments within the budget and schedult limitations established by the program management staff.
The Health, Safety & Environment Department is a staff and service organ-ization, within the Human Resources and Communications organization, with responsibilities including criticality safeguards, radiation safety, indus-trial hygiene, and industrial safety.
In this staff capacity, the department is responsible for the development of an overall safety program and serves as advisors to the functional organization and all levels of Rocketdyne manage-ment to ensure that programs are implemented in a satisfactory manner in all operations throughout the Division.
While the responsibility for safety is delegated to all operational man-agement, members of HS&E have the authority to halt any unsafe operation.
Approval to restart a halted operation comes from the designated Approval Authority, acting with the concurrence of HS&E.
The Nuclear Safety and Licensing Department is a staff and support organ-ization within the Atomics International organization with responsibilities including nuclear licensing, nuclear material safeguards, accountability, and control, and management and administration of the Nuclear Safeguards and Review Panel (NSRP).
The department director serves as Chairman of the NSRP.
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The key positions by title with safety-related responsibilities in con-nection with licensed activities leading to top management are:
Manager, Radiation and Nuclear Safety
- Manager, Industrial Hygiene and Safety Staff for Criticality Safeguards Fuels Committee, NSRP
-Director, Health, Safety & Environment Radiation Safety Committee, NSRP- -Vice President, Human Resources and Com-munications Chairman, NSRP President, Rocketdyne The key positions by title with programmatic responsibilities in licensed activities for R&D and nuclear fuel operations leading to top management are:
-Project Manager, Fuel Decladding
-Director, Land-Based Power Projects
-Division Director, Atomics International
-President, Rocketdyne The key positions by title with operational safety responsibilities for f
licensed activities in Building 020 (RIHL) at the Santa Susana Field Labora-tories (SSFL) are:
pAssistant Manager, RIHL
-Manager, Nuclear Operations
-Manager, Development and Test
-Chief Engineer, Atomics International
-Division Director, Atomics International
-President, Rocketdyne These diagrams reflect the current organizational arrangement. While titles and the organization structure are subject to changes as the business involvement of Rocketdyne changes, the principal lines of authority remain unchanged.
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2.3 SAFETY REVIEW COMMITTEE r*
Department directors are supported by a staff safety organization--the
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Health, Safety & Environment (HS&E) Department--and by an advisory group, the Nuclear Safeguards Review Panel (NSRP).
NSRP is headeo by the. director of the Nuclear Safety and Licensing (NS&L) department.
In his capacity as NSRP chairman, the director of'NS&L reports directly to the president of Rocket-dyne. This arrangement provides for the resolution of dif ficult questions by top management when necessary.
FSRP maintains continuing safety reviews of all planned and existing nuc-lear fuel handling operations, and utilization of radioisotopes and radiation devices.
The primary purpose of NSRP is to provide management with the addi-tional assurance of an independent evaluation team for the protection of the general public, personnel, and equipment. The panel is composed of officers, the members of the Fuels Safety Review Committee and the Radiation Safety Review Committee. The officers include a chairman, who is the director of the NS&L department, and the Radiation Safety Officer, who is a member of the HS&E department.
Committee meetings are called by the respective chairmen as necessary to fulfill committee responsibilities without unnecessary delay to Rocketdyne programs. Meetings may also be held at the recuest of project personnel.
Each committee is responsible for the review of designs; operations with nuc-lear fuels; and activities involving radioisotopes and radiation devices, Reviews cover both proposed and existing designs and procedures, as well as
'i changes therein. The Fuels Committee reviews nuclear fuel activities, par-I,,h ticularly-in operations that may involve potential conditions of accidental O
criticality, such as fabrication, manipulation, storage, movement, and dis-posal of fissionable material. All committees perform an annual inspection of working areas scheduled for a maximum of 14 months between inspections.
At these times, the committees also review the organization of the operating groups and the assignment of local responsibility for operations.
Meetings are usually conducted with project personnel present and must have a quorum, which consists of a majority of the committee members.
Committee recommenda-i tions f or action arising f rom the meetings or reviews are approved by the NSRP chairman or his designee and transmitted to the appropriate department direc-tor (called the approval authority in committee correspondence).
The executive secretary provides a pertinent, written summary of each meeting, along with recommendations of the committee, for transmittal to each member, the appropriate approval authority, and others as might be designated.
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The approval authority is obligated to implement committee recommenda-tions and to ensure that the action taken is reported to appropriate NSRP personnel.
In the event that the approval authority disagrees with any recom-mendations, the committee's recommendations, together with a report prepared by the approval authority, are reviewed for a final decision by the superior of the NSRP chairman in the NSRP organizational channel, i
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- C The written approval of the approval authority is initially necessary for all' operations involving licensed activities under SNM-21. Subsequent inde-r"'
pendent information as to the degree of compliance with the safety procedures which.the approval authority is obligated to enforce is supplied by HS&E field personnel reporting through established channels and by the appropriate com-mittee of NSRP.
2.4 APPROVAL AUTHORITY FOR PERSONNEL SELECTION All NSRP members are appointed by the NSRP chairman with the approval of his superior in the NSRP organization channel; the use of alternates is not authorized.
Adequate time and funding for committee work are, by directive, budgeted by each project involved in activities under the jurisdiction of the NSRP.
Each committee consists of a chairman and several engineers, physi-cists, and other professionals, from the operating departments.
The chairman is not organizationally associated with the operations or programs for which the committee is cognizant.
Approval authorities are appointed by the president of Rocketdyne. They are, in general, the. department directors who have management responsibility for the operations under cognizance of the committees of NSRP.
2.5 PERSONNEL EDUCATION AND EXPERIENCE REQUIREMENTS The selection of safety committee personnel is based on demonstration of a high degree of technical competence in a discipline (s) associated with the work reviewed by the committee.
The number of members is determined by the range of the fields required to perform a satisfactory safeguards review.
NSRP officers are ex officio voting members of each committee.
At least one member of the Fuels Committee, other than the criticality Safeguards advisor, must be qualified in nuclear safety to perform independent verification of the methods of nuclear safety assessment used in the evalua-tion of criticality problems.
It is required that this member concur that the method of analysis used is appropriate and that the results of the calcula-tions establish saf e nuclear parameters for the operation reviewed. The extent of the t evie and the basis for concurrence are documented.
The minimum qualification requirements for operational personnel with safety-related management and staff positions are:
Operational Personnel 1)
Department Director and Hicher Manacement Positions - Bachelor's degree or equivalent in one of the physical sciences or engineering disciplines.
Eight years of experience in a technical discipline related to the technology associated with department operations, ESG-82-33 1.2-4
including 1 to 3 years of management experience.
Demonstration of the management and technical capabilities tc crganize and direct rmd technical groups responsible for conducting the operations for which the department is responsible.
2)
Manager - Bachelor's degree or equivalent in' one of the physicai sciences or engineering disciplines.
Five years of experience in a technical discipline related to the technology associated with Group operations, including i to 3 years of technical management experi-
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Demonstration of the management and technical capabilities to ence.
organize and direct the operations.
Safety Personnel l
1)
Department Director - Nuclear Safety and Licensing - Bachelor's degree in one of the physical sciences or in engineering.
Eight years of experience in a technical discipline related to nuclear technology, plus.1 to 3 years of management experience.
Demonstra-tion of the managerial and technical capabilities required to orga-nize and conduct the ESG safety program.
2)
Criticality Safequards Advisor - Bachelor's degree in one of the physical sciences. Two years of experience in criticality analysis, including criticality analysis associated with operations with fis-sionable materials outside of reactors.
Demonstration of sufficient judgment and analytical capability to establish and maintain a technically sound and effective criticality safeguards and control g
program.
3)
Chairman. Fuels Committee. Nuclear Safeguards Review Panel -
Bachelor's degrec in one of the physical sciences.
Two years of experience in reactor engineering or nuclear hazards evaluation, ~
including one year of service on the Fuels Committee.
Demonstration of sufficient judgment and analytical capability to direct the ef forts of the committet and to make independent nuclear safety evaluations.
4)
Physicist, Fuels Committee. Nuclear Safeauards Review Panel -
Bachelor's degree in one of the physical sciences.
Two years of experience in criticality analysis, including evaluations of opera-tions with fissionable materials outside of reactors.
Demonstration of sufficient judgment and analytical capability to review and evaluate a criticality saf eguards and control program.
5)
Criticality Safeauards Coordinator - Two years of college education or equivalent in one of the physical sciences. One year of experi-ence in nuclear fuel handling--operational or safety.
Demonstration of suf ficient initiative and conscientiousness to implement and audit criticality centrol procedures throughout the company's operations.
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Management Personnel - Health. Safety iL Environment - Bachelor's degree in one of the physical sciences. Two years of experience in
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radiation protection operations.
Demonstration of sufficient Judg-(d ment and analytical capability to establish and maintain a tech-nically sound and ef fective program for the function to be supervised.
7)
Radiation Safety Officer - Bachelor's degree in one of the physical sciences. Two years experience in radiation protection.
Demon-stration of suf ficient judgment and analytical capability to estab-lish and maintain a technically sound and effective radiation safety program.
The Bachelor's degree requirements listed for safety personnel with the exception of the de;;artment director may be waived with explicit review by the Health, Safety & Environment Department director.
In addition, appointments to the positions are subject to review and approval by the next level of man-agement above the department director.
In walving the degree requirements, specific consideration is given to a potential appointee's formal education, which must include college-level courses in scientific areas applicable to the requirements of the position, along with his on-the-job training, technical and administrative knowledge, work experience, and demonstrated capability and performance in job assignments associated with and/or related to the parth.a-lar position.
In the cases of the Criticality Safeguards Advisor and the -
chairman of the Fuels Committee, for example, demonstrated knowledge and cap-ability during assignments prior to the appointment in areas such as reactor physics, criticality analysis, process control, materials properties, and reactor fuel fabrication techniques are required.
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2.6 TRAINING It is the policy of Rocketdyne to provide the necessary training to all employees who are assigned to work in areas where their personal safety or the safety of operation requires that they receive special training.
Addition-ally, the policy requires retraining of these employees at appropriate inter-vals to assure continued safe practices.
Individuals who are to work unsuper-vised with radioactive materials complete formal training in radiation safety prior to beginning that work.
New employees and other personnel whose rt. gular assignments include exposure to radiation or radioactive matnrials must com-plete a formal training course covering the general aspects of working with these hazards.
This training includes all aspects of radiation safety ard criticality prevention that are appropriate for the work assignments.
Management establishes the necessary training requirements for personnel assigned to licensed activities.
The Rocketdyne Training function initiates the training program, schedules training sessions, provides instructors, dis-tributes attendance notices to selected trainees, and maintains records of fC ESG-82-33 1.2-6
i attendance and results of any required written examination tests and quizes.
The Training function periodically audits the training packages of personnel
.(
to assure conformance to the established requirements and informs appropriate management, including the director of HS&E, of any deficiencies.
Supplementary training in specialized aspects of safety is provided to saf ety personnel.
Retraining in radiation safety and criticality control is performed every 2 years.
Radiation and Nuclear Safety personnel are provided
!l' with other appropriate forms of continuing education.
2.7 OPERATING PROCEDURES Prior to initiating a project involving potential hazards, authorization must be obtained f rom the appropriate department director and from designated safety and/or Criticality Safeguards personnel in the HS&E department.
i A technical evaluation is performed by HS&E personnel for operations involving radiation and nuclear safety.
Written safety procedures are required for any new operations or programs.
At the discretion of the HS&E director or his designee, these procedures may refer to safeguards and pro-cedures developed for applicable conditions under other programs only if the potential for personnel exposure is less than the quarterly limits specified in 10 CFR 20.
These procedures are prepared by HS&E personnel and the operating engi-for the programs.
The operations and associated procedures are neers approved, in advance of starting program activities, by the responsible line supervisors and by HS&E supervisory personnel.
Technical evaluation and approval requirements for operations involving nuclear safety are discussed N
under Criticality Safeguards in Section 4.
Reviews of operations for radiation' safety and criticality control, and approvals based on these reviews, are documented.
Backup information for analyses is maintained by the reviewers.
Changes in established polidies, prc:edures, f acilities, and equipment af fecting radiation and nuclear safety must be authorized in advance by the HS&E director.
Policies and criteria governing radiation and nuclear safety are estab-lished by the HS&E department and documented in Rocketdyne Doerating Policies; new policies and changes in existing policies are issued with the approval of the HS&E director.
Revised Rocketdyne Operating Policies are distributed to all managers.
Changes in the general methods of implementing these policies and criteria must be authorized by the HS&E director or his designee.
Changes in radiation safety procedures developed for specific operations must be authorized by the responsible operations and HS&E managers following a technical evaluation by HS&E personnel.
Also, unless previously approved, ESG-82-33 1.2-7
9 changes in nuclear safety procedures must be authorized by the Criticality Safeguards Advisor following a technical evaluation by Criticality Safeguards
. ('~~
personnel.
Technical Data Operations distributes revised procedures to all 3
personnel directly af fected, including the appropriate department director and the appropriate NSRP committee.
These revised procedures, which ata available in the work area, replace superseded documents as soon as changes are approved.
Written operating procedures are prepared for all routine or recurring operat u s.
These operating procedures must confort. to the provisions of the applicable Radiation Safety Plan and Nuclear Safety Analysis, and are approved in writing by the appropriate manager.
Approved procedures must be available to the user prio-to the implementation of new or modified activities or processes involving SNM.
When an approved operating procedure does not exist, a Controlled Work Permit must be prepared for work on or involving entry into an area or system containing SNM, where a potential exists for release of contamination, for entry into an airborne conta:nination area (10 CFR 20.203(d)), and for entry into a high radiation area (10 CFR 2.202 (b)(3)).
Each Controlled Work Permit must be signed by a representative of Radiation and Nuclear Safety, Unless previously approved, changes in nuclear fuel handling and radio-isotope and radiation device safety procedures developed for specific opera-tions must be authorized by responsible linv management and approved by the Criticality Safeguards Advisor or chairman of the appropriate safety review committee after their technical evaluation. All such changes are documented, either in memoranda or in revisions of ree tired of ficial documents, and are
[
distributed to all directly affected pers nnel.
N Facilities and equipment changes aff ecting radiation and nuclear safety are submitted to NSRP for review.
Recomr.endations resulting f rom such reviews are acted on by the appropriate approval authority.
Changes approved by these reviews are inspected prior to startup.
l 2.0 AUDITS AND INSPECTIONS Internal inspection and review is performed by the HS&E department staff and by the NSRP, under the direction of the NS&L and HS&E directors. Addi-tional effort of this nature is performed by the Employee Health and Safety Committee.
Inspections performed by the HS&E staff include daily radiation safety and weekly criticality control inspections in areas with significant radiation, contamination, or criticality hazards.
Audits of the radiation safety and criticality control programs are performed quarterly by individuals not directly responsible for the activities audited.
ESG-82-33 1.2-8 i
Quarterly reviews of the radiation safety program and the radiological environmental monitoring program are performed by the Radiation Safety manager O
or his designate.
U 2.8.1 Reporting Requirements l
Day-to-day operational health and safety conditions are observed by the I
health and safety staff in the HS&E department.
The staff is maintained at a level sufficient to provide essentially daily surveillance in active opera-tional areas.
The technicians are instructed to report any hazardous cc1di-tion or instance of noncompliance with regulatory directives to appropriate operational supervision and to HS&E supervisory personnel.
If for any reason the condition is not expeditious'y correctea, it is reported by HS&E personnel I
to the director of the HS&E den'.r cment, or his designated alternate, who in turn notifies the appropriate cepartment director.
The results of quarterly audits by the Criticality Safeguards Coordinator l
l and the Criticality Safeguards Advinor are documented and reported to the I
director of the NS&L department.
Radiologie., data (film badge and bioassay results, radiatien and con-tamination se ey results, effluent concentration results, environmental moni-
.toring resul a for each facility are reviewed at a minimum frequency of once l
per quarter by the health physicist responsible for radiation protection work in the facility.
Each health physicist is instructed to contact the appro-priate operational supervisor. and to arrange for any corrective action indicated by the data as being necessary.
If for any reason the corrective action is not taken expeditiously, the situation is reported to the HS&E department director, or his designated alternate, who in turn notifies the appropriate department director.
An annual report is prepared for the Radiation Safety Committee of the NSRP reviewing personnel exposure and effluent release data.
2.8.2 Periodic Safety Review Provisions Periodic reviews and evaluations of operational health and safety requirements and conditions are conducted by Criticality Safeguards, by the NSRP, and by the Employee Health and Safety committee.
The intent of the periodic saf ety reviews is to provide audit functions beyond the day-to-day operational surveillance provided by the HS&E department in the discharge of its duties as a staf f safety organization.
The Employee Health and Safety committee is described below.
2.8.2.1 Criticality Safecuards ESG-82-33
.v 1.2-9
Criticality Safeguards personnel (normally the Criticality Safeguards Coordinator)' periodically inspect working areas to ensure that all criticality
{) '
safety criteria and regulatory directives are being followed.
Inspection frequencies are established by the Criticality Safeguards Advisor. Minimum v
frequencies are:
1)
Major fuel handling areas, with large quantitles and frequent trans f ers-weekly 2)
Criticality control areas where the fuel handling load is small, such as analytical laboratories or storage areas---semiannually.
In areas where substandard practices are observed, the inspector is authorized and directed to discuss such practices with operational personnel to ensure that they have sufficient knowledge of approved procedures; the responsible manager normally accompanies the inspector during the next inspec-tion, which must be made within 1 week of the initial observation.
If prac-tices which could result in a definite hazard are observed during any inspec-tioc line supervision is immediately notified to suspend operations until corrective action is taken.
The Criticality Safeguards Advisor will also inspect the working area under the following minimum conditions:
1)
Before the start of a new project 2)
Within 1 week af ter the start of a new project
.p) t' 3)
Monthly in areas where the coordinator performs weekly inspections 4)
Annually in other areas.
These inspections will determine that the equipment, facility layout, and all variables associated with the criticality safety analyses conform to the con-ditions required by the analyses and any applicable license conditions. The advisor's inspections are documented.
2.8.2.2 Nuclear Saf eguards Review Panel Each committee of NSRP performs an inspection of every working area under L
its cognizance at frequencies' determined by the committee chairman.
Compre-l hensive reviews of safety conditions are also conducted at these times. The minimum inspection and review frequency is once per year.
Recommendations are approved by the NSRP chairman or his drsignee and transmitted to the appro-priate approval authority. Safety rev ews are required before potentially hazardous operations and are performed as the need arises.
Recommendations are transmitted to the appropriate approval authority following these reviews.
ESG-D2-33 I.2-10
2.8.2.3 Employee Health and Safety Committee
- ('
The Employee Health and Safety Committee is composed of workers from
\\
each operational department. The committee members are appointed by opera-tional supervision and serve for specified perious that normally do not exceed 1 year.
Each committee member is taught the basic safety subjects relevant to Rocketdyne operations.
He is required to make a bimonthly safety inspection of his area and to report his findings in writing to his supervisor for cor-rective action. This program results in specialized safety training for a large percentage of the company's employees.
2.8.2.4 Executive Safety Council The Executive Safety Council has prime responsibility for the Division -
health and safety policy consistent with the operations to be conducted at all facilities.
Chaired by the President and composed of top functional execu-tives, the Council provides management oversight of the.. Division's overall safety posture to ensJre that programs consistent with established policies are in effect. An inmediate convening of the Council may take place should it be indicated by the seriousness of a problem.
2.8.2.5 Senior Managemant Incident Review Committee
~Any incident involving a problem with people or systens that is indica-tive of inadequate Rocketdyne performance is reviewed by the Senior Management Incident Review Committee.
This review is to insure that Division personnel, products, and facilities are provided optimum protection.
" Lessons Learned" f rom-incidents are applied. across operations to minimize the chance of occur-O rences of the same nature in the future.
The Vice President, Quality Assurance & Systems Safety, conducts the affairs of the Committee which is composed of all functional executives, with the Director of Health, Saf ety and Environment serving as secretary.
Root causes of incidents are addressed to ensure corrective action taken meets the system deficiencies identified.
Action items are assigned to responsible functions and tracked until completion.
2.8.3 Transmittal of Findings to Manacement Inspection and review findings resulting f rom day-to-day surveillar:ce and f rcm periodic safety inspections are usually transmitted through personal dis-j cussion with af f ected personnel. At the discretion of the inspector or his 1
superior, internal memoranda may be used.
Internal memoranda are requred f or
'the transmission of all NSRP findings; standard distribution lists are estab-lished by the NSRP chairman. For the transmission of recommendations, these lists always include the appropriate approval authority. The approval signa-I' ture of the NSRP chairman or his designee is required for these memoranda.
l ESG-82-33 I.2-11 l
2.8.4 Enforcement of Safety Rules and Practices
/m Each department director is responsible for enforcing compliance with
-(
internal safety rules and operating procedures and additional conditions imposed by regulatory. directives (e.g., NRC and California licenses and regu-lations and DOE orders). This responsibility is implemented by directives to subordinate line supervision.
The degree of complisnce is audited by the HS&E department and by NSRP.
Instances of noncompliance are reported to line man-agers or directly to the department director, as necessary, to obtain correc-tive action.
2.9 INVESTIGATING AND REPORTING OFF-NORMAL OCCURRENCES Rockwell International will investigate and rep 3rt all events that sig-nificantly threaten or lessen the ef festiveness of h :alth, safety, or environ-mental protection.
It is the policy of Rockwell International (ROP M-501) to report all incidents internally to responsible of fice rs, including the Health, Safety & Environment and the Nuclear Safety and Licersing Department. TSis R0P requires NS&L to make the necessary reports to governmental regulatory agencies.
I The deportability of'each event will be determin!d on a case-by-case basis by the director of NS&L or his designee. The de.:ision will be based on the significance of the event, the requirements of the regulations and license conditions, and the guidance given by Regulatory Guidt 10.1, " Compilation of Reporting Requirements Subject to NRC Regulations."
2.10 RECORDS b
Minutes of meetings of the Nuclear Safeguards Review Panel and of its subcommittees and correspondence involving committee recommendations and actions are maintained by NS&L for a period of at leas: 6 months after term-ination of the project or for 2 years af ter their date of preparation, which-ever is longer.
Correspondence pertaining only to panel membership is spe-cifically exempted from this requirement.
Nuclear safety analyses and criticality studies ard other written reports prepared by the Criticality Safeguards Coordinator or (riticality Safeguards Advisor regarding radiation and nuclear safety are alst maintained by HS&E for a period of at least 2 years af ter the date of preparation.
Source leak test records are kept for 5 years.
The reports and records of reportable of f-normal occurrences are main-
)
tained in HS&E files for a period of at least 2 years.
Records of audits and i
inspections perf ormed by or for the HS&E Department are kept for at least 2 years.
ESG-82-33 I.2-12 i
l
l
}
Personnel exposures are maintained and stored until disposition is authorized -by the Commission (10 CFR 20.401C(c)(1)).
ALARA findings, records of routine radiation surveys, and environmental surveys are reported and distributed in annual reports by HS&E and are kept for at least 2 years.
The data are not stored.
Training and retraining records required by Section 2.6 are maintained by the Training Department for a period of at least 2 years.
Instrument calibra-tion records are kept for at least 2 years.
i 1
Alterations or additions to the plant are made on the file copies of the facility structural drawings and descriptions and are maintained for the life of the plant as altered, by the Facil~lties & Industrial Engineering Department.
0248Y/clh O
I i
l l
ESG-82-33 I.2-13 l
l
)
3.0 RADIATION PROTECTION Oh.
3.1 SPECIAL ADMINISTRATIVE REQUIREMENTS 3.1.1 Controlled Work Permit Entry into posted areas where there is likely to be external radiation or airborne radioactivity in excess of acceptable levels for continuous work is cMtrolled. Varying degrees of control, consistent with the hazard involved, are exercised over posted creas by Radiation and Nuclear Safety.
The Con-trolled Work Permit (Form 719-L) is a means of restricting access to posted areas on the basis of types of personnel and potential hazards.
The highest degree of hazard is associated with an area in which the active contamination and radiation levels are of such significance that special rigid entry con-trols and safety precautions are necessary.
This form requires approval of the managers controlling areas where work is to be performed, the personnel who are responsible for performing the work, and the Radiation and Nuclear j
Safety representative.
3.1.2 ALARA Policy Along with its other responsibilities in maintaining an effective program of industrial and environmental safety, Rocketdyne is concerned with minimiz-ing any adverse effects due to operations with radioactive materials. This cone.ern is consistent with NPC requirements for maintaining radiation expo-sures as icw as reasonably achievable (ALARA) and is expressed in the form of cn ALARA program for minimizing and controlling radiological hazards.
tO^
The major management directive relative to the ALARA program is contained in Rocketdyne Operating Policies (ROP) M-500 and M-504.
Rocketdyne Operating Policies are available to all members of management and form the basic policy statements for operations at Rocketdyne.
These specific R0P's delineate responsibilities of operating supervision, the Nuclear Safeguards Review Panel, project organizations, Protective Services, and the Health, Safety &
Environment Department toward controlling and minimizing radiation doses.
The direct goal of radiological safety procedures, including design, review, operations, training, and monitoring, is contained in the policy statement- "Rocketdyne's policy is to hold all exposures to ionizing radiatici to the minimum possible."
These directives are expanded f urther tri R0P M-508, " Areas Requiring Special Saf ety Precautions" (radiological and nonradiological), and HS&E Procedure G-01, " Radioactive Materials and Ionizing Radiation."
I
)
ESG-82-33
- 1. 3 -1 1
E 3.1.3 Radiation Safety Plans 1
[~'T Radiation Safety Plans are prepared to cover all activities with SNM, C /.
addressing all appropriate aspects of radiation safety, and are approved by the NSRP and the HS&E Department.
These plans are revised when necessary to I
ef fectively reflect current operating conditions.
3.2 TECHNICAL REQUIREMENTS 3.2.1 Access Control At SSFL, there are controlled areas in the Rockwell International Hot Laboratories (Building 020).
These areas are posted in accordance with 10 CFR 20.203.
The procedures for exiting and entering these areas, including stepof f pads, change facilities, and protective clothing requirements, are provided in Section 12.
3.2.2 Ventilation Requirements The efficiency of f acility exhaust filter bank systems is assessed by in-place testing,- utilizing " cold" dioctylsebecate (DOS) and a forward light scattering photometer, or equivalent, at filter installation / replacement or, as a minimum for those filter systems required to limit offsite consequences of a radiological accident, at a testing f requency of once per year.
This testing requirement applies only to the final stage of filters.
Exhaust sys-tem flow rates are checked annually and whenever any process changes are made
-,,)
that may alter flow rates.
(.q Minimum Performance Specifications, Efficiency The filter bank ef ficiency for particles of 0.8 pm diameter is 99% for uranium and low-toxicity radionuclides areas using a standard " cold" 00S test.
For plutonium and high-toxicity radionuclides areas, the filter media ef f1-ciency is 99.97% for particles of 0.3 um diameter and the filter bank ef fi-ciency is 99.95% f or particles of 0.8 pm diameter.
Minimum Performance Specifications. Fire Resistance All filters are constructed of fire-resistant materials.
For uranium and low-toxicity radionuclides areas, the filters cre capable of continuous opera-tion at 250*F with no loss in filtration effi:iency.
For plutonium and high-toxicity radionuclides, the filters are capable of operation for 5 min at 700 150*F with no loss in filtration ef ficiency.
Filters used specifically for mater'el recovery do not need to satisfy these requirements if utilized in conjui tion with filters for radiological control purposes.
ESG-82-33 I.3-2 l
M,_iyimum Installation Specification A
The filters are located at a sufficient distance from the working areas V
to assure the maintenance of filter integrity under anticipated accident con-ditions.
The filters are protected by prefilters or systems adequate to remove entrained incandescent particles.
Exhaust from enclosures and rooms containing unencapsulated enriched uranium or radionuclides of similar or lesser radiotoxicity may be filtered by a single-stage HEPA filter.
Exhaust from enclosures containing unencapsulated plutonium or radionuclides of similar or greater radiotoxicity must have three stages of HEPA filtration.
Rooms containing such enclosures must have two stages-of HEPA-filtration These filtration requirements are not applied to the use of small quantities of plutonium or similar material for gamma spec-troscopy and other analyses.
3.2.2.1 Enclosures The specifications for enclosures are listed below.
1)
Enclosures such as fume hoods and glove boxes must be provided for all operations involving radioactive materials or nuclear fuels that under normal conditions, generate radioactive mate-rial concentrations in air it, excess of occupational stand-ards.
Refer to Tables 3-1 and -2 and Specification 14 for criteria governing the type of enclosure to be used for the more radiotoxic nuclides.
O 2)
The minimum average f ace velocity for a fume hood with the sash d
in its proper operating position, and for an opening in a special enclosure, is 100 ft/ min.
For dusty operations, the minimum is 150 ft/ min.
3)
For machining or other operations that could impart high speed to a dust particle, the average face velocity at the enclosure opening must be at least 150 ft/ min, l
j l
4)
Glove boxes are required for work with any radioactive material for which the quantity, specific activity, radiotoxicity, physical and chemical form, and existing or potential environ-ment (thermal, pressure, chemical, etc.) could, in combination,
)
create potential hazards of such magnitude that fume hoods and special enclosures wculd offer insufficient personnel protection.
5)
Separate minimum requirements are established for glove boxes according to use.
High-level glove boxes are required for work with large quantities of plutonium, and other radionuclides of similar radiotoxicity, and for less radiotoxic nuclides in such i
ESG-82-33 1.3-3 l
A
TABLE 3-1 i
' ~ '
ENCLOSURE REQUIREMENTS FOR HIGHLY RADIOT0XIC NUCLIDES Quantity Enclosure 2.0.12100 pCi Fume hood
>100 uti but 110 mci Low-level glove box
>10 mci High-level glove box TABLE 3-2 MODIFICATION FACTORS FOR ENCLOSURE REQUIREMENTS Operational Procedure Factor Normal operations 1
Precipitation Filtration or centrifuging Solvent extraction in mixer settlers Chromatography Pipetting (not by mouth) or titrating active solutions Cleaning and degreasing q
' i,y Storage (temporary) 0.01 Simple wet operations 0.1 Diluting stock solutions for use Washing precipitates Complex wet operations with risk of spills 10 Distillation procedures Solvent extraction in pulsed column Sampling and transfer of solutions Evaporation to dryness using heat Simple dry operations Fusing procedures for preparation of solution 10 Fluorination Transfer of dried precipitates Complex dry and dusty operations 100 Machine or hand crushing Machining or sawing metals Sieving Vigorous mixing by machine Melting operations 1000 ESG-82-33 g
- L I.3-4
L large quantities that similar potential hazards exist.
Low-level glove boxes are permitted for. work with uranium and with
-k
).
suf ficiently small quantitles of other radionuclides.
e o
6)
Glove boxes must be maintained at negative pressure with i
respect to:the room housing them. The minimum pressure differ-ential requirements are -0.3 in. of water for high-level glove l
boxes and --0.1 in, of water for low-level boxes.
L An exception to this specification is that if the use of the glove box is for reasons other than control of airborne con-tamination, the negative pressure requirement need not be met.
7)
All glove box exhausts must be connected with exhaust systems meeting -all specifications listed in Subsection 3.'2.2 immedi-ately preceding.
8)
Humidity-resistant, high-ef ficiency filters are also' required at the inlet and outlet of high-level glove boxes.
9)
An emergency exhaust capability, automatically actuated by loss of. pressure differential, must be provided for high-level glove boxes to assure a minimum airflow rate of 150 f t/ min into a box through any two, completely open glove ports. This requirement may be waived if the exhaust system routinely provides a con-tinuous exhaust capability meeting the flow rate criterion.
. 4
- 10) A manually actuated emergency exhaust capability must be provided for low-level glove boxes to assure a minimum airflow rate of 100 f t/ min into a box through one completely open glove port.
This requirement may be waived if the exhaust system routinely provides a continuous capability meeting the flow rate criterion.
- 11) The maximum glove box leak rates, with the boxes at -0.6 in. of water with respect to the room, are 1% of the glove box volume in 24 h.
- 12) An inert atmosphere is required in a glove box if necessary to.
prevent hazardous chemical reactions.
13)
Glove-box shielding, either structurally mounted or internally arranged, is required if necessary to ensure that regulatory standards for external dose are not exceeded.
Remote systems are required if the hand dose is limiting and the radiation levels preclude glove work without exceeding regulatory external dose standards for extremities.
t 1.3-5
l'
- 14) Radioactive materials for which the RCG in air is less than 4 x 10-10pti/cc'are considered to be highly radiotoxic if the rN specific activity is sufficiently high that the RCG, exprested in terms of mass, is less than 4 x 10-8pg/cc.
For these radionuclides, types of enclosures are specified according to the radioactive material quantities involved and according to the operatiocs to be performed.
These specifications appear in Table 3-1 and-Table 3-2. : Table 3-1 provides basic quantity limits, while modification factors for these limits appear in Table 3-2.
The enclosure modification factors are contingent upon the type of operation involved.
To establish the appli-t cable modification factor:
(1) the particular operation (s) to be conducted, and the quantity of material to be involved in the operation (s),'are first established, and (2) the type of enclosure required is then determined by entering Table 3-1 with the product of the quantity of material to be involved and the modification factor. The modification factor is selected from Table 3-2, on the basis of the type of operation to be performed.
15)
Plutonium processing operations involving more than 2 grams, where the material is not encapsulated, will be performed in enclosures providing three stages of high-efficiency filtra-tion. However, only the final stage of filtration will require testing upon initial installation and upon each replacement.
- 16) Concerning the radioactive material inventory limits at work stations, the amount of radioactive material which may be pres-s ent at a work station is a function of the radiotoxic aspects of the material and of criticality considerations.
The cri-teria listed below also consider the physical form of the mate-rial in determining the quantity which may be present at a work
- station, a)
Normal Form--The amount of plutonium and uranium in normal form that is permitted at a work station within a facility is a function of the type of enclosure used for the opera-tion to be performed.
The criteria for determining this amount of material is given in Specification 14 for radio-logical consideration and in Section 4-D for nuclear safety considerations.
b)
Encapsulated Form or Special Form--The amount of plutonium and uranium encapsulated or in special form that is per-mitted at a work station or for storage within a facility is a function of the environment to which it would nor-mally be subjected. The limitations imposed on the amount of material permitted by radiotoxic considerations are exempted, provided the encapsulation or special form is i
ESG-82-33 I.3-6 l
4
maintained within the anticipated environmental s inditions.
Nuclear saf ety ' considerations become the limiting f actor for p-determining the amount of-material permitted at the work Q
station.
3.2.2.2 Hot Cells Safety requirements for hot cell operations are listed below.
1)
Hot cells are required for work with radioactive materials if radiation levels and potential air contamination levels pre-clude work accomplishment in manipulator Slove boxes without exceeding regulatory radiation protection standards, 2)
Cells must be maint.ained at a negative pressure with respect to adjacent areas if potential airborne radioactive material haz-ards exist.
The rormal pressure dif ferential requirement is
-0.5 to -0.1 in. st water; under certain operating conditions the dif f erential pressure can be administratively controlled f rom -0.05 to 0.1 in. of water.
This condition exists when an attempt is being made to establish an inert atmosphere for operation with pyrophoric materials and it is necessary for operations personnel to feed nitrogen continuously.
These requirements may be waived during filter changing, ventilation system maintenance, and personnel entry into a cell.
3)
The ventilation system for a cell in which potential air con-I~
tamination nazards exist may be shut down for filter changing and/cr maintenance only if cell operations are discontinued and all cell openir.gs are sealed.
4)
During. personnel entry into 6 cell in which potential air con-tamination hazards exist, a linear airflow rate of 100 f t/ min is required across the door face and intc the cell.
5)
Hot cells are to be used for work with materials, such as irradiated plutonium, which, in the absence of severe radiation levels, would require enclosures to minimize the spread of contamination.
6)
Cell exhausts inust be connected with exhaust systems meeting all specifications listed in Subsection 3.2.2.
l-7)
An inert atmosphere is required for a cell to prevent hazardous I
chemical reactions.
8)
Shielding is required in all directions to ensure that (1) per-sonnel working full-time in adjacent areas are not subjected to an annual whole body dose greater than 5 rem; (2) the dose rate outside the f acility or room in which the cell is housed is l
suf ficiently low that personnel on Rocketdyne premises could ESG-82-33 1.3-7
_ = _
not be subjected to a dose rate greater that 2 mrem /h; and (3) personnel offsite could not receive a dose greater than 2 mrem (m)
-in 1 h, or 100 mrem in 7 consecutive days, or 0.5 rem in 1 yr.
wJ 9)
Electrical and mechanical interlocks are required as necessary to ensure that personnel are not exposed to inadequately shielded sources within a cell.
3.2.3 Instruments (Survey. Countino. Criticality Monitors)
The kinds of equipment used in radiation detection by the HS&E depart-ment, available f or portable survey, air sampling, criticality monitoring, and sample counting, include the following:
DC Powered (portable instrumentation)
Category 1 0 y ionization-type survey meter; sensitivity -'O.1 to <5000 mR/h Category 2 Thin-window pancake Geiger-type survey meter; sensitivity - background to 5 x 105 counts / min Category 3 a scintillation-type survey meter; sensitivity - background to 105 counts / min Category 4 6 y Geiger-type survey meter; sensitivity - 0.1 to 200 mR/h Ad Category 5 07 Geiger-type survey meter; sensitivity - 0.1 to 1000 R/hr Category 6 n 1.il (scintillation). neutron monitor; sensitivity - 0.1 to 103 n/s Category 7 y scintillation low-range survey meter; sensitivity - background to 3 mR/h AC Powered Category 8 Counting scalers and sample changing system; o, 8 sensitivity to a and B radiation -
0.1 to 106 counts / min Category 9 Count rate meter for a or 8 radiation Category 10 Air monitor 0-y Category 11 Air monitor a Category 12 SAM-II special nuclear materials assay monitor Category 13 Multichannel analyzer MI (Tl) or Ge detectors ESG-82-33 I.3-8
Calibration ~and maintenance of this equipment are controlled by the Instrument Calibration Laboratory and performed periodically to manufacturer f~
and/or user recommendations.
Instruments are calibrated after repair and on a regular schedule.
Battery-powered instruments are calibrated every 13 weeks; AC-powered instruments are caliDratti every 6 months.
If the manufacturer recommends more frequent servicing, that interval is used for routine calibra-tion.
Calibration sources are traceable to the National Institute of Stan-dards and Technology.
3.2.4 Internal and External Exposure Performance criteria for each of the several systems or programs impor-tant to control of personal internal or external exposures are discussed in the following subsections.
3.2.4.1 Ventilation A ventilation system is provided to prevent inhalation exposures in excess of occupational standards established in 10 CFR 20* or CAC Title 17.*
Action levels based on releases f rom the ventilation system are covered in the Onsite Radiological Contingency Plan, RI/R0 88-206, July 25, 1988.1 i
Action l
levels for normal operation are set at the maximum permissible concentration (MPC) limits for the most restrictive radionuclides present.
3.2.4.2 Air Samplinq 1)
Low-volume particulate air samplers (about i ft3/ min) or lapel air samplers (about 5 f t3/h) are required in all loca-(
tions for special situations where there is a potential for eirborne radioactive material in concentrations in excess of 25'I of the occupational exposure standards, based on knowledge of the material, facility, and operations.
(The only signifi-ccnt gaseous activity encountered in Rocketdyne operations is the fission product, Kr-85.
Exposure to this material, which is released only when a fuel element is cut open or when unciad fuel is heated or disrupted, is monitored by use of beta-sensitive survey instruments and film badges.
The MPC for this radionuclides is based on limitirg skin dose.)
l 2)
To the extent practical, sampler inlets are located at posi-tions representative of the air to which workers are exposed.
- 10 CFR 20 is an abbreviation for Title 10, Code of Federal Regulations, Part 20; this regulation is applicable to work with special nuclear materials, i
CAC Title 17 is an abbreviation for California Administrative Code, Title 17; this regulation is applicable to work with radioactive materials other than special nuclear materials.
ESG-82-33 1.3-9 l
l
l l
\\
3)
Samples are renmed for evaluation in counting systems et the end of each shift cr weekly as appropriate to monitor for p
potential airborne exposures.
G 4)
In areas used for work with plutonium and other nuclides of similar radiotoxici.y (having an activity PPC less than 4 x 10-10pCi/ml and a mass MPC less than 4 x 10-8pg/ml),
l samplers are required at each glove box, fume hood, or similar containment equipment '.,i which the radioactive material is l
handled.
5)
Air must be drawn through filters by use of vacuum pumps (Gast Manufacturing Model 0211-V103-GB or equivalent) equipped with filter holders (Gelman Instruments Model 4170 or equivalent),
or by use of house vacuum lines equipped with filter holders (Gelman Instruments Model 4170 or equivalent). A typical lepel t
sampler would be Mine Safety Appliance Monitair or equivalent.
i r
6)
Filter paper ef ficiency is at least 99% for particles of 0.3-pm diameter or greater.
7)
Provisions are made for the determination, with 20% accuracy, of the volume of air drawn through the filter.
8)
Sampler flow rate calibration and adjustment are performed every 6 months 3.2.4.3 Bioassay
+
i Bicassay, including urine and fecal analysis, is provided for personnel working in posted areas. Samples are collected and analyzed at regular fre-quencies. The sampling analyses performed are determined for each individual
?
according to his frequency of contact with radioactive material, the nature of the tasks performed, the quality and quantity of radioactive material in-volved, and the physical and chemical form of the material.
Collections are normally at annual, semiannual, or quarterly intervals.
Fecal specimens and additional urine specimens are collected if urine specimens indicate a sig-nificant internal deposition may have occurred.
In addition, fecal specimens are collected as appropriate when employees may be exposed to radioactive materials that are known to be highly insoluble in the lungs.
The bioassay program ccqforms te Regulatory Guide 8.11 for uranium and is consistent with its goals for other radionucl1 des.
I 3.2.4.4 Protectiva Clothing Protective clothing is required for all who inter posted areas via change rooms or stepoff areas.
Required protective clothing generally consists of
. j' laboratory smocks and shoe covers worn over personal clothing. However, addi-tional specialized clothing is used when warranted by the nature of the opera-l ESG-82-33 I.3-10
._-_--_--_________________-_________-_-___--_-__--____-_____--____a
tion, quantity and quality of material involved, and/or chemical and physical form of the material involved. Protective clothing requirements are estab-f lished by HS&E and may include any or all of the following:
laboratory
\\
smocks, coveralls, surgeon's caps, hoods, surgeon's gloves, cloth gloves, shoes, canvas shoe covers, plastic booties, and respiratory _ protective devices. As'an example, use of gloves is mandatory for operations that r.equire hand manipulation of unencapsulated radioactive materials.
Depending upon the chemical and physical form of the material, the gloves required'may be rubber, canvas, leather, asbestos, etc., or combinations thereof.
Laboratory smocks and coveralls are red-trimmed, or blue-trimmed for use
'in plutonium areas. This identification is used only for smocks and coveralls to be used exclusively in radiologically posted areas. Smocks and coveralls are pocketless and have no openings to permit reaching into the pockets of personal clothing 3.2.4.5 R'espirctory Protection Respiratory protective devices are normally used for nonroutine opera-tions (e.g., cutting an exhaust duct containing residual radioactive mate-rial) or for emergencies that might produce airborne radioactive material in excess of the limits specified in Appendix B, Table 1, Column 1,10 CFR 20, and subject to the conditions specified in 10 CFR 20.103c.
3.2.4.6 Surf ace Contamination Monitoring Measurements of surf ace contamination are required in all areas where work with unencapsulated radioactive material is done at Rocketdyne. Such measurements are also required for materials and equipment prior to removal f rom areas in which work with unencapsulated radioactive materials is conducted.
For all measurements, the use of instrumentation appropriate for the type of emitter involved is required.
In the case of uranium, alpha measurements are required for enriched uranium and also for normal and depleted uranium that has been subjected to a melting process within the previous 120 days.
Beta measurements are permitted for normal and depleted uranium that has not been subjected to recent melting.
Action Guides and Upper Limits for removable surface contamination are shown in Table 3-3.
The Action Guide is that level of contamination that indicates the need for prompt decontamination without interrupting the work schedule.
It is considered an ALARA goal in all decontamination operations.
The Upper Limit is that level of contamination that indicates the need for an immediate halt of klork, with appropriate decontamination of the area. The area must be decontaminated to levels below the Upper Limit and as close to or below the Action Cuide as is reasonably achievable, considering the time and effort required, the interference with productive work, the potential for 3
recontamination, the amount of waste to be produced, personnel exposure, and similar concerns.
ESG-82-33 1.3-11 1
1 l
1 TABLE 3-3 (j
ACCEPTABLE REMOVABLE SURFACE CONTAMINATION Action Guide Upper Limits 2
l (dpm/100 cm )
Restricted Areas l
Contamination Areas /
Radiation Areas /
Airborne Restricted Type of Unrestricted Radioactive Radioactivity
. Access Contaminant Areas Material Areas Area Areas Transuranic, Ra-226, Rs-228, Th-230, Th-228, h
y
[
Unspecified Pa-231, Ac-227, I-125, I-129 Th-nat, Th-232, Sr-90, Ra-223, 20 100 200 Ra-224, U-232, EU TOM T60UU Unspecified 1-126, I-131, I-133 U-nat, U-235, 20 100 200 (V
)
U-238, and associ-7gg TGM 200000 Unspecified a
a ated decay products i
Beta-gamma emitters I
(nuclides with decay modes other than alpha emission or spon-100 1000 5000 6T taneous fission)
TOM TJ0u SY SY Unspecified 200000 except Sr-90 and others noted above.
Increase by 3X for tritium The classification of contaminants is that provided by the NRC " Guide-lines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material," and the Upper Limits shown for Unrestricted Areas are exactly those provided by the " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses i
for Byproduct, Source, or Special Nuclear Material." Upper limits for the 5
other areas were then derived by first considering a factor of 30 increase for Restricted Areas relative to Unrestricted Areas, based on the ratio of the occupational MPC for airborne radioactivity to the nontccupational MPC.
Because additional precautions are taken for Contamination Areas and Airborne ESG-82-33 1.3-12
I p
> Radioactivity Areas, relative to Radiation Areas and Radioactive Material Areas, such as protective clothing, exit surveys, reduced occupancy, and C-f requent monitoring, an additional f actor of 10 was applied for limits in these areas. - The actual Upper Limits adopted are reduced somewhat f rom these s
arithmetically derived values, as can be seen by multiplying the Unrestricted Area values by 30, and by 300.
Restricted Access ' Areas' are those areas, such as the interior of a hot cell or similar special enclosure, or a radioactive exhaust system filter plenum, that by their nature may have very high levels of contamination and have inf requent entries of short duration under special procedural control to minimize personnel exposure.
Therefore, neither Upper Limits nor Action l
Guides are specified for these areas.
The values adopted as Action Guides were chosen as those value; that can, in most cases, be achieved with an acceptable amount of time and effort, interference with productive work, and not excessive quantities of waste.
An additional limit is applied for surface dose rate of 0.2 mrad /h aver-age (1.0 mrad /h maximum) at 1 cm through 7 mg/cm2 absorber, where appropri-ate. Soil contamination limits are generally set at 100 pCi/g gross detecta-ble beta activity (including naturally occurring radioactivity) for mixed fission products-and activation products, and 30 pCi/g alpha activity (above natural background) for enriched uranium. A limit cf 10 pCi/g alpha activity (above natural background) for hazardous alpha emitters (plutonium, U-233, Am-241, etc.) is used.
For packages offered for shipment, and vehicles used in transportation,
(
the limits provided in 49 CFR 173.443 are used.
I 3.2.4.7 Decontamination Decontamination of personnel is accomplished as needed on a case-by-case basis and is initiated as required by HS&E.
The decontamination is then per-formed in a manner specified by HS&E and/or Medical, as appropriate.
3.2.4.8 Emeraency Evacuation The performance requirements for the protection of personnel in the event of an incident requiring evacuation is covered in the Onsite Radiological Con-tingency Plan, RI/RD88-206, July 25,1988.1 3.2.4.9 Personnel Monitoring (External Radiation) 3.2.4.9.1 External Dose Measurement 1)
Personnel external dose measurement is made by film dosimetry or an equivalent method.
2)
Measurement devices af ford a means of assessing gamma, beta, and, if appropriate, fast-neutron dose to the wearer.
ESG-82-33 1.3-13
r_______
3)
Whole-body dose measuring devices include metal foils and other materials as necessary to provide a means of assessing' neutron V(1-dose in the event of a criticality accident for those personnel routinely assigned to fuel handling operations.
3.2.4.9.2 External Dose Rate Measurement 1)
Beta, gamma, and mixed beta-gamma radiation intensities are evaluated as appropriate to the magnitude of the radiation intensity, with instruments capable of measurement to at least 5 R/h. Measurements at low intensities (0 to 3 mR/h) are per-formed with a Ludlum Model 12S or equivalent survey instrument.
2)
Low-energy gamma or X-ray radiation is evaluated with an ion chamber of suitable energy response or equivalent.
3)
All portable battery-powered radiation intensity measuring instruments are serviced at least once every 3 months. Main-Lance includes replacement of batteries, if necessary, check-out and repair of circuits and components, as necessary and calibration, if required.
4)
Calibration is performed with sources of known energy that' represent the energy of the radioisotopes being measured.
3.2.5 Sealed Source Leek Tests g
Sealed sources of radioactive material containing activities greater thcn that specified in Appendix B of the California Radiation Control Regulations (CAC17, Section 30356), except for gaseous tritium and krypton-85, are tested for contamination and leakage prior to initial use and every 6 months there-after. The test consists of a smear taken from the surface of the source or appropriate accessible surfaces of a storage container or other sucn device.
Sources that are stored and not being used are not leak tested until ready for use or transfer.
If the test shows the presence of 0.005pti of removable contamination, the source is withdrawn from use for decontamination and repair or for disposal as radioactive waste. A report is made to the proper regulatory agency.
0249Y/ reg ESG-82-33 1.3-14
F
(
p 4.0 NUCLEAR CRITICALITY SAFETY 4.1 SPECIAL ADMINISTRATIVE REQUIREMENTS 4.1.1 Criteria for Establishing Plan of Action Before a project involving out-of-reactor handling of fissile materials can be started, project personnel must contact a Criticality Safeg'uards rep-resentative to determine the appropriate plan of action for obtaining the necessary approvals. The criteria used in establishing the requirements for this plan of action are presented in Table 4-1.
TABLE 4-1 PLAN-0F-ACTION CRITERIA Amount of Fissile Material (9)
Nuclear Safety Enrichment Range Analysis Only
( f.)
Required 0.1 - 100 Pu-238, Pu-239, Pu-241, 2100 Pu-233, Cm-244 e
30 - 100 U-235 2200 10 - 30 U-225 1250 5 - 10 U-235 1300 2 - 5 U-235 1400 0.7 - 2 U-235 1900 0 - 0.7 U-235
- Although restrictions for fuels in this enrichment range are not required when fuel is handleo alone, Criticality Safeguards must be notified to deter-mine whether restrictions must be placed on the handling of fuels of this enrichment range because of the possible presence of fuels of higher enrich-i ments in the same area.
4.1.2 Verbal Approval Any project involving less fuel than that given in Table 4-1 must 1
receive approval of the Criticality Safeguards function. These verbal approvals are necessary for the control of fissile materials. Prior to giv-ing approval in such cases, Criticality Safeguards must first determine that ESG-82-33 l
O I.4-1 w__-_______---
y
-there is no interaction, affecting nuclear safety, with any other fuel
. handling projects, regardless of their nature, in any of the proposed fuel handling areas.
" Interaction" as used here is defined as the condition that would~ exist-if there were any additional quantity of fissile material or reflector material more effective than water in either the room or the criticality control area where the material is to be handled.
If inter-action may affect nuclear safety, a Nuclear Safety Analysis.is required, detailing the safe handling procedures to be utilized for the proposed project.
4.1.3 Huclear Safety Analysis-y All projects involving the handling of fuels in quantities greater than those listed.in Table 4-1 require the completion of a Nuclear Safety Analy-l sis. To obtain the required approvals, including the one recommended by the chairman of the Fuels Committee of the ESG Nuclear Safeguards Review Panel, a formal presentation to the committee may be required. This presentation is required whenever there is a new process step involving safety practices not previously approved. The project can be started only after the chairman of the Fuels Committee recommends approval of the Nuclear Safety Analysis and after it is approved by the approval authority. After its review by project supervision and Criticality Safeguards, the review of the Nuclear Safety Analysis by the chairman of the Fuels Committee also constitutes an independent verification of the safety of the analysis..Before. recommending approval, the chairman may call for the advice of the physicist on his com g f
mittee, of the antire committee, or of any other appropriate expert. The i
chairman seeks the advice of the abysicist when he does not know that the
-basis for the nuclear safety of tie operation has been approved or when he desires confirmation of its safety.
4.1.4 Approval The project must prepare a Nuclear Safety Analysis when the quantity of fuel involved is greater than that given in Table 4-1.
If the project involves criteria, operating philosophy, or methods of analysis not approved in the license or amendments to it, approval of a license amendment is required in addition to the' Nuclear Safety Analysis. NRC approval is required for all projects where the safe handling criteria are beyond those contained herein or in subsequent amendments.
4.1.5 Inspection Prior to use with significant amounts of fissile material, all equip-n,ent, containers, and storage racks that provide criticality prevention functions, shall be inspected and tested. Use of such equipment shall be i
approved by the Criticality Safeguards Advisor only after his determination that the equipment is satisfactory. This requirement also applies to modification of such equipment.
4 ESG-82-33 1.4-2 I
___-...--..-__.__.___.------.-L.-____._-u.._.--
-.. - _ - - -. - - -. _ _ _ _. - _ - - - _ - _ - ~ _. - - - - - _ _. _
_-_-.__-.2--
a
I.
l 4.2 TECHNICAL REQUIREMENTS l
The following criteria are used in nuclear safety calculations at the Energy Systems Group. Additionally, the safety factors used for various i
t configurations and arrays are presented. Specific approval is required from j
NRC for changes in these criteria.
I 4.2.1 Single-Unit criteria The' single-unit criteria are listed below:
1)
For batch control (mass limiting), the maximum safe batch size shall be 45% of critical where double-batching is pos-sible.
Where double-batching is not possible (e.g., the con-tainer cannot hold two batches), the maximum safe batch size may be increased to 75% of critical.
2)
For volume control (independent of container shape), the maximum safe volume shall be no more than 75% of critical.
3)
For cylinder diameter control, the maximum safe cylinder dia-meter (independent of cylinder height) shall be no more than 85% of critical.
4)
For slab thickness control, the maximum slab thickness (inde-A pendent of slab cross section) shall be no more than 80% of critical.
5)
For a coll ution of long fuel rods or similar linear arrange-ment, the maximum safe mass per unit length shall be no more than 45% of critical.
I 6)
For a single slab, the maximum safe mass per unit area shall be nc more than 45% of critical.
7)
For concentration control (independent of container size),
the maximum safe concentration shall be limited to an H/Pu-239 atomic ratio of.tB200. For aqueous solutions, this corresponds to a concentration of f_3.5 g plutonium per liter.
It may be desirable to store dilute aqueous solutions (i.e., pickling solt:t tons).
Pu solutions having an H/Pu-239 ratio.18200 are safe. The maximum allowable Pu content of a container (independent of container geometry) will be 450 g.
Under Jostulated accident conditions, however, precipitation may taie place, resulting in an increase in concentration.
The 450 g of contained Pu is no more than 90% of critical, independent of concentration or container geometry, f*b d
ESG-82-33 I.4-3
8)
For concentration control (independent of container size),
f~3 the maximum safe con:entration shall be limited to an H/U-235 s_ )
or H/U-233 atomic ratio of15200. For aquecus solutions, this corresponds to a concentration of 15.0 g of U-235 or U-233 per liter.
It may be desirable to store dilute aqueous solutions- (i.e., pickling solutions). Uranium solu ions hav-ing an H/U-235 or H/U-233 ratio of 15200 are safe.
The maximum allowable U content of a container (independent of container geometry) will be 700 and 503 g for U-235 and U-233, respectively. Under postulated accident conditions, however, precipitation may take place, resalting in an increase in concentration. The given quantities of contained U are no more than 90% of critical, independent of concentra-tion or container geometry.
9)
For safe assemblies based on keff, maximum safe keff shall be no more than 0.95, taking into account the uncer-tainties in the calculations.
The following formulation is used in calculating the keff of a single 2 2 1+MB k,
g eff
- 3,g B r
2 2 2 2
),gB G
G p
h These formulations are used with data tabulated for highly enriched uranium in Y-1272 to yield values of keff that are conservative when compared with experimental data,3 for well moderated (H:U1100) solution or metal-water systems.
I Alternative methods of calculation, including computer programs such as KENO, may only'be used when fully validated for each application. This validation must be done according to the procedures described in ANSI Stan-dard N16.9-1975, " Validation of Calculational Methods for Nuclear Critical-ity Safety," as specified by Regulatory Guide 3.41. This validation must l
assure sufficient similarity between the physical configuration being cal-culated and the calculational model, and between the various experimental systems used for correlation to minimize systematic errors. This must include geometry, materials significant characteristics.present and concentrations, density, and other Linear and nonlinear trends must be deter-mined, and particular care must be taken to assure that neither interpola-tion nor extrapolation is outside the range of applicability as demonstrated by the experimental correlations. Small biases mcy be adjusted; large biases indicate a possibly unreliable calculation.
Uncertainties must be determined from the experimental data, the calculational results, the accur-acy of the representations, and other factors. These.are added to the adjusted calculated value for ke 0.95 at a 95% confidence level. ff so that the upper bound must be below Prior to any use of computer codes for
!L ESG-82-33 1.4-4
criticality control, a demonstration and validation must be submitted to NRC
.hc ~for review and approval.
Details of all calculations are documented and presented to the physi-cist or other qualified member of the Fuels Committee for second-party review.
These criteria are based on normal process conditions (with or without moderation as the process requires) but always assume full water reflection unless eflectors more effective than water are present, in which case the more,ffective reflector is considered.
Under accident conditions (i.e.,
wat w flooding due to sprinklers), the system would remain soberitical. ~
For normal process cor.ditions, individual units are analyzed with a full reflector. For storage arrays, analyzed by the density analog method, actual reflection of each unit is assumed; full reflection is assumed for the array. For accident conditions resulting in partial moderation, such as spray by a fire sprinkler, nominal reflection is assumed unless full reflec-tion may be provided by surrounding materials. Data corresponding to these classes of reflection are presented in TID-7016 (Reference 4, Subtritical Parameters), ARH-600 (Reference 5, Reflector Savings), and TID-7028 (Refer-ence 2, Equivalence of Various Reflector Materials).
Safety criteria are based on experimental data or on criticality analyses utilizing calculational methocs that produce conservative results (i.e., calculations indicate greater reactivities than do the experimental O
data).
V The use of experimental data on delta-phase plutonium-depleted uranium systems for alpha-phase plutonium-depleted uranium systems (see Appendix C) is an example of the application of a conservative calculational method to experimental data.
The methods and results of each criticality study are jointly reviewed by the Criticality Safeguards Advisor and a physicist on the Fuels Com-mittee.
It is required that the physicist on the Fuels Committee agree that the method of analysis used is appropriate and that the results of the cal-culations establish safe nuclear parameters for the operations reviewed.
This concurrence may be based on consideration of experimental data, alter-nate methods of analysis, and other published safety limits and standards, as appropriate. This review will also consider limits of error in the analysis and will confirm the conservatism of the results.
Details of this independent review are documented, including the con-ditions considered, possible alternative methods, and potential sources and significance of possible nonconservative assumptions. Sources of conserva-tion are identified. This documentation is kept with the supporting docu-ment for nuclear safety for the project, to allow inspection.
For R&D efforts involving liquids in the area of fuel development tech-nology and scrap recovery, the criteria described below are esed.
9 O
ESG-82-33 I.4-5
i e -
Batches in noninterconnected solution-containing vessels will 1) be limited to 350 g of U-235 or 230 g of Pu.
For composite l
fuels, the following formula will be used:
350 g U-235 + 230 g Pu where X = wt of U-235 in composite fuel (g)
Y = wt of Pu in composite fuel (g) 2)
Should the situation exist where solution-containing vessels are interconnected, the totai mass limit for all intercon-nected vessels will not exceed 350 g of U-235 or 230 g of Pu..For composite fuels, the limit is specified by the formula listed in (1) above.
4.2.2 Array Criteria The array criteria are listed below:
1)
Arrays of f uel elements shall be no more than 451 of critical when the analysis is based on calculative methods.
- However, the number of units may be increased to 75% of critical when-ever the number of units to achieve criticality has been
.(
determined by experiment.
- 2) Two arrays may be considered isolated f rom each other when separated by a layer of concrete (density not less than 3
140 lb/f t ) at least 12 in. thick.
The above criteria will not be applied to slabs.
- 3) Two arrays may also be considered isolated when separated by the larger of the following:
(a) 12 f t or (b) the largest dimension of the f acing surf aces of the arrays.
If the spac-ing is less, they are regarded as a single array.
4)
The maximum safe number of units (based on these criteria).in an array is determined by accepted solid angle criteria,42 sur-face density,8 or density analog f ormulas9 that have been proved conservative when applied to experimental data.
In the density analog method, correction f actors are explicitly used to account for optimum moderation (as a result of reduced water density from a sprinkler spray) and full array reflection.
Criteria for arrays set forth in nuclear safety guides 4 are included.
In no case will two units be closer to each other ESG-82-33 1.4-6
Q p
L than two diameters (edge to edge) or 12. in., whichever is
~
larger. These criteria are based on standard process condi-tions.
In the event of.an accident 1(e.g., water flooding of units caused by activation of a sprinkler system due to a fire),
the array will be. no 'more than 90%_ of critical. Array safety is based on 45% of the number of units in a critical array rather than on a given maximum keff.
For use of the surface density method, the average surface l-density of the array'is limited to 25% of the critical sur-L
- face density of a fully reflected uniform slab of the appro-priate fissile composition.
Individual units in the array are' limited to 30% of the bare critical mass. This latter requirement is generally satisfied by limiting the unit mass to 45% of the fully reflected critical mass. These criteria are consistent with data presented in LA-3366 (Revised)8 and are conservative as a result of applying infinite-array data to arrays of. limited extent.
5)
When calculational methods are used in the n l ft methodgion of arrays,* solid angle megheds4 'yc ear sa e y
-l evalua
, or the density analog method are used. The.
i solid angle method is only used when the eguivalent U or Pu element density is no greater than 10 g/cm.
In using the o
solid angle method, the Rev. 2, is used exactly. procedure specified in TID-7016, Since these procedures are' limited
/~~ 1
"\\
to moderated units, no adjustment for interspersed moderation is made. The effectiveness of the array reflector must be considered and adjustment made if it is more effective than thick water at the boundary of the array. Shadowed con-tainers are included in the calculation of the solid angle.
- 6) -Safe spacing between storage or transportation units may also' be determined by using the criticality indicator system of Reference 4 Unless experimental data are available for the array of interest, no two units in an array will have an edge-to-edge spacing of less than two diameters or 12 in.,
whichever is larger.
i l
- See discussion of solid angle calculation following this listing.
l ESG-82-33 s
I.4-7
- 7) Two dissimilar stations (work or'in-process storage) that V
contain dissimilar quantities of fissile materials are safe if they are separated by a distance that is not less than the average of the corresponding distances by which each would be safe if separated from a station' identical to itself.
The minimum distance is 3 ft, unless otherwise defined. Storage units of dissimilar types within an array are also safe if 4
they are separated by a distance that is not less than the average of the corresponding distances by which each would be safe if separated from a storage unit identica'l to itself.
If the units differ greatly in moderation, the allowable spacing must be demonstrated by a method that accounts for moderation, such as the density analog method with inter-spersed moderation.
In this case, however, the minimum edge-to-edge distance between units is two diameters or 12 in., whichever is larger. This criterion also applies to a single birdcage and to any other single, safe batch, as well as to an array of natural uranium on only one side of the work station.
- 8) Arrays of natural er depleted uranium may be stored within 8 in. of one side only of an array of enriched fuel.
9)
If infinite arrays of a number of ty the units in each array are similar, pes of fuel are r e when the various typ
-f 7,_')
arrays are safe when combined.
If the arrays consi n of V
units which differ greatly in moderation, further analysis is required.
- 10) Any array of packages licensed pursuant to 10 CD 71 for transport outside the confines of Energy Systems Group (the i
safety of such packages based on criteria outside of those presented above) may be stored in the prescribed arrangement meeting the licensed criteria and may be separated from any other array by the isolated array criteria described above.
Moreover, individual packages approved for shipment undcr special conditions may be stored at Energy Systems Group under the same condition.
In establishing the basis for safety of an array by the solid angle method (see Item 5 in above listing), the solid angle, ip*geradians, is calculated using the formulations in referenced reports.
l Unless specific studies have been made to evaluate the keff of a given container, it will be assumed that the keff (for interaction pur-poses) is equal to 0.65 when the maximum mass of fissile material per con-i tainer is no more than 45% of critical under optimum conditions of water moderation, reflection, and container geometry. However, when the maximum allowable mass per container depends on container geometry, the assumed f
ESG-82-33 1.4-8 i
l keff is increased to 0.80.
This'has been suggestedII for highly en-riched uranium solutions and has been found valid for uranium fuel slugs of 7v'k}
low enrichment.
Based on this consideration, the allowable solid angle for t
the most central unit in arrays with a k yf = 0.65 and keff = 0.80 will e
be 2.5 and 1.0 sr, respectively.
In the use of the solid angle criteria, the minimum edge-to-edge spacing between adjacent units in the array will be two' diameters. or 12 in, and the U or Pu density will be <10 g/m3 The solid angle method will be used only for well moderated ur.its, with k eff calculated for optimum moderation.
Criteria for accept-able solid angles are found in the Nuclear Safety Guide 4 The density analog method can also be used to calculate the size of a critical array as a function of the fraction of critical of an individual unit in the array and the average spatial density of the fissile material in the array.
The following factors and equations are used in this analysis.
F >- V /Vf (I) c where Vc = volun.e of a unit cell of the array Vf = actual volume occupied by fissile material (fissile container volume).
4 The f raction of critical mass of an individual unit, f, is O
f = M /M (2) e cb
.. (f where Me = equivalent mass of a sphere having the same geometric buck-ling as the actual container of fissile material.
(Accounts for the reflector materials surrounding the fuel.)
Mcb = mass of a critict1 bare sphere having the same composition as the fuel.
The lumping exponent f actor, s, is s = 2(1 - f)
(3) and the critical nunber of bare units, N, is given by b
Nb " f5Mcb/Na (4) where M, = actual mass of fissile material in a unit.
ESG-82-33 1.4-9
((.)
The critical number of water-flooded units (No) must take into account interspersed moderation as well as an equivalent infinite water reflector surrounding the array.
- t has been found9 that the addition of interspersed moderation can reduce the miminum critical number of units by as much as 2.5; the magnitude of the riiuction factor depends on the degree of moderation within a unit.
The factor becomes smaller as the degree of moderation increases and is unit when the fissile thterial in each package of the array is at optimum moderation.
The critical ns 3r of m its in a reflected array is cmaller than in a bare array by an add! '.caal factor that depends on the form of the fissile material and the degree of moderation. This reflection factor is presented in Table 4-2 for several systems.
TABLE 4-2 REFLECTION FACTORS FOR FISSILE ttATERIAL SYSTEMS Fissile Material System
- Reflection Factor Pu metal 20
~)
U(93) metal
- 13
(./
U(93)0 8
2 U(93)F 6.0 2
U(93)-H 0, H/U-235 = 60 7.0 2
U(93)-H 0, H/U-235 = 400 4.0 2
U(93)C 2.7 30 U(4.9)0 F 4.0 23
- Values in parentheses are tne percentage enrichment of U-235.
These values were determine a experiments using actual or simulated water reflectors relatively close to compact arrays.
Calculations for simi-lar systems and some experiments have shown that concrete can be a more effective reflector material than water and, therefore, consideration of this must be included in MonteCarlocalculationslgheuseofreflectionfactorsderivedforwater.
indicate that the critical number of fissile units in a concrete-reflected array may be about 0.6 times that in a water-reflected array.
t%
V ESG-82-33 1.4-10
i lhese systems have reflectors that are about 3 to 3-1/2 in, from the
[]
surfaces of adjacent fissile units.
In a vault, the reflector surfaces V
(walls, floor, and ceiling) are generally more distant and are, therefore, less effective. Experimental data showing the variation of reflector effectiveness with separation distance have been presented.13 These data show that a concrete reflector at a separation distance of 8 in. is less than one-half as effective as one at a separation distance of 3-1/2 in.
Thus, a separation distance of 8 in, makes a concrete reflector less effec-tive than a water reflector at 3-1/2 in.
Additional experimental data are also available, simulating the most highly reflected-units in a vault--those units located in a corner.14 These units are reflected by three slabs, the floor and two adjacent walls.
It was found that a thick, three-sided reflector was less effective than a 1-in.-thick reflector completely surrounding the array.
The listed reflection factors are used with the density analog method only if the minimum separation distance between storage units and concrete reflectors is 8 in. or more.
While both these factors decrease with increasing internc1 moderation, only the maximum values (2.5 for interspersed moderation, 20 for reflection of Pu, and 13 for reflection of U-235) are used in the analysis of storage arrays. This provides a conservative evaluat'.on of vault storage and reduces the dependence on material compo it'.on.
A 4.2.3 Limited Moderation Criteria b
Uncontrolled water moderation is assumed in establishing maximum safe batch sizes or maximum safe geometry, unless both of the following con-ditions exist:
1)
The fuel is in a nonflammable enclosure in which there is no source of water.
2)
The enclosure is impervious to water under foreseeable l
accident conditions, or entry of water is not inconsistent with the double contingency principle.
i A condition of no moderation is assumed when the process-required moderation of the fuel (e.g., the addition of stearate compound agglomerat-ing agent) has an H/X atomic ratio
- no greater than 0.5 or a C/X ratio
- no greater than 500 (e.g., in a mold for the casting of fissile material plates).
- X is the number of atoms of fissile material.
l g
ESG-82-33
(
I.4-ll
However, when the process-required moderation of the fuel is greater g
J than for a condition of no moderation and when it is desired to base safety
>b on limited moderation, the maximum safe batch size is no greater than 45% of critical, based on the optimum moderation with combinations of fissile mate-rial and available moderator material under normal operating conditions.
The quantity of moderator in the enclosure is measured before its addition and is controlled in the same manner as the addition of fissile material is controlled or otherwise limited. Under credible accident conditions, the ratio of moderator to fissile material is such that the allowable batch size is no greater than 75% of critical.
For operations in which the normal quantity of moderating material is less than about one-half the applicable limit, nuclear safety may be based on limited moderation without the stringent accounting required for the con-ditions of controlled limited moderation. Adherance to this limit is veri-fied in the normal periodic inspections and review of operations, i
0004Y/sjd ESG-82-33 7
I.4-12 4
l L
D..
5.0 ENVIRONMENTAL FROTECTION 5.1 EFFLUENT CONTROL SYSTEMS Provisions for aqueous effluent control are listed below:
1)
For fuel fabrication areas, hot cells, and other posted areas in which unencapsulatt.d radioactive materials and/or nuclear fuels are used in large quantities, holdup tank (s) must be pro-vided to allow radioanalysis prior to discharge without inter-ference with operations. Liquid level in a tank is determined by means of liquid level indicators, riarms, or periodic inspection.
2, Where a dual tank system is provided, the capacity of each holdup tank must be sufficient for ef fluent collection during the mr,ximum period anticipated for radioanalysis and discharge.
3)
Provisions must be included to protect the tanks from sludge, sludge buildup, and highly alkaline tnd/or acidic materials.
4)
Provisions for obtaining representative samples must be included.
'5)
Provisions for the controi of spills from the tanks must be 6
included.
!V) 6)
The capability for discharge to all required receptors must be provided.
7)
For effluents from high-level glove boxes, a sampling tank must be provided to allow a check of the activity level and pH prior to transf er to a holdup tank, with transfer capability to drums or to a truck if necesssry.
8)
Piping used for effluents from posted ereas must be entirely separate from, and located at a suitable distance from, water supply piping.
9)
Piping used for effluents from posted areas must be suffi-ciently durable and corrosion resistant to preclude laakage to the soil during the anticipated t seful life of the f acility.
10)
For areas in which radioactive materials and/or nuclear fuels are used only in laboratory-scala quantities, an automatic sampling tank and device may be employed rather than holdup tanks, provided that all effluents (other than sanitary) from the posted areas flow through the tank and are representatively ESG-82-33 O(.)
I.5-1
sampled in a manner subsequently permitting the determination of the average ratio of radioactivity to ef fluent volume, V
- 11) All ef fluents from posted areas must be discharged in confor-mance with applicable regulations and license requirements.
12)
Liquid ef fluents potentially contaminated with plutonium are analyzed prior to discharge.
Gaseous eaission control safeguards are listed below:
- 1) All gaseous emissions from posted areas must be discharged in accordance with the all specifications listed below.
Radio-logical measurement must be provided.
- 2) Continuous representative sampling of stack emissions for par-ticulates is required, using filter media 99% ef ficient for particles of 0.3 pm diameter and air volume measurement devices accurate to 20%.
- 3) Radiometric analysis of samples is required at a f requency suf ficient to ensure that regulatory standards for radioactive material concentrations in unposted areas are not exceeded.
Radiometric analysis of continuous. air sampler filters is per-
' formed at least as often as every 2 weeks.
- 4) For facilities from which radioactive gases may be discharged in significant quantities, the gases must be collected in hold-up. tanks or other cor. tainers and radioanalyzed prior to dis-charge; or stack gas monitors must be provided with recording capability and sufficient sensitivity to permit compliance with regulatory standards for radioactive materials in unposted a reas.
- 5) Operations with encapsulated-plutonium are performed in enclo-sures providing three stages of HEPA filtration.
Exhaust ven-tilation f rom areas in which these enclosures are located passes through two stages of HEPA filtration.
These filtration requirements are not applied to the use of small quantities of plutonium for gamma-spectroscopy and other analyses.
l l
- 6) A continuous particulate alpha air monitor is provided for stacks exhausting areas in which unencapsulated plutonium is used.
This monitor has an audible alarm and is set to alarm at 40 x 10-12 uCi-hhm3 above background.
Emission control is the responsibility of the manager, Radiation &
)
Nuclear Safety.
1 f
ESG-82-33 j
I.5-2
5.2 : ENVIRONMENTAL MONIIGRING
'd --
The radiological and nonradiological monitoring program is described in
'.j '
'the environmental monitoring annual reports.
The two most recent of these s
. reports, RI/RD88-14415 and RI/RD89-13916, are included in Appendix B.
.These reports describe the program as it currently exists.
Changes are made to maintain a program suitable and consistent with past, present, and antici-pated operations, radionuclides, inventories, and environmental regulations.
Four semiannual ef fluent monitoring reports' for 1987 and 198820,21,22,23 l were also transmitted to the USNRC Region V Of fice, Office of Inspection and Enf o rcement.
5.3 LIQUID RADI0 ACTIVE WASTE Liquids that are highly contaminated are solidified for disposal, Liquids with lesser contamination are collected in a hold-up tank and trans-ferred to a DOE facility for evaporation.
No contaminated liquids are dis-charged to the environment.
i l
}
5135Y/ reg ESG-82-33 I.5-3 1
i 1
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6.0 SPECIAL PROCESS COMMITMENTS l
u All processes used in licensed operations are standard and routine.
No special procedures or actions are required for unique processes or operations now planned or scheduled.
All operations are monitored and audited by a Safety Review Committee and Panel, as described in Sections 2 and 11, which have in their charter the responsibility for review of all processes and pro-cedures. Should there be a unique or special process presented to them for review that they would consider to be an unapproved safety question under the conditions of the license, this process and/or procedure would be submitted to the NRC for review and approval.
.r t(
0006Y/srs ESG-82-33 o
I.6-1 1
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V 7.0 DECOMMISSIONING PLAN The decommissioning plan, AI-78-10,17, has been updated and is sub-mitted in' Attachment 7, as part of this license renewal. 'A currerit cast estimate in 1988 dollars has been made.
This information, together with a commitment to the availability of the funds for decontamination of the facilities used for licensed work in order to release them for unrestricted use, is in Attachment 8 to this application.
i
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ESG-82-33 I.7-1 l
1
. i.
M. p' 8.0 RADIOLOGICAL CONTINGENCY PLAN RI/RDBB-206, "Onsite Radiological Contingency Plan for Rockwell Inter-national Operations Under Special Nuclear Material License No SNM-21,"
1 revised July 25, 1988, has been submitted to and approved by the NRC.
1 0008Y/gib l.
I
(
ESG-82-33 1.8-1 l
y.
9.0 OVERVIEW 0F OPERATION 9.1 CORPORATE INFORMA110N L
Fuel examination, decladding, and other activities licensed under SNM-21 are carried out at the Rocketdyne Division.
Rocketdyne reports to the Aero-space Operations of Rockwell International.
Rockwell International is a stockholder-owned corporation incorporated in the State of Delaware. The cor-parate headquarters is a 2230 East Ircperit.i Highway, El Segundo, CA 90245.
l The corporate directors and of ficers are identified in the annual report s which is included as Appendix A.
9.2 FINANCIAL QUALIFICATION The annual report ( Appendix A) preserts the information demonstrating financial capability to carry out the requirements of the license.
9.3
SUMMARY
OF OPERATING OBJECTIVES AND PROCESS At the Santa.Susana Field Laboratories (SSFL) site, Rockwell Interna-tional. Hot Laboratories (RIHL) (Building 020) are involved in 00E-funded activitles decladding spent DOE-owned fuel elements and other commercially funded activities requiring hot-cell facilities.
9-.4 SITE DESCRIPTION Complete site information on location and demographical and topographical inf ormation are included in RI/RD88-206, "Onsite Radiological Contingency
\\
Plan," July 25,'1988.I Geographical, meteorological, hydrological, seis-mological, and geological characteristics are included in Environmental Assessment, AI-76-21 and supplements.IB The information on these subjects is up to date in that issue of the document, except for deletion of the decommissioned fuel fabrication facilities at the Headquarters site.
9.5 LOCATION OF BUILDINGS ON THE SITE Descriptive information regarding the facility structures where licensed activities are under'., and their locations on the site are included in RI/R088-206 and in the Physical Security Plan, ESG-80-17, revised May 26, 1987.I9 9.6 MAPS AND PLOT PLAN Maps of the site, location maps, and photographs defining the site bound-ary and other site features out to a 50-mile radius are included in RD/RI88-206,8 E SG-80-17,19 and AI-76-21.lB ESG-82-33 11.9-1
r -.
9.7 LICENSE HISTORY The first special nuclear materials license was issued by the AEC in
' (A.
April 1956 to North American Aviation, Inc.
North American Aviation, Inc.,
)
became North American Rockwell in 1967 and Rockwell International in 1973.
Application for license renewal and an amendment for a broad license in its present form were first submitted in March 1967. The broad license, SNM-21, Docket 70-25, was issued in July 1970 with an expiration date of July 31, 1975. Application for renewal of this license was made in June of 1975. The current License SNM-21 was issued September 15, 1977. The expiration date is September 30, 1982.
Recently, following termination of uranium fuel f abrication at the Fead-quarters site, the facilities identified as Buildings 001.and 131, and Yards 507, 508, and 511 were decontaminated and decommissioned. Release of these f acilities for unrestricted use has been approved by NRC.
Since renewal of the license on June 28, 1984, corporate organizational changes have resulted in merging the Energy Systems Group and the Rocketdyne Division, with the Division Director of Atomics International reporting to the President of Rocketdyne.
In addition, the license has been amended to remove Building 055 as an authorized facility for licensed activities. The remaining licersed activiti.es at Building 020 were unaffected; the management and con-trol of day-to-day activities at the Hot Laboratory (Building 020) are under the same personnel in Atomics International and the Rocketdyne HS&E Department as previously in the Energy Systems Group organization.
9.8 CHANGES OF PROCEDURES, FACILITIES, AND EQUIPMENT The administrative procedures and controls that are in place at Rocket-dyne to ensure that independent safety review of all activities, whether new or continuing, is performed and documented, are described fully in Section 11.3, Organizational Procedures.
Activities covered by these procedures include change:. in operations, associated procedures, and f acilities and equipment utilized in these operatioris. Also included in Section 11.3 are those procedures that assure that a safety review is done; the responsibili-ties for requesting safety analysis and reviews; the analysis review, approval and verificatlou procedures; and the documentation requirements.
The organi-1 Iational structure responsible f or carrying out these organizational proce-dures is described in Section 11.1.
ESG-82-33 11.9-2
10.0 FACILITY DESCRIPTIONS
[f 10.1 SSFL FACILITIES 10.1.1 Hot Laboratory 10.1.1.1 Plant layout The Rockwell International Mot Laboratories (RIHL) is constructed of both reinforced normal and dense concrete.
The floor plan is shown in Figure 10-1.
The laboratory consists of four rectangular hot cells, backed by decon-tamination rooms for examining high-level radioactive material, and of a building structure surrounding the cells to provide office space, an operating I
gallery, operations support, a mockup area, and a service gallery.
The examinations are conducted from the operating gallery, where the in-cell equipment is remotely operated.
The manipulators, analytical equip-ment, and controls for the various cell operations are located in this area.
l The cells are serviced f rom the service gallery, located to the rear of the hot cells.
Separating the cells and service galiery are the decontamination rooms, where equipment is decontaminated prior to removal from the cells to the hot-storage area.
The decontamination rc, oms also serve as contamination control areas between the cells and the service gallery. A hot-storage area is provided for contaminated equipment.
Also connected with the service gal-1ery is a hot-manipulator repair room for servicing low-level, radioactively contaminated equipment.
In addition, controlled-environment glove boxes are available for use with radioisotopes and low-level radiation operations.
A machine shop and mockup area allow fabrication and mockup of remotely operated equipment prior to installation in the cells.
The facility also includes a O
change room, a photographic laboratory, and of fice areas.
%)
The main walls are 42-in.-thick, high-density (4.4 g/cm3) magnetite concrete.
With 1 x 106 Ci of 1-MeV gamma emitters located at the inside surf ace of the wall, this thickness is suf ficient to reduce the radiation level to less than 0.1 mR/h at the outside surf ace. The cell walls have been surveyed with a fuel element f rom the Organic Moderated Reactor Experiment, which was reading 700,000 R/h at 1 f t, to determine voids.
Voids were ex-posed, packed with lead wool and lead shot, resealed, and rechecked with the fuel element source. Cell dimensions are shown in Table 10-1.
TABLE 10-1 HOT-CELL INTERNAL DIMENSIONS Physical Dimensions Cell Length Depth Height j
1.
Metallography 16 ft 10 f t 16 ft 9 in.
2.
Materials testing 16 ft 10 f t 16 f t 9 in.
3.
Disassembly and sample cutting 20 ft 10 f t 16 ft 9 in.
4.
Disassembly and nondestructive examinations 33 ft 6 in.
10 f t 16 f t 9 in.
ESG-82-33 11.10-1
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The viewing windows in the main cells are 42 in. thick.
They consist of eight plates of glass as shown in Table 10-2.
The windows in each cell have I
been located to provide convenient work stations well adapted to criticality
~r control..
TABLE 10-2 VIEWING WINDOW PARAMETERS Plate Densigy Thickness Number Location Material (g/cm )
(in.)
1 Outside Tempered plate 2.5 1
2 Outside Lead glass 6.2 4
3 Outside Lead glass 6.2 4
4 Outside Lead glass 6.2 4
5 Outside Lead glass 3.27 8
6 Outside Lead glass 3.27 8
7 Cell side Lead glass 3.27 8
8 Cell side Nonbrowning tempered plate 2.7 1
Transfer drawers with steel-encased lead sliding doors at each end are provided in the walls between each of the cells.
Storage drawe rs are also provided in the wails between the main cells and the decontamination rooms.
These drawers are used for storage and have been modified so that they can be opened only from the main cell side except for Cell 1, "hich opens from Decon-tamination Room 1.
Large doors are provided in each of the decontamination rooms and main cells.
The cell doors are 21-in.-thick Meehanite, with 16-in. dense concrete for the decontamination rooms.
The doors provide access for moving materials and equipment in and out of the main cells.
In addition, a 21-in.-thick Meehanite door is provided at the end of Cell 4 adjacent to a mockup and large cask handling area.
This door was provided to allow entrance of very large reactor components.
A 12-in.-diameter port with a 12-3/4-in.-thick sliding lead door allows for loading and unloading of transfer casks.
All doors are provided with inflatable seals to minimize leakage around the doors and between the areas.
Under normal operating pressures, leakage around the doors is essentially prevented.
With all doors shut and sealed, each cell and decontamination room can be operated independently.
For example, the oxygen content of the atmosphere in Cell 3 can be reduced by using the nitrogen purge while normal cell ventilation is used in Cells 2 and 4; or Cells 2 and 4 can be entered while an examination is proceeding in Cell 3.
l-ESG-82-33 11.10-3 C-_--------_----.----_------------
~
Handling equipment is provided fnr in-cell movement of materials and operation of test apparatus.
One pair of master-slave manipulators is located l
at each cell window.
In addition, each cell is equipped with a 3-ton bridge f-i trane.
General Mills. manipulators are provided in Cells 2, 3 and 4.
Pro-visions have been made for future installation of a rectilinear manipulator in l
Cell 1.
L i
A raised floor is used in all the cells to provide a better working
' height during examinations.
These floors are made from 1/4-in.-thick plywood completely covered with a 1/32-in. stainless-steel plate and sealed so that liquids cannot penetta'.e the wood.
The floor is supported by an aluminum angle-and-channel 91:umre bolted together.
Although the structure can be i
removed, it is rigid arc cannot fall once in place.
10.1.1.2 Utilities, Including Emergency Power l
There is no pressurized water supply in the maia cells.
A nitrogen atmosphere (less than 5% oxygen) is used to suppress, control, and extinguish l
fires.
In Cell 1,1 water for the cleaning of metallog aphic supplies is pro-l vided by a flexible tube through a shielded cell access port.
All pressure fittings and controls are mounted external to the cell.
In-cell cooling uvits are provided to reduce and control the temperature of the cell atmosphere.
These units have explosion-proof electrical connec-tions and use a standard refrigerant (not water) for cooling purposes.
A recirculation cooling system is used so that f ailure of a unit af fects only the temperature 'of the cell atmosphere, not the amount, type, or pressure.
/^
The normal power for the facility is supplied from the SSFL 4160-V power distribution system.
Transformers in the facility convert the power to 480/277/208/120 V as needed.
A 480-V, 200-kW diesel engine-powered generator provides emergency power for the f acility in the event normal power is lost.
Power for starting the
. diesel is supplied by batteries that are charged with normal source current.
There is a delay of about 5 s between the time normal power is lost and the time the emergency generator is on the line.
This period is sufficiently short to prevent positive pressures in any of the cells, although there is a slight decrease in the negative pressure.
10.1.1.3 liea ting, Ventilation, and Air Conditioning The building ventilation systems were designed principally to control airborne contamination.
These systems direct the leakage of air f rom the out-side of the building into the main cells.
The airflow is always from an area of lower contamination to an area of higher contamination within the build-ing. This flow pattern is accomplished by successively reducing the pressure of the atmosphere at each area of higher level of contamination.
The direc-tion of tir leakage is shown in Figure 10-2.
ESG-82-33 11.10-4
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ESG-82-33 II.10-5
10.1.1.3.1 High-Volume Cell Ventilation
(^
Ventilation for the four hot cells and the decontamination room is pro-3
(
vided by a 12,540-ft / min constant-volume blower.
A second, identical blower is located in parallel and is automatically actuated in the event of a failure of the primary blower.
Both blowers are on an emergency power system.
The exhaust f rom the cells passes through prefilters located in each cell exts)t for Cell 3, which has prefilters and high-ef ficiency filters. The fil-ter system located in the basement is determined to be 99.95% effective for particles >0.8pm in size, using a standard " cold" DOS test.
The exhaust from the north exhaust in Decommissioning Room 4 is through prefilters and high-efficiency filters.
Low-volume cell ventilation is controlled by pressure instruments located in the operating gallery.
Under normal conditions (high-volume exhaust), with
'all cell doors closed and sealed, the pressure in the cells is maintained at
>0.6 in. of water, negative with respect to adjacent areas, controlled by pressure controllers modulating the makeup air valve located in the basement.
3 De pressure differential results in ~100- to 400-f t / min leakage f rom the operating gallery into each cell.
Since the blower is a constant-volume 3
blower, makeup air to total the 12,540-f t / min blower capacity is auto-matically added by a valve in the basement of the f acility.
A schematic of the system is.shown in Figure 10-3.
Suf ficient ventilation-system capacity is provided to create large flow rates into the cells when any cell door or other opening is made. When a cell 3
door is opened, about 4000 f t / min are exhausted f rom the cell.
This cor-g.
(
responds to a flow rate of about 200 linear f t/ min through the opening into the cell.
This flow rate is adequate to prevent the release of contamination from the cell into the adjacent decontamination room.
10.1.1.3.2 Low-Volume Cell Ventilation To supply an inert atmosphere in the cells for fire prevention and the protection of pyrophoric materials, a low-volume ventilation system is pro-vided.
This system can maintain a negative pressure in a cell of 0.02 to 0.05 in. of water with all doors closed and sealed, although the normal range is a negative 0.1 to 0.5 in, of water.
Less-than-normal pressure dif ferential results in less air leakage into the cells and reduces the amount of inert gas makeup required to maintain the oxygen content below 5%.
Nitrogen is the only inert gas currently in use.
The supply of dry nitrogen to the cell is automatically controlled to prevent the cell dif fereatial pressure f rom dropping below 0.05 in. of water -
with respect to adjacent areas.
This saf eguard prevents cell pressurization in the event of temporary system imbalances.
ESG-82-33 11.10-6 b
.I i
D PRE AND ABSOLUTE FILTERS IN SERVICE GALLERY HOT CHANGE ROOM hhfh*T ND SOLUTE OPERATING GALLERY FILTERS MANIP. ROOM (HOT)
HT R
CELL 1 CELL 7 CELL 3 CELL 4 PRE FILTER $
(LOCATED IN CELL 3)
CENERAL EXHAUST ~
BLOWERS 22,890 cfm l
3 f
LOW VOLUME C
VALVES P
=
~
HIGH VOLUME /
HIGH VOLUME CELL VENTILATION VALVE!
12.540 c!m M
ABSOLUTE FILTERS DILUTION BLOWER:
16,575 cfm MAKE U' AIR VALVE 2 3 67 UN;
%.2536 Figure 10-3.
Ventilation Schematic for the RIHL The low-volume ventilation system is used for extended cell operation with a nitrogen atmosphere.
In this phase, air inleakt.ge for Cell 3 has been measured to be:
Cell Atmosphere Pressure AirLegkageRate Relative to Adjacent Areas (ft / min)
(negative inches of waterl 15 0.05 24 0.10 60 0.25 f
i, j
ESG-82-33 11.10-7
Cell 3 was selected for these measurements because it has an additional open-ing for fuel element transfers and, therefore, has higher leak rates than h
Cells 2 and 4.. Based on these measurements and normal nitrogen usage, tanks
, (.)
to hold 3600 gal (12 tons) of liquid nitrogen and a 5000-gal tank have been installed. Although the average nitrogen usage is about 15 tons per 3-month period, usage during a nitrogen purge is 0.29 ton /h. Thus a capacity is pro-vided to allow 96 h of continuous usage or approximately five 8-h shifts. The liquid nitrogen tanks will serve as a backup to the installed gaseous nitrogen system.
10.1.1.3.3 Gaseous Nitrog_en Supply Building 020 is provided with gaseous nitrogen directly from the 3500-psi SSFL main. A hand valve is installed in case the system must be shut down.
Pressure Regulator 1 will reduce the pressure f rom 3500 to 300 psi. The second regulator will reduce the pressure to 30 psi, the working pressure.
3 Two pressure relief valves and a burst disc are used to prevent high-pressare 3
gas from reaching the hot cell. This system is the primary system in pro-viding a nitrogen atmosphere in the cell and decontamination rooms.
10.1.1.3.4 General Posted Area Ventilation The general posted area ventilation blower provides exhaust for the hot change room, hot side of the manipulator repair room, hot laboratory, hot shop, airlock, hot equipment storage, service gallery, and operating gallery.
3 a
Two identical 22,890-ft / min constant-volume blowers tre located in parallel to provide this exhaust (Figure 10-3).
One blower is normally in operation, and the second or standby blower is automatically actuated if the first blower
[
fails. Only one of the two blowers is on the building emergency power sys-I b
tem. However, unless the blower on the emergency power system is inoperative, the electrical circuit sequence ensures the provision of general building c
ventilation during a loss of line power.
t Figure 10-4 is a floor plan showing the location of all filters on the main floor. The prefilters on all ventilation systems are changed frequently to extend the life of the high-efficiency filters.
Iri-cell prefilters are changed during every cell cleanup (NB-week intervals) or when the negative pressure in the cell is greater than 0.5 in. of water.
10.l.1.3.5 ventilation Alarm System The warning system for the ventilation system is centered on two control panels. The instruments of the hot gas control panel actuate alarms in each l
cell. Each cell has an individual alarm to indicate a deflated door seal, loss of vacuum, or high-temperature atmosphere, t
Stack sampling is performed to permit th) measurement of particulate
(
radioactive material discharged from the facility.
A gas monitor is installed f
in the. stack to measure radioactive gas discharges and to indicate accidental d
criticality of' low-energy release occurring within a cell.
F 3
ESG-82-33 11.10-8 j
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'10.1.1.4 Room 139 Alpha Boxes
- ,5
_Two alpha glove boxes are located in Room 139.
The glove boxes are ex-
.t'j hausted through three stages of HEPA filtration, one within the glove bux and two within a dual containment filter rousing (located in Room 139).
From the third stage of HEPA filtration the gasses are exhausted through redundant blowers and released to the atmosphere f rom the facility stack.
The three stages of HEPA filtration are used to filter the atmosphere f rom the box.
All gas from the box is monitored by an alpha monitor before reaching the stack.
Pressure dif ferences between the box and Room 139 are maintained at -0.5, plus or minus 0.2-in, water pressure.
Pressure in the box is automatically controlled by an exhaust valve operated by a pressure controller.
In the event a glove is lost, a pressure sensor will sense.he loss and open a 4-in. valve.
The exhaust will provide for 150 f t / min :hrough two 3
open glove ports.
Entrance into the glove box room is f rom passage 143 into Air Lock Room 140 and f rom Air Lock Room 140 into Room 139 (reference Figure 10-1).
The doors into Air Lock Room 140 and Room 139 are controlled by electrically actuated locks that permit the opening of only one door at a time.
10.1.1.5 - Waste Handling 10.1.1.5.1 Liquid Wastes Radioactively contaminated liquid wastes from all hot facility drains are collected in one 3000-gal waste tank in the holdup tank building, which is s
located to the east of the RIHL at the perimeter fence line.
An attempt is made to absorb or solidify all highly contaminated waste ia the cell at the time of generation.
Thus, most of the tank centents is generated during decontamination, using water.
A weir box is used to catch large particles prior to their entering the holdup tanks.
In Cell 1 (the metallographic cell), highly acidic ana basic solutions are used.
A S-gal baf fle tank is provided to separate coarse particles f rom the liquid prior to release to the holoup tanks.
The sanitary sewer system drains to the Rocketdyne sewage treatment pl a r,t,
Discharge from this plant is to a 3-million-gal retention pond.
10.1.. 5.2 Solid Wastes Solid wastes remaining af ter decladding of spent reactor f uel are sealed in drums and shipped to the DOE-owned disposal f acility for handling and ultimate disposal at an authorized burial site.
ESG-82-33 11.10-10
. Contaminated solid wastes f rom other licensed activities are accumulated in drums and disposed of by shipment to authorized burial sites or retur:ad to L / '
.the customer for burial under his license.
(,
10.1.1.6 Chemical System No chemical processing of-nuclear fuel materials is per'ormed at the RIHL 10.1.1.7 Fire Protection The entire building is under a fire protection system.
Except for the hot cells and decontamination rooms, the system is a fully supervised, ordi-nary hazard, preaction sprinkler system with thermopneumatic devices to sound an alarm, indicate the fire zone, and open a valve filling the pipes to the sprinkler heads (standard pendulum, 165'F heads).
Fires in the hot cells or decontamination rooms are detected by the same type of sensing device, provid-ing both visual and audible alarms; a fire in these two areas would be con-trolled by means of a manual nitrogen purge system, which essentially floods only those two areas with nitrogen and reduces the oxygen content to an extent that combustion cannot continue.
Use of flammable materials in the hot cells is minimized and the storage of unenclosed flammables when the hot cells are unattended is not allowed.
The final stages of HEPA filtration are protected f rom fires in the cells primrily by distance and dilution of any heated air by cooler air.
This is shown in ar, analysis presented,to NRC in 1977 and submitted as a separate attachment to this application.
O
- Fire extinguishers are provided throughout the f acility to supplement and back up the fire protection system.
'l 0251Y/crp ESG-82-33 11.10-11 E________.__________________________________________________________________________________
11.0 ORGANIZATION AND PERSONNEL 11.1 ORGANIZATIONAL RESPONSIBILITIES As a separate operating group of Rockwell International Corporation, the Rocketdyne Division is under the direct management of the Division presi-dent.
Research, engineering, manuf acturing, and other support activities are accomplished within separate subdivisions organized into functional organiza-tions.
The president delegates to the department directors the responsibil-ity for ensuring that all operations are conducted in a safe manner and in conformance with the provisions of applicable licenses and regulations.
In addition to the functional organization described above there is a program management organization through which program managers,are assigned the responsibility f or planning, controlling, and directing specific pro-g rams.
The actual performance of the various associated tasks is accomplished by the functional departments within the budget and schedule limitations established by the program management staff.
The Health, Safety, & Environment Department is a staf f and service organization, within the Human Resources and Communication organization, with several responsibilities including, criticality safeguards, radiation safety, industrial hygiene, and industrial safety.
In this staf f capacity, the department is responsible for the development of an overall safety program and serves as advisors to department directors and all levels of Rocketdyne man-agement to ensure that programs are implemented in a satisfactory manner in all operations throughout the Division.
(
The Nuclear Safeguards function maintains a continuing safety review of
\\
all planned and existing nuclear fuel operations, and radioactive isotope utilization.
This activity, which is implemented through the Safety staff anc a Nuclear Safeguards Review Panel, chaired by the director of the Nuclear Safety and Licensing Department, provides management with the assurance of an independent evaluation of the adequacy of protection for Rocketdyne personnel and equipment and the general public.
The Criticality Safeguards function is responsible for the development ard administration of a criticality saf eguards and control progcam for all fabrication, storage, and handling operations with "out-of-reactor" nuclear fuels.
This function includes development and establishment of standards atid criteria for nud ear safety, establishment and implementation of a prog am r
control, and review and approval of operational procedures f or f uel material processing, f abrication, handling, and storage.
The Radiation and Nuclear Saf ety, Industrial ifygiene, and industrial Safety f unctions are responsible for development and establishment of Divisicn tafety p ro g rams, including design and opera tional safety criteria and standards; participation in the development of opt. rational plans and pro-l cedures; review and approval of plans for radioactive material and radiation ESG-82-33 11.11-1 L__________-_--------_
machine' acquisitions; review and approval of experimental uses and disposition of radioactive and toxic materials; implementation of safety standards and
/
procedures required by local, state, and federal regulations; and the surveil-k.
lance of all operations to ensure satisfactory implementation of the safety program throughout the Division, 11.2 ORGANIZATIONAL CHARTS The organizational charts showing the organizational arrangement of the plant with the personnel occupying each position with safety responsibility as indicated in Section 2.2 are included as an attachment to the letter of trans-mittal for this document.
11.3 ORGANIZATIONAL PROCEDURES The organizational procedures recorded below describe the management con-trols employed at Rocketdyne to assure an adequate safety program. They de-fine management lines of responsibility and the methods for maintaining con-tinuity in matters relating to safety in an evolving program.
They include methods for adopting new operating procedures and for changing existing oper-ating pror.edures.
They also include the methods used to evaluate supervision and the safety program.
The format of the organizational procedures discussion is as follows:
(1) Subsection 11.3.1, " Responsibility Assignments," is a description of the-responsibility for safety, including the relationship to management of an advisory safety group and the designation of the management position bearing prime ree. possibility for safety in licensed operations; (2) Subsection 11.3.2, f.
i
" Internal Inspection and Review Program," discusses the methods by which man-
\\
agement is apprised of day-to-day operational safety conditions, the provi-sions for periodic reviews, the methods by which management is informed of review findings, and the methods by which compliance with internal operating procedures, license conditions, and applicable regulations are enforced; (3) Subsection 11 3.3,
" Initiation of Projects Involving Potential Hazards,"
explains the requirements for technict.1 evaluation (safety) prior to starting 3 project, as well as the methods employed to inform management, the technical staff, and the advisory group of these requirements, and closes with a pre-sentation indicating the management positions authorind to initiate new pro-jects that have been previously evaluated f rom the safety viewpoint; and (4) Subsection 11.3.4, is entitled " Changes in Policies, Procedures, Fr. cili-f ties, and Equipment."
Management positions having the authority to authorize changes are indicated, along with the requirement for technical evaluation prior to the authorization of changes, and the methods of informin5 cansge-ment, the technical staff, and the advisory group are explained and described, R sponsibility Assignn.ents I
11.3.1 f
1 The president of the Rocketdyre Division bears the prime responsibility for the safety of employees and the public as af fected by Rocketdyne opera-tions. The successful discharge of this responsibility requires:
ESG-82-33 i
11.11-2
[
l 1)
Identification of potential and existing hazards
[]
2)
Development and provision of adequate desig. and procedural V
safeguards 3)
Enforcraient of the utilization of safeguards and of compliance with internal operating procedures.
Due to the size of the organization and to the complex nature of the work performed, delegation of safety authority is necessary.
la addition, the com-plexity of modern safety technology necessitates employing technical personnel who specialize in safety. The administrative procedures through which the L
delegation of safety authority is performed and by which the services of the safety staff are employed are described below.
11.3.1.1 Delegation of Authority The president delegates safety authority by assigning definite safety responsibilities to appropriate personnel. Enforcing the utilization of safeguards and compliance with internal operating procedures is delegated through the line organization to the individual unit managers, i.e., to the first-line managerial position. However, the department director position is the lowest level of management to which is delegated full responsibility for the review and approval of all activities conducted under a license or con-tract. The degree of inclination toward compromise in the interest of com-pleting a job increases with organizational proximity to the work area, and it becomes necessary to counterbalance this tendency by emphasizing the safety
<~
responsibility of managers on the third supervisory level. This emphasis takes the form of a requirement for safety review and approval by department directors for all potentially hazardous tasks under their cognizance. With respect to safety for these tasks, department directors, therefore, accept full management responsibility.
'i.3.1.2 Safety Staff Department directors are supported by a staf f safety organization--the health, Saf ety & Environmer,t (HS&E) Department--and by an aditif,ory group, the Nuclear Sa eguards Rev 3w Panel (NSR9). *he NSRP is headed by the director of r
the Nuclear Safety and Licensing Department. The director of HS&E reports to the presider;t through an organizational char'nel separate froN the channels of department directors having responsibility fer work with radioact1ve and nuclear fuels materials. The NSRP chairman reports directly to the president in his capacity as NSRP chairman. This arrangement provides for the ret.olu-tion of difficuh questions by top management when necessary.
The responsibilities of the HS&E safety staff are to:
establish standards of safety; examine proposed operations f or hazards; determine the saf ety measures that are necessary; evaluate the degree of compliance with established safety measurcs, licenses, and regulations; establish nuclear materW management control and waste management; and advise the department ESG-82-33 11.11-3
- directors and their line supervisors accordingly. With regard.to radiation safety.these responsibilities, along with those of the NSRP, are performed as
/'
described below.
11.3.1.2.1 Health. Safety & Environment l
The HS&E department is assigned several staf f responsibilities that l
include industrial hygiene and safety, health physics anc radiation safety, criticality safeguards and safety engineering.
The Radiation and Nuclear Safety, Industrial Hygiene, and Industrial Safety functions (1) establish design and operational safety standards; (2) review and approve designs and operational procedures; (3) approve radio-active material and radiation machine acquisitions; (4) approve the final dis-position of radioactive and toxit materials; (5) identify and designate posted areas for radioactive and toxic materials and control their use in these areas; (6) designate applicable safety procedures; (7) perform field measure-ments of radiation and of radioactive and toxic material contamination levels; (8) control the use of radioisotopes used as sources; (9) implement safety standards and procedures; (10) provide independent information as to the degree of compliance with the safety procedures that operations supervision are obligated to enforce; and (11) perform measurements of internal and exter-nal radiation doses and of radioactive and toxic material concentrations in effluents released to the environment.
The Criticality Safeguards function is responsible for the development and administration of a criticality safeguards and control program covering all out-of-reactor operations with fissile material, including fuel fabrica-tion, storage, and movements under Rocketdyne jurisdiction. The program is conducted by the Criticality Safeguards Advisor with assistance of the Criti-cality Safeguards Coordinator.
The Criticality Safeguards Advisor develops and establishes standards and criteria for nuclear safety; establishes and operates a pror, ram of control; approves all operational procedures for out-of-reactor prote: sing, handling, and storage of nuclear fuels and ensures that these procedures are consistent with the criticality safeguards and control program; approves or performs all nuclear rafety calculations feauir/d for out-of-reactor criticality control; serves as a consultant to the Fuels Com-mittee of the NORP; periodically inspects and audits cuarterly the criticality sateguards and control program. The Criticality Safeguards Cotirdinator estab-lishes criticality control areas; maintains an accurate record of the location of fissile materials; reviews and approves ficsile material transf ers between criticality control areas; approves procedures for the handling and storage of fissile materials within limits delegated by the Criticality Safeguards Ad-visor; performs quarterly audit and periodic inspecticas of all criticality control areas; and informs project personnel cf the criticality cor, trol pro-gram as it af f ects their operations.
I ESG-82-33 11.11-4
11.3.1.2.2 Nuclear Safeguards Review Panel Q
NSRP maintains coatinuing safety reviews of all planned and existing L'
nuclear fuel handling operations, and utilization of radioisotopes and radi-ation devices.
The. primary purpose of NSRP is to provide management with the additional assurance of an independent evaluation team for the protection of the general public, personnel, and equipment. The panel is composed of offic-ers, a Fuels Saf ety Review Committee, and a Radiation Saf ety Review Com-mittee.
The of ficers include a chairman, who is the director of the HS&L department, and a Radiation Safety representative, who is a member of the HS&RS department.
All NSRP members are appointed by the NSRP. chairman with the approval of his superior in the NSRP organization channel; the use of alternates is not authorized.
Adequate time and f unding for committee work are, by directive, budgeted by each project involved in activities under the jurisdiction of NSRP.
Each committee consists of a chairman and several engineers, physi-cists, and other professional personnel f rom the operating departments. The chairman is not organizationally associated with the operations or programs for which the committee is cognizant.
The selection of each member is based on the demonstration of a high degree of technical competence in a disci-pline(s) associated with the work reviewed by the committee.
The number of members is determined by_ the range of the fields required to perform a satis-f actory safeguards review.
NSRP of ficers are ex of ficio voting members of each committee.
At least one member of the Fuels Committee, other than the Criticality Saf eguards Advisor, is qualified in nuclear safety to perform independent verification of the methods of nuclear safety assessment used in the evalua-tion of criticality problems.
It is required that this member concur that the method of analysis used is appropriate and that the results of the calcula-tions establish safe nuclear parameters for the operation reviewed.
The extent of the review and the basis for concurrence are documented.
Committee meetings are called by the respective chairmen as necessary to fulfill committee responsibili*cies without unnecessary delay to Rocketdyne p rog rams.
Meetings may also be held at the request o? project personnel.
Each committee is responsible f or the review of operations with nuclear fuels and activities involving radioisotopes and radiation devices.
Reviews cever both proposed and existing designs and procedures, as well as changes 1
therein. The Fuels Committee reviews nucinr fuel activities, particularly in operations that may involve potential conditions of accidental criticality, such as fabrication, manipulation, storage, novement, and disposal of fission-able material.
All committees perform an arnual inspection of working areas scheduled for a maximum cf 14 maaths between inspections. At these times, the committees also review the organitatbr. of the operating groups and the as-signment of locel rescensioility for operations.
Meetings are usually con-ducted with projen perscar,el present.
Committee recommendations for action I
arising f rom the meetings or reviews are approved by the NSRP chairman or his designee and t ransmi t ted to the approval authority responsible for that operation.
ESG-82-33
%s 11.11-5
l' l
The executive secretary provides a pertinent, written summary of each meeting, along with recommendations of the committee, for transmittal to each member, the appropriate approval authority, and others as might be designated.
i
.y Approval authorities are appointed by the management official to whom the panel chairman reports in the NSRP organization channel.
They are generally l
directors who have management responsibility for the operations under cogni-zance of the committees of the NSRP.
The approval authority is obligated to implement committee recommendations and to ensure that the action taken is reported to the appropriate NSRP personnel.
In the event that the approval l
authority disagrees with any recommendations, the committee's recommendations, together with a report prepared by the approval authority, are presented to the NSRP chairman for resolution.
If the approval authority finds the Panel Chairman's decision unacceptable, he can appeal to the Division President.
The written approval of the approval authority is initially necessary for all operations involving reactors, critical assemblies, nuclear fuels or radioactive isotopes, and radiation devices.
Subsequent independent informa-tion as to the degree of compliance with the safety procedures the approval authority is obligated to enforce is supplied by HS&RS field personnel report-ing through established channels and by the appropriate committee of the NSRP.
11.3.2 Internal Inspection and Review Program Internal inspection and review are performed by the HS&E department staf f and by the NSRP. Additional effort of this nature is performed by the Employ-ee Health & Safety committee.
11.3.2.1 Reoorting Requirements Day-to-day operational health and safety conditions are observed by the health and safety staf f in the HS&E department. The staff is maintained at a level sufficient to provide essentially daily surveillance in active opera-tional areas. The technicians are instructed to report any hazardous condi<
tion or instance of noncompliance with regulatory directives to appropriate operational supervision and to HS&E supervisory personnel.
If for any reason the condition is not expeditiously corrected, it is reported by HSSE person-nel to the director of the HS&E department, or his designated alternate, who in turn notifies the appropriate department director.
The results of quarterly audits b1 the Criticality Safeguards Coordinator and the Criticality Safeguards Advisor are documented and reported to the director of the NS&L department.
Radiological data (film badge and bioassny results, reoiation er.d ton-tamini. tion survey results, effluent concentration results, environmental moni-toring results) for each facility are reviewed ac a minimum frequency of once per quarter by the health physicist responsible for radiation protection work ESG-82-33 11.11-6
in the facility.
Each health physicist is instructed to contact the appro-priate operational supervisor and to arrange for any corrective action
.b.O indicated by the data as being necessary.
If for any reason the corrective action is not taken expeditiously, the situation is reported to the HS&E department director, or his designated alternate, who in turn notifies the appropriate department director.
11.3.2.2 Periodic Safety Review Provisions Periodic reviews and evaluations of operational health and safety requirements and conditions are conducted by Criticality Safeguards, by the NSRP, and by the Employee Health and Safety committee, and by the Nuclear Safeguards Review Committee.
The intent of the periodic safety reviews is to provide audit functions beyond the day-to-day operational surveillance provided by the HS&E depart-ment in the discharge of its duties as a staff safety organization.
11.3.2.2.1 Criticality Safequards Criticality Saf eguards personnel (normally the Criticality Saf eguards Coordinator) periodically inspect working areas to ensure that all criticality safety criteria and regulatory directives are being followed.
Inspection fre-quencies are established by the Criticality Safeguards Advisor. Minimum fre-quencies are:
1)
Major fuel ha,dling areas, with large quantities and f requent transfers--weekly 2)
Criticality control areas where the fuel handling load is small such as analytical laboratories storage areas-semiannually.
In areas where substandard practices are observed, the inspector is authorized and directed to discuss such practices with operational personnel to ensure that they have sufficient knowledge of approved procedures; the responsible manager normally accompanies the inspector during the next inspec-tion, which must be made within 1 week of the initial observation.
If prac-tices that could result in a definite hazard are observed during any inspec-tion, line supervision is immediately notified to suspend operations until corrective action is taken.
The Criticality Safeguards Advisor also inspects the working area under the following minimum cond'itions; j
1)
Gefore the start of a new project 2)
Within 1 weet af ter the start of a new project ESG-82-33 11.11-7 L___ _ _ _ _
f~
3)
Monthly in areas where the coordinator performs weekly inspections-4)
Annually in uther areas.
These inspections determine that the equipment, facility layout, ar.d all vari-ables associated with the criticality safety analyses conform to the condi-tions required by the analyses and any applicable license conditions.
The advisor's inspections are documented, 11.3.2.2.2 Nuclear Safequards Review Panel Each committee of NSRP performs an inspection of every working area under its cognizance at frequencies determined by the committee chairman.
Compre-hensive reviews of safety conditions are also conducted at these times.
The minimum inspection and review f requency is once per year.
Recommendations are approved by the NSRP chairman or his designee and transmitted to the appropri-ate approval authority.
Safety reviews are required before potentially haz-ardous operations and are performed as the need arises.
Recommendations are transmitted to the appropriate approval authority following these reviews.
11.3.2.2.3 Employee Health and Safety Committee The Plant Safety committee is composed of workers from each operational department.
The ' committee members are appointed by operational supervision and serve for specified periods that normally.do not exceed 1 year.
Each com-mittee member is taught the basic safety subjects relevant to Rocketdyne oper-ations.
He is required to make a bimonthly safety inspection of his area and n'y to report his findings in writing to his supervisor for corrective action.
This program results in specialized safety training for a large percentage of the company's employees.
11.3.2.2.4 E_xecutive Safety Council The Executive Safety Council has prime responsibility for the Divisien health and saf ety policy consistent with the nperations to be conducted at all facilities.
Chaired by the President and composed of top f unctional execu-tives, the Coutcil provides management oversight of the Divis toa's overall safety posture to ensure that programs cone.1 stent with established policies are in effect.
An immediate convening of the Council may take place should it be indicatect by the seriousness cf 6 problem.
11.3.2.2.5
- enior Manacement Incident Review Committee Any incident involving a problem witn people or systems that is indica-tive of inadequate Rocketdyne performance is reviewed by the Senior 16dgement Incident Reviw Committee.
This review is to insure that Division personnel, products and f acilities are provided optimum protection.
" Lessons Learned" d
f rom incidents are applied across operations to minimize the chance of occur-rences of the same nature in the future.
l l
ESG-02-33 11.11-8
The Vice President, Quality Assurance & Systems Safety conducts the affairs of the Committee which is comprised of all functional executives, with
./T the Director of Test and Flight Safety Assurance serving as secretary.
l Root causes of incidents are addressed to ensure corrective action taken meets the system deficiencies identified.
Action items are assigned to responsibic functions and tracked until completion.
11.3.2,3 Transmittal of Findings to Management Inspection and review findings resulting from day-to-day surveillance and f rom periodic safety inspections are usually transmitted through personal dis-cussion with affected personnel.
At the discretion of the inspector or his superior, internal memoranda may be used.
Internal memoranda are required for the transmission of all NSRP findings; standard distribution lists are estab-lished by the NSRP chairman.
For the transmission of recommendations, these lists always include the president, the superior of the NSRP chairman in the NSRP management channel, and the appropriate approval authority.
The approval signature of the NSRP chairman or his designee is required for these memoranda.
11.3.2.4 Enforcement of Safety Rules and Practices Each department director is responsible for enforcing compliance with internal safety rules and operating procedures and additional conditions imposed by regulatory directives (e.g., NRC and California licenses and regulations and DOE Orders).
Thisresponsibilityisimplementedbydirectivesl to subordinate line supervision.
The degree of compliance is audited by the HS&E department and by the NSRP.
Instances of noncompliance are reported to line managers or directly to the department director, as necessary, to obtain i
corrective action.
11.3.3 Initiation of Proiects Involvina Potential Hazards Prior to initiating a project involving potential hazards, authorization must be obtained from the appropriate department director and from designated safety and/or Criticality Safeguards personnel in the HS&E department.
11.3,3.1 Requirements for Technical Evaluations 11.3.3.1.1 Health. Safetv j_ Environment A technical evaluation is performed by HS&E personnei fnr operetions involving radiation and nuclear safety.
Written safety procedures are required for any new operations or programs.
At the discretion of the H%E director or his designee, these procedures may refer to procedures developed for applicable conditions under other programs only if the potential for per-sonnel exposure is less than the quarterly limits specified in 10 CFR 20.
L These procedures are prepared by HS&E personnel and the operating engineers for the p reg rams.
The operations and associated procedures are i
approved, in advance of starting program activities, by the responsible line ESG-82-33 II.I1-9 l
lL___-_-----------------------
managers and by HS&E management personnel.
Technical evaluation and approval requirements for operations involving nuclear safety are discussed under V)
(
Criticality Safeguards in Sections 4 and 15.
-11.3.3.1.2 Nuclear Safeguards Review Panel Prior to initiating operations involving either (1) radioactive isotopes and radiation devices or (2) nuclear fuels in excess of the quantities listed in Table 4-1 of the Criticality Safeguards subsection, extensive safety analyses are performed by engineers from the operating department, and operating procedures are prepared.
The resulting documents are approved by responsible management and then transmitt?d to NSRP. Technical evaluations of these analyses and procedures are performed by the appropriate NSRP committee chairman and, at his discretion, by designated engineers from NSRP, including the Radiation Safety representative.
( Additional detail regarding the Fuels committee is provided in the criticality Safeguards subsection.)
Following a favorable recommendation f rom NSRP, the operations are authorized by the appropriate approval authority.
The Radiation Safety representative is responsible for ensuring that operations involving radiation safety are evaluated and approved by supervisory Radiation Safety personnel as stated in 11.3. 3.1.1, above.
11.3.3.2 Transmittal Methods f or Radiation Saf ety Requirements Generalized radiation safety c ri teria are published as Rocketdyne Operating Policies and distributed to all supervision.
In addition, when pro-jects are initiated, the following me%ods are used to inform management, con-cerned technical personnel, and NSRP of radiation safety requirements:
C 1)
All f acility and equipment designs (construction, f abrication, modification) are reviewed by HS&E engineers.
Safety require-ments are coordinated with the designer and with operational l
personnel and are transmitted to the designer by memoranda.
Final designs are approved by management personnel within the US&C department.
l 2)
Associated radiation safety procedures are prepared and approved as described in 11.3.3.1.1, above, and then trans-mitted to all direct'.y affected personnel, inciuding the l
appropriate department director.
3)
When planned operations are not covered by an existing radia-tion safety procedure, the HS&E department h contacted by the operating departme9t and saf ety penr.edures are developed and distributed as described immediately acove.
11.3.3.3 Task Approval'Within Previousiv Authnrized Projects The management positions authorized to initiate tasks using radioactive mater 4als in the performance of previously authorized projects are specified ESG-82-33 11.11-10
l below.
These considerations are presented in greater detail, regarding work
' with nuclear fuels, in the "riticality Saf eguards subsection.
f~
1 11.3.3.3.1 Health, Safety & Environment Operations involving radiation safety are approved in advance by both the responsible operations and HS&E management.
The appropriate HS&E manager may, at his own discretion, either (1) authorize certain tasP, stipulating that each specific task be performed under the personal direction of the operations supervisur (or an engineer-in-charge reporting directly to the operations supervisor) and under the periodic surveillance of a health physics techni-cian, or (2) delegate his authorization for operations to qualified per-sonnel under his supervision.
This authority may be delegated only for tasks to be performed in conjunction with a project previously approved by the HS&E director.
11.3.3.3.2 Health. Saf ety & Environment--Criticality Saf equards Operations involving quantities of nuclear fuel less than those listed in Table 4-1 of the Criticality Safeguards subsection are approved by the respon-sible operations unit manager and by Criticality Safeguards personnel.
The approval of the Criticality Saf eguards Coordinator concerning fuel transfers within or into and out of a single material balance area (MBA) may be delegated to the operations unit manager responsible for the MBA if either of the conditions listed below exists:
r 1)
The entire MBA and all work performed within the MBA are under
(
the supervision of one line supervisor, or 2)
The unit manager demonstrates to the Criticality Safeguards Advisor an acceptable procedure for ensuring t$ at the total 1
quantity of plutonium, U-233, and U-235, or ar.y mixture of these materials, present in the MBA at any time does not exceed 350 g.
11.3.3.3.3 huclear Safequards Review Panel R&D operations with nuclear fuels are initially approved through the line supervision channel to, and including, the appropriate approval authority as previously described.
Operations authorized by the approval authority have f avorable recommendations f rom NSRP.
All manipulations with nuclear fuels are initially approved by the Criticality Safeguards Advisor.
The operations are described in cae or r or.2 f ormal documents.
Thess documents specify the opera-tions (or in seme cases, types of operations) that can be initiated by the uperations or production workers, by the operations unit manager, and by his imrnediate superior.
Operations not covered in these documents, and not pre-viously approved, must be reviewed by NSRP and approved by the appropriate approval authority.
For work witn nuclear fuels, review and approval by the Criticality Safeguards Advisor is also required. The resulting authorizations ESG-82-33 11.11-11
are in memorandum form, and they have the same internal status and are used in the same manner as the documents centioned above.
t i
V 11.3.4 Changes in Policies, Procedires, Facilities, and Equipment Changes in established policies, procedures, facilities, and equipment affecting radiation and nuclear safety must be authorized in advance by management as indicated below.
11.3.4.1 Policy Changes Policies and criteria governing radiation and nuclear safety are estab-lished by the HS&E department and documented in Rocketdyne Operating Poli-cies; changes are issued with the approval of the HS&E director.
Revised standard operating policies are distributed to all managers.
Changes in the general methods of implementing these policies and criteria must be authorized by the HS&E_ director.
11.3.4.2 Procedure Changes 11.3.4.2.1 Health. Safety, & Environment Changes in radiation safety procedures developed for specific operations must be authorized by the responsible operations and HS&E managers following a technical evaluation by HS&E personnel.
Also, unless previously approved, changes in nuclear safety procedures must be authorized by the Criticality Safeguards Advisor following a technical evaluation by Criticality Safeguards e
personnel.
Revised procedures are distributed to all personnel directly
(
af f ected, including the appropriate department director and the appropriate NSRP committee.
11.3.4.2.2 Nuclear Safecuards Review Panel linless previously approved, changes in nuclear fuel handling and radio-isotope and radiation device safety procedures developed for specific opera-tions must be authorized by responsible line management and approved by the Criticality Saf eguards Advisor and the chairman of the appropriate safety review committee af ter their technical evaluation [cf,, 11.3.3.3.3 regarding cnanges within previously authorized projects].
All such changes are docu-mented, either in memoranda or in revisions of required official documents, and are distributed to all directly affected personnel.
11.3.4.3 Facilities and Ecuipment Changes Fat lities and equipment changes af fecting radiation and nuclear safety are be.,ubmitted to NSRP for review.
Recommendations resulting f rom such reviews are acted on by the appropriate approval authority in accordance with the provisions of this chapter.
ESG-82-33 11.11-12
__---.1-
-w
.x--L.-----_-.--.-.-_--------.-_---------------_--,--,--,---a----a_----,_--
s 11.3.5 Records of Authorization r~S Appropriate documentation of the review function for the authorization of h
a program or project is maintained by the appropriate NSRP committee chair-man.
Changes in policies, procedures, facilities, or equipment that have not received prior approval are reviewed as specified in 11.3.4.
Records of the review and authorization of such changes are also maintained by the NSRP com--
mittee chairman for a minimum period of 6 months af ter program termination.
11.3.6 Records Retention Minutes of meetings of the Nuclear Safeguards Review Panel anu of its subcommittees and correspondence involving committee recommendations and actions are maintained by NS&L for a period of 5 years af ter their date of preparation.
Correspondence pertaining only to panel membership is specifi-cally exempted from this requirement.
Written reports prepared by the Criticality Safeguards Coordinator or Criticality Safeguards Advisor regarding radiation and nuclear safety are also maintained by HS&E for a period of 5 years af ter the date of preparation.
11.4 FUNCTIONS OF KEY PERSONNEL The key personnel with health and safety functions and responsibilities to carry out the organizational responsibilities stated in Section 11.1 are director of NS&L, director of !!S&E, the manager c' Radiation and Nuclear Saf ety, the Criticality Safeguards staf f, the Industrial Hygiene and Safety staff, and the chairmen of the several safety review committees. These s
responsibilities are to carry out the procedures of the organizations that t
they lead or are af filiated with as detailed in Section 11.3 above.
11 5 EDUCATION AND EXPERIENCE OF KEY PERSONNEL The qualifications of the key personnel identified by title in 11.4 above, together with the qualifications of other personnel responsible for safety of operations, are included as an attachment to the letter of trans-mittal of this document.
11.6 TRAINING The following instructions are given to personnel who are assigntd to work with SNM under the license:
-1)
Pr',or to beginning work with radioactive materials and/or nuclear fuels, personnel are indoctrinated with regard to radiation and nuclear safety rules.
Personnel whose regular assignments, for tne first time, include work in posted areas will complete a training course covering the general aspects of working with these materials, i
ESG-82-33 11.11-13 i
)
i t
t p
2)
This course _must include (a) a description of the nature and l-hazards of radiation; (b) basic principles of radiation and fG-nuclear safety; (c) requirements of Rocketdyne Standard Oper-V ating Policies and NRC and California licenses and regulations; (d) safe handling practices, including considerations of acci-dental criticality; and (e) emergency procedures.
3)
Following indoctrination but before beginning work in posted areas, personnel are given instruction and must participate in
" dry run" practices in using the equipment, techniques, and procedures associated with their individual jobs.
a)
Particular emphasis is placed on changes in operating pro-cedures and conditions and on emergency procedures, b)
Emphasis is also given to explanations of potential con-sequences resulting from accidents due to deviations from operating procedures that have been noted during the pre-ceding year.
4)
Instructors for all training courses are qualified by education and/or experience to provide such instruction.
5)
Refresher training in emergency procedures is accomplished by means of annual emergency evacuation drills.
Refresher train-ing in radiation safety and criticality control is given at 2-year intervals.
Appropriate continuing education is provided to Radiation and Nuclear Safety personnel instead of the radia-tion worker refresher training.
6)
All appropriate female personnel and interested male personnel are provided supplemental information related to radiation exposure during pregnancy.
The eniergency training required by personnel is described in the NRC-approved "Onsite Radiological Contingency Plan for Rockwell International Operations Licensed Under Special Nticlear Materials License No. SNM-21,"
RI/R088-206, August 25, 1908.1 Line management is ultimately responsible for ensuring that th? estab-lished criticality control procedures are followed.
To assist in this respon-sibility, it is required that, af ter approval has been received for any new operation, line management meet with their perscnnel to review and discuss the approved fuel h.5ndling procedures.
Representatives f rom Criticality Saf e-guards and Radiation Safety participate in the review and discussion at the first complete meeting.
Approval to begin operations on any new pro,iect can be given only after the first of such discussions is held and after the under-standing by operations personnel of the Criticality Safeguards and Radiation Safety requirements of the job is assured.
Additional personnel cannot be added to the crew working on an approved project until line management has reviewed the criticality program for the project with these new people and has
(
ESG-82-33 11.11-14
so advised the Criticality Safeguards Coordinator, giving the names of the new people and the date of their indoctrination.
Criticality Safeguards personnel
[_}
discuss procedures with new employees during routine inspections to determine N/
their understanding of the nuclear safety procedural requirements.
(By definition, a "new person" is one who was not in attendance at the initial review meeting for that specific project.)
In addition to the discussion with operations personnel before each new project begins, there.is a continuous training program to remir.d older em-ployees, as well as to educate new employees, of the current criticality safe-
~
guards program and practices. This training program consists of both formal and informal instruction.
Criticality Safeguards program talks are given to operations personnel of existing projects when inspections indicate lack of compliance witn, or lack of understanding of, existing procedures. These talks may be either requested by project personnel or initiated by Criticality Safeguards.
Training requirements for the ALARA program are contained in the policy statement contained in the Health and Safety Procedure G-01.
This program is administered by llS&E personnel trained / certified as health physicists, l'
00llY/crp ESG-82-33 11.11-15 1
l
II 12.0 RADIATION PROTECTION PROCEDURES AND EQUIPMENT l' Q 12.1 PROCEDURES
.,g The procedures for conducting radiation surveys are covered in Subsec-tion '12.4, and the measures for ent.cring that the occupational radiation expo-sures will be ALARA are covered in Subsection 13.2.
Personnel radiation monitoring and the necessary direct-reading dosimetry equipment are covered in Subsection 12.3, and the attendant bioassay program is covered in Subsection 13.3.
The surface contamination monitoring program is covered in Subsection 13.5.
The records systems for exposure control are covered in Subsection 12.5.
12.2 POSTING AND LABELING All areas at Rocketdyne are established as restricted areas as defined in 10 CFR 20.3.
Radiologically controlled areas that are required by these regu-lations, to be designated with signs, are referred to in this document as
" posted areas" in compliance with 20.203, " Caution Signs, Labels, Signals, and Controls," of 10 CFR 20.
tr utdition, areas in which a significant hazard exists are identified by a sign u aring the words:
" WARNING--RESTRICTED ACCESS AREA" or the equivalent. Access to these areas requires prior HS&E approval and compliance with required safety procedures.
Each criticality control area is divided into a number of work stations.
A work station is described by the location of a specific piece of equipment (lathe, sandblaster, plating bath, etc.) required for a given process opera-tion. The allowable batch limit for each type of fuel to be processed at a given work station is entered on a work station limit card.
In addition, approved material of only one type, form, or enrichment is to be present at one time, no matter how small the quantitles involved.
The only exception to this rule is when the process being performed at the work station is intended to change the type, form, or enrichment of the material or when it is neces-sary to assemble fuel of more than one enrichment (e.g., assemble a fuel rod with fuel pieces of more than one enrichment).
When a give1 work station is not in use for processing, a sing'te allow-able batch of anj material may be stored at that station, if the material to be stored is of a type not listed on the (.ard, the fuel must be in a container ESG-82-33 11.12-1
i
{
i-and have a special Source and Special Nuclear Materials (SS) label attached to it in a conspicuous position. These containers are to contain no more than
('
one allowable batch, based on the particular material form and enrichment
-(
involved.
Work station limit cards (for storage only) are conspicuously posted at each storage location. The only work permitted at the storage station is that of unloading and loading fuel _from a single storage unit at a time.
12.3 PERSONNEL MONITORING Personnel exposure to radiation and radioactive matierials is limited to the occupational exposure standards published in 10 CFR 20, CAC Title 17, and DOE Order 5480.lA Chapter XI.
The values are maxima, and every effort is made to maintain exposures as f ar below these values as practicable.
Persons under 18 years of age are not permitted to work in posted areas, and, when visiting, their external exposures are limited to 10% of the quarterly dose limits for workers. Any exposure in excess of the occupational exposure standards neces-sitates immediate notification of HS&E management, and the exposed person is restricted from further work with radioactive materials until the exposure level / period limits are again satisfied.
Personne. external dose is measured with film badges, thermoluminescent dosimeters (TLDs), and pocket chambers. These devices af ford a means of assessing gamma, beta, and, when appropriate, fast-neutron doses to the wearer.
Gamma and/or beta doses of 10 mrem to 1000 rem and whole-body f ast neutron ('l to 10 MeV) doses of 10 mrem are measurable with film dosimeters.
Gamma and/or beta doses of 5 mrem to ~105 rem can be measured using TLDs.
(
Use of a TLD is normally restricted to monitoring extremities. Although a commercial TLD and film badge supplier is employed, TLD dosimetry capability is maintained at Rocketdyne for immediate evaluations when required.
Film badges or thermoluminescent dosimeters are routinely processed once i
per month or once per quarter, as determined by HS&E. Whenever it is sus-pected that an individual has received an excessive radiation exposure, the individual's film badge or thermoluminescent dosimeter is processed immedi-ately. Also, if it is anticipated or suspected that a person will receive a larger-than-normal radietion exposure, special film badges and TLD devices will be issued as required by Radiation Safety.
R. S. Laudauer and Co. is under contract to Rocketdyne to provide the instrumentation. 6nd to read and provide the exposure readings for the per-sonnel dosimeter as required, which is described above.
12.4 SURVEYS The routine surface contamination survey program is covered in Subsec-tion 13.5.
The routine air sampling program is covered in i b::ction 13.4.
{
l y
ESG-82-33 11.12-2
L L
Alpha, beta, gamma, and mixed beta-gamma radiation intensities 'are measured with Eberline Model R0-2 or R0-2A, Ludlum Model 12 with pancake GM or
.(
alpha scintillator, or equivalent survey instruments.
Instruments used pro-
\\
vide a measurement range of zero to several hundred R/h as appropriate to the dose rate ent-ountered.
Maintenance operations in restricted-access areas are controlled via the controlled work permit that requires that radiation surveys be made and assessed during work in these areas.
This is described in Subsection 3.1.1.
L The requirements of the ALARA program must be takta into consideration. The ALARA program is covered in Subsection 13.2.
12.5 REPORTS AND RECORDS Reports conform to the reporting commitments made in Subsection 2.9.
Minutes of the meetings of the Nuclear Safeguards Review Panel and its subcommittees and correspondence involving committee recommendations and actions are maintained for a period of 5 years after the date of preparation, Written reports prepared by the Criticality Safeguards Coordinator or the Criticality Safeguards Advisor regarding radiation and nuclear safety are maintained for a period of 5 years after the date of preparation.
Records of alterations in or additions to facilities are kept up to date to reflect the current facility construction.
Maintenance and calibration records for instrumentation used by HS&RS g
radiation monitoring as required by Subsection 3.2.3 are kept for the life of each instrument.
Personnel dosimetry records and records of the bioassay program are stored permanently.
Results of air monitoring surveys, surface contamination, and other routine monitoring for radiation control are also maintained permanently.
Environmental survey reports are prepared and distributed annually (refer to Chapter 14) as a permanent record.
12.6 INSTRUMENTS Portable and laboratory technical equipment and instruments for per-forming radiation and contamination surveys, sampling airborne radioactivity, and monitoring area radiation have been selected on the basis of known per-formance and acceptability in the industry. The equipment selected has per-
)
formed properly, is maintainable, and is caoable of being properly calibrated within the ranges of consistency required for their individual applications.
)
ESG-82-33 11.12-3 1
t Equipment in each'of the twelve categories listed in Subsection 3.2.3 are provided for the HS&E~ staff for normal and emergency operations in the fol-p lowing quantities:
XJ Ca',,ory Number of in Instrument Category 1
45 l.
2 18 3
24 4
49 5
4 6
2 7
7 8
13 9
29 10 10 11 11 1:
2 13 6
Most of these instruments are available to the HS&E personnel in the controlled areas.
All of these instruments are available for use in HS&E laboratory facilities.
The HS&E laboratory facilities have capabilities for performing identi-fication of -emitting radionuclides, counting " swipe" samples for surface
/'
contamination studies, counting air samples, counting in vivo, counting gross
\\,
samples from the bioassay program, and analyzing the results of the bioassay program performed at outside laboratories.
Calibration, maintenance, and repair of this equipment are controlled through the computerized calibration, result, and inventory system (CRIS).
This system provides notification to the using function when each equipment item is scheduled for service and calibration. The system provides for noti-fication to HS&E management when these schedules are not met.
The frequency and method of calibration are in accordance with the manufacturer's recom-mendation, metrology standards, and HS&E requirements.
The calibrations are traceable to the National Institute of Standards and Technology (NISI, formerly NBS).
I 12.7 PROTECTIVE CLOTHING Protective clothing is used by all who enter posted areas via change rooms or stepof f areas.
Required protective clothing generally consists of -
i 1
2 ESG-82-33 I I.12-4
laboratory smocks and shoe covers worn over personal clothing.
However, addi-tional specialized clothing is used when warranted by the nature of the opera-(
tion, quantity and quality of material involved, and/or chemical and physical
\\
form of the material involved.
Protective clothing requirements are estab-lished by HS&E and may include any or all of the following:
laboratory smocks, coveralis, surgeon's caps, hoods, surgeon's gloves, cloth gloves, shoes, canvas shoe covers, plastic booties, and respiratory protective devices. As an example, use of gloves is mandatory for operations that require hand manipulation of unencapsulated radioactive materials.
Depending upon the chemical and physical form of the material, the gloves required may be surgeon's, canvas, leather, etc., or ct.mbinations thereof.
Laboratory smocks and coveralls are red-trimmed (blue-trimmed for pluto-nium facilities).
This identification is used only for smocks and coveralls to be used exclusively in radiologically posted areas. Smocks and coveralls are pocketless and have no openings to permit reaching into the pockets of personal clothing.
12.8 ENTRY AND EXIT PROCEDURES In preparation to enter a posted area, a worker puts on the specified protective clothing and other devices, nnd checks to be sure he has the appro-priate monitoring devices, such as film badge, dosimeter, lapel air sampler.
Upon exiting, he performs a self-survey with appropriate radiation detection instruments, stores or disposes of protective clothing, or sets it aside for laundering, steps over the line and washes his hands.
He reads and records his dosimeter reading.
r"N Q
12.9 ADMINISTRATIVE CONTROL LEVELS The administrative action levels and alarm set-point frequency of mea-surement and action taken for several radiological conditions are specified in the Health & Safety operating procedures and instructions.
12.10 RESPIRATORY PROTECTION The issuance and use of respirators are under the direct control of the HS&E department except for self-contained breathing apparatus (SCBA) utilized by Protective Services for emergency response situations. A minimum of a medical review plus 2 h of general training in the use of the respirator is given to any person who requires respiratory protective devices.
Additional training is given for special devices such as airline respirators or SCBA.
The equipment itself may be issued only by HS&E personnel.
The proposed usage is reviewed by a HS&E representative who determines the appropriate device to be used. A schedule of equipment issuance and return for maintenance, clean-ing, and sanitizing is established and recorded.
The niinimum f requencies for cleaning at.d servicing are shown in Table 12-1.
Each time a respirator is returned, it is serviced, cleaned, and sanitized before being reissued.
ESG-82-33 11.12-5
f I
p TABLE 12-1 l
-p RESPIRATOR ISSUANCE SCHEDULES Use Period Immediate emergency use One day or until emergency is over Single short-term use*
One month Routine daily operations
- One year and as required in radio-active areas **
Self-contained breathing Six months ***
apparatus
- The total time any one canister containing gas or vapor adsor-bents is worn in each of the above situations will vary. New canisters will be issued when sensing odor or irritation occurs.
Particuate filter cartridges should be exchanged when breathing resistance is experienced.
- Equipment shall be issued for this period if the conditions anticipated are sufficiently defined so that HS&E may issue the proper equipment.
- HS&E shall issue Self-Contained Breathing ApparatLn fSCBA) on a month-to-month basis without requiring that equipment be returned each month provided all of the following conditions exist:
- 1. Each SCBA is completely inspected at least once a month by Operations.
- 2. The SCBA facepiece is returned to the respirator lab every 6 months and exchanged for a freshly sanitized facepiece.
NOTE: Air-purifying respiratory protective devices shall not be utilized for facility emergency reentry.
Additional rest.rictions are imposed on the reuse of chemical canisters.
These units are removed from returned masks and stored for reuse only if all of the conditions listed below exist:
1)
The expiraticn date on the canister has not passed.
2)
The canister -is not damaged or contaminated.
3)
The canister has not been unsealed or used.
ESG-82-33 11.12-6
1 A respirator is issued only to the employee who will use it, and sub-sequent use by other personnel without reissue is prohibited.
Field testing
(
of the facepiece is performed by the issuing HS&E representative. A qualita-(]f.
tive fit test using irritant smoke or organic vapor such as isoamyl acetate may be performed if required due to the degree of hazard anticipated.
For nonroutine and emergency operations in areas of unknown but poten-tially high radioactive concentrations, pressure-demand SCBA are generally used.
In all other cases, protection is attained by proper engineering safe-guards; all areas have air concentrations well below the occupational standard under normal conditions. When respiratory protection is considered advisable for precautionary purposes (e.g., changing HEPA filters), the maximum poten-tial air concentration is considered, based on radioacthe materials that may be present, past experience, etc. The respiratory protective device then selected is that which will afford adequate protection against the highest estimated pctential airborne concentration of radioactive or toxic aerosols.
Prior to reuse, dual inspections of respirators are performed, first by servicing nersonnel and second by the issuing HS&E representative.
A respirator, including its filter cartridge, may be reused only if both the conditions listed below exist:
1)
The surface contamination level is less than 1 dis / min cm2, beta emitters, and/or 0.1 dis / min cm2, alpha emitters.
2)
The cose rate f rom fixed contamination is less than 1 mrad /h at the respirator or cartridge surf ace.
i 0012Y/c1h
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J ESG-82-33 11.12-7 J
13.0 OCCUPATIONAL RADIATION EXPOSURES
(
13.1' OCCUPATIONAL' EXPOSURE ANALYSIS An annual assessment is made to review and analyze employee exposures and i
effluent release data. The three most recent reports, for the years 1986, 1987, and 1988 are contained in Appendix D.
13.2 MEASURES TAKEN TO IMPLEMENT ALARA I
The general objective of the program to implement ALARA is to minimize radiation doses received by both individuals and groups by eliminating J
unnecessary exposures and closely controlling exposures necessary to the proper performance of the work.
The primary. responsibility for identifying the need and the method for achieving this objective is assigned to the Radia-tion and Nuclear Safety Unit.
Individual doses are scnitored in terms o' millirems per job, week, month, quarter,. and year.
Group doses are monitored in terms of man-rems in similar time intervals.
Individual doses per quarter and year are directly compared with appropriate dose standards as established in 10 CFR 20 and with prorated standards for weekly and monthly doses.
Value judgment considera-tions are applied to engineering designs and operating procedures to assure ef fective -operations and control of radiological hazards. As a result of ef fective engineered.and operational controls, the exposure of the general public is so low as to be undeterminable, and thus the cost-benefit analysis is applied to the workers most directly affected.
Additional guidelines for implementation are taken f rom U.S. NRC Regula-tory Guides 8.8 and 8.10 and DOE Order 5480.11.
Tasks involving radiation exposure that offer the greatest potential for a significant reduction in individual and group doses are identified by operating supervision and by Radiation and Nuclear Safety.
Exposures that, when added to other exposures normally expected for an individual, may result in a yearly dose approaching or exceeding i rem generally fall in this category.
While the maximum personnel dose limits are those stated in the regula-tions (10 CFR 20), a planning guide of 1.0 rem per calendar quarter whole-body dose is applied in practice. Where necessary and allowable, control of doses up to 3 rem in a calendar quarter may be applied.
1 The use of pocket dosimeters and special film badges that are processed following such work provides special control during operation in high-exposure situations.
I ESG-82-33 11.13-1 I
i I
l.
,h Radiation and contamination surveys and determination of airborne
.w. radioactivity. concentrations are made by trained and experienced Radiation
~O Safety technicians and engineers.
Allowable working times and the need for
! - V~
special precautions, such as protective clothing and respiratory protective devices, are determined by Radiation Safety personnel. Radiation and Nuclear Safety, operating as a separate organizational function within the Health, Safety & Environment Department, is independent of the operations groups and has the responsibility and authority to halt unsafe operations.
13.2.1 Documentation Prior to beginning operations with the potential for radiation exposure, facility and equipment design and procedures are reviewed by the appropriate committee of the Nuclear Safeguards Review Panel.
These reviews, recommenda-tions, and responses are documented ii. the panel files.
Personnel doses, as recorded'by the external dosimetry and bioassay pro-grams, are documented, reviewed, and reported as part of the function of these programs.
Unusual or high exposures are investigated, and, if warranted, changes in working conditions are recommended.
In severe cases, changes may be required.
Doses indicated by pocket dosimeters are recorded at the f acil-ity and continually reviewed by the assigned health physicist.
Minor adjustments of operations or f acility de,ign are recorded in log-books maintained individually by the health physicists. Major modifications and considerations are documented in safety plans.and internal letters.
13.2.2 Reassessment A continuing review of operations, equipment, and personnel doses is per-formed by the facility health physicist in conjunction with the manager and other members of Radiation and Nuclear Safety.
13.2.3 Internal Review External doses are reviewed and analyzed to determine the distribution of doses and to identify the highest dose groups.
Comparisons are made against past operating experience and against comparable data from similar operations, where available.
Positive indications of internal contamination are so low and so infrequent that such analysis is not considered to be useful and the major effort is directed at reviewing facility contamination surveys.
While the continuing operation of the radiation protection program, in a manner consistent with ALARA goals, is the responsibility of the Radiation Safety Of ficer, reviews of the results and functioning of this program are conducted by the Radiation Safety Committee of the Nuclear Safeguards Review Panel. These reviews are held as p'.rt of the committee's regular annual ESG-82-33 11.13-2
comprehensive review of radiological operations and cover exposures and doses to personnel, facility safeguards and procedures, and ef fluent control. Spe-
.s cific exposure situations are considered, and recommendations for improved tA control may be made.
The major operation of the Radiation Safety Committee is reviewing and auditing operations with radioactive material and radiation-pro-ducing devices, and the membership is experienced in analyzing radiological safety problems.
13.3 B10 ASSAY PROGRAM Internal ext)osures to radioactive materials are determined by Dioassay
-measurement techniques, including urinalysis and in vivo methods.
Urinalysis is performed for all. workers potentially exposed to airborne radioactive materials.
This program is outlined below.
It conforms to the provisions of NRC Regulatory Guide 8.11.
1)
Specimen Collection Criteria a)
Baseline Specimen--A baseline specimen is obtained from each worker prior to his first assignment to work in a posted area subject to airborne radioactive material, b)
Pilot-Program Specimen--During the progress of a pilot program leading to a new and extensive fabrication, pro-cess, or research ef fort, specimens are collected from each worker at f requencies of -l week to 1 month until the effectiveness of contamination controls and engineering
/
safeguards employed by the program are verified for the material quantities involved.
c)
Initial Project Specimen--When the full-scale project begins, specimens niay be collected more frequently (depending on the nature of the work) for all personnel until the effectiveness of contamination controls and engineering safeguards employed by the program are veri-fled for the larger quantities involved, d)
Routine Project Specimen--Following verification of the effectiveness of contamination controls and engineering safeguards, specimens are normally collected for all per-sonnel on a quarterly basis at staggered intervals.
If results are consistently negative over a long period of time, the frequency may be reduced to semiannually or annually.
e)
Special Specimen--Specimens are collected each time a significant inhalation exposure is suspected and following each instance of a contaminated wound. Special follow-up specimens are collected after every ir.3tance of high uri-nary or fecal excretion of radioactive material.
ESG-82-33 11.13-3 r
2)
Specimen Evaluation Criteria L /9 a)
Routine specimens are analyzed by a commercial certified L V laboratory. Spiked specimens may be used to evaluate the accuracy of the commercial laboratory results.
Evaluation of bioassay results to determine internal exposure is per-formed by HS&E.
b)
Special specimens may be evaluated at Rocketdyne by HS&E for screening purposes in addition to the commercial laboratory analysis.
c)
If a specimen indicates excretion of radioactive material equivalent to more than 10% of the appropriate investiga-tion level, a second specimen is analyzed.
If the second specimen result verifies the first, reporting is performed as outlined:
(1)
Positive excretion bioassy results are reported to the HS&E representative assigned to the facility or.
project at which the exposed individual is employed; corrective action such as additional engineering or administrative safeguards is taken as necessary to control subsequent exposures.
(2)
Excretion rates greater than 50% of the maximum per-missible body burden (MPBB) equivalent are reported n-to HS&E management.
.V (3)
For excretion rates greater than 50% MPBB equivalent, corrective action is taken under the direct surveil-lance of HS&E supervision.
(4)
For excretion rates greater than 50% of the MPBB equivalent, personnel may be restricted from further work in posted areas, normally until the rate is less than 25% of the standard.
(5)
For excretion rates greater than 100% of the.4PBB equivalent, personnel are restricted f rom work as described in (4) immediately above, d)
Both material mass and activity excretion may be measured as necessary to determine exposure details.
Activity measurements are made when exposure to radionuclides of sufficiently high specific activity is involved. Mass measurements are made in the case of radionuclides of such low specific activity that activity measurements are impractical (e.g., normal and depleted uranium).
Both Q) f ESG-82-33 II.13-4 4
(m) activity and mass measurements are performed when war-V' ranted, as when exposures to both enriched and normal or depleted uranium are involved.
e)
Urinary excretion rate guides are conservatively appropri-ate to the radionuclides (s) involved.
(For example, stan-dards used are 100 dis / min. day for enriched uranium, 0.4 dis / min day for plutonium-239, and 500 dis / min day for mixed fission products.)
13.4 AIR SA!!PLING PROGRAll Continuous air monitors with audible alarms are installed in radioactive material handling areas when there is a potential for accidental airborne radioactive material concentrations of sufficient magnitude to cause, within 1 h, inhalation exposures exceeding the standards for posted areas. !!onitors with sufficier.t sensitivity to warn workers in time to prevent overexposure are located in prevailing room air currents near work areas.
Air is sampled for shift-length periods, for subsequent evaluation for radioactivity by councing systems, by use of breathing-zone air samples, house vacuum lines, or small vacuum pumps equipped with filter holders (Gelman Instruments 11odel 1220 or equivalent).
Such samplers are required when there is a potential for airborne radioactive material concentrations greater than 25% of the standards for posted areas. These long-term samplers are mounted p
near breathing haight in the handling areas. Air volume measurements are made d
using flow-rate indicators or samplers equipped with limiting orifices or flowmeters. Other air monitors are listed as Category 10 instrumentation in Subsection 12.6.
Large air volumes are sampled over short periods, for subsequent activity evaluation in counting systems, with ur samplers (Staplex liodel TFlA0110V or equivalent). These large-volume samplers are used to evaluate particular operations by locating the sampler at appropriate heights in the immediate area of the operation.
Action '.evels and actions to be taken if these levels are exceeded are described in Subsection 12.9.
1 13.5 SURFACE CONTAMINATIDM There is a potential for the spread of contamination in all arets where the licensed activities are underway as identified in Subsection 9.5.
Change rooms and stepoff pads are normally provided between the administrative areas and areas where airborne radioactivity could exist. Protective clothing (as described in Subsection 12.7) is provided in these change rooms, as are some operational survey meters. The policy on the use of protective clothing is also described in Subsection 12.7.
n ESG-82-33 l
11.13-5
g 13.5.1 Surveys V
The normal frequency for surface contamination surveys in areas in which work is being conducted with unencapsulated radioactive materials is once per week. Additional surveys are required at any time there is an indication of excessive contamination. Surveys are performed less frequently where experi-ence indicates that there is little contamination potential. However, the minimum frequency is once per month.
In unposted areas, and in posted areas in which work with unencapsulated radioactive materials is not permitted, the minimum frequency is semiannually.
13.5.2 Snear Test Limits and Action Guides Nonfixed contamination at sufficiently high levels may be subject to resuspension in air, creating a potential inhalation exposure problem. When contamination levels exceed the Upper Limits given in Table 13-1, immediate cessation of operations is required and decontamination procedures are initiated. These values are called smear test Upper Limits.
Action is taken when contamination levels considerably below the smear test limits are detected. Smear test Action Guides are specified as the levels of contamination at which decontamination or other corrective action is promptly effected but cessation of operations is not required. For low-MPC, ligh-SA contaminants, detection of contamination at any level is considered to be indicative of trouble, and steps are taken promptly to identify the con-tamination source, correct it, and effect decontamination. The action guides O
are shown in Table 13-1.
Every practical effort is made to maintain levels V
below these guides.
l Radionuclides groupings are specified according to the NRC " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material."
13.5.3 Survey Instrument Limits Fixed contamination of sufficiently high levels may create an external dose rate problem. The generation of nonfixed contamination from fixed con-l tamination is also a potential problem.
l These limits appear in Table 13-2; they are employed as absolute maxima.
Every practical effort is made to maintain levels below these maxima, and decontamination or return to an appropriate area is required when the maxima are exceeded.
l l
l 3
i ESG-82-33 II.13-6
i
(
TABLE 13-1 ACCEPTABLE REMOVABLE SURFACE CONTAMINATION Action Guide Upper Limits (dpm/100 cm2)
Restricted Areas
' Contamination Areas /
Radiation Areas /
Airborne Restricted Type of Unrestricted Radioactive Radioactivity Access Contaminant Areas Material Areas Area Areas Transuranic, Ra-226, Ra-228, Th-230, Th-228, h
h
[
Unspecified Pa-231, Ac-227, I-125, I-129 Th-nat, Th-232, Sr-90, Ra-223, 20 100 200 Ra-224, U-232, YOU T60U TOUOU Unspecified I-126, I-131, 1-133
(]-
U-nat, U-235, 20 100 200 U-238, and associ-T6UU" M"
200000 Unspecified s
c ated decay products Beta-gamma emitters (nuclides with decay modes other than alpha emission or spon-100 1000 5000 8
taneous fission)
TOT 6 Y TUDUSY BY Unspecified 200000 except Sr-90 and others noted above.
Increase by 3X for tritium l
ESG-82-33 11.13-7
TABLE 13-2 a
SURVEY INSTRUMENT SURFACE CONTAMINATION LIMITS
~
Type of
~
Contaminant Limit liigh-MPC 0.2 average,1.0 maximum mrad /h 1 cm from surface through 7 mg/cm2 absolute-Low-MPC, low-SA 500 average, 2500 maximum dpm/100 cm b 2
Low-MPC, High-SA 20 aAveraged of 1 m bThis limit is applicable to uranium enriched in U-235 and to natural and depleted uranium that has been subjected to a recent melting process at Rocketdyne.
For other natural and depleted uranium,-the limit for high-MPC contaminants applies.
13.5.4 Removal of Material and Equipment and Release of Facilities Contamination measurements are required prior to removing material.and equipment f rom areas in which work with unencapsulated radioactive materials is conducted.
For release for unrestricted use, material, equipment, and facilities must satisfy the requirements of Regulatory Guide 1.06, Section 4.
(]
Where this guide does not provide numerical limit, those shown in Table 13-3
(/
are appl.ied.
13.5.5 Contamination Limits for Radioactive Materials Shipments and Radioactive Waste Contamination measurements are required prior to the release of shipping containers and packages containing radioactive materials from ESG premises and prior to the release into normal waste channels of solid waste generated in posted areas in which work with unencapsulated radioactive material is con--
ducted. The release employed is the same in both cases end is shown in Table 13-4.
13.6 SHIPPING AND RECEIVING Contamination controls and limits established for removal of equipment f rom the controlled area, whether for storage, shipment, or other uses, are covered in Subsection 13.5.
Radioactive materials being shipped to or received from offsite are sub-ject to f urther controls.
Radioactive materials are packaged for shipment in containers that satisfy all state and federal regulations governing the ESG-82-33 11.13-8
TABLE 13-3 ACCEPTABLE LIMITS FOR RESIDUAL RADI0 ACTIVITY
[
(ROCKWELL INTERNATIONAL /ROCKETDYNE DIVISION)
Surface Contamination Total Removable (Average)
(Maximum) 2 2
Beta-gamma emitters 5000 dpm/100 cm 15000 dpm/100 cm 100 dpm/100 cm 2
2 2
Alpha emitters 100 dpm/100 cm 300 dpm/100 cm 20 dpm/100 cm Surface Dose Rale, At 1 cm through 0.1 mrad /hr 0.5 mrad /hr 7 mg/cm2 Ambient Exposure Rate At 1 meter from 5 R/hr above background surfaces Soil Contamination Alpha emitters (above background)
Ra-226, Ra-228, Th-230, Th-232 5 pCi/g averaged over upper 15 cm, 15 pCi/g averaged over p
15-cm-thick (j
layers deeper than 15 cm.
Transuranic (TRU) 25 pCi/g Th U 30 rCi/g Beta-gamma emitters (gross detectable
/ beta activity including background) 100 pCi/g (below 3 meters)
Average Maximum 1000 pCi/g 3000 pCi/g WLter Radioactivity Released to 10 CFR 20, Appendix B unrestricted areas Table II, Column 2 Released to sanitary 10 CFR 20, Appendix B sewer system Table I, Column 2 Notes:
- 1) Surface contamination is to be averaged over an area not to exceed 1 m2
- 2) Maximum surface contamination is not to cover an area exceeding 100 cm2
- 3) Soil contamination is to be averaged over a volume not to exceed 1 m3
- 4) Maximum soil contamination is not to involve more than 10,000 cm3 in each 1 m3 11.13-9
TABLE 13-4 CONTAMINATION LIMITS FOR
. ("]
RADI0 ACTIVE MATERIALS SHIPMENTS AND WASTE
\\
).
Aloha Emitters Beta Emitters Smear Te!,t Survey Instrument Smear Test Survey Instrument Limit Limit Limit Limit (dpm/100 cmh (dpm/100 cm, surface)
(dpm/100 cm )
(mrad /h, surf ace) 2 2
220 500 average, 2200 0.4 average 220 2500 maximum 2.0 maximum shipment of such materials. Health Physics personnel then monitor the closed containers and attach a tag thereto that describes 1.he radiation levels, external contamination levels, primary contaminants, and total activity con-tained, assuring that all of these values are within limits prescribed by the state and f ederal regulations. The Nuclear Materials Management Unit '(NMM) ensures that all necessary notifications are made and that internal records are maintained.
If the material is fissile, Criticality Safeguards ensure',
that spacing, etc., are adequate.
Other approval signatures are obtained f rom Radiation and Nuclear Saf ety, NMM, and Criticality Saf eguards. Traffic per-forms a final inspection to verify that the packaging is satisfactory, affixes shipping labels to the container, and completes the shipment.
Incoming shipments of radioactive material are monitored by HS&E person-nel prior to transfer within a f acility. The shipments are then transferred l
to the SNM vault or other suitable area where contents of the shipment are D
verified (NMM for fissile material and Radiation Safety personnel for non-fissile material). Transfer from the vault to the person who ordered the material is accomplished only af ter verification that the person is equipped and ready to receive the material.
13.6.1 Release of Exclusive Use Transport vehicles Af ter unloading a vehicle used for transport of packages as " Exclusive Use of Vehicle," a survey for total and removable contamination is performed.
Total radiation must not exceed 0.5 mrem /h, removable contamination must not exceed those limits shown in Table 13-3.
0013Y/sj h ESG-82-33 11.13-10
{
14.0 ENVIRONMENTAL SAFETY--RADIOLOGICAL AND NONRADIOLOGICAL-l The. Rocketdyne Division' Environmental Monitoring and Facilits Effluent-
.k Annual Reports for 1987'and 1988 RI/RD88-14415 and RI/RD89-1391 -respec-tively, are included in Appendix B.
Also four Semiannual Ef fluent Monitoring Reports. for 1987 and 198820,21,22,23 were transmitted to the U.S.
NRC
' Region V.0f fice, Of fice of -Inspection and Enforcement. These reports give the results and an analysis of these results f rom the' nonradiological and radio-
'. logical environmental monitoring program.
(~,
!A 0014Y/ reg r
k)%
ESG-82-33 11.14-1
___________-___________-_______-___Q
15.0 NUCLEAR CRITICALITY SAFETY V]
(
15.1 ADMINISTRATE'.'E AND TECHNICAL PROCEDURES 15.1.1 Documentation Requirements When the project involves less fuel than that given in lable 4-1, verbal approval of Criticality Safeguards is sufficient for project startup. When a project involving quantities of fuel equal to or greater than those given is L
planned, the engineer involved must inform Criticality Safeguards of the plan of action for the project.
This plan of action includes a brief description of every step in the fuel handling procedure under Rocketdyne jurisdiction, from receipt of the material up to and including storage in the Source and Special Nuclear Materials (SS) vault and delivery to a customer, or delivery and first storage at another Rocketdyne f acility.
(If all or. portions of the specif:c procedures to be followed are contained in an existing approved Nuc-lear Safety Analysis, the document may be referenced rather than rewritten.)
The plan of action may include a process flow sheet indicating the form and quantity of the material in each step in the process f rom receipt until ster-age (see above).
If the safety of a particular process step depends on the geometry of the equipment, the geometry must be specified as well as the pro-posed handling procedures.
15.1.2 Nuclear Saf ety Analysis Requirements l
In instances where a Nuclear Safety Analysis is required, it is normally
)
prepared by the group requesting the approval for fuel handling. The cover-page format for this analysis is shown in Figure 15-1.
In addition to the n
information required in this form, the analysis must contain a step-by-step description of each operation in the process, the room in which the operation is to.be performed, and the maximum safe batch size at each step, or the con-t,ainer geometry indicating the degree of safety involved when tbe safety depends on the geometry of the process vessel. The rcom number, the process steps to be performed therein, the amount of fissile material (or number of f uel rods, plates, etc.) at each work station, and the number of work stations in a given room are included.
The Nuclear Safety Analysis also contains a room layout (or as many room layouts as are required to include all work stations used in the fuel handling process) showing the exact physical loca-tion of every work station within a room. Where storage cabinets or racks in a particular room are involved, their location is also included.
The spacing permitted between stored units, as well as the minimum spacing allowed between storage stations and/or work stations, is also included in the Nuclear Safety Analysis.
The co.aplete Nuclear Safety Analysis requires the review of the chairman of the Fuels Committee.
Before recommending approval, the chairman must request that this Nuclear safety Analysis be reviewed by the entire committee i
I i
ESG-82-33 11.15-1 k
neckeidyne Divi
. HockwellInternational Corpora'l
' NUCLEAR SAFETY ANALYSIS NSA NO
,k REFERENCED REPORTS:
PAGE OF _
5 AUTHOR:
FUEL DESC 'IPTION:
i DEPT /G ROUP:
EXT:
A33Ay:
REQUESTER'S MANAGER G.O. NO:
MA53 OF FiZ:LE '30TCPE:
RESPONSIBLE ENGINEER:
j SUBACCOUNT NO:
1 PHYSICAL FORM:
}
SCCPE OF WORK COVERED:
j estimATso ouR ATtDN:
MANAG ( FOR EACH AREA IN WHICH MATERIAL WILL BE RECEIVED:
MANAG DATI:
AREA:
MANAC CATg AR E A:
{
i MAN 4G;M DATf:
AREAS,
MANAGER LAYt:
, AREA:
b MANAGER DATE:
AR E A:
RADIATION AND NUCLEAR SAFETY REVIEW CRITICALITY $AFEGUAROS COORDINATOR:
C ATE; RADIATION $AFETY MANAGER ;
...__ O ATE:
N CRIT CALITY $AFfGUARCS ADVISOR:
DATE:
FUELS SAFEGUARDS COMMITTEE REVIEW:
CH AIRMAN'S SIGNATURE:
DATE:
APPROVAL BY NUCLEAR SAFEGUARDS REVIEW PANEL (NSAP) APPROVAL AUTHORITY:
DATEa l
RECOMMENDATj0NS AND RESTRICTIONS BY FUELS SAFEGUARDS COMMITTEE AND NSRP APPROVAL AUTHORITY:
SIGN ATU RE DATE SIGNATURE DATE Figure 15-1.
Nuclear Safety Analysis Cover Page ESG-82-33
\\
II.15-2 L_i____________--_____--_.--
l I
if (1) the chairman is concerned about a control procedure that has not been approved by earlier analysis and reviews or (2) an element of safety is n
involved that requires further evaluation.
The Nuclear Saf ety Analysis must
-i) receive the signatures of the requestor's immediate manager, the managers of all other areas into which it is proposed that the material be moved during the course of the project, the Nuclear Materials Management manager in cases where vault storage is involved, the Criticality Safeguards Coordinator, the i
l Radiation Safety Manager, the Criticality Safeguards Advisor, the chairman of the Fuels Committee, and the approval authority.
15.1.3 Modifying Existing Procedures During the course of a project, it is sometimes necessary to modify the existing operations. These modifications may involve changes in precedures, work station relocations or additions, or changes in h. clear safety proce-dures. The request for such modifications is made in writing to the Criti-cality Safeguards Advisor by either the person who prepared the original report or his manager. A copy of this request is also sent to the chairman of the Fuels' Committee. These modifications are reviewed and approved by the i
Rocketdyne Criticality Saf eguards staf f.
Criticality Saf equards Approval--If the Criticality Saf eguards Advisor and the chairman of the Fuels Committee agree that only a change in procedure is involved or new work stations are to be added or old ones deleted or relocated (no change in approved safety criteria), their approval is sufficient. Such l
approval is also suf ficient when the nuclear criteria to which the modifica-l tion conforms already appear in the previously approveu Nochar Safety Analysis.
Prior to recommending approval, the chairman of the Fuels Committee
. []
requests that the committee review the modification (s) when (1) c control V
procedure that is associated with the modification has not been approved by
}
earlier analysis and reviews and hence requires review or (2) a new element of safety is involved that requires evaluation. This includes processes or facilities not previously reviewed.
Changes in equipment are also reviewed when the basis for safety is changed (i.e., batch process to a continuous process, safety based independent of moderation to that based on the absence of or on controlled moderation). The review of new facilities includes a physical inspection before startup with fissile materials (e.g., enriched l',
Pu, U-233).
All fuel handling facilities are under an annual inspection, scheduled so there wil be no more than 14 months between inspections.
Reviews of existing facilities include inspections before change from those handling fissile material in solid form to those involving solutions containing fissile material, when it is desired to change process control independent of the degree of moderation to that based on controlled limited moderation, or when a batch process is changed to a continuous one.
Approvals are in writing.
I ESG-82-33 11.15-3 H
l
15.2. PREFERRED' APPROACH TO DESIGN f'T No' design changes to ex~, sting process lines are required for current
('j ^
programs. 'The present design has been affirmed and approved, based on the requirements and criteria contained in Section 4.
15.3 BASIC ASSUMPTIONS This 'information is contained in Appendix C.
15.4 ANALYTICAL METHODS AND VALIDATION REFERENCES This-information is contained in Appendix C.
15.5 DATA SOURCES This information is contained in Appena1x C.
15.6 FIXE 0 POISONS 2
Fixed poisons are not required in processes covered by this license.
15.7 STRUCTURAL INTEGRITY POLICY AND REVISION REQUIREMENTS There are no plans for additional structures in the present process lines.
If any were to be added, structural integrity to assure an adequate i
margin of safety would be reviewed in the internal approval process as described in Sections 4.1 and 15.1.
7
-tk 15.8 SPECIAL CONTROLS There are no special controls beyond those already described above.
j 000lY-srs/sjh i
i l
(
ESG-82-33 11.15-4 l
1
j 16.0 PROCESS DESCRIPTION AND SAFETY ANALYSIS 16.1 ROCKWELL INTERNATIONAL HOT LABORATORY 16.1.1 Operations Description l
16.1.1.1 Research and Development Operations The specific operations performed in the cells vary depending upon cur-rent needs and changing program requirements; the work is principally research I
and development.
Thus, only general techniques and types of operations typi-l cal of hot-cell operation are listed.
The various capabilities for the cell I
block area are listed below:
1)
Cell 1 a)
Preparation of samples of irradiated material for i
I metallography b)
Microhardness testing c)
Microscopic measurements d)
Preparation and replication of samples for electronmicroscopy e)
Autoradiography on mounted samples.
2)
Cell 2 a)
Materials testing Tensile testing Stress-rupture and creep testing Fatigue testing b)
NaK and sodium distillation c)
Visual examination d)
Density measurements e)
Dimensional measurements f)
Minor component disassembly g)
Fission gas collection f
(,
ESG-82-33 11.16-1
l-h)
Isotope encapsulation.
h'"
3)
. Cell 3
\\
a)
Disassembly cell for irradiated materials b)
Sample preparation c)
Elox equipment d)-
Cutoff wheel e)
Waste packaging f)
Visual examination g)
Stereomicroscopic examination h)
Dimensional measurements 1)
Cask unloading.
j)
Laser cutting 4)
Cell 4 a)
Hydrogen analysis
(
-b)
Profilometer measurements c)
Annealing studies d)
Permeation testing e)
Major component disassembly and repair f)
Visual examination g)
Stereomicroscopic examination l
h)
Fuel canning i)
Dimensional measurements j)
Waste packaging k)
Cask unloading and loading 1)
Density measurements
('s t.
j ESG-82-33 11.16-2
m)
Gamma spectrometry A
n)
Autoradiography on capsule assemblies.
Q 16.1.1.1.1 Procedures and Equipment A flow diagram for a typical hot-cell examination is shown in Figure 16-1.
16.1.1.1.1.1 Materials Post-irradiation examinations are conductd on irradiation experiments containing such solid reactor fuel materials as U-Mo, UO, U-IrH, 2
plutonium-bearing fuels, etc. Also, cladding materials examined include aluminum, sintered aluminum powder (SAP), stainless steel, zirconium, and Hastelloy alloys. Sodium and NaK bonded capsules are examined.
Irradia tion experiments are received for examination f rom sodium, organic, and water-cooled reactor systems and tests.
16.1.1.1.1.2 Disassembly Many different types of equipment are used for remote disassembly of fuel-element and reactor components, including, for example, a remotely oper-ated pipe cutter used to make tubular cuts on irradiation test capsules.
i Several hacksaws are used for disassembly operations and for cutting radio-active scrap into lengths suitable for processing. An abrasive cutoff wheel j
is used for sectioning fuel and cladding.
Disassembly of a variety of reactor hardware and components is done by a remote milling machine. An electrical
(
discharge machine (Elox) is extensively used for precision disassembly and sectioning-fuel and cladding. All types of cuts can be made with this ma-4 chine, such as plug cut, coring, and longitudinal cuts for slitting fuel assembly cladding.
1.aser cutting can also be used as a fuel decladding tech-nique. Many specialized jigs and fixtures for remote operation and dis-assembly of components, such as an electric arc cutting device, have been designed and fabricated.
1 i
16.1.1.1.1.3 Analytical 1
Nondestructive testing equipment currently in use includes a gamma activ-ity scanning spectrometer equipped with a movable stage used to record the relative activity profile of fuel assemblies.
An eddy current device can be t rrd nondestructively to detect fuel cracks and cladding defects in irradiated Tuel rods.
Permeation appa atus can readily detect and record the helium or hydrogen diffusion rates through metal cladding materials.
A pin-hole camera is used to take autoradiographs of fuel assemblies to determine fuel integrity before disassembly operations are initiated.
il l
ESG-82-33 11.16-3
UNLOAD FROM CASK NORMAL FLOW.
VISUAL EXAMINATIONS
/~
PHOTOGRAPHY
-- --- ALTERN ATE FLOW t
? '
DIMENSIONS 1\\
CECMETRY DETERMIN ATION PolNT
___--.___-_~~--
GAMMASCAN F13ll0N CAS CUTER CAPSULE Di$ ASSEMBLE OUTER CAPSULE F11110N CA$ OUTER CAPSULE Di$A3SEMBLE OUTER CAPSULE CUT SPECIMENS FROM HARDWARE }-- - -.
~.
I I
METALLOGRAPrIY CHEMICAL ANALYS13 OLMENSIONS CLADDING DETER 5t[AbON 70iST~~~~ ~
~~~~~~~~~~7 VISUAL EXAMINATION PHOTOGRAPHY (UNDER NITROGEN ATMO3PHERE)
HYDROGEN PERMEATIONilf REQ'D)
VISUAL EXAMINATION PHOTOGRAPHY FISSION GAS INNER CAPSULE CL ADDING DIMENSION!
DISASSEMBLE INNER CAPSULE Dl3 ASSEMBLE INNER CLA*) DING l
CUT SPECIMENS FROM H ARDWAREh _ - - - - - -
)i f
I j
i METALLOGRAPHY CHEMICAL ANALYSIS FUEL DIMENSIONS UNDER FUEL EXAMINATION UNDER FUEL PHOTOGRAPHY NITROCEN FUEL PHOTOGRAPHY NITROGEN DISTILLATION OF Na OR NaK ATMOSPHERE' FUEL DiMEN5 IONS Ol!TILLAT10N OF Na OR NaK I
.- ---_--.........i FUEL INTEGRITY POINT PRELIMINARY $PECIMEN CUTTING (IF REQ'DJ -y DENSITY ELECTRICAL RE3ISTIVITY I
HEAT TREATINGIPRE AND POST DENSITY AND DIMEN310N!' CYCLE) 1 l
CUT $PECIMEN$
i j
a I
I 1
HYDROGEN I
. ANALY!!!
METALLOGRAPHY BURNUP CHEMICAL TOTAL FISSION ANALY313 ANALYSIS CAS RELEASED u
I f DATA REQUCTION )
l l REPORT WRITING l
)
Figure 16-1.
RIHL Process Description 52s 4
CSG-82-33 11.16-4
l l
Equipment used in destructive testing includes such items as a fission g:ss apparatus, for collecting fission gases produced in irradiated fuel
(]
assemblies.
Q)
A special annealing apparatus is used for the post-irradiation testing of fuel samples at various temperatures and conditions to determine irradiation damage effects.
A remote gla'.s-blowing operation is used in conjunction with this apparatus for sealing samples in quartz capsules.
The vacuum extraction hydrogen analyzer collects and measures the volume of hydrogen gas evolved f rom a f uel sample at elevated temperatures. The hydrogen analyzer is used primarily on hydride fuel experiments.
Radiochemistry and mass spectrometry analyses are conducted on fuel samples to determine fuel burnup.
Density measurements are obtained using the Archimedes liquid-displacement method. Two types of density systems--an Ainsworth remote readout balance and a direct-reading Mettler analytical balance--are employed.
Operation of micrometers, tri-mikes, calipers, pi-tapes, and dial indicators remotely can provide accuracies to 0.001 in.
Two remotized profilometer systems, utilizing pressure transducers, are cur-rently in use to accommodate a wide variety of experiments. This equipment is primarily used to determine changes in length and diameter af ter irradiation.
All measuring equipment is designed for continuous calibration while in service.
16.1.1.1.1.4 Materials Testinq and Metallography Remote tensile testing is performed on an Instron Model TT-C universal
-m()
machine, or equivalent, capable of subjecting test specimens to 20,000-lb force in tension or compression. An extenslometer attachment allows strain determinations to be made to an accuracy of 0.0002 in.
In addition, equipment is available for the performance of stress-rupture testing of hollow cladding tubes by internal pressurization and for fatigue testing of strip specimens by alternate bending.
Metallographic samples can be mounted in epoxy resins or thermosetting plastics.
Vacuum impregnation mounting is employed for porous or fragile materials.
Two automated grinding and polishing stations allow simultaneous processing of several specimens.
Vibratory polishing equipment is also avail-able. A microscope is used to check the quality of the metal surf ace before examination.
A standard Bausch and Lomb Model A-1001 metallograph has been shielded in order that radioactive specimens emitting up to 50,000 R/h can be examined.
Examinations can be made at magnifications from 50 to 1000X, using bright-field, dark-field, or polarized light.
Photomicrographs are routinely prepared in color as well as in black and white. With a Unitron Model TM measuring microscope, precise measurements can be made in three dimensions, to an accuracy of 0.0001 in. over a 1-by 1-in. range.
f n()
ESG-82-33 11.16-5
E<
l' p
l-Microhardness measurements can be taken with 100- to 1000-g loads on a L
Wilson "Tukon" hardness tester. Electron microscopy of irradiated materials
! - (~'.~
is performed by taking faxfilm replicas of the sample and decontaminating the
\\,
resulting replicas to an extent that allows examination in a 1 uncontrolled area.
l 16.1.1.1.2 Hazards Control Cells 2, 3, and 4 are divided into criticality work stations and in-process storage stations with a minimum of 3 f t between fuel in adjacent stations. Only one allowable batch at a time may be in a single station (except under special approval for material in transit). The allowable amount of nuclear fuel materials at each work station must be posted near the viewing window on the operating gallery side of the hot cell.
Inside the cells, the station areas are outlined on the cell floors and walls to serve as guidelines for material location and movement.
Fuel material is kept within these bound-aries. A log of all material movements and material in storage is maintained for each cell.
In addition, a master log is maintained by the materials bal-ance accountability custodian.
Explicit records of size, weight, description of fuel material, and descriptions of all work and storage stations are con-tained in the master log.
Prior to movement of fuel material within the cell or between cells, approval must be obtained f rom the lead operator. The approval of this nove is based on the records for both cells, in addition to the master log.
If, in addition to any batch that is stored in the sample storage drawer at the back of the cell, only one other batch exists within the cell, this batch can be moved without prior approval to any station within the cell.
However, prior to adding or removing material f rom this cell or the g
sample storage drawer, the final station location must be recorded and
(
approval obtained.
(Most of the work performed falls into this category.
Allowing movement within a cell eliminates a large number of unnecessary I
approvals and makes the necessity for approval more meaningful.)
In-process fuel is stored in storage racks.
The quantity of fuel stored in each rack is no more than the maximum safe batch size as determined by a nuclear safety analysis for the particular fuel involved, its enrichment, and its composition.
Storage racks are made for each specific experiment; these or other storage racks can be located in the storage stations at the back of the cells or against the 6-f t-high lead shields that project f rom the back of Cells 3 and 4.
Ilowever, no more than one allowed batch can be moved into or stored in one station at a time.
The inter-cell f uel transfer drawers, located in the shielding walls between the cells and decontamination rooms, are 42 in. long and can extend as much as 36 in. into the cell. When the drawer is housed within the cell wall, the presence or absence of fuel cannot be visually ascertained. When the a
drawer is extended, the fuel in any work station within the cell must be at least 3 f t f rom the f uel in the transf er drawer. Therefore, in the event a fuel transfer drawer is to be moved into a cell, all in-cell fuel must be I
ESG-82-33 11.16-6
located at 'least 3 f t f rom the maximum drawer extension and approval obtained from the lead operator. He will' determine if the operation is permissible, O~
based on the master records of the amount of material within the drawer. All k) inter-cell fuel transfers are limited to 350 g of U-235 or an allowable batch. The drawers between the cells and the decontamination rooms are f or.
storage-only service.
However, when the drawer is to be opened, the same pro-cedure for in-cell f uel location and approval applies.
(This description is related to operations with U-235; similar precautions would be developed for other fissile materials.)
Cell 1 is normally used for preparing irradiated fuel and cladding samples for metallographic examination. Temporary storage for fissile material samples is provided in 2-in.-diameter stainless steel tubes placed.
a rack in the storage drawer located at the back of the cell. The material coming into the cell is primarily irradiated fuel samples in the form of slug sectior.s. Although the average weight of these samples is about 3 g, no material is accepted without approval of the MBA custodian, who checks the master log prior to the move and changes the master log af ter the move. The criticality control limit for Cell 1 is 350 g of U-235, exclusive of the stor-age station at the back of the cell, which is also limited to 350 g except as provided by special analysis The samples in temporary storage are regulated so that the total weight of the U-235 is not more than 350 g, based on their acceptable weight as received. The 350-g limit also includes the portion of the fuel that has been abraded into fine particles and is in storage in a shielded sump.
For the purpose of criticality control, it is assumed that all fissile material in the samples being ground is collected in this sump. Therefore, the sump contents are removed when a total of 350 g of fissile material has been processed through the grinding wheel, or when the radiation level exceeds 5 R/h at 1 f t f rom either the sump or connections, or when the sump becomes plugged. The radiation level is normally limiting.
An unshielded sump receives an amount of fissile material from polishing (micron range per specimen).
The contents of this sump are removed when the radiation level exceeds 10 R/h at contact.
The radiation level is measured approximately once a inonth.
The limit of fuel samples in process is computed according to their full received weight. A log of these weights and sump removals is maintained in the Cell 1 logbook. All fuel material movements into and out of the cell are recorded in the master log. Also, a master log of the fuel specimens in the storage drawer is maintained, i
16.1.1.1.3 Special Operations An example of operations in the RIllL involving a large fuel element is typified by a Sodium Reactor Experiment Core III UC f uel element (6.5'.' U-235 enriched). The element is an eight-rod cluster containing 2.06 kg of U-235.
The rods are 0.60 in. in diameter by 66 in. long.
Although this element con-tains more than 350 g of U-235, it may be present at a given work station and ESG-82-33 II.16-7
\\
I l
subdivided into individual rods. The maximum allowable number of rods is 16
-(45% of critical), independent of container geometry and degree of water
' I '.
flooding.25 However, only one rod (257 g of contained U-235) may be at a single work station if it is to be subdivided into a number of smaller pieces.
The mass limit is no more than 45% of critical for the fuel type and piece size of interest, independent of the degree of water flooding and spacing between pieces.
If the rods are in the form of assembled elements, the maxi-mum allowable number of elements at a given work station will also be no more than 45% of critical for the element configuration of interest, independent of the degree of water flooding or spacing between elements.
16.1.1.2 Decladding Operation Typical of this kind of operation is the decladding of the irradiated Southwest Experimental Fast Oxide Reactor fuel, mixed Pu0 /UO2 pellets.
2 This operation is carried out in glove boxes within the decontamination rooms or hot cells of the f acility. The fuel-element cladding is cut open to allow the removal of the core fuel and depleted uranium oxide pellets and the inter-nal hardware. The fuel and depleted uranium oxide pellets are placed in a transfer canister for shipment to the DOE f acilities.
The movement of fuel into, within, and out of RIHL and the decontamina-tion rooms is controlled. The maximum quantity of core fuel pellets allowed in RIHL is limited to an amount that presents no criticality problem without modera tion.
Radiation surveys are made at each work station prior to initiat-ing operations to substantiate exposure control plans.
The glove box is exhausted through three stages of HEPA filtration.
The s
decontamination room is exhausted through two stages of HEPA filtration.
16.1.2 Safety Analysis of Each Step Each step with significant hazards, such as criticality, radiation exposure or radioactivity release, and fire, is examined as part of the Criticality Safeguards and Fuels Committee reviews, i
16.1.3 Safety Features at Each Step The adequacy of the facility safety features, such as shielding, containment, ventilation, and filtration is considered during reviews.
Special safety f eatures, such as mass or geometry control for criticality prevention, are specified as needed.
0016Y/cih ESG-82-33 II.16-0 I
o
17.0 ACCIDENT ANALYSIS (7
' A summary of. postulated accidents that could occur associated with
/
)
Rocketdyne's licensed activities is included in RI/RD88-206, "Onsite Radio-logical Contingency Plan for Rockwell International Operations Licensed Under Special Nuclear Material License No. SNM-21," August 25, 1988.1 The acci-dent studies in this report have been considerably refined from those included in the environmental report and the previous license support document, AI-75-46.31 73 I
)
'\\_/
i 0017Y/clh L
1 ESG-82-33 11.17-1 L
4 a
i t
I APPENDIX A 1988 ANNUAL REPORT - ROCKWELL INTERNATIONAL 5
k
's
e
- a.
APPENDIX B ENVIRONMENTAL REPORTS e
APPENDIX B ENVIRONMENTAL REPORTS
[
The following environmental reports are included:
RI/RDB8-144, "Rocketdyne Division Environmental Monitoring and facility Effluent Annual Report - De Soto and Santa Susana Field Laboratories Sites,1987."
RI/RD88-139, "Rocketdyne Dhision Environmental Honitoring and facility Effluent Annual Report - De Soto and Santa Susana Field Laboratories Sites, 1988."
O l
l l
0019Y/ reg ESG-82-33 B-1
d
[.
O h
l 1
I i
l l
ed APPENDIX C CRITICALITY STUDIES FOR FUEL HANDLING 6
9 b
s G
4 4
1 APPENDIX C
/
l
(
CRITICALITY STUDIES FOR FUEL HANDLING The f uels currently of interest at Rocketdyne Division are uranium metal, UAl, U/PuC, U0, and Pu02 in the form of powder, pellets, or rods, or x
2 l
I dispersions in aluminum. The uranium enrichments vary from depleted to highly enriched.
Fluorides or other compounds may also be handled.
Safe parameters considered are those. that are independent of container geometry as well as those that are dependent on the geometry of the con-tainer.
Both water-moderated and unmoderated systems have been considered.
However, reflectors equivalent to those for infinite water have been assumed in all cases.
l Although examples of the application of criticality data presented below are directed to specific fissile materials, the methods of applying critical-ity data to other materials are similar. When larger batch sizes than those to be presented become' desirable, it will be necessary to make additional l
criticality studies for the fuels of interest. This demonstration is intended to present methods of analysis rather than actual values for limiting parame-ters. Such values are determined f rom current data.
A.
METHODS OF.",NALYSIS Nuc1 car safety criter;a for all fuel handling operations are based on experimental data,2,32 on conservative interpretations of experimental O
data,4 or on calculative methods that result in conservative criteria when U
compared with experimental data.
Safety of arrays is based on criteria established in the Nuclear Safety Guide,4 either the surface density method or the density analog method.9 To determine the maximum safe number of fuel rods in a single batch, the max 1 mum-saf e-mass-per-unit-length principle is applied.
Nuclear codes used in making criticality studies are checked against experimental data for water-moderated and water-reflected systems or other appropriate tystems.
The calculated critical parameters (mass, volume, and dimensions) must be smaller than the experimental ones over a range of systems similar to that being studied to establish the method as conservative. Such calculations must be done with codes that are generally recognized as suitable for out-of-reactor calculations.
For each use, the code must be validated against well-established experimental or theoretical results, with well-defined uncertainties.
This validation must demonstrate that the range of applicability is appropriate in terms of geometry, materials, composition, densities, reflectors, moderation, and other pertinent features.
In general, nuclear safety limits for process and storage operations are determined by the use of relatively simple calculational methods or previously established criteria.
However, situations may arise in which it is necessary to utilize the maximum safe capacity of a storage vault or work station.
In such a case, recourse to computer codes may be essential. While no computer ESG-82-33 C-1
codes are currently being applied to determining nuclear safety criteria, a b)'
wide variety of codes has been used in reactor and nucle.ar safety analysis in f
'the past. Continuing reactor analysis activities and a modern computer facility provide the functionisl capability for utilizing computer codes as required.
Use of a computer code would require selection of the most appropriate
)
type, whether diffusion, transport, or Monte Carlo.
Following this, a par-l ticular code would be selected, depending upon appropriateness and the effectiveness of applying it to the specific problem. The code would be made i
operational on the Reckwell International coa.puter system and thoroughly checked out by means of all the test cases provided with the package.
Once operational, the code would be validated by calculating specific configurations related to the process or storage operation being analyzed.
)
The validation calculations would be chosen from configurations for which I
experimental data or other validated calculational results are available.
{
The final results would be checked to the extent practical by comparison i
with simple calculations or other safety criteria.
Limits of error would be i
. established by the validation calculations and by calculations made with alternative options, variations in operating parameters, and modifications of input data. Prior to any use of computer codes for criticality control, a demonstration and validation must be submitted to NRC for review and approval.
B.
SAFE PARAMETERS Criticality control limits are established for all operations involving fissile materials. These limits are based on published data and methods sub-ject to the following criteria:
Controlling Fraction of Critical Parameter Value Allowed Mass 0.45 Volume 0.75 Cylindrical diameter 0.85 Slab thickness 0.80 Mass per unit length 0.45 Mass per unit area 0.45*
Mass per unit volume 0.45 Multiplication constant 0.95
- For analysis of arrays of fissile units, this limit shall be 0.25 with individual units limited to 0.30 of critical for i
bare units.
\\
ESG-82-33 C-2
In many cases, data for these controlling parameters are directly avail-able; however, in some cases, the limits must be derived.4 The limits for mass per unit length and mass per unit area are examples, being derived from the limits on cylinder diameter and slab thickness, respectively.
In general, it is the practice to base criticality control on a single limited parameter, such as mass or volume. When this is done, the unlimited parameters are assumed to have their most reactive values. Often it is pos-sible to improve process efficiency while retaining safety by using a combina-tion of limited parameters. Two examples will be discussed here.
In the graphs related to these examples and others following, certain dotted portions are included for correctness but are not applicable unless homogeneity can be guaranteed.
Combined Mass-Volume Control--Under conditions of combined mass-volume control, the allowable mass may be increased if the volume is limited, or the volume may be ihcreased if the mass is limited. A graph of the allowable limits, as derived from Reference 4, is shown in Figure C-1.
6
- g g
g l
i g
g
/
~~~
O id 4 W
8 3
5zg2 g_
s i
u
,1 1
I I
I I
0 O
1 2
3 4
5 6
7 8
9 10 11 235 U
gg33 gg,3 Figure C-1.
Combined Volume-Mass Control Limits for Highly Enriched Uranium--Uncontrolled Water Moderation and Full Water Reflector ESG-82-33 C-3
Controlled Limited Moderation--Limited amounts of moderating materials may safely be mixed with fuel materials as long as adequate controls are placed on the quantities.
These limits are shown in Figure C-2.
A double batch'of fuel and moderator, using the solid linE limit, is less than 90% of critical. -If it is possible to accidentally add excess moderator independent
'of fuel quantity, a'. lower limit is required.
The limit shown as a dashed line applies if double batching of moderator alone is possible, for example. This would result in less than 75% of a critical mass. When these limits are used, moderation must be controlled as closely as fuel is.
^e f
i l
i l
l i
l l
l l
Es E3 E
iC l2
$g INDEPENDENT MODERATOR ooUBLE BATCHING I
I I
I l
/,
n__
i o
1 2
3 4
5 6
7 8
9 to 11 O.
230 U
MASS Ikg) 00-10337 3
Figure C-2.
Controlled Limited Moderation Limits for Highly Enriched Uranium--Full Water Reflector For operations in which the normal quantity of moderating material is f ar below the applicable limit, nuclear safety may be based on limited moderation without the stringent accounting required for the conditions of contro11e'd limited moderation.
Combinations of Fissile Materia 1--Combinations of two or more fissile I
materials are treated by estalbishing the equivalence of one material to the other and basing limits on data for a single material. For example, fuel ele-
)
ments may be fabricated using a blend of enriched uranium and plutonium.
Where this method of analysis is used for plutonium / uranium blends, all U-235 will be considered, regardless of the uranium enrichment.
It may conservatively be assumed that each gram of U-235 is equivalent to 0.7 g of Pu-239 (when there can be no moderation, at optimua water moderation, and for homogeneous mixtures when maximum concentrations can be guaranteed).
Table C-1 indicates the conservatism of this assumption under the conditions l
for which this factor can be used.
3
)
f~
ESG-82-33 C-4 1
1
TABLE C-1 U-235 TO Pu-239 CONVERSION Experiment Pu-239 Equivalence Experiment
(
Parameter for U-235
. Conversion Method)'
for Pu-239 Suberitical 11.5 g/l 8.0 g/l 7.0 g/l concentration Suberitical mass Thermal 760 g 532 g 510 g Fast 20.1 kg 14.1 kg 4.9 kg In all cases (see Table C-1) the equivalent Pu-239 (calculated by the 0.7-conversion factor) values are greater than those for the corresponding experimental critical Pu-239 parameters. Therefore, the conversion methcd is conservative. For safe geometries of mixtures, the safe Pu parameters will be used. This is conservative, as all maximum Pu parameters are smaller than the corresponding ones for U-235.
-C.
SAFE PARAMETERS DEMONSTRATED FOR Pu 1.
Introduction In comparing reported criticality data,2,32 it was noted that there is no agreement for all fuel concentrations between the data in the two refer-enced reports. Where this discrepancy was found, the smaller of the critical parameters was used. The figures presented are based on. conservatively drawn, smooth curves of the maximum safe parameter versus fuel concentration or other independent variable. These curves are based on Pu systems. Similar curves may be developed for U-235 and U-233 systems. Current analyses emphasize the use of. subcritical limits.provided in TID-7016, Rev. 2.
All plutonium criticality analyses (where the principal isotope is 239) are based on total plutonium mass. Normally the Pu-240 content will exceed the Pu-241. content. This will make the results conservative, with respect to TID-7028, for example, where the plutonium is approximately 97% Pu-239.
Only in cases in which the normal distribution of isotopes.(239 > 240 > 241) does not hold would it be necessary to analyze specifically for the effects of Pu-241.
O ESG-82-33 C-5 l
q/
C 2.
Effect of Density 1
Since criticality is affected by fuel density, it is necessary to estab-lish the relationship between fuel density (grams of Pu-239) and the H/Pu-239 l
atomic ratio.
Figure C-3 contains plots of the grams of Pu-239 versus H/Pu-239 100.000 LOOD i
i 10.000 iI I
800
. Pu. s Pu H O 600 60.000 7
2 6.
40.000 - SQL TION
~ #
Ij 20.000 200 6
Q 10.000 100 8.000
- a Pu H O 30 2
6.000
~
Pu H O -
2 B'
4.000 h
40 2.000 20 1.000 19 1
2 4 6 810 20 40 60 80100 30
,s\\
!O ATOMIC RATIO (H Pu239) zw x w.:uu Figure C-3..
Fuel Concentrations--Pu-239 Systems atomic ratio for solutions, as well as for a-Pu ( o = 19.6 g/cm3) and 6-Pu 3
(o = 15.65 g/cm ) -water mixtures.
The concentration decreases with increasing H/Pu-239 atomic ratio for each system, being greater for the more dense fuels at the lower H/Pu-139 atomic ratios. All three systems have the same Pu-239 concentration for H/Pu-239 atomic ratios greater than N100.
Unless there are specifically approved data indicating lar er safe criteria i
for fuel-water mixtures having intermediate concentration Pfuel greater than the density o corresponding to a given H/Pu-239 atom c ratio), the data presented will be corrected as follows to obtain maximum safe param-container volume limits by (c/c )3, educed by the ratio of (p/po)2, the eters.
The mass limits should be r and the container linear dimensions by o
(p/pg).
If p is less than the o for a given H/Pu-239 atomic ratio, how-g ever, limits must not be increased by these ratios.4 ESG-82-33
(
C-6 l
3.
Safe Parameters (Moderated Systems)
)
The maximum safe batch size (45% of critical) as a function of Pu con-centration is plotted in Figure C-4.
The dotted portions of the curves (in L
0.1 0.2 0.4 06 0.8 1 2
4 6 8 10 20 40 50 ' 80 100 4
i e
i >
a r q['s,.,s',--
g g-s' x
2
.5 51 y 0.8 50.6 b
y D.4 5
3 SPHERICAL GEOMETRY, E
{ 0.2 4P. 0F CRITICAL l
I
,0.01 0.C2 0.04 0.06 0.080.1 0.2 0.4 0.6 0.3 1 2
4 6 8 10 PLUTONIUM CONCENTRAT10N'(g-Pu239 3
/cm )
4-147 UNC 00 25d Figure C-4 ' Maximum Safe Batch Size Versus Pu-239 Concentration the region indicating maxima) are in areas where there are greater uncertain-ties in the data. The dotted portion of the curve at concentrations less than that for which the maximum safe hatch is safe, independent of concentration, can be used only-if homogeneity of the mixture can be guaranteed.
If not, the solid portion of the curve must be used.
Criteria indicated by the solid line i
curves will always be followed to ensure nuclear safety.
It will be seen that
'the maximum safe batch size at optimum water moderation and reflection is
-0.23 kg contained Pu-239. When it is possible to exclude water moderation, the maximum-safe batch size may be increased to 2.52 kg contained Pu-239.
These safe batch sizes are independent of container geometry.
Figure C-5 contains a plot of the maximum safe sphere volume as a func-tion of Pu-239 concentration for Pu-239-water mixtures or solutions.
It is seen, from this figure, that the maximum safe volume increases with decreasing i
concentration. Thus, if the maximum Pu-239 concentration of a fuel is 5.0 g
-(the maximum safe volume for 19.6 g Pu-239/cmPu-239/cm,themaximumsafevol at 1.151 rather than 0.191 ESG-82-33 C-7 I
i
w; fh y
0.1
' 0 '2 0j 0.6'O.81.0 2.0 4.0 60 8.0 10
' 20 40 60 80 100 10 s
i.
i i
l i
e i
e i
i e
i i
I I
13 6
g g2 5
N h 1.0 4
- y 0.s
^-
. l 0.6 e*
0.4 67*. OF CRITICAL 0.2 0
I 0,
0.01 0.02 0.04 0.06 0.0: 0.1 0.2 04 0.6 0.8 1.0
~
2.0 4.0 6.0 8.0 10 i
PLUTONIUM CONCENTRATION (g Pu239 3
/cm,
4147 UNC
. Figure C-5.
Maximum Safe Sphere Volume Versus Pu-239 Concentration Figure C-6 contains the maximum safe batch size (45% of critical) of Pu-239 versus container volume.
This is independent of the degree of water moderation and reflecton.
It is seen, from this figure, that the maximum safe batch size becomes as small as 0.23 kg contained Pu-239 when the volume it.
213.5 1.
As the volume decreases, the maximum safe batch size approaches.
2.52 kg contained Pu-239 (maximum safe batch size for all geometries when water or other moderation can be excluded).
Figure C-7 contains plots of the material buckling of Pu-239-water mix-tures as a function of Pu concentration.
These curves'are based on Refer-ences 32 and 2 criticality data, as discussed-above for critical spheres using 8 cm as the water-reflected extrapolation length for the sphere radii.
It is used-in determining the maximum safe batch size (45% of critical) when it is desired.to take advantage of finite specific geometry containers.
ESG-82-33 C-8
__.z_._______z_
l
e
.i 2.6 i
i i
2.4 6
2
} 2.0
-4% OF CRITICAL
~
3
~
B g 1.6 li g
= 1.2
'. S
.g f-00.8 5m 0.4 0
0 2-4 6
8
. 10 u
14 16 18 20
' $PHERE VOLUME (hters) 4+67 UNC -
00 25M4 Figure C-6.
Maximum Safe Batch Size Versus Sphere Volume 0 022 0.026 0.030 0.034 0.028 0.c42 0.046 0.050,g 3
10 a B 9,
y6
- 0.5 2
4
- u
. _n -
g.2
- 02 e
u-8 -;
0.1 Figure C-7.
5'O.a 0.08
- g 3
_ a,c3 Material Buckling Versus O
Pu-239 Concentration
- 5 ' O.4
- 0.04 i
5g 0.2
- 0.32
=>
W 0.1 O.01 0.006 0.010 0.014 0.01B 0.022 0.026 0.030 0.C34 MATERIAL BUCKLING (cm'2) 4.w a:
n: sus
- 6.,
. Figures C-8 and C-9 contain plots of the maximum safe cylinder diameter (85% of. critical) and maximum safe slab thickness (80% of critical, respec-tively, as a functon of Pu-239 concentration. These safe parameters increase ESG-82-33 C-9
q U
g 10
.g
.g.
,g E-, g s U$4 dl2
- 5W 8% OF CRITICAL g2
' k 1.
'I
~
0.01 0.02 0.04 0.06 0.1 0.2 0.40.6 1
2 4
6 10 20 PLUTONIUM CONCENTRATION (g Pu239am3) 4147 UNC 00 25336 Figure C-8.
Maximum Safe Cylinder Diameter Versus Pu-239 Concentration 4
.g
.g ii nE2 5
=
- 1
@ 0.8 d.6 0
p d54 80*, OF CRITICAL 0
a 0.2 l
.l
,,I g
- s 0.01 0.1 1
10 20 PLUT0NIUM CONCENTRATION (g Pu239,tm3) 4-147 UNC 00 05337 Figure 9.
Maximum Safe Slab Thickness Versus Pu-239 Concentration with decreasing Pu-239 concentration. As was the case for maximum safe sphere volume, the maximum safe cylinder diameter and slab thickness may be based on the maximum density possible for the fuel being analyzed. The safe slab thickness (Figure C-9) is especially conservative at Pu concentrations 3
>0.1. g/cm, in order to give due consi(Jeration to the experimental point at 3
N1.1.g/cm in Figure 30 of Reference 2.
Figure C-10 contains a plot of the maximum safe mass / unit length (45% of critical) versus Pu-239 concentration for an assembly of fuel rods.
The dotted portions of the curves (in the region indicating maxima) are in areas i.
where there are greater uncertainties in the data. The dotted portion of the curve i
ESG-82-33 C-10
,1
_ _ _ _ _ _ _ - - - ~ - - - - - - - - ~ - - - - - - - - ' ^ ' - ^ ~ " ' ' ' ~ ~ ~ ~
3 --:
L
[
'i O.01
~ 0.02 0 04 0.060.080.1 0.2 0.4 0.60.8 1 2
4 6 8 ~ 10
\\
6 i
i i
4 i i g
. i 4
c 72
-5 l
IEgi 0.8
@ 0.6 N
$ 0.4 45', OF CRITICAL R
E
$ 0.2
.a~
- i 0.1 O.!
0.2 0.4 0.6 0.8 LD 2
4 jN 8 10 20 40 60 80 100 Pu CONCENTRATION (g-Pu
/ liter) mnn wans Figure C-10.. Maximum Safe Mass per Unit length g
Versus Pu-239 Concentration t
at concentrations less than that for which the maximum safe parameter is safe, independent of concentration, can be used only if homogeneity of the mixture can be guaranteed.
If not, the solid porion of the curve must be = cd. To ensure nuclear safety over the ' entire range of concentrations, safe criteria are always based.on the solid-line curves. The critical parameter as a func-tion of concentration-was obtained by multiplying the cross-sectional area of the critical cylinder by.its corresponding concentration.
It is seen from this figure that the maximum safe parameter goes through a minimum when plotted as a function of Pu-239 concentration.
This figure may be used in establishing the maximum safe number of fuel rods in an assembly.
If this is based on the minimum of the curve, the safe number of rods is independent of the degree of water flooding or container geometry.
Larger safe masses / unit
-assembly may be obtained for specific geometries.
Figure C-11 contains a plot of the maximum safe mass / unit length con-I tainer (45% of critical) as a function of container cross-sectional areas.
Here, too, safe criteria are always based on the solid-line curve. The dotted portion of tge curve for. containers having cross-sectional areas larger than about 60 in, can-.only be-used if homogeneity of the mixture can be guar-anteed Since the cross-sectional area is based on cylindrical containers, it is conservative to apply this curve to containers of other shapes.
i ESG-82-33 C-11 l
_______________________a
r Nf I
4 ma -
i 1
i s
g
-@n2 l
}
45*. 0F CRITICAL li -,L l 50.8-D.6 0.4
,r s
w a'
v'n
' E 0.2 E
Q s'
l l
.,,, 2 4
6 8 10 20 40 60 60 100 200 400
' CROSS SECTIONAL AREA (in.2)
.._..!$f!
Figure C-11.
Maximum Safe Mass per Unit length Versus Container Cross-Sectional Area
(
The maximum allowable keff of an assembly of. fuel roas may also be used in establishing its safety.
For example, it has been determined that a group-
. ing of 0.30 in.-diameter, 5.0% U-235 enriched, metal rods placed in a cylinder having a diameter of 6-5/8 in.'has maximum keff of 0.90 independent of the degree of water moderation or water reflection.
This is safe (keff < 0.95).
The-keff was calculated using the following formulation:
2 2 1+MBN k
eff = 1 + M B 2 2 G
- whereBkandBG are the material buckling and geometric bucklina, respectively, 2
and M is the migration areas.
ThenuclearparametersBkandM2 have been Eextracted from published data.26 ESG-82-33 C-12
1 I
(,aj In using the maximum-safe-mass-per-unit-area principle, the allowable surf ace density depends on (1) the maximum allowable quantity of fuel in a single unit or package, (2) the presence of significant moderation (H/X >0.5, C/X >5M, or Be/X >50), and (3) the presence of close-fitting reflectors about the a'tuj.
1 The maximum safe mass / unit area (based on water-flooded arrays) is limited to 25% of critical when each unit contains a maximum quantity of fuel equal to no more than 30% of critical for bare assemblies, based on the con-tainer geometry, the fuel composition, and the degree of water moderation of interest.
The maximum safe mass / unit area (25% of critical) for infinite slabs is given as a function of Pu-239 concentration in Figure C-12.
The dotted portion R
0.01 0.02 0.040.060.081 0.2 0.4 0.60.81 2
4 6 8 10
~
1 l 6 0.8 8
2 0.6 6
0
$ 0.4 4
i o 0.2 2
5 o(
l 25*' OF CRITICAL l G,
c
[0.08%
s 0.8 E 0.06 l-0.6 2
/
0.04 O.4 E
E
$ 0.02 0.2 3
h 0.01 0.1 0.1 0.2 0.4 0.60.81 2
4 6 8 10 20 40 60 80 8
PLUTONIUM CONCENTRATION (g-Pu
/cm )
04 25366 Figure C-12.
Maximum Safe Mass Per Unit Area Versus Pu-239 Concentration 3
of the curve at concentrations <0.02 g/cm can only be used if homogeneity of the mixture can ta guaranteed.
If not, the solid portion of the curve must be used. The critical parameter as a function of concentration is obtained by multiplying the critical thickness of an infinite slab by its corresponding concentration.
It is seen, from this figure, that the maximum safe parameter goes through a minimum ethen plotted as a function of Pu-239 concentration.
/3 ESG-82-33 V
C-13
p This curve may be used in establishing the maximum safe mass of fuel / unit area of floor space or wall space for a single-plane infinite array of safe group-ings of fuel.
If this is based on the minimum of the curve, the safe mass /.
unit area is independent of slab thickness or the degree of water flooding.
Where the thickness of the infinite slab can be controlled, larger safe masses / unit area are allowable.
Figure C-13 contains the maximum safe mass / unit area versus slab thick-The dotted portion of the curve for slab thickr. asses greater than ness.
N5-in. can only be used if homogeneity of the mixture can be guaranteed.
If not, the solid portion of the curve must be used.
E 5E 1
E 0.B O
to 8.
0.6 6
g 0.4 4
w-25% oF CRITirAL l 0.2 2
b E 0.1 O
b 0.08
/
1
)
0.8 h 0.06 S
O.6 0.04 0.4 i
y 0.02 0.2 5
E 0
.01
^
o,1
)
0.1 0.2 0.4 0.60.81 2
4 6 8 top 20 40 60 80 SLA8 THICKNESS Ond 00-25367 Figure C-13. Maximum Safe Mass per 1
Unit Area Versus Slab Thickness The following is an example of the use of Figure C-13.
Consider a process for which the safety of Pu fuel in an array of 12 x 12 x 2-in.-deep pans must be established.
The cross section of these pans has the same area as that for a 13.54-in.-diameter cylinder. The geometric buckling of a
. cylinder 2 in. in height and having this diameter is the same as that for a 3.88-1 sphere (based on a conservative extri.polation length of 8 cm).
The r,
minimum critical mass of Pu distributed uniformly throughout this volume (bare
!)
sphere) is 21 kg contained Pu-239.2 However, if the Pu were at optimum distribution in the bare pan (or in equivalent sphere volume), the minimum critical mass of a bare assembly would be 10 kg contained Pu-239.2 This t
would allow the use of 2.5 kg contained Pu-239/ pan (25% of critical in a bare pan).
However, another criterion requires the maximum safe batch size per ESG-82-33 C-14
l L
bq pan to be no more than 45% of critical based on a water-reflected pan. Based on Figure C-6, the maximum safe batch size in the pan (equivalent to a 3.88-1 sphere) is 2.06 kg contained Pu-239.
From Figure C-13, the maximum safe mass / unit area for a singlg-plane infinite array of pans having a height of 2in.ig0.12kgPu-239/ft.
Thus, each pan requires a surface area of 17.2 ft to satisfy this criterion.
The pans of fuel (placed in the same plane) must, therefore, be spaced at a center-to-center distance of at least 4.15 f t (3.15 f t edge-to-edge) when loaded to their full capacity.
1 With the exception of the data presented in Figures C-12 and C-13 (safe mass / unit area), all the above criteria are' used for isolated individual units. Where the application of Figures C-12 and C-13 to interaction problems becomes too restrictive, solid angle criteria, safe array criteria for close-fitting reflectors,4 or the density analog method 9 will be used to estab-lish the basis for safe arrays.
4.
Safe Parameters a.
Unmoderated Systems increases in safe parameters may be realized by considering the oxide instead of the metal (particularly for unmoderated systems), metal of reduced density, Pu-240 content, and U-238 dilution. The following analysis presents a conservative method for establishing safe batch sizes for plutonium metal systems based on dilution with natural or depleted urar am.
Obviously, the (yd alloy analysis must be known before credit can be taken for the dilution of V
the Pu.
Allowance factors are presented in Reference 4 for uranium metal of dif-ferent U-235 enrichments, as well as for metals or compounds of fissionable isotepes and elements having atomic numbers between 11 and 83 (from Na to Bi).
Figure C-14 is a plot of the allowance factor on mass limits as a func-tion of U-235 enrichment. This applies only to ur. moderated systems.
For enrichments 15% U-235, there is no limit to the allowable mass. Figure C-15 contains plots of the allowance f actor on mass limits for U-235, Pu-239, or U-233 metals mixed homogeneously with elements having atomic numbers between 11 end 83.
The allowance f actors for oxides, carbides, nitrides, and fluo-rides of the fissile isotope mixed with elements having atomic numbers between 11 and B3 are also included.
Figure C-14 has been extended to mixtures of Pu-239andnaturaluraniumbyassignir.ganequivalenceof1.7gofU-235gr every gram of Pu-239. This factor has been estimated to be conservative.
Although this factor of 1.7 is based on aqueous homogeneous systems, it can be conservatively applied to Pu-natural U alloys containing from 25 to 54% Pu by weight.
The conservatism is apparent when calculated safe batch sizes for alloysofPgforPu-depletedV(0.28%U-235) alloys.and natural U are compared wi mental data Pu-natural U alloys should not have significantly lower critical masses than those for Pu-depleted U systems of corresponding Pu content over the range of Pu concentrations ESG-82-33 C-15
b I
1 i
l V
100 1,
i i i
i 80 j
)
60
't2
}
5 40
-i M
20 E
i g
i 6 10 Figure C-14.
Allowance Factor
<0 on Mass Limits for Uranium
~
[6 Metal Versus U-235 Enrichment
)
a i
E8-4
-i<
2 j^s
' f,
,}
f f
f f
f f
6810 20 40 60 100 ENRICHMENT (Percent U235) 2-5-67 UNC 00-25342 specified. Figure C-14 should not be applied to Pu-natural U alloys contain-ing less than 25 wt % Pu until such applicability has been established by experiment or the results of suitable calculational methods. Neither should the figure be applied to alloys containing more than 54 wt % Pu, as the cor-responding U-235 percentage equivalence would then equal or exceed the 93% for which data exist. The maximum safe batch size (kg contained Pu) is the prod-uct of 2.52 and the allowance f actor given in Figure C-14.
For alloys of plutonium and depleted uranium (0.28% U-235), there are 2
experimental data that can be used to establish safe batch sizes for unmoderated alloy systems. An alloy in a graphite mold will be considered unmograted, provided the atomic ratio of C/Pu-239 is no greater than 500.
The graphite (reflector) surrounding the Pu in the mold will be considered in establishing the maximum allowable graphite in the mold.
j ESG-82-33 C-16
i
~
6 gegtumt ra,mc.,
l i
i 4
El[EtcEtIss m!'
ran ccurounts or rissitt isorces axo c.a.o.r.
j an"IsMi,LTittenisfd n2 i
e l
i ruction$ rutt ce sity or nisia iset$t '
41O UNC 00 5313 Figure C-15.
Allowance Factor for Fissile Isotope (Metal)
Mixed Homogeneously With Elements Listed In determining the maximum' safe batch size for unmoderated assemblies, al'l: isotopes-of Pu in the alley material,will be considered as Pu-239. Assum-ing'that the total Pu content ~in the alloy is 28 wt %, it is equivalent to 47.6% U-235. When added to the U-235 content of the natural uranium, the total equivalent U-235 becomes 48.1% U-235 enrichment.
From Figure C-14, the allowance factor for this fuel is 1.6.
This will be used for the. allowance
~
factor,in-' increasing the safe batch size of contained Pu in the alloy. There-fore, the maximum safe batch becomes 4.03 kg total Pu (1.6 x 2.52).
Credit may be taken for dilutions with elements havina mass numbers between 11 and 83.
Assume the alloy of interest contains 28% Pu, 4% Mo, a..d 68% natural U.
'The U-Pu content of this alloy is 96% (29.2% Pu).
From Figure C-15, the metal allowance factor'is 1.03. The allowance factor for " enrichment" is 1.56 (Figure C-14, based on 29.2% Pu), and the total allowance factor becomes.
1.60... The maximum safe batch size for this alloy is, therefore, 4.03 kg total Pu. This formulation will also be used for plutonium-depleted uranium alloys when there 'are not data (see below) for the specific depleted uranium enrich-ment of interest.
1 3
Thergareexperimentalcriticalitydata5 for 6-phase Pu (density of 14.2 g/cm ) and depleted U (0.28% U-235) alloys. The date. are plotted in-the referenced report as a function of U-reflector thickness and volume frac-tionofPy). These data can~be converted to data for e-phase Pu (density of 19.6 g/cm3 by using the inverse square law, whicn states that the critical mass gs inversely proportional to the square of the fuel density (14.2/
N 19.6). Since a 3-in.-thick " reflector is more effective than an infinite d
l ESG-82-33 C-17
)
L
water reflector,2 the allowance f actor (minimum critical mass + 5.6) for j
a-phase Pu-depleted U (0.28% l)-235) alloys for. f ast systems reflected by this I
i
-thickness U have been ' determined and plotted in Figure C-16 as a function i
G u
2.4
. Pu DEPLETED U(0.28', U235) SYSTEMS 1
82.0 E
C W l.6 5'
d
" 1.2 0.8 g,4 0
0.2 0.4 0.6 0.8 1.0 j
v0LUME FRACTION, s. PHASE Pu AT 19.6 shmb d
4447 UM:
CH5344 Figure C-16.
Allowance Factor Versus Plutonium Volume Fraction of volume fraction of Pu in the alloy. The calculated porton of the curve for i
volume f ractions 20.6 is dotted. 0bviously, it is not possible for the Pu at
(
reduced density to have an ullowance factor less than unity. Therefore, the 1
allowance f actor'will be unity for all f ractional densities 20.6.
This is j
also a measure of the conservatism in the method of analysis used.
Since there are no data -for volume f ractions <0.25, no increases in allowance f actor I
are made for lower Pu densities (Figure C-16).
This method of analysis will be used for alloys having this composition, or in which the depleted U-235 enrichment is <0.28/% when the allowance f actors are graater than those given by the U-235 equivalent enrichment method described abc.ve.
j Based on the alloy contining 28% Pu and 72% depleted U (0.28% U-235), an allowance factor of 2.08 may be applied to the unalloyed maximum safe.baten size. This corresponds to a maximum safe batch size of 5.2 kg total Pu.
~
b.
Limited Moderator Systems Available moderator material under normal operating conditions is defined s
as moderator material, inside an enclosure, that is either used in the opera-d tional process or is available in an open contaner where it can accidentally
)
become mixed with the fissile material (e.g., the fissile material can fall J
into the moderator material). Under these conditions, the 45%-of-critical criterion constitutes protection against double-batching of either the fissile material or the moderator material.
ESG-82-33 C-18
For example, assume that it is desired to blend 3.0 kg contained Pu (in Pu0 ) with water and/or an organic moderator whose moderator equivalence to 2
water can be determined. The maximum safe quantity of water, or its equiva-
- lent in organic moderator, is 2.01.
If as much as 3 kg of the contained Pu content was accidentally homogeneously mixed with 2.0 1 of water, the result-ing mixture would be no more than 45% of the critical mass.
If the water is
~
double-batched instead of the Pu, the system is still less than 75% of criti -
cal. The presence of moderator material under " credible accident condi-tions," as used in this. discussion, includes moderator material (e.g., water) that is not normally subject to mixing with the fissile material but is con-tained within a sealed system (e.g., part of an instrument used in measuring particle size of the fissile material, or containeC in a furnace cooling coil).
Under " credible accident conditions," the moderator material may leak I
out of the instrument or out of the cooling coil and become mixed with the f.issile material.
The following assumptions are utilized for this discus-sion:
(1) There is' no moderator material in the enclosure that is normally used in the process with the fissile material; (2) there is no open vessel containing moderator into which the fissile material may accidentally fall.
However, it is assumed that there is water available within an iristrument or in a cooling-coil system and' that this water does not nonnally come in contact with the fissile material. The maximum safe batch size is 4.8 kg contained Pu (in Pu0 ), based on no more than 45% of geometry in the absence of a moderator.3gritical independent of container 2
l l
If the capacity of the instru-ments and cooling-coil system within the enclosure is no more than 3.01 of water, the system cannot be made critical (i.,e., it would be <75% of critical) independent of the manner of mixing with the fissile material.
In the discussion, no credit is taken for the Pu-240 isotopic content because of the absence of data on the effect of Pu-240 content on the criti-cality of Pu0 -water mixtures.
The only available Pu-240 data are for Pu 2
metal-H O mixtures.
2 c.
Unlimited Moderation It may be desirable to store dilute aqueous solutions (i.e., pickling solutions).
Pu solutions having an H/Pu-239 ratio 28200 are safe. The maxi-mum allowable Pu content of a container (independent of container geometry) will be 450 g.
Under normal processing conditions, an infinite quantity would be safe under these conditions.
Under postulated accident conditions, how-ever, precipitation may take place, resulting in an increase in concentra-tion.
The 450 g of contained Pu is no more than 90% of critical, independent of concentration or container geometry.
ESG-82-33 C-19
n
(;
~F:
ce m
f:
y
- 5.. Solution Handlina The data and the methods of analysis presented are' also used in the handling and processing of solutions. Among the additional variables.that
.must be further evaluated for solutions are the following:
- 1). Transfer lines in and out of vessels 2)
Precipitation-(when not required by process step)
- 3) Maximum density as a function of moderation for solutions as well as precipitated' solids
- 4) Fissionable materials, in equipment not safe by' geometry, where material accumulations are not normally expected but where they might occur as a result of equipment failure, misoperation,. or other abnormal conditions 5)
Point in time and'in process when homogeneity can be assured (e.g., mixing of solutions)
-6)
Piping to be_sure.that it is not possible to transfer solutions from containers of safe to non-safe geometry
- 7) Safety of solution spillage on floor or inside glove box
- 8) Leakage of solutions into vessel jackets
- 9) Procedures for equipment maintenance
- 10) Administrative controls on making vessel, pipi'ng, or other pro-cess modifications.
6.
Fuel Storace There will be no more than the maximum safe batch size allowable (no more than 45% of critical) in a given storage container.
The maximum safe batch size in a given volume is shown in Figure C-5.
For long fuel elements, the maximum safe batch size can be determined using Figure C-10 or C-11.
The maximum safe batch size determined from Figure C-10 is independent of the con-tainer geometry; that determined from Figure C-ll is dependent on container However, all allowable ba'ch sizes are safe independent of the geometry.
t degree of water moderation and reflection as well as the fuel piece size.
Unless specific studies have been made to evaluate the keff of a given
. container, packaged as described above, it will be assumed that the keH (for interaction purposes) is equal to 0.65, when the maximum mass of fissile ESG-82-33 M
C-20
.,--__---,__.-----.--.---,__,,-----.-_-.-_____.-,____.___._--------,__-.---.__.__.----------___.__m,-,_-
material per container is no more than 45% of critical under optimum condi-tions of water moderation, reflection, and container geometry. However, when the maximum allowable mass per container depends on containar geometry, the assumed keff is increased to 0.80.
This has been suggestedIl for highly enriched uranium solutions and has been found valid for uranium fuel slugs of low enrichment. Based on this consideration, the allowable solid angle for the most central unit in these arrays with a keff =f the solid angle criteria 0.65 and keff = 0.80 will be 2.5 and 1.0 sr, respectively.
In the use o
-the minimum edge-to-edge spacing between adjacent units in the array will be
' two diameters and the U or Pu density will be 510 g/cm3,
- 7.. Low Enrichment Uranium As shown in Figure C-14, uranium at enrichments below 5% is safe against criticality in.the absence of moderation. However, with moderation, two dif-ferent types of. situation must be considered. Aqueous solutions of low-enriched uranium require increasingly greater U-235 mass for criticality as
.the enrichment is decreaged, with the critical mass approaching infinity near 1%. 'An allowance factor for mass cf U-235 ir solution as a function of enrichment for fully reflected systems is shown in Figure C-17.
100; r
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50 I
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\\
U l
$10 s.
[7
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< 5
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1
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i t i; A x 1
b%
1 3. 5 7 10 30 50 70 100 U235 ENRICHMENT (%)
00 25368 Figure C-17 Allowance Factor for Aqueous Homogeneous Solutions of U-235
= - -
ESG-82-33 C-21
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5 U235 ENRICHMENT (%)
' 00-25371 U235 ENRICHMENT (%)
Figure C-20.
Safe Diameters of Figure C-21. Safe Thicknesses of Infinite Cylinders for Water-Infinite Slabs for Lattices of Reflected Uranium Rods in Water-Reflected Uranium Rods in Light Water Light Water D.
APPLICATION OF NUCLEAR SAFETY FORMULATIONS t'
This section provides a demonstration of the formulations given in the nuclear safety criteria. Also, calculational techniques used in establishing the nuclear safety for the fuel element f abrication process are presented.
f 1.
__ Demonstration of Use of Density Analoo Method The array being evaluated is given for the fuel vault storage.
The 5.0-in.-diameter cans are stacked to a maximum height of 21 in. (four cans high). The spacing between shelves is 36 in.
The spatial density factor F, given by F = V /V, = 46.772, c f where Vc = (27)(20)(36) = 19,440 in.3 i
f, (-)(5.02)2(21) = 415.63 in.3 y
(4)
ESG-82-33 C-23
)
1
l 1
The f raction of critical mass of an individual unit, f, is given by l
f = M /Mcb-n
)
e i
C/
The equivalent volume of a unit, used in calculating M, is calculated I
e to be 4.3541 by converting the geometric bucklir@ of a cylinder (5.02-in.
diameter by 21 in. tall) to spherical geometry, using 8 cm as the extrapola-tion length (conservative because each unit is bare). The effective fuel density in a unit is 10 kg + 6.811 1 (volume of a unit of four cans) or 3
1.468 g U-235/cm. The equivalent mass of a sphere, Me, is, therefore, 6.392 kg U-235.
Based on the distribution of fuel in the unit, it is 0.0834 3
that of full-density U-235 metal (17.6 g U-235/cm ).
The minimum cirtical mass of a bare highly enriched U-235 metal sphere is 48.0 kg contained U-235.2 The allowance f actor for highly enriched U compounds (smaller than for U metal or alloys) is 3.0.4 Therefore, the minimum critical mass of a bare sphere of uranium having this density is 144 kg contained U-235 and f =
6 392/144 = 0.044389. The lumping factor, s, is, therefore, 1.91122 and Fs is 1554.9, from which Nb = (1554.9) 0
= 22,390.
With a reflector f actor of 13 and an interspersed moderator f actor of 2.5,9 the minimum critical number of units in the array is 689, from which the maxi-mum safe number is 310 (45% of critical).
2.
Demonstration of Use of Solid Anale Method 4 The formulation used in cale';lating the solid angle between two
[
\\
interacting units is I
(
\\
LD/ h (L/2)2 + h2
\\
/
where D is the projected " diameter" of a container, to its closest edge, and h is the distance f rom the center of the most central container to the pro-jected diameter of the other container.
The length is L.
Figure C-22 is an end view of the container being evaluated.
The fuel length considered in this sample problem is 48 in. The calculated solid angles are shown in Fig-r ure C-22.
3.
Reactivity of a Sinale Water-Reflected Container The following is an example for calculating the keff of a single water-reflected container containing 0.30-in. diameter uranium metal fuel rods having a 5% U-235 enrichment.
It is assumed that the container has water, so 3
that the assembly of rods and water is under conditions of optimum l
ESG-82-33 C-24
r-
/
1
'u~J LDj Oi=
HjY(L/2)2 + H 2 i
S in.
[ 33/2 in.
~
5 1
l 2b 3b h
y Y
D i
keff (unit) = 0.55 x
1 T 3/4 in.
_! 17-M I
2a 3a T
---18 in.+
Q TO Q
(,_)
FROM MOST
\\~d CENTRAL PROJECTED Oi UNIT UNIT DIAMETER h
SOLID ANGLE NO.
(in.)
(in.)
(in.)
(STERADIANS) 1 18 3.5 15.5 0.380 2a 17.75 5.0 16.00 0.520 2b 17.75 5.0 16.00 0.520
)
3a 25.3 6.10 22.25 0.402 3b 25.3 6.10 22.25 0.402 i
4 39.8 6.10 36.75 0.182 5
35 5.0 33.25 0.176 TOTAL SOLID ANGLE FOR RACK 2.581 ALLOWABLE OT = 9 - 10 keff = 3.50 I
12-18-69 UNCL
~ bO-25364A Figure C-22.
End View of Rack
)
ESG-82-33 C-25 i
,/
moderation. A container having a diameter' of 6-5/8 in. and a height of 108 in. is used in the demonstration.
The formulation given in Section 4.2 was used in this evaluation.
2 2 1+MB g k
=
eff 2 2 34g3 G
The constants used were taken from Reference 26.
The sai,rple calculations are repeated as a function of fuel concentration to determine that the maximum reactivy has been calculated.
A fuel concentration of 189 g U-235/1 was used in the calculation, for which keff = 0.90.
M2 = 26.49 cm2 1 = 6.50 cm*
B 2 = 0,019998 cmd g
2 g, = 0.026133 cm-2 B
(O 4.
Metal Picklina b)
In the fabrication of uranium or plutonium fuels, it may be necessary to pickle metal buttons to remqxe excessive oxide coatings or otherwise clean the surface.
It was recognizedao that it is possible to start dissolution in a system that is subcritical at both the starting configuration of metal in water and the fully dissolved configuration of uniform solution, but may achieve a super-critical configuration as metal surrounded by fissile solu-tion. Therefore, it is possible to dissolve some fissile material in the course of the pickling operation. This increase in reactivity must be con-sidered in establishing the allowable batch sizes.
Goertzel analytically predicted 39 that a nonuniform distribution of fuel may result in a minimum critical mass as much as 30% smaller than that with a uniform distribution of fuel. On this basis, allowable batch sizes based on uniform fuel distribution will be reduced by 30% for application to
- The extrapolation length, 1, used must be consistent with that used in developing material buckling data.
O ESG-82-33 C-26
processes involving the dissolution of fuel.
Comparison of the reduced hatch sizes with criticality data reported in Reference 40 for plutonium metal-solutions systems verifies the validity of this approach. Only in the volume region of about 3 1 do the Reardon and Czerniejewski calculations indicate a slightly more conservative ma s.
This method of establishing allowable batch sizes is used for all process steps (e.g., button pickling or bare fuel plate cleaning) where fissile material is simultaneously present in both solid form and solution in a container.
5.
Calculation of Extrapolation Lenoth for Graohite The extrapolation length for graphite was calculated on the basis of the critical masses of infinite water-reflected and infinite graphite-reflected spheres of highly enriched U-235 metal.2 The material buckling for both systems is identical.
The extrapolation length for the water-reflected system was taken as 8.00 cm.41 2
2 2 " (I
/
\\
r B
M Rg+ ig Rg+ig where Ry and RG are the radii of the critical water-reflected and raphite-reflected spheres, 6.76 and 6.09 cm, respectively, and ig (8.00 cm and tg are the infinite water and infinite graphite extrapolation lengths, respectively.
0021Y/sjh C-27
I 1
O APPENDIX D ANNUAL REVIEW OF RADIOLOGICAL CONTROLS O
FOR ',.19 g 6,1987,1983 O
Y APPENDIX D ANNUAL REVIEW OF RADIOLOGICAL CONTROLS FOR-1979, 1980, AND 1981 L-N001T1000285 Annual Review of Radiological Controls--1985 l
N001T1000287 Annual Review of Radiological-Controls--1987 N001T1000301 Annual Review of Radiological Controis--1988 9
9 ESG-82-33 D-1
h O
4 REFERENCES O
I l
1.
F i
i e
O u_
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\\ )-
(
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