DOE G 421.1-1, Criticality Safety Good Practices Program Guide for DOE Nonreactor Nuclear Facilities

NOT MEASUREMENT SENSITIVE DOE G 421 .1-1 DOE GOOD PRACTICES GUIDE CRITICALITY SAFETY GOOD PRACTICES PROGRAM GUIDE FOR DOE NONREACTOR NUCLEAR FACILITIES u.s. Department of Energy Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. This document has been reproduced directly from the best available copy. Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; (615) 576-8401. Available to the public from the U.S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161; (703) 487-4650. Order No. xxxxxxxxxx DOE G 421.1-1 8-25-99 FOREWORD iii This DOE Good Practices Program Guide is approved for use by all Departments and Contractors of the U.S. Department of Energy responsible for the criticality safety of fissionable materials. It is DOE policy to use National and International Consensus Standards (e.g., ANSI/ANS Standards, ISO Standards, ECS Standards) when such standards are available to meet DOE needs. This DOE Good Practices Program Guide is a comprehensive guidance document to assist in developing a criticality safety program to implement the DOE Order (or Rule) on nuclear criticality safety, and the invoked ANSI/ANS standards, through use of good practices. Its comprehensiveness precludes its full applicability to all sets of conditions, since a good practice for one set of conditions may be an unnecessary, or a poor, practice for a similar, but not identical, set of conditions. DOE Good Practices Guides generally should not be used to develop audit check-lists. This DOE Good Practices Guide is not a requirements document and shall not be used as an auditing standard. It is intended only to provide guidance. Requirements for DOE nuclear criticality safety programs are found in higher level documents, e.g., Policy, Rule, Order, and Manual. These documents, e.g., the Order or the Manual, may invoke National and International Consensus Standards. Use of the word "shall" in this DOE Good Practices Guide is only to try to maintain consistency with higher level documents. This DOE Good Practices Guide was developed, over a nine-year period, by Calvin Hopper of the Oak Ridge National Laboratory under contract to the U.S. DOE, with the consensus of a representative work group of DOE Headquarters personnel and DOE site, i.e., laboratory and contractor, personnel. This development included 15 revisions, partly based on a general review and on 6 meetings with the work group. Additional major editorial revision was provided by James Mincey of the Oak Ridge National Laboratory. Because the work performed at the different DOE sites is diverse (viz., hands-on unshielded fissionable material operations at some sites and remote shielded operations at other sites), this DOE Good Practices Guide is diverse. Hence, it is comprehensive and covers most of the areas of responsibility pertaining to conducting a nuclear criticality safety program. To this end, information in this document has been gathered eclectically, therefore it is inappropriate to use this document in its entirety for anyone site or for any single application. Its intent, therefore, is to present a comprehensive text of good practices for nuclear criticality safety, and to depend on good judgment in both engineering and management to be the principal determinant for applicability of these good practices. While even a comprehensive text of good practices cannot address every need, it can serve as a source of ideas to address differing needs as they arise.

Beneficial comments (recommendations, additions, deletions) and any pertinent data that may be of use in improving this document should be addressed to Burton M. Rothleder (project manager for this document). Mr. Rothleder can be reached at 301-903-3726, fax 301-903-6172, or email burton .rothleder@hq.doe.gov. iv ACKNOWLEDGMENTS DOE G 421.1-1 8-25-99 The U.S. Department of Energy wishes to acknowledge the following persons as significant contributors to major parts of the technical content of this document: Calvin Hopper (Oak Ridge National Laboratory) James Mincey (Oak Ridge National Laboratory) Kenneth Yates (Westinghouse Safety Management Solutions) *** Charles Barnett (Lawrence Livermore National Laboratory, retired) Leslie Davenport (Battelle Pacific Northwest Laboratories, retired) Howard Dyer (Oak Ridge National Laboratory) Ivon Fergus (U.S. Department of Energy) Ronald Knief (Ogden Environmental and Energy Services) William R. Waltz (Savannah River Site, retired) The U.S. Department of Energy also wishes to acknowledge the supportive contributions to this document and to the review process from the many experienced members of the nuclear criticality safety community. DOE G 421.1-1 8-25-99 CONTENTS v 1. SCOPE ............................................................ 1 1.1 SCOPE ....................................................... 1 1.2 APPLICATION .................................................. 1 1.3 STRATEGY AND INTENT ........................................... 1 1.4 DOCUMENT REFERENCES. ......................................... 2 1.5 ANSI/ANS Series-8 Standards. ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 2 1.6 MAINTENANCE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 2 2. APPLICABLE DOCUMENTS .............................................. 3 2.1 DOE DOCUMENTS ............................................... 3 2.1.1 DOE 1300.2A ............................................. 3 2.1.2 DOE 1324.2A ............................................. 3 2.1.3 DOE 5000.3B ............................................. 3 2.1.4 DOE 5480.3 .............................................. 3 2.1.5 DOE 5480.4 .............................................. 3 2.1.6 DOE 5480.11 ............................................. 3 2.1.7 DOE 5480.18B ............................................ 3 2.1.8 DOE 5480.19 ............................................. 3 2.1.9 DOE 5480.20A ............................................ 3 2.1.10 DOE 5480.21 ............................................ 3 2.1.11 DOE 5480.22 ............................................. 4 2.1.12 DOE 5480.23 ............................................ 4 2.1.13 DOE 420.1. ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 4 2.1.14 DOE 5484.1 ............................................. 4 2.1.15 DOE 5500.2B ............................................ 4 2.1.16 DOE 5500.3A ............................................ 4 2.1.17 DOE 5700.6C ............................................ 4 2.1 .18 DOE 6430.1 A ............................................ 4 2.1.19 DOE-STD-3007-93 ......................................... 4 2.1.20 DOE-STD-3013-94 ......................................... 4 2.1.21 DOE/TIC-11603-REV. 1 ...................................... 4 2.2 OTHER FEDERAL DOCUMENTS ...................................... 5 2.2.1 Code of Federal Regulations (CFR) ............................... 5

2.2.1.1 Title 10, Part 70 of the CFR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 5 2.2.1.2 Title 10, Part 71 of the CFR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 5 2.2.1.3 Title 10, Part 830 of the CFR .............................. 5 2.2.2 Nuclear Regulatory Commission (NRC) Documents . . . . . . . . . . . . . . . . . . . .. 5 2.2.2.1 NRC Regulatory Guide 3.1 ................................ 5 2.2.2.2 NRC Regulatory Guide 3.4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 5 2.2.2.3 NRC Regulatory Guide 3.34 ... . . . . . . . . . . . . . . . . . . . . . . . . . . .. 5 2.2.2.4 NRC Regulatory Guide 3.35 ............................... 5 2.2.2.5 NRC Regulatory Guide 3.68 ... . . . . . . . . . . . . . . . . . . . . . . . . . . .. 5 2.2.2.6 NRC Regulatory Guide 8.12 ... . . . . . . . . . . . . . . . . . . . . . . . . . . .. 5 2.2.2.7 NUREG/BR-0167 ...................................... 6 2.2.2.8 NUREG/CR-1278, SAND80-0200,RX,AN ...................... 6 vi DOE G 421.1-1 8-25-99 2.2.2.9 NUREG/CR-4639, EEG-2458 .............................. 6 2.3 NON-GOVERNMENT DOCUMENTS .................................... 6 2.3.1 American National Standards Institute (ANSI) . . . . . . . . . . . . . . . . . . . . . . .. 6 2.3.1.1 ANSI/ANS-8.1-1998 ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 6 2.3.1.2 ANSI/ANS-8.3-1997 ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 6 2.3.1.3 ANSI/ANS-8.5-1996 ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 6 2.3.1.4 ANSI/ANS-8.6-1983,R95 ................................ 6 2.3.1.5 ANSI/ANS-8.7-1998 .................................... 6 2.3.1.6 ANSI/ANS-8.9-1987,R95 ................................ 6 2.3.1.7 ANSI/ANS-8.10-1983,R88 ............................... 6 2.3.1.8 ANSI/ANS-8.12-1987,R93 ............................... 7 2.3.1.9 ANSI/ANS-8.15-1981,R95 ............................... 7 2.3.1.10 ANSI/ANS-8.17-1984,R97 .............................. 7 2.3.1.11 ANSI/ANS-8.19-1996 .................................. 7 2.3.1.12 ANSI/ANS-8.20-1991 .................................. 7 2.3.1.13 ANSI/ANS-8.21-1995 .................................. 7 2.3.1.14 ANSI/ANS-8.22-1997 .................................. 7 2.3.1.15 ANSI/ANS-8.23-1997 .................................. 7 2.3.1.16 ANSI/ANS-10.3-1986 .................................. 7 2.3.1.17 ANSI/ANS-10.4-1987 .................................. 7 2.3.1.18 ANSI/IEEE-Std-500-1984 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 7 2.3.2 Industry Related Reference Documents ............................ 7 2.3.2.1 ANS-9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 8 2.3.2.2 LA-11627-MS ........................................ 8 2.3.3 Journal Articles, and Meetings and Conference Proceedings . . . . . . . . . . . . .. 9 3. TERMS AND DEFINITIONS ............................................. 10 4. GENERAL GUIDANCE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 25 4.1 APPLICABILITY ................................................ 25 4.2 INTERPRETATION OF THE GOOD PRACTICES PROGRAM GUIDE .............. 25 4.3 MAINTENANCE OF THE GOOD PRACTICES PROGRAM GUIDE ... . . . . . . . . . . . .. 25 4.4 DOCUMENT ARCHITECTURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 25

4.4.1 Topical Structure .......................................... 25 4.4.2 Content Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 25 4.4.3 Section and Paragraph Numbering .............................. 27 5. DETAILED GUIDANCE ................................................ 28 5.1 ADMINISTRATION .............................................. 28 5.1.1 Contractor President/Chief Executive Officer ....................... 28 5.1.1.1 Responsibility........................................ 28 5.1.1.2 Policy ............................................. 28 5.1.1.3 Organization ........................................ 28 5.1.1.4 Program Oversight .................................... 28 5.1.1.5 Corrective actions. .................................... 29 5.1.1.6 Resources .......................................... 29 5.1.1.7 Stop Work Policy ..................................... 29 5.1.2 Facility Operations Managers .................................. 29 5.1.2.1 CAS Management .................................... 29 5.1.2.2 Procedures development and maintenance . . . . . . . . . . . . . . . . . . .. 29 DOE G 421.1-1 8-25-99 vii 5.1.2.3 Staff training ........................................ 30 5.1.2.4 Design and procedure reviews ............................ 30 5.1.2.5 Configuration control program ............................ 30 5.1.2.6 Self-assessments ..................................... 30 5.1.2.7 Compliance ......................................... 30 5.1.2.8 Audit response approval ................................ 30 5.1.2.9 Safety documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 30 5.1.2.10 Contingency analysis documentation . . . . . . . . . . . . . . . . . . . . . .. 30 5.1.2.11 Facility shutdowns ................................... 30 5.1.2.12 Maintenance of NCS controls ....................... ; . . .. 30 5.1.2.13 Fire safety plans . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 30 5.1.2.14 Operational postings .................................. 31 5.1.2.15 Delegation of responsibilities ............................ 31 5.1.2.16 Development of criticality accident evacuation routes ........... 31 5.1.2.17 Monitoring for process accumulations ...................... 31 5.1.2.18 Facility access and other NCS controls ..................... 31 5.1.3 Line/Production Management .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 31 5.1.3.1 Acceptance of authority and responsibility . . . . . . . . . . . . . . . . . . .. 31 5.1.3.2 Standards compliance .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 31 5.1.3.3 Operational approvals .................................. 32 5.1.3.4 Staffing and training ................................... 32 5.1.3.5 Configuration control. .................................. 32 5.1.3.6 Procedures ......................................... 32 5.1.3.7 Maintenance ........................................ 32 5.1.3.8 Self-audit.... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 32 5.1.3.9 Emergency planning ................................... 32 5.1.3.10 Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 32 5.1.3.11 Notifications ....................................... 32 5.1.3.12 Review requests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 32 5.1.3.13 Delegation of authority and assignment of responsibilities ........ 33

5.1.4 First Line Supervision ....................................... 33 5.1.4.1 Responsibility........................................ 33 5.1.4.2 Training. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 33 5.1.4.3 Provision of training ................................... 33 5.1.4.4 Procedural development ................................ 33 5.1.4.5 Safety practices .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 33 5.1.4.6 Operational reviews ................................... 33 5.1.4.7 Operational approvals .................................. 34 5.1.4.8 Process monitoring .................................... 34 5.1.4.9 Recovery and deviation evaluations. . . . . . . . . . . . . . . . . . . . . . . .. 34 5.1.4.10 Labeling and posting .................................. 34 5.1.4.11 Access control .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 34 5.1.5 Fissionable Material Operations Personnel ......................... 34 5.1.5.1 Responsibility........................................ 34 5.1.5.2 Operational procedures ................................. 34 5.1.5.3 Terminate operations .................................. 34 5.1.5.4 Inquiries ........................................... 34 5.1.5.5 Training requirements .................................. 35 5.1.5.6 Notification ......................................... 35 5.1.5.7 Emergency response .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 35 viii DOEG421.1-1 8-25-99 5.1.6 Facilities Maintenance Organization. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 35 5.1.7 Engineering and Projects Organization (E&PO) ...................... 35 5.1 .7.1 Responsibilities....................................... 35 5.1.7.2 Design requirement compliance ........................... 35 5.1.8 Criticality Safety Organization ................................. 35 5.1.8.1 Technical Direction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 35 5.1.8.2 Quality assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 36 5.1.8.3 Criticality Safety Organization personnel qualification ............ 36 5.1.8.4 Maintenance of familiarity ............................... 37 5.1.8.5 Consultation ........................................ 37 5.1.8.6 Obtaining consultation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 38 5.1.8.7 Performing self-assessments and audits . . . . . . . . . . . . . . . . . . . . .. 38 5.1.8.8 Operational reviews ................................... 38 5.1.8.9 Procedural reviews .................................... 38 5.1.8.10 Incident reviews . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 38 5.1.8.11 Process event/fault tree analyses ......................... 38 5.1.8.12 Quality and configuration control of software and data sets ....... 38 5.1.8.13 NCS evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 38 5.1.8.14 NCS analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 38 5.1.8.15 NCS recovery actions ................................. 39 5.1.8.16 CAS and CDS sensors, Nuclear Accident Dosimeter (NAD), and

evacuation zone boundary shielding and deployment evaluation ........ 39 5.1.8.17 Peer review ........................................ 39 5.1.8.18 Records retention .................................... 39 5.1.8.19 NCS procedures ..................................... 40 5.1.8.20 Design reviews. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 40 5.1.8.21 Selection and approval of effective controls. . . . . . . . . . . . . . . .. 40 5.1.8.22 Assistance in shutdowns ... . . . . . . . . . . . . . . . . . . . . . . . . . . .. 40 5.1.8.23 Accident yield estimation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 40 5.1.8.24 Participate in CAS evacuation drills . . . . . . . . . . . . . . . . . . . . . . .. 40 5.1.8.25 Technical training support .............................. 40 5.1.8.26 Reviews of fire safety plans . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 40 5.1.8.27 Operational experience feedback. . . . . . . . . . . . . . . . . . . . . . . . .. 40 5.1.8.28 INCSRC support ..................................... 40 5.1.9 Installation Nuclear Criticality Safety Review Committee (INCSRC) ........ 40 5.1.9.1 Composition ........................................ 41 5.1.9.2 Guidance to management ............................... 41 5.1.9.3 Investigation of incidents. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 41 5.1.9.4 INCSRC Annual reviews ................................ 41 5.1.9.5 Response to requests .................................. 42 5.2 PERSONNEL SELECTION, QUALIFICATION, TRAINING, AND STAFFING PROGRAM 43 5.2.1 Program for Operations and Support Personnel . . . . . . . . . . . . . . . . . . . . .. 43 5.2.2 Installation Nuclear Criticality Safety Staff . . . . . . . . . . . . . . . . . . . . . . . .. 43 5.2.3 Visitors and Clerical Employees ................................ 44· 5.2.4.1 Compliance ......................................... 45 5.2.4.2 Quality Audit ........................................ 45 5.3 OPERATING, STORING, AND TRANSFERRING - PLANS, PROCEDURES, REQUIREMENTS, AND CONTROLS ............................................... 46 5.3.1 General Requirements for Operating Plans and Procedures .............. 46 5.3.1.1 Start-up, operations, and modifications ...................... 46 DOE G 421.1-1 8-25-99 ix 5.3.1.2 NCS parameter identification ............................. 46 5.3.1.3 Single failure safety assurance ............................ 46 5.3.1.4 Procedural convenience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 46 5.3.1.5 Procedural reviews .................................... 46 5.3.1.6 Supplementation ..................................... 47 5.3.1.7 Operational deviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 47 5.3.2 Processing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 47 5.3.2.1 Plans, flowsheets, and layouts . . . . . . . . . . . . . . . . . . . . . . . . . . .. 47 5.3.2.2 Procedural description of material composition ................. 47 5.3.2.3 Procedural description of allowed material quantities . . . . . . . . . . . .. 47 5.3.2.4 Procedural description of spacing requirements . . . . . . . . . . . . . . . .. 48 5.3.2.5 Procedural specifications for material collection and transport. . . . . .. 48 5.3.2.6 Procedural specifications for administrative controls ............. 48

5.3.3 Receiving and Inspecting Fissionable Material . . . . . . . . . . . . . . . . . . . . . .. 48 5.3.3.1 Verification ......................................... 48 5.3.3.2 Material placement .................................... 48 5.3.4 Storing Fissionable Material ................................... 48 5.3.4.1 Container design ..................................... 48 5.3.4.2 Container criteria ..................................... 48 5.3.4.3 Container descriptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 49 5.3.4.4 Container identification and closure. . . . . . . . . . . . . . . . . . . . . . . .. 49 5.3.4.5 Container venting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 49 5.3.4.6 Containerization of plutonium or 233U • • • • • • • • • • • • • • • • • • • • • • •• 49 5.3.4.7 Plutonium storage monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 49 5.3.4.8 Facility design criteria .................................. 49 5.3.4.10 Storage facility plans and layouts ......................... 50 5.3.4.11 Admonitions about moderating and reflecting materials .......... 50 5.3.4.12 Removal and return of materials .......................... 50 5.3.4.13 Exclusion of superfluous materials ........................ 50 5.3.4.14 Readiness inspections ................................. 50 5.3.4.15 Postings .......................................... 50 5.3.4.16 Instructions ........................................ 50 5.3.4.17 Emergency planning .................................. 51 5.3.4.18 Exclusion from storage requirements .. . . . . . . . . . . . . . . . . . . . .. 51 5.3.4.19 Use of shipping containers .............................. 51 5.3.4.20 Material constraints .................................. 51 5.3.4.21 Pyrophoric materials .................................. 51 5.3.4.22 Heat removal ....................................... 51 5.3.5 Fissionable Material Transportation .............................. 51 5.3.5.1 Onsite transfers ...................................... 51 5.3.5.1.1 Onsite transport safety analysis ...................... 51 5.3.5.1.2 Operating procedures ............................. 51 5.3.5.2 Offsite transfers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 52 5.3.6 Posting and Labeling ........................................ 52 5.3.6.1 Posting of Fissionable Material Handling, Storage, and Work Areas .,. 52 5.3.6.1.1 Postings for presence of fissionable materials . . . . . . . . . . . .. 52 5.3.6.1.2 Symbol ....................................... 52 5.3.6.1.3 Storage postings ................................ 52 5.3.6.1.4 Process limits. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 53 5.3.6.1.5 Uniformity of postings . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 53 x DOE G 421.1-1 8-25-99 5.3.6.1.6 Administrative control postings ...................... 53 5.3.6.1.7 Precautions or prohibitions . . . . . . . . . . . . . . . . . . . . . . . . .. 53 5.3.6.1.8 Visibility of postings, operator aids .................... 53 5.3.6.2 Labeling Requirements for Fissionable Material ................. 53 5.3.6.2.1 Label......................................... 53 5.3.6.2.2 Attachment .................................... 53 5.3.6.2.3 Contained information . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 53 5.3.6.2.4 Specialized needs .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 53 5.3.6.2.5 Unirradiated reactor fuel/targets ...................... 54 5.3.6.2.6 Irradiated reactor fuel/targets .,. . . . . . . . . . . . . . . . . . . . .. 54

5.3.6.3 Empty containers ..................................... 54 5.4 CRITICALITY ACCIDENT ALARM AND DETECTION SYSTEMS ................ 55 5.4.1 Conditions for CAS Coverage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 55 5.4.2 Conditions for CDS Coverage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 56 5.4.3 Conditions Not Requiring CASs and CDSs ......................... 56 5.4.3.1 Shielded Operations ................................... 56 5.4.3.2 Licensed/certificated packages ..... . . . . . . . . . . . . . . . . . . . . . .. 56 5.4.3.3 Incredibility ......................................... 57 5.4.4 Design Requirements ....................................... 57 5.4.4.1 Characteristic radiation detection .......................... 57 5.4.4.2 Alarm logic ......................................... 57 5.4.4.3 Trouble warning ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 57 5.4.4.4 Alarm purpose ....................................... 57 5.4.4.5 Resistance to detector saturation .......................... 57 5.4.4.6 Alarm resets ........................................ 58 5.4.4.7 Automated alarming ................................... 58 5.4.4.8 Response testing ..................................... 58 5.4.4.9 Minimization of false alarms and system vulnerability ............ 58 5.4.4.10 Backup power supply ................................. 58 5.4.4.11 Seismic resistance ................................... 58 5.4.4.12 Response time ...................................... 58 5.4.4.13 Detection criteria .................................... 58 5.4.4.14 Sensitivity ......................................... 59 5.4.4.15 Spacing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 59 5.4.4.16 Signal ............................................ 59 5.4.4.17 Reliability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 59 5.4.5 Testing ................................................. 60 5.4.5.1 Initial ............................................. 60 5.4.5.2 Post repair . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 60 5.4.5.3 Radiation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 60 5.4.5.4 Periodic. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 60 5.4.5.5 Corrective Action ..................................... 60 5.4.5.6 Procedures ......................................... 60 5.4.5.7 Records... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 60 5.4.6 Location Analysis .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 61 5.4.6.1 Shielding and location analysis ... . . . . . . . . . . . . . . . . . . . . . . . .. 61 5.4.6.2 Yield estimation ...................................... 61 5.4.7 Familiarization with Operation ................................. 61 5.4.7.1 CAS alarm response ................................... 61 5.4.7.2 Emergency procedures ................................. 61 DOE G 421.1-1 8-25-99 xi 5.4.7.3 Signal familiarization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 61 5.4.7.4 Signal demonstration .................................. 61 5.4.7.5 Periodic alarm signaling .... . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 61 5.4.7.6 Annual evacuation drills ................................. 62 5.4.7.7 Visitor training ....................................... 62

5.5 EMERGENCY PREPAREDNESS ...................................... 63 5.6 NUCLEAR CRITICALITY SAFETY CONTROL PRINCIPLES AND METHODS ........ 64 5.6.1 Double-Contingency Principle (Application) . . . . . . . . . . . . . . . . . . . . . . . .. 65 5.6.1.1 Requirements of contingencies . . . . . . . . . . . . . . . . . . . . . . . . . . .. 66 5.6.1.2 Scope of contingencies .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 69 5.6.1.3 Double-contingency analysis ............................. 69 5.6.1.4 Contingency controls reliability . . . . . . . . . . . . . . . . . . . . . . . . . . .. 70 5.6.1.5 Contingency control margin of subcriticality . . . . . . . . . . . . . . . . . .. 70 5.6.1.6 Treatment of dependent contingencies ...................... 70 5.6.1.7 Identification of engineered or administrative controls ............ 70 5.6.1.8 Exception from Double-Contingency Principle .. . . . . . . . . . . . . . . .. 70 5.6.1.9 Preferred hierarchy of controls ............................ 70 5.6.1.10 Avoidance of administrative controls . . . . . . . . . . . . . . . . . . . . . .. 70 5.6.1.11 Exemption from double-contingency principle ................. 71 5.6.2 General Nuclear Criticality Safety Control Principles and Practices . . . . . . . .. 71 5.6.2.1 Safety assurance ..................................... 71 5.6.2.2 Potential criticality assessment. . . . . . . . . . . . . . . . . . . . . . . . . . .. 71 5.6.2.3 Burn-up credit ....................................... 71 5.6.2.4 Storage... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 71 5.6.2.5 Special actinide element evaluations ........................ 71 5.6.2.6 Subcritical neutron multiplication measurements . . . . . . . . . . . . . . .. 71 5.6.2.7 Criticality accident alarm systems . . . . . . . . . . . . . . . . . . . . . . . . .. 71 5.6.2.8 Process and equipment design ............................ 71 5.6.2.9 Process analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 71 5.6.2.1 0 Bases for su bcritical ity ................................ 71 5.6.2.11 Operating procedures ................................. 72 5.6.2.12 Exempt quantities of fissionable materials ................... 72 5.6.3 Nuclear Parameters Important to Nuclear Criticality Safety and Their Control . 72 5.6.3.1 Geometry control ..................................... 72 5.6.3.1.1 Equipment design reliance .......................... 72 5.6.3.1.2 Fissile solution transfers ........................... 72 5.6.3.1.3 Allowances with geometry control .................... 72 5.6.3.1.4 Control monitoring ............................... 72 5.6.3.1.5 Thermal insulation concerns . . . . . . . . . . . . . . . . . . . . . . . .. 72 5.6.3.1.6 Sump designs .................................. 73 5.6.3.1.7 Floor drains .................................... 73 5.6.3.1.8 Inadvertent transfers . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 73 5.6.3.1.9 Backflow prevention .............................. 73 5.6.3.2 Spacing (interaction) control ............................. 73 5.6.3.2.1 Storage and transfer .... . . . . . . . . . . . . . . . . . . . . . . . . .. 73 5.6.3.2.2 Storage rack integrity ............................. 73 5.6.3.3 Neutron absorber (poison) control . . . . . . . . . . . . . . . . . . . . . . . . .. 73 5.6.3.3.1 Suitability ..................................... 73 5.6.3.3.2 Raschig rings ........................ . . . . . . . . . .. 73 5.6.3.3.3 Representative samples . . . . . . . . . . . . . . . . . . . . . . . . . . .. 74

xii DOE G 421.1-1 8-25-99 5.6.3.3.4 Minimum soluble poison concentration ................. 74 5.6.3.3.5 Soluble poison monitoring .......................... 74 5.6.3.4 Concentration (density) control ........................ 74 5.6.3.4.1 Process changes in density ......................... 74 5.6.3.5 Moderation control .................................... 74 5.6.3.5.1 Monitoring neutron moderation . . . . . . . . . . . . . . . . . . . . . .. 74 5.6.3.5.2 Consideration of interstitial moderation ................. 75 5.6.3.5.3 Consideration of non-aqueous moderation ............... 75 5.6.3.5.4 Installed fire protection systems ...................... 75 5.6.3.5.5 Exclusion of moderating materials . . . . . . . . . . . . . . . . . . . .. 75 5.6.3.5.6 Use of water in fire fighting ......................... 75 5.6.3.6 Reflection control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 75 5.6.3.6.1 Assumptions about neutron reflection .................. 75 5.6.3.6.2 Avoidance of reflection control . . . . . . . . . . . . . . . . . . . . . .. 75 5.6.3.6.3 Use of water in fire fighting ......................... 75 5.6.3.7 Mass control ........................................ 75 5.6.3.7.1 Over-batching .................................. 75 5.6.3.7.2 Double batching . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 76 5.6.3.7.3 Material form ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 76 5.6.3.8 Volume control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 76 5.6.3.8.1 Volumetric limits ................................ 76 5.6.3.9 Enrichment or isotopic control ............................ 76 5.7 NUCLEAR CRITICALITY SAFETY DESIGN AND ANALYSIS GUIDELINES . . . . . . . . .. 77 5.7.1 Scope .................................................. 77 5.7.2 Overview of the Nuclear Criticality Safety Control Design Process . . . . . . . .. 77 5.7.2.1 Majorprojects ....................................... 77 5.7.2.1.1 Preliminary process hazards review .. . . . . . . . . . . . . . . . . .. 77 5.7.2.1.2 Design process and design reviews .................... 77 5.7.2.1.3 Preoperational process hazards review. . . . . . . . . . . . . . . . .. 78 5.7.2.2 Projects involving modifications to existing equipment ............ 78 5.7.2.2.1 Screening process hazards review. . . . . . . . . . . . . . . . . . . .. 78 5.7.2.2.2 Preliminary process hazards review .................... 80 5.7.2.2.3 Preoperational process hazards review. . . . . . . . . . . . . . . . .. 80 5.7.3 Six Basic Nuclear Criticality Safety Control Design Objectives. . . . . . . . . . .. 80 5.7.3.1 Objective 1 - to control criticality probability using a preferred hierarchy of controls .............................................. 80 5.7.3.2 Objective 2 - to identify potential criticality scenarios ............ 80 5.7.3.3 Objective 3 - to eliminate potential criticality scenarios to the extent practical .............................................. 80 5.7.3.4 Objective 4 - to demonstrate that criticality risks are acceptably low .. 80 5.7.3.5 Objective 5 - to evaluate the operability of criticality safety controls .. 81 5.7.3.6 Objective 6 - to document the nuclear criticality safety control design . 81 5.7.4 Means of Controlling the Criticality Risks . . . . . . . . . . . . . . . . . . . . . . . . .. 81 5.7.4.1 Three basic means of criticality safety control ................. 82

5.7.4.1.1 Passive-engineered control . . . . . . . . . . . . . . . . . . . . . . . . .. 83 5.7.4.1.2 Active-engineered control .......................... 83 5.7.4.1.3 Administrative control . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 83 5.7.4.2 Criticality safety control methods .......................... 83 5.7.4.2.1 Geometry control ................................ 83 DOE G 421.1-1 8-25-99 xiii 5.7.4.2.2 Spacing control ................................. 87 5.7.4.2.3 Neutron poisons ................................. 88 5.7.4.2.4 Fissile concentration control . . . . . . . . . . . . . . . . . . . . . . . .. 90 5.7.4.2.5 Moderation control ............................... 90 5.7.4.2.6 Reflection control ................................ 91 5.7.4.2.7 Mass control ................................... 92 5.7.4.2.8 Enrichment or isotopic composition control .............. 92 5.7.4.2.9 Density control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 92 5.7.5 Identifying Potential Criticality Scenarios .......................... 93 5.7.5.1 Four measures contributing to successful identification ........... 93 5.7.5.2 Approaches for criticality scenario identification ................ 94 5.7.5.2.1 Using logic models to identify accident scenarios .......... 94 5.7.5.2.2 Postulating accident scenarios directly. ................. 95 5.7.5.2.3 Deductive logic tree example ........................ 96 5.7.6 Eliminating Potential Criticality Scenarios . . . . . . . . . . . . . . . . . . . . . . . . .. 98 5.7.7 Judging Acceptability of a Potential Criticality Scenario ............... 101 5.7.7.1 Double-contingency analysis meaning and application ........... 101 5.7.7.2 Basic steps in implementing double-contingency and performing a contingency analysis .................................... 102 5.7.7.3 Qualifications for a contingency barrier ..................... 103 5.7.7.3.1 Quantitative guidelines for acceptable contingency barrier failure probabilities ....................................... 103 5.7.7.3.2 Guidelines for independency . . . . . . . . . . . . . . . . . . . . . . .. 105 5.7.7.3.3 Quantification of the simultaneous collapse frequency of two controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 105 5.7.7.4 Conspicuous and prominent identification of double-contingency means of control .................................... . . . . . . . . .. 105 5.7.8 Operability of Criticality Safety Controls ......................... 105 5.7.8.1 Identifying controls important to nuclear criticality safety . . . . . . . .. 105 5.7.8.2 Examining the operability of the set of controls ............... 106 5.7.8.3 Incorporating good human factors practices . . . . . . . . . . . . . . . . .. 106 5.7.8.4 Incorporating uniformity into the design . . . . . . . . . . . . . . . . . . . .. 106 5.7.8.5 Facilitating sampling .................................. 106 5.7.8.6 Facilitating inspection and maintenance . . . . . . . . . . . . . . . . . . . .. 106 5.7.8.7 Facilitating flushing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 107 5.7.8.8 Anticipating process changes . . . . . . . . . . . . . . . . . . . . . . . . . . .. 107 5.7.8.9 Accommodating fire control systems. . . . . . . . . . . . . . . . . . . . . .. 107 5.7.9 Documenting the Nuclear Criticality Safety Control Design . . . . . . . . . . . .. 108 5.7.9.1 Documentation of the nuclear criticality safety control design proposal

5.7.9.2 Documentation of the nuclear criticality safety control design ..... . 5.7.10 Examples ............................................. . 5.8 SOFTWARE QUALITY ASSURANCE AND VALIDATION ................... . 5.8.1 Software Requirement ..................................... . 5.8.2 Verification of Calculational Method ........................... . 5.8.3 Software Configuration Control .............................. . 5.8.4 Validation of Calculational Method ............................ . 5.8.5 Code user corroboration ................................... . 5.9 NUCLEAR CRITICALITY SAFETY EVALUATION (NCSE) GUIDELINES .......... . 5.9.1 Personnel Requirements for Performing NCSEs .................... . 110 110 111 112 112 112 112 112 113 114 114 xiv DOE G 421.1-1 8-25-99 5.9.2 Performance and Documentation of the NCSE ..................... 114 5.9.2.1 Peer review ........................................ 116 5.9.2.2 NCSE documentation ................................. 117 APPENDIX A. PERSONNEL SELECTION, QUALIFICATION, TRAINING, AND STAFFING PROGRAM .............................................. 118 A.1 Program for Operations and Support Personnel . . . . . . . . . . . . . . . . . . . . . . . . .. 118 A.1.1 Continuing proficiency of personnel ............................ 118 A.1.2 Nuclear fission chain reactions and accident consequences. . . . . . . . . . . .. 118 A.1.3 Neutron behavior in fissioning systems. . . . . . . . . . . . . . . . . . . . . . . . . .. 119 A.1.4 Criticality accident history. .................................. 11 9 A.1.5 Response to criticality accident alarm signals. ..................... 119 A.1.6 Nuclear criticality safety parameters. ........................... 119 A.1 .7 Policy and procedures. ..................................... 119 A.1.8 Evaluations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 119 A.2 Installation Nuclear Criticality Safety Staff. ............................ 119 A.2.1 Qualification. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 120 A.2.2 Functional specialties. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 121 A.2.2.1 Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 122 A.2.2.2 Evaluation. ........................................ 123 A.2.2.3 Implementation. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 123 A.2.2.4 Confirmation ....................................... 123 A.2.2.5 Continuing competence. ............................... 123 A.2.3 Qualification process. ...................................... 123 A.2.3.1 Principles. ......................................... 124 A.2.3.2 Classification levels. .................................. 125 A.2.3.2.1 Entry. ....................................... 125 A.2.3.2.2 Apprentice. ................................... 126 A.2.3.2.3 Specialist. .................................... 126 A.2.3.2.4 Senior. ...................................... 127 A.2.3.2.5 Lead. ....................................... 127 A.2.3.3 Functional specialization. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 127 A.2.3.4 Continuing competence. ............................... 128 A.2.3.5 Requirements. ...................................... 129

A.2.3.5.1 Education. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 130 A.2.3.5.2 Experience. ................................... 131 A.2.3.5.3 Training. ..................................... 131 A.2.3.5.4 On-the-job training. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 132 A.2.3.5.5 Professional development. .... . . . . . . . . . . . . . . . . . . . .. 133 A.2.3.5.6 Personal characteristics. .......................... 134 A.2.3.6 Overall qualification. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 134 A.2.4 Classification-specific qualification programs. . . . . . . . . . . . . . . . . . . . . .. 135 A.2.5 Functional-specific qualification programs. . . . . . . . . . . . . . . . . . . . . . . .. 135 A.2.6 Continuing competence. .................................... 135 A.2.7 Documentation and records. ................................. 136 A.2.8 Evaluation and Documentation. ............................... 137 APPENDIX B. GRADED APPROACH ........................................ 138 B.1 Graded Approach to Criticality Safety Analyses and NCSEs. ................ 138 B.1 .1 Levels of analyses and evaluations. ............................ 138 DOE G 421.1-1 8-25-99 xv B.l .1.1 Level A. .......................................... 138 B.l .1.2 Level B. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 139 B.l .1.3 Level C. .......................................... 139 B.l.2 Complexities of facility fissionable material operations. ............... 139 B.l.2.1 Class I. ........................................... 139 B.l.2.2 Class II. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 139 B.l.2.3 Class III. .......................................... 140 B.l.2.4 Class IV ........................................... 140 B.l.3 Analysis Content. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 140 B.l.3.1 Operational description. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 140 B.l.3.2 Fissionable material forms. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 140 B.l.3.3 Credible operating condition changes. ...................... 140 B.l.3.4 Analysis of accident scenarios. . . . . . . . . . . . . . . . . . . . . . . . . . .. 140 B.l.3.5 Need for CAS or CDS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 141 B.l.3.6 Safety controls description. ............................. 141 B.l.3.7 NCSE summary. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 141 B.l.4 Performance of nuclear criticality safety analyses and NCSEs. .......... 141 B.l .5 Results of the Graded Approach. .............................. 141 APPENDIX C. ESTIMATING THE WAITING TIME UNTIL THE SIMULTANEOUS COLLAPSE OF TWO CONTINGENCIES ................................................. 143 C.l Introduction.................................................. 143 C.l.1 DCP Review ............................................ 143 C.l.2 Markov Model ........................................... 143 C.l.3 Probabilistic Description .................................... 143 C.2 General Markov Model .......................................... 143 C.3 Symmetric Case . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 146 APPENDIX D. EXAMPLES OF DESIGN OF NUCLEAR CRITICALITY SAFETY CONTROLS .... 148 D.l Double-contingency analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 148

D.l.l Example # 1 ............................................ 148 D.l .1.1 Identifying potential criticality scenarios - logic diagram . . . . . . . . .. 148 D.l.l.2 Evaluation against the Double-Contingency Principle .. . . . . . . . . .. 149 D.l.l.2.1 Identifying the two barriers for double-contingency ........ 149 D.l.l.2.2 Qualification of the barriers for double-contingency . . . . . . .. 149 D.l.l.3 Identifying the means of control for each contingency barrier . . . . .. 154 D. 1. 1 .4 Review, relative to the other nuclear criticality safety objectives . . .. 154 D.2 Examples of eliminating unnecessary criticality scenarios ............ : . . . . .. 156 D.2.1 Example # 1 ............................................ 156 D.2.2 Example # 2 ............................................ 156 D.2.3 Example # 3 ............................................ 156 D.3 Examples of passive-engineered features and devices ..................... 157 D.3.1 Air break. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 157 D.3.2 Barometric seal leg . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 157 D.3.3 Criticality drain. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 157 D.3.4 Nuclear safety blank . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 159 D.3.5 Large line sizes .......................................... 159 D.3.6 Restricting orifices ........................................ 159 D.3.7 Relative elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 159 D.4 Examples of active protective devices . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 160 xvi DOE G 421.1-1 8-25-99 0.4.1 Rupture disk ............................................ 160 0.4.2 Backflow prevention devices ................................. 160 0.4.3 Radiation monitoring systems ................................ 160 APPENDIX E. SOFTWARE CONFIGURATION CONTROL PROCEDURE ................. 161 E.1 Specific responsibilities .......................................... 161 E.1 .1 Contractor safety organization manager . . . . . . . . . . . . . . . . . . . . . . . . .. 161 E.1.2 Software System Team . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 162 E.1.3 Functional System Manager . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 162 E.1.4 System Administrator ...................................... 163 E.1.5 Installation nuclear criticality safety organization . . . . . . . . . . . . . . . . . . .. 163 E.1.6 Software developer. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 164 E.2 Software identification .......................................... 164 E.3 Software control. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 165 E.4 Software change procedure .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 165 E.5 Nonconformance Report procedure .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 166 E.6 Software testing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 166

APPENDIX F. EXAMPLE COMPUTATIONAL TECHNIQUE VALIDATIONS ............... 175 F.1 Selection and description of critical experiments . . . . . . . . . . . . . . . . . . . . . . . .. 175 F.2 Selection and description of the computational method .................... 175 F.3 Description of similarities and differences. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 177 F.4 Input variables ................................................ 178 F.5 Acceptance criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 178 F.6 Areas of applicability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 179 F.7 Example validation ............................................. 180 APPENDIX G. BIBLIOGRAPHY OF JOURNAL ARTICLES AND MEETING AND CONFERENCE PROCEEDINGS ................................................... 195 INDEX ............................................................. 210 DOE G 421.1-1 8-25-99 AEG ANS ANSI CAS CDS CFR DBA DOE DOT E&PO FEM FMCA FMEA HEU INCSRC JTA LCO LEU LTB NAD NCS NCSE NCSS NCSSST NRC OJT PAG PRA PVC SME SNR USL V&V Acronyms Average Energy Group American Nuclear Society American National Standards Institute Criticality Accident Alarm System Criticality Accident Detection System Code of Federal Regulations Design Basis Accident US Department of Energy US Department of Transportation Engineering and Projects Organization Fissionable Equivalent Mass Fissionable Material Control Area Failure Modes and Effects Analysis High Enriched Uranium Installation Nuclear Criticality Safety Review Committee Job/Task Analysis Limiting Condition for Operation Low Enriched Uranium Lower Tolerance Band Nuclear Accident Dosimeters Nuclear Criticality Safety Nuclear Criticality Safety Evaluation Nuclear Criticality Safety Specialist Nuclear Criticality Safety Software System Team US Nuclear Regulatory Commission On-the-Job Training Protective Action Guide Probabilistic Risk Analysis PolyVinyl Chloride Su~ectMatterExpert Software Nonconformance Report Upper Subcritical Limit Verification and Validation xvii xviii List of Figures OOE G 421.1-1 8-25-99 Figure 5.3.6.1.2-1. Fissionable material symbol ................................. 52 Figure 5.7.2.1.2-1. Illustration of iterative process for development of nuclear criticality safety controls. ........................................................ 79 Figure 5.7.5.2.3-1. Illustration of deductive logic tree ............................. 97 Figure 0.1.1. Schematic of dry location and scrubber system ...................... 150 Figure 0.1.2. Logic diagram for potential criticality via liquid addition to dry location ...... 151 Figure 0.1.3. Logic diagram for potential criticality via liquid addition to dry location - two barriers added ......................................................... 152 Figure 0.1.4. Contingency analysis diagram .................................. 153 Figure 0.3.1. Schematic of air break. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 158 Figure 0.3.2. Illustration of barometric seal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 158 Figure 0.3.3. Criticality Orain. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 159 Figure F. 7.1. Typical results for the single-sided, uniform-width, closed-interval,

LTB technique. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 191 Figure F.7.2. Example results for the single-sided, uniform-width, closed-interval, L TB technique . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 194 DOE G 421.1-1 8-25-99 List of Tables xix Table 5.7.4-1. Criticality Safety Control Methods and Typically Associated Means of Control . 82 Table 5.7.5.2.3-1. Examples of Phenomena and Initiating Events Leading to the Mechanism for a Potential Criticality Event ..................................... . . . . . . .. 99 Table 5.7.7.3.1-1. Guidelines for Performing Quantitative Risk Analysis of Double-Contingency Control Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 104 Table 5.7.9-1. Summary of Nuclear Criticality Safety Control Design Documentation Objectives 109 Table A.2.3.3-1. Matrix of Qualification Levels for Each of Four Primary Functional Specialties 128 Table A.2.3.5-1. Education and Experience Requirements from DOE Order 5480.20 for Reactor and non-Reactor Facility Personnel Compared to Those for NCSS Personnel ......... 130 Table B.l .4-1. Resources required for performance of the NCSE .................... 142 Table D.l.l. Contingency Analysis - Summary Sheet . . . . . . . . . . . . . . . . . . . . . . . . . . .. 155 Table F.2. Partial Listing of Computer Codes, Models, and Hand Calculational Methods .... 176 Table F. 7.1. Calculated D Values . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 185 Table F. 7 .2. Input Data for Example Problem ................................. 192 Table F. 7 .3. Calculated Terms for Example Problem . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 193 xx List of Forms DOE G 421.1-1 8-25-99 Form E.1 Software Revision Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 167 Form E.2 Software Nonconformance Report .................................. 169 Form E.3 NCS Software System Version No.1 Catalog . . . . . . . . . . . . . . . . . . . . . . . . . .. 170 Form E.4 Request for User Access ......................................... 171 Form E.5 NCS Software Disaster Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 172 Form E.6 Software Labeling Protocol ....................................... 173 Form E.7 NCS Software System Team (NCSSST) Sample Charter . . . . . . . . . . . . . . . . . . .. 174 DOE G 421.1-1 8-25-99 1. SCOPE. The SCOPE applies to the entire document. 1.1 SCOPE. This DOE Good Practices Program Guide illustrates and suggests practices and procedures for conducting a nuclear criticality safety (NCS) program at U.S. Department of Energy (DOE) nonreactor nuclear facilities having significant quantities of fissionable materials. These DOE practices and procedures are relevant to NCS program administration and oversight, NCS personnel selection and training, performance of NCS evaluations and analyses, emergency response, and programmatic control of processes, storage, procedures, hardware, and software. Throughout the text of this document, the term "fissionable material," used for concision, refers to the term that is specifically relevant to criticality safety concern, "significant quantity of fissionable material."

1.2 APPLICATION. This DOE Good Practices Program Guide applies to government contractors operating DOE nonreactor nuclear facilities having significant quantities of fissionable materials (unirradiated and irradiated) in process, storage, or transport outside a nuclear reactor core. This application extends to processing and burying of fissionable material wastes and handling or processing and storing of reactor fuel in storage pools. 1.3 STRATEGY AND INTENT. It is DOE policy to use National and International Consensus Standards (e.g., ANSI/ANS Standards, ISO Standards, ECS Standards) when such standards are available to meet DOE needs. This DOE Good Practices Program Guide is a comprehensive guidance document to assist in developing a criticality safety program to implement the DOE Order (or Rule) on nuclear criticality safety, and the invoked ANSI/ANS standards, through use of good practices. Its comprehensiveness precludes its full applicability to all sets of conditions, since a good practice for one set of conditions could be an unnecessary, or a poor, practice for a similar, but not identical, set of conditions. DOE Good Practices Guides generally should not be used to develop audit check-lists. This Good Practices Program Guide is not a requirements document and shall not be used as an auditing document. This Good Practices Program Guide shall not be incorporated in a Contract. It is intended only to provide guidance, but to meet this intention nuclear criticality safety professionals are expected to be familiar with its content. Requirements for DOE nuclear criticality safety programs are found in higher level documents, e.g., Policy, Rule, Order, and Manual. These documents, e.g., the Order or the Manual, may invoke National and International Consensus Standards. Use of the word "shall" (i.e., as the statement of a requirement) in this Good Practices Program Guide is only to try to maintain consistency with its direct and implicit use in higher level documents and shall not be used to impose requirements beyond those in higher level documents, except for use of the term "shall" in this paragraph. Therefore, no additional requirements, i.e., requirements that cannot be found in higher level documents, shall be imposed by users of this document. Because the work performed at the different DOE sites is diverse (viz., hands-on unshielded fissionable material operations at some sites and remote shielded operations at other sites), this Good Practices Program Guide, being relevant to such work, is also diverse. Hence, it is comprehensive and covers most of the areas of responsibility pertaining to conducting a nuclear criticality safety program. To this end, information in this document has been gathered eclectically; therefore it is inappropriate to use this document in its entirety either for anyone site or for any single application. Its intent, therefore, is to present a comprehensive text of good practices for nuclear criticality safety, and to depend on good judgment in both engineering and management to be the principal determinant for applicability of these good practices. While even a comprehensive 2 DOEG421.1-1 8-25-99 text of good practices cannot address every need, it can serve as a source of ideas to address differing needs as they arise.

Imbalance of detail is an expected characteristic of a comprehensive document that draws eclectically from diverse sources. In this Good Practices Program Guide, such imbalance, while present, is manifest by too much detail in some areas rather than too little detail in anyone area. As a fault, therefore, it is one of form rather than content, and a conservative fault, at that. 1.4 DOCUMENT REFERENCES. Some DOE documents (e.g., DOE Orders) referenced in this Good Practices Program Guide may have been revised or canceled by the time of issuance. Citation of references in this Good Practices Program Guide is intended to identify information that is relevant to this Good Practices Program Guide whether such information is found in the reference or in its revision or replacement. 1.5 ANSI/ANS Series-8 Standards. The basic elements and control parameters of programs for nuclear criticality safety at DOE must meet the requirements of specified ANSI/ANS Series-8 standards. Therefore, in the interest of clarity, familiar phraseology directly from these standards is sometimes used in this Good Practices Program Guide. Where such phraseology is used without attribution, failure to attribute is unintentional. 1.6 MAINTENANCE. The Good Practices Program Guide represents the desire of the DOE criticality safety community to have available a written account of the good practices of the various members of the criticality safety community so that all may be able to profit from the experiences of each, as applicable, hence from which practices may be culled that can be used in specific circumstances in the spirit of DNFSB Recommendation 95-2. Maintenance actions that are needed to keep the Guide current must be undertaken judiciously so that a contemporaneously valid version of the Guide is always available to the criticality safety community. Since criticality safety is a continuously evolving activity, the Guide cannot include recent events, especially since time­ consuming consensus reviews are necessary for each new draft, with revision reviews that would encompass several drafts estimated to take at least two years. The Guide is published here as Revision 0, with revisions to follow in subsequent years reflecting user experience and new initiatives. Among the new initiatives that affect criticality safety that will appear in Revision 1 are those included in Board Recommendations 95-2 and 97-2. The former affects criticality in general through its intention of achieving integrated treatment of safety management. The latter affects criticality in particular through the initiatives of bounding experiments and analyses; organization of experiment and calculation records; techniques for interpolation, extrapolation, and determination of area of applicability; use of simplified methods of analysis where applicable and defensible; and assignment of criticality safety as a staff function assisting line management. When Revision 1 is issued to include the effects of these, and perhaps other, initiatives, yet newer initiatives can be expected to have been only recently in place in readiness for Revision 2. Another aspect of document maintenance to be considered is the updating of references and the updating of terminology relevant to ancillary subject matter (e.g., relevant to packaging and to transportation). Updating that is deemed nonessential for Revision ° will be reserved for inclusion in Revision 1. DOE G 421.1-1 8-25-99 2. APPLICABLE DOCUMENTS

3 2.1 DOE DOCUMENTS. Information from the following DOE Orders, Standards, and Guides was used in the development of this Guide. Certain Orders have now been superseded by new Orders or have been canceled. These Orders are designated by phraseology using the terms superseded or canceled. 2.1.1 DOE 1300.2A. DEPARTMENT OF ENERGY TECHNICAL STANDARDS PROGRAM, of 5-19- 92, provides requirements for the development and application of technical standards in Department of Energy facilities, programs, and projects. 2.1.2 DOE 1324.2A. RECORDS DISPOSITION, of 9-13-88, contains procedures for the retention and disposition of records. Canceled by DOE 0 200.1, INFORMATION MANAGEMENT PROGRAM, of 9-30-96. 2.1.3 DOE 5000.3B. OCCURRENCE REPORTING AND PROCESSING OF OPERATIONS INFORMATION, of 1-19-93, establishes a system for reporting unusual occurrences having programmatic significance. Canceled by DOE 0 232.1, OCCURRENCE REPORTING AND PROCESSING OF OPERATIONS INFORMATION, of 9-25-95. 2.1.4 DOE 5480.3. SAFETY REQUIREMENTS FOR THE PACKAGING AND TRANSPORTATION OF HAZARDOUS MATERIALS, HAZARDOUS SUBSTANCES, AND HAZARDOUS WASTES, of 7-9-85, describes the requirements for packaging and transportation of hazardous materials, hazardous substances, and hazardous wastes. Canceled by DOE 0 460.1 A, PACKAGING AND TRANSPORTATION SAFETY, of 10-2-96. 2.1.5 DOE 5480.4. ENVIRONMENTAL PROTECTION, SAFETY, AND HEALTH PROTECTION STANDARDS, of 5-15-84, specifies the application of mandatory ES&H standards to DOE operations. Portions canceled by DOE 0 440.1, WORKER PROTECTION MANAGEMENT FOR DOE FEDERAL AND CONTRACTOR EMPLOYEES, of 9-30-95. 2.1.6 DOE 5480.11. RADIATION PROTECTION FOR OCCUPATIONAL WORKERS, of 12-21-88, provides radiation protection standards and program requirements for operations with respect to the protection of the worker from ionizing radiation. Canceled by 10 CFR 835. 2.1.7 DOE 5480.18B. NUCLEAR FACILITY ACCREDITATION TRAINING PROGRAM, of 8-31-94, institutionalizes a performance-based training process for DOE Category A reactors and high-hazard and selected moderate-hazard non-reactor nuclear facilities. Canceled 12-20-96. 2.1.8 DOE 5480.19. CONDUCT OF OPERATIONS REQUIREMENTS FOR DOE FACILITIES, of 7-9-90, which establishes requirements dealing with the conduct of operations for DOE operators. 2.1.9 DOE 5480.20A. PERSONNEL SELECTION, QUALIFICATION, TRAINING, AND STAFFING REQUIREMENTS AT DOE REACTOR AND NON-REACTOR NUCLEAR FACILITIES, of 2-20-91, describes the requirements for personnel involved in the operation, maintenance, and technical support of DOE-owned Category A and B reactors and non-reactor nuclear facilities. 2.1.10 DOE 5480.21. UNREVIEWED SAFETY QUESTIONS, of 12-24-91, establishes the means by which Unreviewed Safety Questions (USQs) are identified and the means of resolution of USQs. 4 DOE G 421.1-1 8-25-99 2.1.11 DOE 5480.22. TECHNICAL SAFETY REQUIREMENTS, of 2-25-92, establishes DOE's nuclear facility technical safety requirements. 2.1.12 DOE 5480.23. NUCLEAR SAFETY ANALYSIS REPORTS, of 4-10-92, which establishes uniform requirements for the preparation and review of safety analyses. 2.1.13 DOE 420.1. FACILITY SAFETY, Section 4.3, Nuclear Criticality Safety, of 10-13-95, establishes DOE's nonreactor nuclear facility nuclear criticality safety program.

2.1.14 DOE 5484.1. ENVIRONMENTAL PROTECTION, SAFETY, AND HEALTH PROTECTION INFORMATION REPORTING REQUIREMENTS, of 2-24-81, establishes the requirements and procedures for reporting and investigating matters of significance for the protection of environment, safety, and health at DOE operations. Canceled by DOE 0 231.1, ENVIRONMENT, SAFETY, AND HEALTH REPORTING, of 9-30-95, and by DOE 0 225.1, ACCIDENT INVESTIGATIONS, of 9-29- 95. 2.1.15 DOE 5500.28. EMERGENCY CATEGORIES, CLASSES, AND NOTIFICATION AND REPORTING REQUIREMENTS, of 4-30-91, establishes requirements for the coordination and direction of planning, preparedness, and response to operational emergencies. Canceled by DOE 0 151.1, COMPREHENSIVE EMERGENCY MANAGEMENT SYSTEM, of 9-25-95. 2.1.16 DOE 5500.3A. PLANNING AND PREPAREDNESS FOR OPERATIONAL EMERGENCIES, of 4-30-91, which establishes requirements for the development of site-specific emergency plans and procedures at nuclear facilities. Canceled by DOE 0 151.1, COMPREHENSIVE EMERGENCY MANAGEMENT SYSTEM, of 9-25-95. 2.1.17 DOE 5700.6C. QUALITY ASSURANCE, of 8-21-91, describes DOE's quality assurance program requirements for DOE and non-nuclear facility contractors. (See also 10 CFR 830.120, QUALITY ASSURANCE). Canceled by DOE 0 414.1, QUALITY ASSURANCE, of 11-24-98. 2.1.18 DOE 6430.1A. GENERAL DESIGN CRITERIA, of 04-6-89, contains the criteria for the design and construction of DOE facilities. (See also DOE 0 430.1, LIFE CYCLE ASSET MANAGEMENT, of 2-28-97.) 2.1.19 DOE-STD-3007-93. GUIDELINES FOR PREPARING CRITICALITY SAFETY EVALUATIONS AT DEPARTMENT OF ENERGY NON-REACTOR NUCLEAR FACILITIES, of December 1993, contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at Department of Energy (DOE) Non-Reactor Nuclear Facilities. 2.1.20 DOE-STD-3013-94. CRITERIA FOR SAFE STORAGE OF PLUTONIUM METALS AND OXIDES, of December 1994 provides for safe storage (for at least 50 years or until final disposition) of plutonium metals, selected alloys, and stabilized oxides that contain a minimum of 50 weight-percent plutonium. 2.1.21 DOEITIC-11603-REV. 1. NONREACTOR NUCLEAR FACILITIES: STANDARDS AND CRITERIA GUIDE, of September 1986, is a source document that identifies standards, codes, and guides that address nuclear safety considerations at nuclear facilities. DOE G 421.1-1 8-25-99 5 2.2 OTHER FEDERAL DOCUMENTS. Information from the following other Federal documents was used in the development of this Guide. 2.2.1 Code of Federal Regulations (CFR). 2.2.1.1 Title 10, Part 70 of the CFR. Domestic Licensing of Special Nuclear Material establishes the procedures and criteria for the issuance of licenses to receive title to, own, acquire, deliver, receive, possess, use, and initially transfer special nuclear materials. 2.2.1.2 Title 10, Part 71 of the CFR. Packaging and Transportation of Radioactive Material establishes the requirements for obtaining Nuclear Regulatory Commission (NRC) approval of packaging and shipment of licensed material. 2.2.1.3 Title 10, Part 830 of the CFR. Nuclear Safety Management establishes requirements for preventing the uncontrolled release of radioactivity to the environment, inadvertent criticality, limiting and monitoring facility staff exposure to radiation and radioactivity, and protecting the public from exposure to radiation and radioactive contamination.

2.2.2 Nuclear Regulatory Commission (NRC) Documents. Information from the following NRC Regulatory Guides and NUREGs was used in the development of this Guide. 2.2.2.1 NRC Regulatory Guide 3.1. Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material describes a method of using borosilicate-glass Raschig rings as a neutron absorber for criticality safety control in plants processing special nuclear materials. 2.2.2.2 NRC Regulatory Guide 3.4. Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors describes acceptable procedures for the prevention of criticality accidents in the handling, storing, processing, and transporting of fissionable materials outside of nuclear reactors. 2.2.2.3 NRC Regulatory Guide 3.34. Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Uranium Fuel Fabrication Plant describes methods used for performing analyses to assess the risk to public health and safety resulting from postulated nuclear criticality accidents in uranium fuel fabrication and processing plants. 2.2.2.4 NRC Regulatory Guide 3.35. Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant describes methods used for performing analyses to assess the risk to public health and safety resulting from postulated nuclear criticality accidents in plutonium processing and fuel fabrication plants. 2.2.2.5 NRC Regulatory Guide 3.68. Nuclear Criticality Safety Training provides guidance on an appropriate nuclear criticality safety training program for the use of special nuclear material for license applicants and operations staff. It is not adequate for training for the nuclear criticality staff. 2.2.2.6 NRC Regulatory Guide 8.12. Criticality Accident Alarm Systems describes the specifications for use of criticality alarms where there is a potential hazard to workers from nuclear criticality accidents. 6 DOE G 421.1-1 S-25-99 2.2.2.7 NUREG/BR-0167. Software Quality Assurance Program and Guidelines, using industry standards, provides guidance in the development and maintenance of software. 2.2.2.S NUREG/CR-127S, SANDSO-0200,RX,AN. Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications provides some useful information for the performance of human reliability analyses. 2.2.2.9 NUREG/CR-4639, EEG-245S. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Summary Description provides some useful human reliability data for the performance of human reliability analyses, primarily plant-specific reactor data from public domain sources. 2.3 NON-GOVERNMENT DOCUMENTS. 2.3.1 American National Standards Institute (ANSI). 2.3.1.1 ANSI/ANS-S.1-199S. Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors provides basic criteria and limits for operations with fissionable materials outside reactors except for critical experiments. The standard also provides requirements for establishing the validity and areas of applicability of any calculational method used in assessing nuclear criticality safety. 2.3.1.2 ANSI/ANS-S.3-1997. Criticality Accident Alarm System provides the performance criteria for detecting nuclear criticality accidents.

2.3.1.3 ANSI/ANS-S.5-1996. Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material describes the chemical and physical environment for usage, properties of the rings and packed vessels, maintenance inspection procedures, and criticality operating limits for solution systems containing 235U, 239pU, or 233U. 2.3.1.4 ANSI/ANS-S.6-19S3,R95. Safety in Conducting Sub critical Neutron-Multiplication Measurements in Situ provides safety guidance for conducting subcritical neutron-multiplication measurements where physical protection of personnel against the consequences of a criticality accident is not provided. 2.3.1.5 ANSI/ANS-S.7-199S. Guide for Nuclear Criticality Safety in the Storage of Fissile Materials provides mass and spacing limits for the storage of uranium containing greater than 30 wt % 235U, for 233U, and for plutonium as metals and oxides. 2.3.1.6 ANSI/ANS-S.9-19S7,R95. Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials provides criteria and data based on experiments and calculations applicable to homogeneous aqueous solutions. 2.3.1.7 ANSI/ANS-S.10-19S3,RSS. Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement provides criteria for the prevention of nuclear accidents in facilities with shielding and confinement and a definition of the adequacy of the shielding and confinement required. DOE G 421.1-1 8-25-99 7 2.3.1.8 ANSI/ANS-8.12-1987,R93. Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors provides single parameter limits for fissionable units of simple shape containing the three principal fissile nuclides. 2.3.1.9 ANSI/ANS-8.15-1981,R95. Nuclear Criticality Control of Special Actinide Elements provides single parameter limits for maintaining nuclear criticality safety of special actinide elements. 2.3.1.10 ANSI/ANS-8.17-1984,R97. Criticality Safety Criteria for the Handling, Storage, and Transportation of L WR Fuel Outside Reactors addresses LWR fuel rods and units outside reactor cores. 2.3.1.11 ANSI/ANS-8.19-1996. Administrative Practices for Nuclear Criticality Safety provides criteria for the administration of a nuclear criticality safety program for operations outside of reactors in which there exists a potential for criticality accidents. 2.3.1.12 ANSI/ANS-8.20-1991. Nuclear Criticality Safety Training provides criteria for the administration of a nuclear criticality safety training program for personnel who manage, work in, or work near facilities where the potential exists for a criticality accident outside of reactors, though it does not apply to the training of nuclear criticality safety staff. 2.3.1.13 ANSI/ANS-8.21-1995. Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors provides guidance for using fixed neutron absorbers integrally in nuclear facilities and fissionable material process equipment outside reactors to provide criticality safety control. 2.3.1.14 ANSI/ANS-8.22-1997. Nuclear Criticality Safety Based on Limiting and Controlling Moderators provides guidance for criticality safety by the limitation and control of moderators in the range from no moderation to optimum moderation for fissile materials. 2.3.1.15 ANSI/ANS-8.23-1997. Nuclear Criticality Accident Emergency Planning and Response provides guidance for minimizing risks to personnel during emergency response to a nuclear criticality accident outside reactors.

2.3.1.16 ANSI/ANS-10.3-1986. Guidelines for the Documentation of Digital Computer Programs presents guidelines for the documentation of digital computer programs prepared for scientific and engineering applications with the objective to facilitate effective selection, usage, transfer, conversion, and modification of computer programs. 2.3.1.17 ANSI/ANS-1 0.4-1987. Guidelines for the Verification and Validation of Scientific and Engineering Computer Programs for the Nuclear Industry provides guidelines for the verification and validation (V&V) of scientific and engineering computer programs developed for use by the nuclear industry with the objective to identify activities that will improve the reliability of scientific and engineering computer programs and reduce the risk of incorrect application. 2.3.1.18 ANSI/IEEE-Std-500-1984. IEEE Guide to the Collection and Presentation of Electrical, Electronic, Sensing Component, and Mechanical Equipment Reliability Data for Nuclear-Power Generating Stations provides data useful for performing equipment reliability analyses. 8 DOE G 421 .1-1 8-25-99 2.3.2 Industry Related Reference Documents. 2.3.2.1 ANS-9, GLOSSARY of Terms in Nuclear Science and Technology, American Nuclear Society Standards Subcommittee ANS-9 on Nuclear Terminology and Units, Harry Alter, Chairman, La Grange Park, Illinois, 1986. 2.3.2.2 LA-11627-MS, GLOSSARY of Nuclear Criticality Terms, Hugh C. Paxton, Los Alamos National Laboratory, Los Alamos, New Mexico, October 1989. 2.3.2.3 PNL-SA-4868, Rev. 5, Anomalies of Nuclear Criticality, E. D. Clayton, Pacific Northwest Laboratory, Richland, Washington, June 1979, provides discussions and explanations of deviations from commonly accepted rules of criticality behavior. 2.3.2.4 LA-3366 (Rev), Criticality Control in Operations with Fissile Material, H. C. Paxton, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, November 1972, provides criticality data and makes them understandable in terms of simple reactor physics concepts to help develop intuition for conditions to be avoided during operations. 2.3.2.5 NUREG-0492, Fault Tree Handbook, U.S. Nuclear Regulatory Commission, January 1981, is a textbook on the fault tree technique for acquiring information about a system. 2.3.2.6 DOE/NCT--04, A Review of Criticality Accidents, William R. Stratton (revised by David R. Smith), Los Alamos National Laboratory, Los Alamos, New Mexico, March 1989, provides discussions of forty-one criticality accidents and the characteristics of their prompt power excursions. 2.3.2.7 Criticality and Fissionability Properties of Selected Actinide Nuclides, N. L. Pruvost, E. D. Clayton, and C. T. Rombough, Los Alamos National Laboratory, Los Alamos, New Mexico, to be issued, provides information concerning the nuclear properties of selected nuclides of the first ten of the fourteen actinide elements -- thorium through einsteinium. 2.3.2.8 LA-10860-MS, Criticality Dimensions of Systems Containing 235U, 239pU, and 233U, 1986 Revision, H. C. Paxton and N. L. Pruvost, Los Alamos National Laboratory, Los Alamos, New Mexico, July 1987, provides a compilation of critical data obtained from experiments performed during the period of 1945 through 1985, supplementing TID-7016 (paragraph 2.3.2.9 of this Guide). 2.3.2.9 TID-7016, Rev. 3, and LA-12808, Nuclear Criticality Safety Guide, Los Alamos National Laboratory, Los Alamos, New Mexico, September 1996, provides general guidance information related to nuclear criticality safety principles, experience, and practice.

LA-2063 (1956), and TID-7016 (1957), and TID-7016, Rev. 1 (1961), and TID-7016, Rev. 2 (1978), Nuclear Safety Guide, Los Alamos National Laboratory, Los Alamos, New Mexico, October 1996, have been reissued under a single cover as historical documents. 2.3.2.10 NUREG/CR-6504, Vols. 1 and 2 (and ORNL/TM-13322N1, V2),An Updated Nuclear Criticality Slide Rule, Oak Ridge National Laboratory, Oak Ridge, Tennessee, April 1997 (Vol. 1) and April 1998 (Vol. 2). DOE G 421.1-1 8-25-99 9 2.3.2.11 NEA/NSC/DOC(95)03/1-VII, International Handbook of Evaluated Criticality Safety Benchmark Experiments, Organization for Economic Cooperation and Development, Nuclear Energy Agency, Nuclear Science Committee, Paris, September 1998 (or most recent version). 2.3.3 Journal Articles, and Meetings and Conference Proceedings. A bibliography of technical journal articles and of technical meeting and conference proceedings relevant to nuclear criticality safety can be found in Appendix G. 10 DOE G 421.1-1 8-25-99 3. TERMS AND DEFINITIONS. The following provides terms and definitions used within this Guide. The word shall is used to denote a requirement, the word should to denote a recommendation, often of a higher level document, and the word may to denote permission, neither a requirement nor a recommendation, except when used in its other context, meaning contingency, i.e., possibility. A more general understanding of the roles of requirements and recommendations in this Guide can be obtained from paragraph 1.3, Section 5.2, and Appendix A. Phrases or words that are italicized are defined or listed elsewhere in the terms and definitions of this section. 3.1 ABSORBER, NEUTRON - A material with which neutrons significantly interact by reactions, resulting in their disappearance as free particles. 3.2 ABSORPTION, NEUTRON - A neutron-induced reaction, including fission, in which the neutron disappears as a free particle. The absorption cross section is designated as. See capture, neutron; cross section, neutron. 3.3 ACCIDENT, CREDIBLE - Those accidents with an estimated probability of occurrence > 10-6/year. 3.4 ACCIDENT, CRITICALITY (also NUCLEAR CRITICALITY ACCIDENT) - The release of energy as a result of accidentally producing a self-sustaining or divergent fission chain reaction. 3.5 ACCIDENT, DESIGN BASIS (DBA) - Accidents that are postulated for the purpose of establishing functional requirements for safety significant structures, systems, components, and equipment. 3.6 ALARM SYSTEM, CRITICALITY ACCIDENT (CAS) - A system capable of providing an immediate emergency evacuation alarm signal (usually audible but may be visible) after detecting (usually by the detection of gamma or neutron radiation, or both) a criticality accident. 3.7 ALBEDO, NEUTRON - The probability under specified conditions that a neutron entering into a region through a surface will return through that surface. 3.8 ALPHA PARTICLE - A 4He nucleus, usually emitted during a nuclear transformation. 3.9 ANALYSIS, NUCLEAR CRITICALITY SAFETY - The work products that contribute to the development of a nuclear criticality safety evaluation. In general, it is the technological term of art, "analysis," specifically applied to nuclear criticality safety. 3.10 AREA, ARCHIVE STORAGE - An area in a computer system's storage that contains copies of source and executable code for superseded versions of the software and the master copy of the source and the current executable version.

3.11 AREA, DEVELOPMENT STORAGE - An area in computer storage in which software is stored during development without the possibility of inadvertent production use. Upon completion of the software change, the software is transferred to the Migration Area. Access to this area is limited to that necessary for development. The Development Storage Area resides in the individual developer's computer space. DOE G 421.1-1 8-25-99 11 3.12 AREA, MIGRATION STORAGE - An area in computer storage in which software verification tests are performed in a simulated production environment. Upon completion of the verification testing and with the approval of the Software System team, the software is transferred to the Production Storage Area. Access to the Migration Storage Area is limited to that necessary for testing. 3.13 AREA, PRODUCTION STORAGE - An area in a computer storage from which the software is invoked by authorized software users. Only the current version of the software will be in the Production Storage Area, and only the system administrator and a designated back-up will have write access to this area if proper software quality assurance procedures are followed. 3.14 AREAL DENSITY - The total mass of fissionable material per unit area projected perpendicularly onto a plane. For an infinite, uniform slab, it is the product of the slab thickness and the concentration of fissionable material within the slab. 3.15 AREAS OF APPLICABILITY - The ranges of material compositions, geometric arrangements and other factors within which the bias and its corresponding uncertainty of a calculational method are established. 3.16 BARN - A unit of area used in expressing nuclear cross sections; 1 barn = 10-24 cm2. 3.17 Be/X - Conventionally, the atomic ratio of beryllium to 235U, 239pU, or 233U in a solution or mixture. Where there is more than one fissile species, the ratios are specified separately. 3.18 BETA PARTICLE - An electron, of either negative or positive charge, that has been emitted by an atomic nucleus or neutron in a nuclear transformation. 3.19 BIAS, CALCULATIONAL - A measure of the systematic disagreement between the results calculated by a method and experimental data. The uncertainty in the bias is a measure of both the precision of the calculations and the accuracy of the experimental data. See DOE Order 420.1, Section 4.3. 3.20 BIRDCAGE - A container and attached cage-like structure for maintaining a safe distance between a body of fissionable material and other objects (including other bodies of fissionable materia!), which, if brought too close, might give rise to criticality. 3.21 BUCKLING - The eigenvalue of the Helmholtz equation (either Bm2 or B/). Algebraic expressions can be used to relate material (Bm 2) or geometric (Bg 2) characteristics of critical, subcritical, or supercritical fissionable material systems. 3.22 BURST, PROMPT - Usually refers to the pulse or spike of energy from fissions produced by a prompt burst reactor. 3.23 C/X - Conventionally, the atomic ratio of carbon to 235U, 239pU, or 233U in a solution or mixture. Where there is more than one fissile species, the ratios are specified separately. 3.24 CALCULATIONAL METHOD - The mathematical equations, approximations, assumptions, associated numerical parameters, such as neutron cross sections, and calculational procedures that yield the calculated results. 12 DOE G 421.1-1 8-25r99

3.25 CAPTURE, NEUTRON - Neutron absorption not leading to fission or other neutron production. The capture cross section is designated Dc, See absorption, neutron; cross section, neutron. 3.26 CENT - A unit of reactivity equal to one-hundredth of the increment between delayed criticality and prompt criticality (a dol/ar). 3.27 CERTIFICATION - The process by which contractor nuclear facility management provides written endorsement of the satisfactory achievement of qualification of a person for a position. 3.28 CHAIN REACTION, NUCLEAR FISSION - A sequence of nuclear fission reactions in which the fissions are induced by neutrons emerging from preceding fissions. Depending on whether the number of fissions directly induced by neutrons from one fission is on the average less than, equal to, or greater than unity, the nuclear fission chain reaction is convergent (subcritica/), self­ sustaining (critica/), or divergent (supercritica/). 3.29 CODE, EXECUTABLE - The machine-language program that is the output after translation (compiling) and linking of the source code. 3.30 CODE, SOURCE - The original mnemonic or high-level statement versions of a program. The starting information or "source" from which the final "object" (machine language or executable code) is derived. 3.31 CONFIGURATION CONTROL TEST - Periodic testing of the production version of software to determine if unauthorized changes have occurred. 3.32 CONFIRMATION - The performance of audits, inspections, surveillance activities, and other assessments of compliance with regulatory or nuclear criticality safety program requirements, analysis/evaluation requirements, and other requirements. 3.33 CONSERVATISM - Simplifying approximations and assumptions in safety analyses and evaluations and their applications that increase the safety margin above the required minimum. 3.34 CONTAINER - See packaging. 3.35 CONTINGENCY - A credible but unlikely change in a conditionlcontrol important to the nuclear criticality safety of a fissionable material operation that would, if it occurred, reduce the number of barriers (either administrative or physical) that are intended to prevent a nuclear criticality accident. 3.36 CONTROL AREA, FISSIONABLE MATERIAL (FMCA) - Fissionable material operating or storage areas where physical and procedural controls are applied to maintain nuclear criticality safety. 3.37 CONTROL - The apparatus, processes, and mechanisms that, when manipulated, could affect the chemical, physical, metallurgical, or any other process of the nonreactor nuclear facility in such a manner as to affect nuclear criticality safety. 3.38 CONTROLLED DOCUMENT - A document whose content is maintained uniform by an administrative control system. DOE G 421.1-1 8-25-99 13 3.39 CONTROLS, ACTIVE-ENGINEERED - Those active means for ensuring nuclear criticality safety control methods. These means of control include active electrical, mechanical, and hydraulic hardware that sense a process variable important to nuclear criticality safety and provide automatic action to secure the system to a safe condition without requiring human intervention. 3.40 CONTROLS, ADMINISTRATIVE - Those administrative means for assuring nuclear criticality safety control methods. These means of control include organization and management, procedures, record keeping, assessment, and reporting necessary to ensure the nuclear criticality safety of a nonreactor nuclear facility.

3.41 CONTROLS, PASSIVE-ENGINEERED - Those means for ensuring nuclear criticality safety control methods that do not require human intervention or electrical or mechanical reaction to off­ specification conditions. These means of control take advantage of natural forces, such as gravity, physical chemistry limitations, and inherent physical characteristics, such as rigidity and structural integrity of cylindrical geometries, and limited compressibility of solids. These means of control include devices to prevent unsafe accumulations of fissionable material within a unit such as siphon breaks, filters, and pipe blanks between process vessels and spacing devices such as birdcages, racks, and stanchions between containers as well as fixed neutron poisons within vessels such as Raschig rings or between containers. 3.42 CORE - That part of a fissionable material system containing most or all of the fissionable material, as distinguished from a reflector. 3.43 CREDIBILITY - See credible. 3.44 CREDIBLE - Offering reasonable grounds for being believed on the basis of commonly accepted engineering judgment. 3.45 CRITICAL - Fulfilling the condition that a medium capable of sustaining a nuclear fission chain reaction has an effective multiplication factor, keff, equal to unity. (A nuclear reactor is critical when the rate of neutron production, excluding neutron sources whose strengths are not a function of fission rate, is equal to the rate of neutron loss.) 3.46 CRITICAL INFINITE CYLINDER - For a specified fissionable material and surrounding reflector, the infinitely long cylinder with a diameter that would be critical. 3.47 CRITICAL INFINITE SLAB - For a specified fissionable material and reflector on each surface, the slab of infinite lateral dimensions with a thickness that would be critical. 3.48 CRITICALITY ACCIDENT SCENARIO, POTENTIAL - A credible sequence of events that could lead to a criticality accident, which starts with an initiating event, such as a process upset, valving error, pluggage of a line, andlor operator error, followed by events involving failure or loss of criticality safety control (preventive) measures. Each potential criticality accident scenario represents a single path of events leading to a criticality accident, and all scenarios together encompass the total probability of criticality accident thereby permitting the evaluation of total risk in the facility. 3.49 CRITICALITY SAFETY, NUCLEAR (NCS) - Protection against the consequences of an inadvertent nuclear fission chain reaction, preferably by preventing the reaction. 14 3.50 CRITICALITY - The condition of being critical. DOE G 421.1-1 8-25-99 3.51 CROSS SECTION (0), NEUTRON MICROSCOPIC - A measure of the probability of a specified interaction between an incident neutron and a target particle or system of particles. It has the dimension of area and may be visualized as the area normal to the direction of an incident particle, which has to be attributed to the target particle to account geometrically for its interaction with the incident particle. It is commonly expressed in barns. Such cross sections include but are not limited to neutron capture (oc), fission (0,), neutron scatter (as), and neutron absorption (as). 3.52 CROSS SECTION (I), NEUTRON MACROSCOPIC - For a pure nuclide, it is the product of the neutron microscopic cross section for a particular reaction and the number of target nuclei per unit volume, giving it units of inverse length; for a mixture of nuclides, it is the sum of such products.

3.53 D/X - Conventionally, the atomic ratio of deuterium to 235U, 239pU, or 233U in a solution or mixture. Where there is more than one fissile species, the ratios are specified separately. 3.54 DECAY, RADIOACTIVE - A spontaneous nuclear transformation in which particles or gamma radiation is emitted, in which x-radiation is emitted following orbital electron capture, or in which the nucleus undergoes spontaneous fission. 3.55 DELAYED CRITICALITY - State of a fissionable material system such that the multiplication factor, k"ff' equals 1 as the steady-state condition. 3.56 DELAYED NEUTRONS - Neutrons emitted when the beta-decay of a fission product leads to a sufficiently highly excited state in the daughter nucleus that neutron emission is energetically possible. The time delay, relative to emission of prompt neutrons, is from somewhat less than 1 second to about 60 seconds. 3.57 DESIGN FEATURES - Active or passive features that are necessary to prevent, or reduce the probability of, a criticality accident. 3.58 DETECTION SYSTEM, CRITICALITY ACCIDENT (CDS) - A criticality accident detection system (usually gamma or neutron, radiation detection, or both) without an immediate emergency evacuation alarm, the purpose of which is to provide sufficient response time to allow for appropriate process-related mitigation, recovery actions, and possible delayed evacuation alarm if radiation exposures could be effectively limited by such actions. 3.59 DOABLE - A doable condition or instruction is one that is capable of being obeyed or performed, respectively. That is, a fissionable material operation, storage, or transport condition, limit, specification, instruction, etc. that can be measured or evaluated, and proper responses taken to obey the condition or properly perform the instruction. 3.60 DOLLAR - A unit of reactivity equal to the increment between delayed criticality and prompt criticality for a fixed chain reacting system. 3.61 DOSE, PROMPT - The total neutron and gamma dose imparted from the first burst resulting from a criticality accident. 3.62 DOUBLE-CONTINGENCY ANALYSIS - A double-contingency analysis is an analysis of potential criticality accident scenarios for the purpose of demonstrating compliance with the DOE G 421.1-1 8-25-99 15 double-contingency principle (application) by identifying appropriate barriers and means of control (see paragraph 5.6.1). This is typically an element of a Nuclear Criticality Safety Evaluation. 3.63 DOUBLE-CONTINGENCY PRINCIPLE (APPLICATION) - Process designs shall incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible. Protection shall be provided by either (i) the control of at least two independent process parameters -- which is the approach that is completely consistent with the Double Contingency Principle as stated in ANSI/ANS-8.1-1998 and which, when practical, is the approach preferred by DOE to be taken to prevent common-mode failure, or (ii) a system of multiple controls on a single process [nuclear] parameter, which shall be the alternative approach to be taken only when the preferred approach is shown to be impractical. The number of controls required upon a single controlled process parameter shall be based upon control reliability and any features that mitigate the consequences of control failure. In all cases, no single credible event or failure shall result in the potential for a criticality accident, except where single contingency operations are permissible, as presented in paragraph 5.1 of ANSI/ANS-8.1 0- 1983,R88. This exception applies to operations with shielding and confinement (e.g., hot cells or other shielded facilities). Double contingency shall be demonstrated by documented evaluations.

3.64 EVALUATION, NUCLEAR CRITICALITY SAFETY (NCSE) - A documented process that demonstrates, by establishing and providing subcritical operating values, the nuclear criticality safety of any part of, or process in, a nonreactor nuclear facility that contains fissionable material. The evaluation provides sufficient descriptions of the facility equipment, fissionable material processes, and operational controls to identify the normal and contingent abnormal operating conditions of the facility. The evaluation contains the technical computational or comparative nuclear criticality safety analysis information that provides the bases of subcritical operating values for the normal and abnormal (contingent) conditions of facility operations or processes. Guidelines for preparing NCSEs are discussed in DOE-STD-3007-93 (paragraph 2.1.19 of this Guide). 3.65 EVENT. ANTICIPATED - Events with an estimated probability of occurrence between 1/year and 1 0-2/year . These events are of moderate frequency and may occur once or more during the lifetime of a facility. 3.66 EVENT. CREDIBLE - Events with an estimated probability of occurrence greater than 10-6 /year. 3.67 EVENT. EXTREMELY UNLIKELY - Events with an estimated probability of occurrence between 10-4/year and 10-6/year. These events are not expected to occur in the lifetime of a facility. 3.68 EVENT. INCREDIBLE - Events with an estimated probability of occurrence less than 10-6 /year. These events are considered to be of extremely low probability of occurrence or non-mechanistic hypothetical events. 3.69 EVENT. UNLIKELY - Events with an estimated probability of occurrence between 10-2/year and 1 0-4/year . These events are not expected but may occur during the lifetime of a facility. 3.70 EXCURSION. NUCLEAR - An episode during which the fission rate of a supercritical system increases. peaks. and then decreases to a low value. Also. see accident, criticality. 16 DOE G 421.1-1 8-25-99 3.71 EXCURSION PERIOD (T) - The reciprocal coefficient of time (t), where fission power in a nuclear excursion increases as eitlTl before a quenching mechanism becomes effective. 3.72 EXCURSION, PROMPT POWER - A nuclear excursion as the result of configuring fissionable material to achieve prompt criticality. In general, a sharp power spike followed by a plateau that may be interrupted by smaller spikes. 3.73 EXERCISE, TABLE-TOP - An event in which re-entry, rescue, etc. actions are simulated that test the emergency management's capability to cope with a nuclear criticality accident. 3.74 EXPONENTIAL COLUMN - A subcritical block or cylinder of fissionable material with an independent neutron source at one end. Under appropriate conditions, the response of a neutron detector decreases exponentially with distance from the source. From the logarithmic rate of this decrease and lateral dimensions of the column, critical dimensions of an unreflected assembly of the material may be deduced. 3.75 EXPOSURE - A measure of the ionization produced in air by x-rays or gamma radiation; the sum of electric charges on all ions of one sign in a small volume of air when all electrons liberated by photons are completely stopped, per unit mass of the air. Note that exposure refers to the environment, not absorbing material. The unit of exposure is the roentgen. Alternatively, exposure is the incidence of radiation on living or inanimate material.

3.76 FACILITY, NON REACTOR NUCLEAR - An operational area (e.g., building, holding, storage, or disposal area) dedicated to activities or operations (handling, storing, or transporting) that involve radioactive or fissionable materials, or both, in such form and quantity that a nuclear hazard potentially exists to the employees or the general public. Included are activities or operations that 1 . produce, process, or store radioactive liquid or solid waste, fissionable materials, or tritium; 2. conduct separations operations; 3. conduct irradiated andlor fissionable materials inspection, fuel fabrication, decontamination, or recovery operations; 4. conduct fuel enrichment operations; or 5. perform environmental remediation or waste management activities involving radioactive materials. Incidental use and generation of radioactive materials in a facility operation (e.g., check and calibration sources, use of radioactive sources in research and experimental and analytical laboratory activities, electron microscopes, and x-ray machines) would not ordinarily require the facility to be included in this definition. Accelerators and reactors and their operations are not included. 3.77 FAVORABLE GEOMETRY - Geometric constraint of fissionable material in which sub criticality is maintained under anticipated conditions. Examples are limited diameter of pipes intended to contain fissile solution, or limited volumes of solution containers. 3.78 FISSIBLE NUCLIDE - A nuclide that cannot sustain a nuclear fission chain reaction with slow neutrons but is only capable of sustaining a nuclear fission chain reaction by interaction with fast neutrons, provided the effective fast neutron production cross section (VO,) exceeds the effective fast neutron absorption cross section (0.). Such nuclides include 231Pa, 234U, 237Np, 238pU, 240pU, 242pU, 241Am, 243Am, 244Cm, 248Cm, 250Cf, and 252Cf. DOEG421.1-1 8-25-99 17 3.79 FISSILE NUCLIDE - A nuclide capable of sustaining a fission chain reaction by interaction with slow neutrons, provided the effective neutron production cross section (vat) exceeds the effective absorption cross section (a .. ). Such nuclides include 232U, 233U, 235U, 239pU, 241pU, 242mAm, 243Cm, 245Cm, 247Cm, 249Cf, 251Cf, and 254Es. 3.80 FISSION, NUCLEAR - The division of a heavy nucleus into two (or, rarely, more) parts with masses of equal order of magnitude, usually accompanied by the emission of neutrons, gamma radiation, and, rarely, small charged nuclear fragments. Although some fissions take place spontaneously, neutron-induced fissions are of major interest in criticality safety. The neutron fission cross section is designated at, and v is the number of neutrons emitted per fission. 3.81 FISSION PRODUCTS - Nuclides produced by nuclear fission or by the subsequent radioactive decay of nuclides formed in this manner. 3.82 FISSION, SPONTANEOUS - Nuclear fission that occurs without the addition of particles or energy to the nucleus. 3.83 FISSION YIELD. EXCURSION - The total number of fissions in a nuclear excursion. 3.84 FISSIONABLE EQUIVALENT MASS (FEM) - That gram mass of a fissionable material having the same mass ratio to its minimum critical mass as that mass ratio of a different fissionable material gram mass to its minimum critical mass. For example, given materials x and y having minimum critical masses of mex and mey, the FEM mass of x, mx' is directly proportional to the grams of y, my, times the minimum critical mass of material x divided by the minimum critical mass of material y (i.e., mx = [my • mexllmey). The FEM enables comparison of a quantity of one fissionable nuclide to a quantity of a second, or reference, fissionable nuclide based on their minimum subcritical mass limits. If more than one nuclide is present, their FEMs (in terms of the reference) can be summed.

3.85 FISSIONABLE MATERIAL - A material of any nuclides capable of sustaining a nuclear fission chain reaction. For nuclear criticality safety purposes, such materials are composed of fissionable nuclides but may include nonfissionable nuclides. Such material may be fissionable material only by virtue of its form, configuration, and environment. That is: natural uranium as mined, processed, and transported in bulk form is not fissionable material; however, natural uranium as fabricated into reactor fuel element pellets or rods may be considered as fissionable material if handled in a processing or operating environment where the pellets or rods could be adequately moderated to create a critical system. This definition is intended strictly for this Guide. 3.86 FISSIONABLE MATERIAL HANDLER - An individual officially designated by installation management to manipulate or handle significant quantities of fissionable materials, or manipulate the controls of equipment used to produce, process, transfer, store, or package significant quantities of such fissionable materials. 3.87 FISSIONABLE NUCLIDE - Any nuclide capable of undergoing neutron induced fission. For nuclear criticality safety purposes, such nuclides include the fissile or fissible nuclides but may also include nuclides such as 227 Ac, 228Th, 229Th, 230Th, 232Th. 233Pa, 236U, 238U, and 239Np. 3.88 GAMMA RADIATION - Short-wavelength electromagnetic radiation emitted in the process of nuclear transition or particle annihilation. 18 DOE G 421.1-1 8-25-99 3.89 GEOMETRY CONTROL - Physically controlling the shape, dimensions, and configuration of fissionable material or of equipment containing fissionable material to maintain such systems safely subcritical. 3.90 GRADED APPROACH - A process of performing a nuclear criticality safety evaluation that acknowledges different levels of effort and documentation are appropriate for different complexities of fissionable material operations and the associated methods and controls applied to maintain subcriticality and safety. 3.91 GRAY - A unit of absorbed dose; 1 Gy = I J/kg = 100 rad. 3.92 GUIDES, PROTECTIVE ACTION (PAGs) - The projected radiological doses or dose commitment values to individuals in the general population that warrant protective action following a release of radioactive material. Protective actions would be warranted provided the reduction in individual dose expected to be achieved by carrying out the protective action is not offset by excessive risks to individual safety in taking the protective action. The PAG does not include the dose that has unavoidably occurred prior to the assessment. 3.93 H/X - Conventionally, the atomic ratio of hydrogen to 235U, 239pU, or 233U in a solution or hydrogenous mixture. Where there is more than one fissile species, the ratios are specified separately. 3.94 HAZARD - A source of danger (i.e., material, energy source, or operation) with the potential to cause illness, injury, or death to personnel or damage to a facility or to the environment (without regard for the likelihood or credibility of accident scenarios or consequence mitigation). "Potentially hazardous" is redundant. Note that a hazardous facility is not necessarily a high-risk facility. 3.95 HIGH ENRICHED URANIUM (HEU) - Uranium having isotopic contents of 235U or 233U greater than or equal to 20 weight percent. HEU generally refers to 93 weight percent 235U.

3.96 INCIDENT, NUCLEAR CRITICALITY SAFETY - A change in process condition or a loss of control beyond the evaluated process variances of the nuclear criticality safety analysis. 3.97 IN HOUR - A unit of reactivity that, when added to a delayed-critical system, would produce a period of one hour; now seldom used. 3.98 INSTALLATION, NONREACTOR NUCLEAR - A contractor-operated DOE site comprising one or more nonreactor nuclear facilities. 3.99 IONIZING RADIATION - Any radiation consisting of directly or indirectly ionizing particles, photons, or a mixture or both. X-rays and the radiations emitted in radioactive decay are examples. 3.100 IRRADIATION - Exposure to ionizing radiation. 3.101 k.". - See MULTIPLICATION FACTOR, EFFECTIVE. 3.102 k. - See MULTIPLICATION FACTOR, INFINITE. DOE G 421.1-1 8-25-99 19 3.103 LIMIT, UPPER SUBCRITICAL (USL) - The limiting value of system reactivity beyond which subcriticality cannot be ensured (e.g., keff). The upper subcriticallimit only allows for uncertainties in the calculations and experimental data used in its derivation. 3.104 LINEAR ENERGY TRANSFER (LET) - The average energy lost by an ionizing radiation per unit distance of its travel in a medium. A high LET is generally associated with protons, alpha particles, and neutrons, whereas a low LET is associated with x-rays, electrons, and gamma rays. 3.105 LOW ENRICHED URANIUM (LEU) - Uranium having isotopic contents of 235U or 233U less than 20 weight percent. LEU generally refers to ~5 weight percent 235U. 3.106 MAINFRAME COMPUTER - For purposes of this procedure, a computer in which file control is achieved by access rules within the operating system rather than by physical control of storage media. 3.107 MANAGEMENT, LlNE/PRODUCTION - The organizational unit that accepts the direct responsibility for, and exercises authority over, the application of nuclear safety to their operations. 3.108 MANAGER, FACILITY OPERATIONS - The person whose facility warrants nuclear criticality safety consideration and controls, and who should, therefore, accept responsibility for the day-to­ day nuclear criticality safety of his/her facility. 3.109 MANAGER, FUNCTIONAL SYSTEM - The person appointed to represent specific Nuclear Criticality Safety Departments in the development and implementation of software configuration control plans for defined software system(s). 3.110 MODEL - A representation of the actual physical parameters used in a calculation. 3.111 MODERATOR - A material which will reduce neutron energy by scattering neutrons without appreciable neutron capture. 3.112 MULTIPLICATION FACTOR, EFFECTIVE (k.,,) - Physically, the ratio of the total number of neutrons produced during a time interval (excluding neutrons produced by sources whose strengths are not a function of fission rate) to the total number of neutrons lost by absorption and leakage during the same interval. Mathematically (computationally), that eigenvalue number (Lagrange multiplier if defined as production-to-Ioss ratio) which, when divided into the actual mean number of neutrons emitted per fission in an assembly of materials, would make the calculated result for the nuclear chain reaction of that assembly artificially critical. 3.113 MULTIPLICATION FACTOR, INFINITE (kJ - The keffof an infinite uniform medium. 3.114 NEUTRON - An elementary particle having no electric charge, a rest mass of 1.67495 x 10-27 kg, and an average lifetime of 1000 s.

3.115 NEUTRON, EPITHERMAL - Neutron of kinetic energy greater than that of thermal agitation, often restricted to energies comparable to those of chemical bonds. 3.116 NEUTRON, FAST - Neutron of kinetic energy greater than 0.1 MeV but not more than a typical Maxwellian distribution with an average energy of about 1.9 MeV. 20 DOE G 421.1-1 8-25-99 3.117 NEUTRON, INTERMEDIATE - Neutron of kinetic energy equal to or greater than 0.1 eV and equal to or less than 0.1 MeV. 3.118 NEUTRON, SLOW - Neutron of kinetic energy less than about 0.1 eV. 3.119 NEUTRON, THERMAL - Neutrons in thermal equilibrium with the medium in which they exist. At room temperature the mean energy of thermal neutrons is about 0.025 eV. 3.120 NONFAVORABLE GEOMETRY - See favorable geometry. 3.121 NON FISSILE FISSIONABLE MATERIAL (see FISSIBLE) - Any composition of nuclides capable. of maintaining a nuclear fission chain reaction with fast neutrons only, provided the effective neutron production cross section (va,) exceeds the effective absorption cross section (as) of the composition. 3.122 NONFISSIONABLE MATERIAL - Any composition of nuclides incapable of maintaining a nuclear fission chain reaction with neutrons of any energy whereby the effective neutron production cross section (va,) is less than the effective absorption cross section (as) of the composition. This definition is intended strictly for this Guide. 3.123 NUCLEAR CRITICALITY ACCIDENT, PERCEIVED - Any presumed nuclear criticality accident as inferred from the observance of physical phenomena (e.g., temperature rises, over-pressures, or others) or the activation of alarm systems indicative of a criticality accident (examples that might be included are CAS, continuous air monitors, and area radiation monitors). 3.124 NUCLIDE - A species of atom characterized by its mass number, atomic number, and nuclear energy state. 3.125 OPERATION, FISSIONABLE MATERIAL - An operation using a significant quantity of fissionable material. An operation includes handling, storage, processing, and transportation. 3.126 PACKAGE - The packaging together with its fissionable material contents as presented for movement or storage. 3.127 PACKAGING - The assembly of components necessary to ensure compliance with specifications for safe containment, storage, and handling of fissionable materials. It may consist of one or more receptacles, absorbent materials, spacing structures, thermal insulation, radiation shielding, vehicle, tie-down systems, auxiliary equipment, and devices for cooling or absorbing mechanical shocks. 3.128 PARAMETER, NUCLEAR - Any physical property whose value affects the nuclear reactivity of a system. Nuclear parameters include the mass, density, and isotopic enrichment of fissionable material; the geometry, reflection, and interaction conditions of the system; and the moderation, composition, and neutron absorption characteristics of the fissionable material mixture and other system materials. 3.129 PARAMETER, PROCESS - Operating or processing variables directly or indirectly affecting nuclear parameters of fissionable materials. Such process parameters may include temperatures, pressures, flow rates, viscosity, elapsed times, heights, rotational velocities, electrical resistivity, electrical potential, electrical currents, pH, color, opacity, etc. DOE G 421.1-1 8-25-99 21

3.130 PEER - An individual who performs peer reviews (e.g., a Criticality Safety Organization member), who has at least equivalent qualifications and standing compared to, and who is independent of, one or more other individuals who perform specific original work. Independent, in this case, means not involved in the performance of the specific original work to be reviewed, to the extent practical, not the immediate supervisor of individuals who performed specific original work to be reviewed, and to the extent practical, having sufficient freedom from funding considerations to ensure that the work is impartially reviewed. 3.131 PEER REVIEW - A review process for appraising and reporting the acceptability of independent and original specific work of others. 3.132 POISON, NEUTRON - A nonfissionable neutron absorber, generally used for criticality control. 3.133 PROMPT BURST REACTOR - A device for producing nondestructive super-prompt-critical nuclear excursions. 3.134 PROMPT CRITICALITY - State of a fissionable material system such that the prompt­ neutron contribution to keff equals unity. 3.135 PROMPT NEUTRONS - Neutrons emitted immediately during the fission process. 3.136 PROTON - A stable elementary particle having a positive charge of 1.60219 x 10-19 coulomb and a rest mass of 1.67265 x 10-27 kg. 3.137 QUALITY FACTOR (QF) - The ratio of dose equivalent to absorbed dose. 3.138 QUENCHING MECHANISM - A physical process that limits an excursion spike. Examples are thermal expansion and microbubble formation in a solution. 3.139 RAD - A unit of radiation absorbed dose; 1 rad = 10-2 J/kg = 10-2 Gy of the medium. 3.140 RADIATION - In context of criticality safety, alpha particles, beta particles, neutrons, gamma rays, and combinations thereof. 3.141 RADIATION MONITOR - A detector to measure the level of ionizing radiation. A purpose may be to give information about dose or dose rate. 3.142 REACTIVITY - A parameter of a fissionable system that is proportional to 1 - 1/keff • Thus, it is zero if the system is critical, positive if the system is supercritical, or negative if the system is subcritical (see effective multiplication factor, keff). 3.143 REACTOR. NUCLEAR - A device in which a self-sustaining nuclear fission chain reaction can be maintained and controlled (fission "reactor," "pile," or "core"). 3.144 RECOVERY - Proposed, evaluated, analyzed, and implemented ameliorative or corrective actions to restore an intended degree of criticality safety. 3.145 REFLECTOR - Material outside a fissionable material system capable of scattering back to the system some neutrons that would otherwise escape. 22 DOEG421.1-1 8-25-99 3.146 REFLECTOR SAVINGS - The absolute difference between a dimension of the reflected core of a critical system and the corresponding dimension of a similar core that would be critical if no reflector were present. 3.147 REFLECTOR, SUPERNORMAL - Any material or combination of materials that offers greater neutron reflector effectiveness than an essentially infinite thickness of water (e.g., about 20 cm of water). 3.148 RELATIVE BIOLOGICAL EFFECTIVENESS (RBE) - A factor used to compare the biological effectiveness of absorbed radiation doses (i.e., rads or grays) because of different types of ionizing radiation; more specifically, it is the experimentally determined ratio of an absorbed dose of a radiation in question to the absorbed dose of a reference radiation required to produce an identical biological effect in a particular experimental organism or tissue. This term is used only in radiobiology, not instead of quality factor in radiation protection.

3.149 REM - A unit of dose equivalent (Roentgen Equivalent Man), replaced by the sievert. The sievert, however, has not appeared in the criticality accident literature. The dose equivalent in rems is numerically equal to the absorbed dose in rads multiplied by the quality factor, and any other necessary modifying factor. 3.150 RISK - The quantitative or qualitative expression of possible loss, usually expressed in dollars or fatalities per year or facility lifetime, that considers both the probability that a hazard will cause harm and the consequences of that event. Not to be confused with hazard. 3.151 ROENTGEN (R) - A unit of exposure; 1 R = 2.58 x 10-4 C/kg in air, where C is coulombs. Strictly, the roentgen applies to x-rays or gamma radiation. 3.152 SAFETY BASIS - The combination of information relating to the control of hazards at a nuclear facility (including design, engineering analyses, and administrative controls), which demonstrates that the facility can be operated safely. 3.153 SHUTDOWN MECHANISM - See quenching mechanism. 3.154 SIEVERT (Sv) - A unit of dose equivalent; 1 Sv = 1 J/kg = 100 rem. 3.155 SIGNIFICANT QUANTITY - The minimum quantity of fissionable material for which control is required to maintain subcriticality under all normal and credible abnormal conditions. 3.156 SITE - See installation, nonreactor nuclear. 3.157 SOFTWARE CATALOG - A list of all software units in operation in a stated software system used for nuclear criticality safety evaluations. It identifies each software unit and states the version approved for use as of a stated date (ordinarily, the date of issue of the catalog). The access control and the scope of the software catalog are determined by the software system team. 3.158 SOFTWARE CONFIGURATION CONTROL - The systematic evaluation, coordination, modification, verification, implementation, and documentation of software. DOE G 421.1-1 8-25-99 23 3.159 SOFTWARE DEVELOPER(S) - The individual(s) responsible for the actual design or modification, or both, of the software. 3.160 SOFTWARE REQUESTOR - The organization for which software is developed and which is responsible for the software requirements definition. 3.161 SOFTWARE SYSTEM - A group of related programs and data which act in concert toward a particular purpose, has a common developing organization, and is suitable for a single set of control mechanisms. 3.162 SOFTWARE SYSTEM TEAM - The persons responsible for the software configuration control. 3.163 SOFTWARE - The instructions that determine and define the operation of an electronic computer or computer system. 3.164 SOFTWARE USER - An organization or person who uses the software. 3.165 SPECIALIST, COGNIZANT NUCLEAR CRITICALITY SAFETY - The qualified nuclear criticality safety specialist who is knowledgeable of specific facility operations, processes, and equipment, and who is assigned by installation management to manage, or to provide directly, nuclear criticality safety analyses, computations, evaluations, reviews, or audits of designs and operations for a specified nonreactor nuclear facility. 3.166 SPECIALIST, NUCLEAR CRITICALITY SAFETY (NCSS) - A professional person who is knowledgeable of nuclear criticality safety issues relevant to facility operations, processes, and equipment, and who is assigned by installation management to provide nuclear criticality safety analyses, computations, evaluations, reviews, or audits of designs and operations for nonreactor nuclear facilities.

3.167 SPIKE (IN A PROMPT-POWER EXCURSION) - The initial power pulse of a prompt-power excursion, limited by the shutdown mechanism. 3.168 STORAGE (Also "External storage") - A portion of a computer system where software and data are stored. Typically, storage is on magnetic disks, but other media may be used where appropriate. Storage may be on-line or off-line (external) to the computer operating system. 3.169 SUBCRITICAL - See nuclear fission chain reaction. 3.170 SUPERCRITICAL - See nuclear fission chain reaction. 3.171 SYSTEM ADMINISTRATOR - An individual responsible for the control of software for a defined software system, including issuance, revision, documentation, and archiving. 3.172 UNCERTAINTY - Lack of absolute precision, accuracy, or sureness of actions characteristic of measurements of data, approximations of results, or execution of procedures. 3.173 URANIUM ENRICHMENT (ENRICHMENT) - The weight percentage of 235U in uranium, provided that percentage exceeds its natural value; if the reference is to enhanced 233U content, n233U enrichment" should be specified. 24 DOEG421.1-1 8-25-99 3.174 VALIDATION. CALCULATIONAL METHOD - The establishment of the bias and calculational uncertainty in the results produced by the combination of the computer software, computer hardware, the data libraries, such as neutron cross sections, and the modeling method employed. The bias is established by correlating the results of criticality experiments with results obtained for these same systems by the method being validated. Commonly the correlation is expressed in terms of the values of kef( calculated for the experimental systems, in which case the bias is the deviation of the calculated values of kef( from the experimentally determined value. However, other parameters may be used. The bias serves to normalize a method over its areas of applicability so that it will predict critical conditions within the limits of the uncertainty in the bias. Generally, neither the bias nor its uncertainty is constant; both should be expected to be functions of composition and other variables. NOTE: Validation is not a required part of a verification test. 3.175 VERIFICATION. SOFTWARE CONFIGURATION CONTROL - The periodic execution of software to determine if unauthorized and undocumented changes thereto have been made. 3.176 VERIFICATION TEST - The testing of new or revised software stored in the migration storage area before the software is transferred into the production storage area. It is coordinated by the software system team assisted by the developer and others as required to test the unit integration, qualification, and acceptance of the software. The extent of the verification test is determined by the software system team based on the magnitude of the change and the consequences of a software failure in service. 3.177 X-RAY - Electromagnetic radiation of wavelength in the range 10-10 cm to 10-6 cm emitted from outside the nucleus. DOE G 421.1-1 8-25-99 25 4. GENERAL GUIDANCE. The DOE Good Practices Program Guide establishes DOE nuclear criticality safety interpretation and guidance to assist in implementation of nuclear criticality safety (NCS) across the DOE complex. This document is not intended to contain an exhaustive compilation of nuclear criticality safety guidance for every situation. However, the document is intended to provide examples for the development of nuclear criticality safety procedures and manuals for DOE contractors. The choice to implement any part of this document is the responsibility of DOE contractor design, operating, technical support, and oversight organizational units, as applicable. The user of this document is expected to follow the precepts and conform to the requirements stated in paragraph 1.3.

4.1 APPLICABILITY. Applicability of this Good Practices Program Guide spans design, construction, operation, maintenance, and decommissioning of covered facilities. It is recognized that the design and as-built configuration of some existing facilities do not meet all of the good practices contained in this document because they were built prior to the development of certain DOE Orders. Where practicable, currently existing operations, systems, and facilities should be upgraded considering the guidance provided by this document. The practicability of such upgrades should consider the cost versus the benefits. 4.2 INTERPRETATION OF THE GOOD PRACTICES PROGRAM GUIDE. The Office of Nuclear Safety Policy and Standards (USDOE Headquarters Environment, Safety and Health - EH-31) is the cognizant organization responsible for the preparation, maintenance, and interpretation of this document. 4.3 MAINTENANCE OF THE GOOD PRACTICES PROGRAM GUIDE. Maintenance and revision of this document shall be in accordance with DOE Order 1300.2A, "Department of Energy Standards Program." 4.4 DOCUMENT ARCHITECTURE. The remainder of the document consists of Detailed Guidance and related appendices. 4.4.1 Topical Structure. The order of presentation of topics proceeds from the general (e.g., administrative topics) to the particular (e.g., criticality safety, and other topic details). More specifically: Sections 5.1, 5.2, and 5.3 discuss administrative topics relating to management (5.1)' qualifications (5.2), and criticality safety procedural matters (5.3); Sections 5.4 and 5.5 discuss the mitigative topics of alarm systems (5.4) and emergency preparedness (5.5); Sections 5.6, 5.7, 5.8, and 5.9 discuss technical topics relevant to prevention of inadvertent criticality, including controls (5.6), design and analysis (5.7), computations (5.8), and evaluations (5.9). 4.4.2 Content Structure. The Detailed Guidance (Section 5) begins with Administration (Section 5.1). This is followed by Personnel Selection, Qualification, Training, and Staffing Program (Section 5.2), serving as a brief introduction to Appendix A 26 DOE G 421.1-1 8-25-99 which bears the same title as Section 5.2 but which contains a detailed discussion of the topic. Following this is Operating. Storing. and Transferring - Plans. Procedures. Requirements. and Controls (Section 5.3). The Detailed Guidance continues with Criticality Accident Alarm and Detection Systems (Section 5.4); and this is followed by a brief discussion of the closely related topic Emergency Preparedness (Section 5.5). The Detailed Guidance continues, more extensively, with Nuclear Criticality Safety Control Principles and Methods (Section 5.6), supplemented by Appendix B, Graded Approach to Nuclear Criticality Safety, and this is followed, in an even more extensive form, by Nuclear Criticality Safety Design and Analysis Guidelines (Section 5.7) where and Appendix C, Estimating the Waiting Time Until the Simultaneous Collapse of Two Contingencies, Appendix 0, Examples of Design of Nuclear Criticality Safety Controls, are referenced for additional details. The Detailed Guidance concludes briefly, initially with Software Quality Assurance and Validation (Section 5.8), which references Appendix E, Software Configuration Control Procedure, and Appendix F, Example Computational Technique Validations, DOE G 421 .1-1 8-25-99 for additional details, and finally with Nuclear Criticality Safety Evaluation {NCSEI Guidelines (Section 5.9).

The entire document concludes with 27 Appendix G, Bibliography of Journal Articles and Meeting and Conference Proceedings. 4.4.3 Section and Paragraph Numbering. In this document, those text units that are specifically numbered 1, 2, 3,4, 5, 5.1, 5.2, 5.3, 5.4, 5.5, 5.6, 5.7, 5.8, and 5.9 are called "Sections" (capitalized word). All other numbered text units (viz., 2.1.13, 3.1, 5.6.3) are called "paragraphs" (uncapitalized word). Either "section" or "paragraph" (uncapitalized words) is used to designate any text unit not expressly numbered. 28 DOE G 421.1-1 8-25-99 5. DETAILED GUIDANCE. 5.1 ADMINISTRATION. Nuclear criticality safety is administered at contractor installations by assigning responsibilities for key nuclear criticality safety requirements and activities. The following position titles and organizations are identified for the purpose of describing key position responsibilities in the administration of a nuclear criticality safety program at an installation. 5.1.1 Contractor President/Chief Executive Officer. The Contractor Installation(s) Corporation/Company President (CEO), through Installation Manager(s), shall' accept overall responsibility for the installation(s) nuclear criticality safety program in a way that demonstrates a continuing interest in safety. This objective should be met by documenting how the following elements of the nuclear criticality safety program are met: 5.1.1.1 Responsibility. Specific responsibilities for the nuclear criticality safety of operations and commensurate authority shall be clearly established. The performance of these responsibilities shall be reviewed on a periodic basis (noting nuclear criticality safety related occurrences, limit­ violations, commendable practices, and others), assigning importance to nuclear criticality safety commensurate with the other aspects of management.2 5.1.1.2 Policy. The corporate/company-level policy for implementing nuclear criticality safety requirements shall be made known to all contractor employees involved in operations with fissionable material through written company policies. 3 5.1.1.3 Organization. Corporate/company-level organizations shall be established as necessary to (1) ensure that supervision is made as responsible for nuclear criticality safety as for production, development, research, or other functions, and (2) to ensure that a Criticality Safety Organization, staffed with personnel skilled in the interpretation of data pertinent to nuclear criticality and familiar with operations, serves as advisors to supervision. The Criticality Safety Organization, to the extent practicable, should be administratively independent of process supervision and be assigned in a manner compatible with that for other safety disciplines.4 5.1.1.4 Program Oversight. A means to periodically evaluate the effectiveness of the nuclear safety program shall be established and corporate/company-level management shall participate in that evaluation.5 An Installation Nuclear Criticality Safety Review Committee (lNCSRC), reporting directly to 1 ANSI/ANS-8.19-1996, section 4.1. 2ANSI/ANS-8.1-1983,R88, section 4.1.1; and ANSI/ANS-8.19-1996, section 4.3. 3A NSI/ANS-8.1-1983,R88, section 4.1.1; and ANSI-8.19-1996, section 4.2. 4ANSI/ANS-8.1-1983,R88, section 4.1.1; and ANSI/ANS-8.19-1996, sections 4.3 and 4.4. 5ANSI/ANS-8.19-1996, sections 4.5 and 4.6. DOE G 421.1-1 8-25-99 29

corporate/company-level management, is one means that has been used to satisfy this requirement.6 The remainder of this document will assume INCSRCs are the means used to satisfy this requirement although there may be other means that are equally acceptable. An INCSRC should be responsible for fostering and monitoring nuclear criticality safety to include, in accordance with the reference in paragraph 2.1.17, an assessment of: (a) the nonreactor nuclear installation nuclear criticality safety program, (b) each nonreactor nuclear facility nuclear criticality safety program, (c) each installation nuclear criticality accident emergency preparedness program including the nuclear criticality accident alarm systems (CASs) and the nuclear criticality accident detection systems (CDSs) for compliance with detection criteria (paragraphs 5.4.1 and 5.4.2), and configuration control of the systems. 5.1.1.5 Corrective actions. Corrective actions resulting from the review process described in paragraph 5.1.1.4 should be promptly completed. 7 5.1.1.6 Resources. Installation(s) shall have adequate resources to maintain an effective NCS program including personnel skilled in the interpretation of data pertinent to NCS and familiar with operations to serve as advisors to supervision.s 5.1.1.7 Stop Work Policy. Provide corporation/company policy and assignment of "stop work" authority to relevant personnel within nonreactor nuclear facility installation staffs. 9 5.1.2 Facility Operations Managers. Facility Operations Managers, whose facilities warrant nuclear criticality safety consideration and controls, shall accept responsibility for the day-to-day nuclear criticality safety of their facility by addressing the following elements:'o 5.1.2.1 CAS Management. For facilities requiring CAS coverage, accept responsibility for, and ensure the administration of, a program for CAS management, including CAS availability and alarm circuit functioning within established limits; maintain a current copy of the CAS location analysis in facility files; and request additional CAS analysis as required by facility and process changes. 5.1.2.2 Procedures development and maintenance. Accept responsibility for preparation and maintenance of procedures (including special procedures as necessary) for facility operation that identify nuclear criticality safety steps/controls and drawings identifying equipment important to 6ANSI/ANS-8.19-1996, section 4.7. 7 ANSI/ANS-8.19-1996, section 4.1; and ANSI/ANS-8.1-1983,R88, section 4.1.5. 8ANSI /ANS-8.1-1983,R88, section 4.1.1; and ANSI/ANS-8.19-1996, sections 4.3 and 4.4. 9ANSI/ANS-8.1-1983,R88, section 4.1.5; and ANSI/ANS-8.19-1996, section 7.7. IOANSI/ANS-8.19-1996, section 5. 30 DOE G 421.1-1 8-25-99 criticality safety, and ensure the use of, and adherence to, such procedures in day-to-day operations. 5.1.2.3 Staff training. Maintain a program of staff training in both the general and facility-specific aspects of criticality safety. 5.1.2.4 Design and procedure reviews. Provide appropriately trained staff to determine when procedures (including special procedures), drawings, and design documents require nuclear criticality safety review, and ensure that such procedures, drawings, and design documents are forwarded to the Criticality Safety Organization for review. 5.1.2.5 Configuration control program. Provide that facility process and equipment configuration control programs that ensure proper nuclear criticality safety review, analyses, approval, and documentation occur prior to implementing or modifying any fissionable material operation within the facility.

5.1.2.6 Self-assessments. Ensure that facility self-audits are performed at least annually, and forward copies of such audit reports to their organization, the Criticality Safety Organization, and the Installation Nuclear Criticality Safety Review Committee (INCSRC). These audits shall ascertain that procedures exist for operations with fissionable materials, that procedures are being followed, and that procedures are consistent with the nuclear criticality safety basis for the operation." 5.1.2.7 Compliance. Accept responsibility for compliance with applicable DOE Orders, safety requirements, Technical Standards, and the nuclear criticality safety basis. Compliance should be documented by the self-audit process, Criticality Safety Organization operational reviews, INCSRC reviews, and other appraisal processes. 5.1.2.8 Audit response approval. Approve the response to NCS review, audit, and appraisal findings. 5.1.2.9 Safety documentation. Ensure that nuclear criticality safety aspects of facility design, construction, and operation are covered by a documented criticality safety analysis, and that operations are documented. 5.1.2.10 Contingency analysis documentation. Ensure that the facility is covered by documented double-contingency analyses. See paragraph 3.62. 5.1.2.11 Facility shutdowns. Accept responsibility for the safe shutdown of their facilities where warranted by actual or indicated criticality safety deficiencies. 5.1.2.12 Maintenance of NCS controls. Ensure that passive engineered, active engineered, and administrative nuclear criticality safety means of control are in place and functioning satisfactorily. 5.1.2.13 Fire safety plans. Accept responsibility for the development of a facility fire safety plan that recognizes, to the extent necessary, both fire safety and nuclear criticality safety considerations as specified in DOE Order 6430.1 A. These considerations should address the possible use of water or other moderator/reflector influences, the possibility of affecting the IIANSI/ANS-8.19-1996, section 7.8. DOE G 421.1-1 8-25-99 31 accumulation of fissionable material, and the required presence of fire fighters in the fissionable material operations area. 12 5.1.2.14 Operational postings. Establish and maintain nuclear criticality safety posting for the facility and labeling of fissionable materials.13 5.1.2.15 Delegation of responsibilities. Delegate nuclear criticality safety responsibilities to lower level facility supervision; however, overall responsibility for facility nuclear criticality safety remains with the facility manager. 5.1.2.16 Development of criticality accident evacuation routes. Ensure that criticality accident evacuation routes provide for timely facility evacuation, that facility changes do not unnecessarily impede or otherwise lengthen evacuation time, and that, to the extent practical, routes do not require personnel to approach potential sites of a criticality accident.14 5.1.2.17 Monitoring for process accumulations. Provide monitoring or surveillance, or both, to forewarn of unacceptable or unsafe accumulations of a significant quantity of fissionable materials in process equipment, storage areas, piping, and ventilation systems, thus permitting normal corrective actions. If unacceptable or unsafe accumulations of a significant quantity of fissionable materials are detected, corrective actions should be taken in conjunction with the area Criticality Safety Organization. 15

5.1.2.18 Facility access and other NCS controls. Provide other nuclear criticality safety features and administration as necessary to provide for the nuclear criticality safety of the facility, including personnel training, familiarization, and qualification for nonreactor nuclear facility access control of both assigned and incidental personnel. 5.1.3 Line/Production Management. Line/production management shall accept the direct responsibility for, and exercise authority over, the application of nuclear criticality safety to their operations by addressing the following program elements:16 5.1.3.1 Acceptance of authority and responsibility. Accept the authority and responsibility for nuclear criticality safety for facility operations under their control to include the implementation of nuclear criticality safety responsibilities as delegated by the contractor President. 5.1.3.2 Standards compliance. Ensure that applicable nuclear criticality safety standards and DOE requirements are applied in the design, modification, and operation of facilities under their control. Means to ensure that this requirement is met potentially include the use of configuration control 12ANSIIANS-8.1-1983,R88, section 4.1.7; and ANSI/ANS-8.19-1996, section 9.5. 13ANSI/ANS-8.1-1983,R88, section 4.1.4; and ANSI/ANS-8.19-1996, section 9.2. 14ANSI/ANS-8.19-1996, section 10.3. 15DOE 420.1, Section 4.3. 16ANSI/ANS-8.19-1996, section 5. 32 DOE G 421.1-1 8-25-99 boards, design reviews, technical reviews, operational readiness reviews, self-assessments/audits, and training, as applicable. Auditable means for demonstrating such compliance should be provided. 5.1.3.3 Operational approvals. Ensure that all operations within the facility are approved by Line/Production Management based upon current nuclear criticality safety analyses and approvals as provided by the Criticality Safety Organization. 5.1.3.4 Staffing and training. Ensure adequacy of staffing and that personnel assigned to work in the facility are adequately trained in specific job tasks and qualified in the procedures for working safely with fissionable materials in accordance with the reference document in paragraph 2.3.1.12. 5.1.3.5 Configuration control. Establish and conduct a configuration control program to ensure that facility modifications to the structure, utilities, operations, or equipment therein that may affect nuclear criticality safety are approved. 5.1.3.6 Procedures. Conduct the testing, start up, operation, emergency control, and corrective/preventive maintenance of the facility in accordance with approved procedures. 5.1.3.7 Maintenance. Ensure that sampling, measurement and control instrumentation, and safety monitoring capabilities are provided and maintained operational. 5.1.3.8 Self-audit. Report, investigate, and document unplanned events and unusual occurrences in accordance with DOE 0 232.1, formerly DOE Order 5000.3B. 5.1.3.9 Emergency planning. Participate in, and concur with, planning for emergency response to fires and criticality accidents. 5.1.3.10 Documentation. Ensure that applicable nuclear criticality safety files are maintained, for example: • operational and equipment approvals, • technical specifications, • auditable records of modifications, • operating reviews, • procedure reviews, • maintenance, • internal audit program, and • internal training.

5.1.3.11 Notifications. Notify the Criticality Safety Organization if any building or process modifications are planned that could interfere with the performance of a CAS, could require a change in the location of a detector, could require any additions to the CAS, or could otherwise affect the system. 5.1.3.12 Review requests. Bring matters requiring INCSRC review to the attention of the Committee, and solicit the Committee's guidance regarding cases where the need for Committee review is uncertain. DOE G 421.1-1 S-25-99 33 5.1.3.13 Delegation of authority and assignment of responsibilities. Delegate the authority and assign the responsibility for the day-to-day nuclear criticality safety of operations at an installation to First Line Supervision. In this regard, Line/Production Management should review the performance of First Line Supervision with respect to nuclear criticality safety on an annual basis (noting nuclear criticality safety related occurrences, limit violations, commendable practices, and others), assigning importance to nuclear criticality safety commensurate with the other aspects of process operations. 5.1.4 First Line Supervision. The First Line Supervision implements safety related responsibilities that are delegated by upper management and has further responsibilities to: 17 5.1.4.1 Responsibility. Accept responsibility for the nuclear criticality safety of operations under their control. 18 5.1.4.2 Training. Be knowledgeable in those aspects of nuclear criticality safety relevant to operations under their control as required by corporation/company training procedures for compliance with the applicable document. 19 5.1.4.3 Provision of training. Ensure that nuclear criticality safety training is provided to personnel under their control in accordance with corporation/company training procedures for compliance with the applicable document and require that these personnel have procedures and operating conditions necessary to perform their functions without undue risk. Records of training activities and verification of personnel understanding shall be maintained.20 5.1.4.4 Procedural development. Develop, or participate in the development of, written procedures applicable to the operations under their control. Maintenance of these procedures to reflect changes in operation should be a continuing supervisory responsibility.21 5.1.4.5 Safety practices. Require conformance with good safety practices including unambiguous identification of fissionable materials and good housekeeping.22 5.1.4.6 Operational reviews. Review all proposed new operations, facility modifications, and process and equipment changes involving significant quantities of fissionable material or nuclear criticality safety. Verify compliance with nuclear criticality safety specification for new and modified equipment prior to its use. 17 ANSI/ANS-S.19-1996, section 5. 18ANSI/ANS-S.19-1996, section 5.1. 19ANSI/ANS-S.19-1996, section 5.2. 20ANSI/ANS-S.19-1996, section 5.3. 21ANSI/ANS-S.19-1996, section 5.4. 22ANSI/ANS-S.19-1996, section 5.6. 34 DOE G 421.1-1 B-25-99 5.1.4.7 Operational approvals. Ensure that all operations within the facility are approved by Line/Production Management based upon current nuclear criticality safety analyses and approvals as concurred by the Criticality Safety Organization. 5.1.4.B Process monitoring. Monitor operations to verify compliance with nuclear criticality safety requirements.

5.1.4.9 Recovery and deviation evaluations. Evaluate all criticality safety specification or procedural limit violations and deviations, and concur with proposed recovery and corrective actions except for emergencies requiring immediate response. 23 5.1.4.10 Labeling and posting. Ensure that appropriate material labeling and area posting are maintained, specifying material identification and all operational/process limits on parameters that are subject to procedural control. Refer to paragraph 2.1.B for posting and labeling details. Posted operational and/or process limits are to be doable by fissionable material handlers.24 5.1.4.11 Access control. Access to areas where fissionable material is handled, processed, or stored shall be controlled. 25 5.1.5 Fissionable Material Operations Personnel. All personnel working with fissionable material shall: 5.1.5.1 Responsibility. Be responsible for nuclear criticality safety of their own actions and the operating systems under their control. 26 5.1.5.2 Operational procedures. Conduct fissionable material operations in strict accordance with approved written procedures and instructions. 27 5.1.5.3 Terminate operations. In the event an unforeseen condition develops and a procedure does not correspond to the operating system, return operations to a known safe stopping point or stop performing the procedure and notify supervision.28 5.1.5.4 Inquiries. Ask supervision for additional training, guidance, instructions, or procedures when uncertain of the nuclear criticality safety of job tasks involving fissionable materials. 23ANSI/ANS-B.1-19B3,RBB, section 4.1.7. 24ANSI/ANS-B.1-1983,RB8, section 4.1.4; and ANSI/ANS-8.19-1996, sections 9.2 and 7.6. 25ANSI/ANS-B.19-1996, section 9.4. 26ANSI/ANS-B.1-1983,RBB, section 4.1.1. 27ANSI/ANS-8.1-1983,R8B, section 4.1.3. 28ANSI/ANS-8.19-1996, section 7.7. DOE G 421.1-1 S-25-99 35 5.1.5.5 Training requirements. Complete and periodically update applicable nuclear criticality safety training in accordance with corporation/company procedures developed to comply with paragraphs 5.1.3.4 and 5.1.4.2 of this Guide. 5.1.5.6 Notification. Communicate information and concerns to co-workers and management as appropriate. 5.1.5.7 Emergency response. Know and follow emergency procedures. 5.1.6 Facilities Maintenance Organization. Through personnel NCS and facility access control training and use of a formal maintenance work permit program, the maintenance organization should ensure that any engineered criticality safety controls have been identified and will not be disturbed or inadvertently deactivated prior to performing work. The maintenance organizations should also verify with facility personnel that it is safe to perform work on an engineered criticality safety control prior to performing such work. 5.1.7 Engineering and Projects Organization (E&PO). The E&PO is responsible for the performance of oversight of design, procurement, and construction of facilities used for the processing, storage, or transport of fissionable material. 5.1.7.1 Responsibilities. E&PO should accept responsibilities in the area of nuclear criticality safety as delegated by the corporation/company President through corporation/company policies, procedures, and practices for nuclear criticality safety design control.

5.1.7.2 Design requirement compliance. In addition to those design responsibilities identified in paragraph 5.1.7.1, E&PO should comply with the nuclear criticality safety design requirements contained in DOE Order 6430.1 A and shall comply with the nuclear criticality safety requirements contained in DOE Order 420.1, Section 4.3 and the ANSI/ANS Standards referenced within DOE Order 420.1, Section 4.3. 5.1.8 Criticality Safety Organization. The Criticality Safety Organization accepts and implements responsibilities delegated by the corporation/company President and as described in corporation/company policies. These responsibilities shall, as a minimum and when applicable and appropriate, include: 5.1.S.1 Technical Direction. Technical direction consists of documenting the nuclear criticality safety program as policies and procedures to implement the elements of the ANSI/ANS standards specified in DOE Order 420.1, Section 4.3. a. ANSI/ANS-S.1-199S, "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," except paragraphs 4.2.2 and 4.2.3, and paragraph 3.3; b. ANSI/ANS-S.3-1997, "Criticality Accident Alarm System," except paragraphs 4.1.2, 4.2.1, and 4.2; c. ANSIIANS-S.5-1996, "Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Materials"; d. ANSI/ANS-S.6-19S3,R95, "Safety in Conducting Subcritical Neutron-Multiplication Measurements in Situ," except paragraph 5.3; 36 e. f. g. h. i. j. k. I. m. DOE G 421.1-1 S-25-99 ANSI/ANS-S.7-199S, "Guide for Nuclear Criticality Safety in the Storage of Fissile Materials, n except paragraph 5.2; ANSI/ANS-S.9-19S7,R95, "Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials"; ANSI/ANS-S.1 0-19S3,RSS, "Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement"; ANSI/ANS-S.12-19S7,R93, "Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors"; ANSI/ANS-S.15-19S1,R95, "Nuclear Criticality Control of Special Actinide Elements"; ANSI/ANS-S.17-19S4,R97, "Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors," except paragraph 4.3; ANSI/ANS-S.19-1996, "Administrative Practices for Nuclear Criticality Safety"; ANSI/ANS-S.20-1991, "Nuclear Criticality Safety Training"; ANSI/ANS-S.21-1995, "Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors"; n. ANSI/ANS-S.22-1997, "Nuclear Criticality Safety Based on Limiting and Controlling Moderators"; o. ANSIIANS-S.23-1997, "Nuclear Criticality Accident Emergency Planning and Response." 5.1.S.2 Quality assurance. Maintain procedures to ensure that accurate nuclear criticality safety analyses for intended fissionable material operations are performed and documented and any identified required controls are implemented for those operations. Ensure that such documentation is consistent with the requirements of the reference in paragraph 2.1.17. 5.1.S.3 Criticality Safety Organization personnel qualifications. Establish criteria (see paragraphs 5.2.2, A.2 and 5.2.2, A.2) for qualifying Criticality Safety Organization staff and qualify and periodically ensure the qualifications of staff performing or reviewing topics such as (a) analyses of fissionable material process NCS event/fault-trees, (b) evaluations of NCS control effectiveness, (c) NCS evaluations, (d) NCS analyses,

(e) analyses of nuclear criticality accident alarm/detection system placement, (f) evaluations of nuclear criticality accident evacuation zones, DOE G 421.1-1 8-25-99 (g) audits of fissionable material processes and violation/procedural reviews, 37 (h) quality and configuration control of software and data sets used for NCS evaluations and for nuclear criticality accident alarm or detection system placement and evacuation zone evaluations, (i) accident or unusual occurrence investigations or root cause analyses, and (j) training for criticality safety. 5.1.8.4 Maintenance of familiarity. Maintain familiarity with (a) current and developing nuclear criticality safety standards and guides, (b) nuclear criticality safety computational codes to the extent that personnel need to use or interpret the codes or their results, and (c) all operations within the corporation/company installation that require nuclear criticality safety controls. Familiarity may be gained through a combination of reading, tours, training, inspections, calculations, and periodic assignments to a facility as appropriate. 5.1.8.5 Consultation. Provide consultation and technical guidance by a. advising corporation/company management of requirements relating to nuclear criticality safety and determining if the nuclear criticality safety program is consistent with this Guide, b. assisting in the development, review, and concurrence of operating procedures, and procedure changes affecting nuclear criticality safety, c. assisting and advising in equipment and process design, review, and concurrence in process and equipment changes affecting nuclear criticality safety, particularly passive engineered controls and active engineered controls, d. concurring in the approval of corporation/company-wide nuclear criticality safety related policies, procedures, manuals, and instructions written for conformance to regulations issued by organizations such as the U.S. Department of Energy (DOE), U.S. Department of Transportation (DOT), U.S. Nuclear Regulatory Commission (NRC) and the U.S. Code of Federal Regulations (CFR), e. assisting with formal nuclear criticality safety related communications between the corporation/company and external organizations such as in requests for exemptions or non-adherences to DOE Orders or other regulations, f. assisting in the development and execution of nuclear criticality safety training programs, and g. assisting in evaluation of unusual occurrences. 38 DOE G 421.1-1 8-25-99 5.1.8.6 Obtaining consultation. Obtain consultation with knowledgeable individuals to obtain technical assistance as needed. 5.1.8.7 Performing self-assessments and audits. Conduct or participate in installation self­ assessments or audits (at least annually) of nuclear criticality safety practices and compliance with procedures to ascertain that procedures are being followed and that process conditions have not been altered so as to affect the nuclear criticality safety of operations and the relevance of nuclear criticality safety evaluations to the operations. 29 5.1.8.8 Operational reviews. Perform walk-through inspections of facilities with nuclear criticality safety approvals. This walk-through may be done in conjunction with other facility audits, if implemented, and should include consideration of nuclear criticality safety practices and compliance with procedures. The frequency of such walk-throughs should be increased in relation to the potential for criticality accidents, degree of controls required, and level of activity ongoing, and should be specified by facility-, site-, or installation-specific procedure or policy documents. However, all such facilities should be inspected annually; facilities covered by a CAS should be inspected quarterly.

5.1.8.9 Procedural reviews. Review new or revised procedures affecting nuclear criticality safety. 5.1.8.10 Incident reviews. Review installation operating or procedural NCS incidents for root causes, possible improvement of safety practices, and procedural requirements; report findings to management. (See paragraph 2.1.14.) 5.1.8.11 Process event/fault tree analyses. Review and concur in, or, as trained, qualified, and requested by management, provide fissionable material process event/fault tree analyses in support of, nuclear criticality safety evaluations and analyses. 5.1.8.12 Quality and configuration control of software and data sets. Provide for software and data set verification, validation, and configuration control for criticality and radiation shielding computational methods used in nuclear criticality safety evaluations and, as applicable, for nuclear criticality accident alarm and detection system (CAS and CDS) placement evaluations. 5.1.8.13 NCS evaluations. Use quality controlled, configuration controlled, verified, and validated software and data sets; handbook techniques and data shown to be valid; or direct comparisons with critical and subcritical experiment data. Perform nuclear criticality safety evaluations to demonstrate technically the subcriticality of fissionable material processes, operations, and situations for transportation and storage under normal and credible abnormal conditions. 30 See Appendix B for discussion of a graded approach to NCS evaluations. 5.1.8.14 NCS analyses. Using results of a structured analysis process (e.g., process event/fault tree, HAZOP, what-if, human reliability, failure modes and effects, MORT) and NCS evaluations, perform and document nuclear criticality safety analyses for proposed new or modified fissionable material processes as requested by facility line/production management. The safety analyses shall contain information to demonstrate compliance with applicable requirements for the prevention of 29ANSI/ANS-8.19-1996, section 7.8. 30ANSI/ANS-8.1-1983,R88, section 4.3. DOEG421.1-1 8-25-99 39 inadvertent criticality and mitigation of consequences from a criticality accident. See Appendix B for discussion of a graded approach to NCS analyses. 5.1.8.15 NCS recovery actions. Respond to, and assist in, recovering from contingencies and any other discoveries that the basis for criticality safety is invalid by supplying advice and commenting on proposed actions in addition to performing and documenting NCS evaluations and analyses as described above. Initial advice should focus on whether it is necessary to declare an emergency, and if so, how to deal with the emergency. Generally, loss of all protection against criticality constitutes an emergency; exceptions include instances where shielding protects workers and equipment vital to safety against the dose consequences of a criticality accident. If the situation does not warrant declaring an emergency, advice should first focus on immediate or near term vulnerabilities that, if left uncorrected, could lead to declaring an emergency. 5.1.8.16 CAS and CDS sensors, Nuclear Accident Dosimeters (NADs), and evacuation zone boundary shielding and deployment evaluation. Participate in the performance of evaluations for placement of CASs and CDSs sensors, NADs, and evacuation zone boundaries for credible nuclear criticality accident source terms using appropriate criticality and shielding codes, NRC Regulatory Guides, or other techniques such as charts, nomographs, and hand calculations, as necessary, providing a peer review of results, documenting the results, and maintaining records of the evaluation.

5.1.8.17 Peer review. Provide peer reviews of (a) fissionable material process analyses affecting NCS, (b) NCS evaluations, (c) NCS analyses, (d) CAS, CDS, and NAD deployment evaluations, (e) nuclear criticality accident evacuation zone evaluations, and (f) quality and configuration control of software and data sets used for NCS evaluations and nuclear criticality accident alarm or detection system placement and evacuation zone evaluations. 5.1.8.18 Records retention. Ensure the maintenance of records during the period of their applicability and at least for periods specified in paragraph 2.1.2 for (a) NCS analyses, (b) CAS, CDS, and NAD placement evaluations, (c) nuclear criticality accident evacuation zone evaluations, (d) fissionable material process audits and violation/procedural reviews, and 40 DOE G 421 .1-1 8-25-99 (e) quality and configuration control of software and data sets used for NCS evaluations and nuclear criticality accident alarm or detection system placement and evacuation zone evaluations. 5.1.8.19 NCS procedures. Prepare, maintain, and interpret policy, standards, guidelines, and implementation procedures for installation nuclear criticality safety requirements. 5.1.8.20 Design reviews. Provide independent safety reviews of documents significant to criticality safety for all new facilities and significant modifications that affect nuclear criticality safety in existing facilities. Documents include training plans, design criteria, project design documentation, technical specifications, procedures, and drawings. 5.1.8.21 Selection and approval of effective controls. With support and approval from first line supervision, effectively select controls for nuclear criticality safety as identified by the nuclear criticality safety analysis. 5.1.8.22 Assistance in shutdowns. Technically assist line supervision in safely suspending fissionable material operations that, in the judgment of line supervision or the Criticality Safety Organization, do not have the required level of nuclear criticality safety. 5.1.8.23 Accident yield estimation. Provide for the evaluation of yield of credible nuclear criticality accidents bounding facility or installation fissionable material processes. 5.1.8.24 Participate in CAS evacuation drills. In cooperation with Operations, monitor and comment upon CAS evacuation drills, emergency procedures, re-entry procedures, and practices. 5.1.8.25 Technical training support. Review, approve, and provide technical information, as requested, for general and facility-specific nuclear criticality safety training curricula. 5.1.8.26 Reviews of fire safety plans. Review and concur in the fire safety plan for each facility having significant quantities of fissionable materials. 5.1.8.27 Operational experience feedback. Examine reports of procedural violations and other deficiencies for potential improvements of safety practices and procedural requirements, and report such potential improvements to management. Collect, analyze, and examine for trends of operational experiences relating to criticality safety to determine if the assumptions in NCS evaluations and analyses are valid, too conservative, or otherwise need revising. Report priority needs for revisions to management.31 5.1.8.28 INCSRC support. Provide nuclear criticality safety experienced personnel to serve on, and assist, the Installation Nuclear Criticality Safety Review Committee, as requested.

5.1.9 Installation Nuclear Criticality Safety Review Committee (lNCSRC). The INCSRC should provide the installation site manager with appraisals of the programmatic effectiveness of the installation operating and nuclear criticality safety organizations. 31ANSI/ANS-8.19-1996, section 6.7. DOE G 421.1-1 8-25-99 41 5.1.9.1 Composition. Should be chaired by an individual who reports to the Installation Manager or a sufficiently high level of management, and should be independent, to the extent practicable, of fissionable material Operations and the Nuclear Criticality Safety Organization. Committee membership may include individuals representing the installation for (a) fissionable material operations and waste management, (b) fissionable material operations development and engineering, (c) fissionable material accountability and security, (d) emergency preparedness, (e) installation maintenance and support, and (f) Criticality Safety Organization. 5.1.9.2 Guidance to management. Provide guidance to management for principles and policy of the installation nuclear criticality safety program including the resolution of any conflicting interpretations of NCS policies and procedures. 5.1.9.3 Investigation of incidents. Participate, as requested by line management, in the investigation of criticality safety incidents that are classed as "violations," in accordance with the reference in 2.1.3. (See also 2.1.14.) 5.1.9.4 INCSRC Annual reviews. INCSRC, or an alternative means of review shall conduct periodic reviews of the Installation NCS program.32 These reviews should be conducted at least annually and include applicable items from the following elements: a. Presentations by the installation NCS Organization regarding the status of (1) any proposed nuclear criticality safety program policy changes; (2) activities of the NCS Organization with respect to (i) pertinent aspects of continuing facility operations and of major new programs or modifications, (ii) applicable audit recommendations and suggestions resulting from audits of the nuclear criticality safety program, (iii) participation in installation personnel education and training efforts, and (iv) development, training, and qualification of NCS personnel; (3) compliance with DOE Orders and other federal, state, or municipal regulations regarding NCS as evidenced by documented periodic reviews and random surveillances; and (4) configuration control program for the conduct of Criticality Safety Organization operations. b. Presentations by operating line management responsible for, or affecting, the installation NCS program regarding 32ANSI/ANS-8.19-1996, sections 4.5 and 4.6. 42 DOE G 421 .1-1 8-25-99 (1) inspections, audits, and self-assessments including follow-up on corrective actions and recommendations from incident investigations, (2) personnel training, qualification, and certification, (3) administration of nuclear criticality safety controls including methods, accomplishments, and procedures, (4) installation and facility access control, (5) process and facility engineering design development, construction, and maintenance activities, (6) configuration control of processes, equipment, and facilities important to NCS, (7) ancillary support equipment and utilities affecting fissionable material operations, storage, or transportation, (8) operations, storage, and transportation of fissionable materials, (9) emergency preparedness and drills, and (10) compliance with statutory requirements regarding NCS as evidenced by documented

self appraisals. c. A physical inspection of the fissionable material control areas for the purpose of focusing on specific program review topics. d. Reviews of installation and facility operating experiences, including operating anomalies and NCS incidents along with incident investigation and prior review responses and follow-up by line management and the NCS organization. e. Review of criticality accident alarm and detection systems performance. f. The documentation of the annual review in a report provided to the Installation Manager that contains the items of review, findings, and recommendations. g. A summary presentation of the annual review report to the Installation Manager. 5.1.9.5 Response to requests. Review on an as-needed or requested basis (a) any proposed criticality safety program policy change, making recommendations to management, and (b) any issues or concerns that should properly come before the INCSRC. DOE G 421.1-1 8-25-99 43 5.2 PERSONNEL SELECTION, QUALIFICATION, TRAINING, AND STAFFING PROGRAM. The purpose of the program is to establish (develop and document) the selection, qualification, training, and staffing requirements for personnel such that persons are qualified to carry out their assigned responsibilities, that they have a broad understanding and acceptance of the inherent risks involved with the operations, and that they maintain a job performance proficiency consistent with effective control of the hazards and risks associated with the operations. Three broad categories of the program are considered in this Guide. The categories are (1) the operations and support personnel associated with fissionable material operations outside of reactors, (2) the installation nuclear criticality safety staff, and (3) visitors and clerical employees. The personnel selection criteria and depth and breadth of nuclear criticality safety training are necessarily variable, depending on the work assignments of personnel. A detailed discussion of this program is found in Appendix A, which is intended to provide guidance for organizations establishing new programs or improving current programs. This guidance is presented in an a posteriori form, expressly to emphasize that the specificity of structure and nomenclature for personnel selection, qualification, training, and staffing is illustrative and suggestive rather than recommendatory. 5.2.1 Program for Operations and Support Personnel. The category of operations and support personnel includes fissionable material handlers and their supervisors, operations support, design, maintenance, technical support (including the members of the Nuclear Criticality Safety Organizations) and emergency response personnel, managers and other administrative personnel, and persons who enter areas where fissionable material is processed, stored, or handled. As consistent with job assignments and personnel acknowledgement of job hazards and risks, the following items should be considered for inclusion in the training and qualification program. • Establish the training and qualification program to provide continuing proficiency of personnel. • Discuss the concept of a nuclear fission chain reaction. • Describe neutron induced fission, neutron capture, and neutron scattering and leakage. • Review and describe selected criticality accidents. • Train personnel in the recognition of, and the response to, criticality accident alarms.

• Explain and illustrate the influence of various nuclear criticality safety parameters on process safety. • Describe the facility management's nuclear criticality safety policy. • Periodically, perform and document evaluations of the training program and trained personnel. 5.2.2 Installation Nuclear Criticality Safety Staff. This category includes the manager and members of the installation Criticality Safety Organization who are responsible for performing computational or comparative evaluations and safety analyses for fissionable material operations; for developing procedural, process, and control requirements; and for providing procedural, process, and equipment/facility reviews and approvals, nuclear criticality safety training program development, and facility operational reviews, appraisals, audits, and investigations. The 44 DOE G 421 .1-1 8-25-99 professional personnel charged with implementing the programs identified in this Guide are designated nuclear criticality safety specialists (NCSS). There are currently only general qualification requirements33, but ongoing and future qualification of individuals should consider developing confirmable documentation that addresses, but is not limited to, the following (see Appendix A for more detail): • A demonstrated capability to perform installation-specific analyses of the NCSS job and its tasks for experienced NCSS personnel. • A qualification checklist, file, card, or other record that identifies each applicable task and the method(s) by which competence has been demonstrated, with performance evaluation based on actual or representative work products. • A baseline education of a baccalaureate degree in engineering or science and a minimum experience in nuclear criticality safety at the facility of one (1) year to independently perform NCSS tasks, and three (3) years to provide independent review and quality assurance of NCS tasks. Equivalencies may be established. • Certification of final qualification by criticality safety management. • Periodic competence confirmations based on practical exercises. First-hand knowledge of the situation of criticality should be the goal of all criticality safety practitioners that comprise the Installation Nuclear Criticality Safety Staff. Such knowledge should be meaningfully obtained through experience in performing criticality experiments, preferably on a routine basis, e.g., as a staff member at a critical mass laboratory for an extended period. When feasible, such knowledge should be obtained in this manner. Nuclear criticality safety specialists are collectively the professional staff with primary responsibility for implementing the activities and programs required to support this Guide. 5.2.3 Visitors and Clerical Employees. As a minimum, visitors and clerical employees entering fissionable material control areas without escort should have been instructed in the identification of the criticality accident alarm system (CAS) signals (audible and/or visible), instruction on the requirement for immediate evacuation if in an area in which a CAS alarms sounds, identification of evacuation routes, and an explanation as to why a CAS alarm is necessary. These persons should also have been instructed to refrain from all actions involving the movement, processing, or storage of fissionable material.34

5.2.4 Auditor Qualifications. Nuclear criticality safety program auditors shall have adequate education, knowledge, training, and experience to review and evaluate the elements of a nuclear criticality safety program in terms of content and adequacy for its intended application. 330 0 E Order 5480.20A, Chapter I, 7.g. 3400E Order 5480.20A, Chapter I, 7.e. DOE G 421.1-1 8-25-99 45 5.2.4.1 Compliance. A compliance auditor of a nuclear criticality safety program should have a demonstrated basic knowledge of DOE Orders, associated statutory requirements, and industry standards and practices in order to recognize the equivalency or adequacy of documented and observed program procedures and practices (i.e., compliance) with the DOE Orders. This knowledge should be demonstrated by either • three years in the administration and management of a non-reactor nuclear facility nuclear criticality safety program, or • education, training, and testing (developed from job/task analyses) in the above subject matter. 5.2.4.2 Quality Audit. An auditor of nuclear criticality safety program quality should have seven years experience in nuclear criticality safety leading to broad knowledge and applications. In addition, such auditors should have specific experience in the particular subject to be audited. Examples of this specific experience may include, but are not limited to, • physicochemical operations and administrative controls used in the processing, handling, transport, or storage of fissionable materials, and typical or experiential upset/contingent conditions of these operations, • computational physics as it relates to computational modeling, use and/or processing of neutron cross sections, and computer code verification and validation, • safety analysis techniques such as failure modes and effects analysis, what-if analysis, management oversight risk tree analysis, etc., • human factors influences on processes, • conduct of inspections, self-assessments, and audits, • emergency preparedness, and • criticality safety training subject matter for operators, supervisors, managers, visitors, etc. 46 DOE G 421.1-1 B-25-99 5.3 OPERATING, STORING, AND TRANSFERRING - PLANS, PROCEDURES, REQUIREMENTS, AND CONTROLS. Operations for which nuclear criticality safety is a consideration shall be addressed by a Nuclear Criticality Safety Evaluation and shall be governed by written plans, written procedures, and controls. 35 5.3.1 General Requirements for Operating Plans and Procedures. 5.3.1.1 Start-up, operations, and modifications. Written plans and procedures for facilities in which nuclear criticality safety is of concern shall cover start-up, operations, and any modifications that may affect nuclear criticality safety. All persons participating in the operation of such facilities shall be familiar with, and understand procedures applicable to, their assigned duties. In this Guide, operating plans and procedures include any set of instructions to do work that can affect criticality safety. They include process operating procedures, storage plans, and modification or maintenance work packages that involve significant quantities of fissionable material, associated materials, engineered safety features, and the CAS.36

5.3.1.2 NCS parameter identification. Procedures shall clearly -specify all nuclear and process parameter limits related to nuclear criticality safety that are intended to be controlled for safety. Nuclear criticality safety steps should be conspicuously identified in operating procedures and should immediately precede the step or group of steps to which they are applicable. New or revised procedures containing such nuclear safety steps, nuclear criticality safety limits, or nuclear criticality safety requirements shall undergo review by the Criticality Safety Organization prior to implementation.37 5.3.1.3 Single failure safety assurance. Procedures should be developed such that no single credible inadvertent departure from a procedure can cause a criticality accident. 5.3.1.4 Procedural convenience. Procedures should be convenient for use by operators and should be free of extraneous material. 5.3.1.5 Procedural reviews. Active procedures shall be reviewed periodically by supervision. The requirement to periodically review active procedures shall itself be a procedure to define the review frequency. Similarly, operations shall be reviewed at least annually to ascertain that procedures are being followed, and that process or facility conditions have not been altered so as to affect nuclear criticality safety adversely. 38 35ANSI/ANS-B.1-19B3,RBB, sections 4.1.2 and 4.1.3; and ANSI/ANS-B.19-1996, sections 7 and B. 36ANSI/ANS-B.1-19B3,RBB, section 4.1.3; and ANSI/ANS-B.19-1996, sections, 5.3, 5.4, and 7. 37 ANSI/ANS-B.1-19B3,RBB, section 4.1.3; and ANSI/ANS-B.19-1996, section 7.5. 38A NSI/ANS-B.1-19B3,RBB, section 4.1.6; and ANSI/ANS-B.19-1996, sections 5.4, 7.4, and 7.B. DOE G 421.1-1 8-25-99 47 5.3.1.6 Supplementation. Procedures should be supplemented as appropriate by posted nuclear criticality safety limits or other appropriate operator aids such as inventory lists, process checklists, flowsheets, and engineering drawings as part of an operator aid program in accordance with references cited in this document.39 40 Suitable allowances may be made for situations where the fissionable material content of products, wastes, or feed materials, or such content under other circumstances, is repetitive or is previously established from the work of others or from process limitations. Examples could include the acceptance of shipper's values for received materials or limitation of material density or concentration because of a specific chemical process. 5.3.1.7 Operational deviations. Deviations from operating procedures and unforeseen alterations in process conditions that affect nuclear criticality safety shall be documented, reported to management, and investigated promptly. Actions shall be taken to prevent a recurrence or to appropriately modify procedures. 41 5.3.1.8 Procedural revisions. Supplementing and revising procedures shall be facilitated as improvements become desirable.42 5.3.2 Processing. Fissionable material processing shall be conducted in an orderly fashion that includes, as appropriate, use of the following: 43 5.3.2.1 Plans, flowsheets, and layouts. Process plans, flow sheets, and layouts should be developed that describe the process, including equipment and facilities in which criticality hazards may exist, using appropriate drawings/sketches, and including dimensions in sufficient detail to permit the development of procedures specified in paragraphs 5.3.2.2 through 5.3.2.6 to evaluate the process.

5.3.2.2 Procedural description of material composition. Procedures shall be developed and used that contain information on the physical and chemical form of fissionable material in the processing operation, including isotopic content, concentrations, densities, and moderation levels of the fissionable material, as applicable and required to ensure criticality safety. This information may be bounded by conservative enveloping assumptions to simplify and eliminate superfluous details. 5.3.2.3 Procedural description of allowed material quantities. Procedures should be developed and used that contain statements of the maximum quantities or concentrations of fissionable material allowed in the process. 39DOE 5480.19. 40ANSI/ANS-8.1-1983,R88, section 4.1.4. 4IANSI/ANS-8.19-1996, section 7.7. 42ANSI/ANS-8.19-1996, section 7.3. 43ANSIIANS-8.19-1996, sections 5.6 and 9.5. 48 DOE G 421.1-1 8-25-99 5.3.2.4 Procedural description of spacing requirements. Procedures should be developed and used that specify required spacing of masses of fissionable material and separation from fissionable material in adjoining areas, as applicable. 5.3.2.5 Procedural specifications for material collection and transport. Procedures should be developed and used that specify safe methods of collecting, handling, and transporting fissionable material. 5.3.2.6 Procedural specifications for administrative controls. Procedures should be developed and used that specify administrative methods to prevent criticality. 5.3.3 Receiving and Inspecting Fissionable Material. The receipt and inspection of fissionable materials shall be controlled by procedures which address:44 5.3.3.1 Verification. Procedures that are consistent with materials controls and accountability requirements should be developed and used for determining, verifying, or noting the contents of each package, including the net weight of the fissionable material therein. 5.3.3.2 Material placement. Procedures should be developed and used for placing fissionable materials in receiving areas and storage facilities. 5.3.4 Storing Fissionable Material. The requirements of this section do not apply (a) when materials are in-process as part of production, analytical and developmental procedures, or transport operations, (b) when an assembly cell is used for assembly and/or storage of weapons components made with these materials, (c) when the number of packages of materials prepared for shipment is limited in accordance with the requirements of DOE 0 460.1 A, formerly DOE Order 5480.3, or (d) to radioactive waste storage or disposal facilities. 5.3.4.1 Container design. Fissionable material container design should be appropriate to the form of stored fissionable material. Criteria for container integrity should be developed in the course of the required safety analysis and the applications of such criteria evaluated by periodic inspection (by facility personnel). For containers involving any significant gas buildup, automatic pressure relief or other venting should be designed to ensure that no personnel exposure to any toxic material will occur under normal storage conditions, or, insofar as practical, under credible accident conditions. Such venting should not permit the spread of contamination.

5.3.4.2 Container criteria. Criteria, such as external and internal corrosion rate for determining the suitability of containers in storage, should be developed as necessary and set forth in writing. These criteria are particularly important in water pool storage of fuel elements or containerized fissionable materials and in the storage of plutonium or 233U. All storage containers should be periodically inspected against the criteria developed. The time between inspections may vary depending upon storage container quality and type. Procedures for conducting these inspections or surveillances should include acceptance criteria for corrosion and other phenomena that can adversely affect criticality safety. 44ANSI/ANS-8.19-1996, sections 7.2 and 9.2. DOE G 421 .1-1 B-25-99 49 5.3.4.3 Container descriptions. Containers of fissile material in areas with sprinkler systems shall be designed to prevent the accumulation of water. 45 Procedures developed and used for storing fissionable material should contain descriptions or identify types of containers in which fissionable materials are allowed to be stored. 5.3.4.4 Container identification and closure. All fissionable material storage containers should be marked and, if practical, coded to indicate the type or category of material, amount, degree of enrichment, moderation, and the radiation level at the outside surface of the container as appropriate to monitor criticality safety parameter limits and controls. Containers should be securely closed and positioned so as to prevent significant displacement and maintain criticality prevention requirements. 5.3.4.5 Container venting. Plutonium containers in which gas buildup can occur should be designed to prevent leakage of gas over the maximum storage period, or vented to prevent an accumulation of explosive gases; however, such venting should not permit the spread of contamination. 5.3.4.6 Containerization of plutonium or 233U. Plutonium- or 233U-bearing, or -contaminated, material should be packaged in a closed metal container. Combustibles within the container should be minimized. Hydrogenous materials ("plastics") should not be used for plutonium packaging. These considerations may also be applicable to 233U. (See also DOE-STD-3013-94, paragraph 2.1.20 of this Guide.) 5.3.4.7 Plutonium storage monitoring. Plutonium storage facilities and containers should be monitored and checked periodically to ensure continued integrity of containment. When required by the form or hazard potential of the stored material, procedures should be developed to detect contamination or loss of primary containment when personnel enter the plutonium storage facility. 5.3.4.B Facility design criteria. The storage of fissionable materials shall be such as to obviate concern with accidental nuclear criticality in the event of fire or flood, earthquake, or other natural calamities.46 In addition, the design of storage structures should tend to preclude unacceptable arrangements or configurations, thereby reducing reliance on administrative controls.47 Where the presence of significant quantities of combustibles cannot be avoided, as in the storage of combustible fissile scrap, a fire protection system shall be installed.48 Where sprinkler systems are installed in fissile storage areas, consideration shall be given to the possibility of criticality occurring in an accumulation of runoff water.49 45ANSIIANS-B.7-1975,RB7, section 4.2.7. 46ANSI/ANS-B.7-1975,RB7, section 4.2.3.

47 ANSI/ANS-B. 7-1975,RB7, section 4.2.4. 48ANSI/ANS-B. 7-1975,RB7, section 4.2.6. 49ANSI/ANS-B.7-1975,RB7, section 4.2.B. 50 DOE G 421.1-1 8-25-99 5.3.4.9 Storage specifications. Procedures should be developed and used for storing fissionable material. 50 These should set forth limits on the total quantity of fissionable material, allowable quantity of individual units, allowable container dimensions, and required spacing of containers in storage areas. 5.3.4.10 Storage facility plans and layouts. Storage facilities and structures shall be designed, fabricated, and maintained in accordance with good engineering practices. 51 Plans and layouts should be developed that contain a description of the storage facility, including dimensions and materials used in construction of the enclosure and shelving, cubicles, cages, and other equipment within the storage area. 5.3.4.11 Admonitions about moderating and reflecting materials. Procedures developed for storing fissile material should contain precautions to avoid entry of water or other moderating materials into a storage area where moderating and reflecting effects of such materials would be unsafe. Nonessential combustible materials should not be stored in a fissionable material storage area. 5.3.4.12 Removal and return of materials. Procedures shall be developed and used that control the removal, or transfer, of fissionable material from storage and the return of such material to storage. These procedures should incorporate means of verifying the weight, isotopic content, chemical composition, and degree of moderation, as appropriate. 5.3.4.13 Exclusion of superfluous materials. Process operations, storage of non-nuclear materials or equipment that is not directly required for fissionable materials storage operations, and all other functions not directly a part of normal fissionable materials storage operations should be excluded from the storage area. 5.3.4.14 Readiness inspections. Documented inspections, in situ tests, and preventive maintenance shall be performed periodically on fissionable material storage areas to ensure that the safety systems and components necessary for criticality safety control are maintained in a state of readiness. 52 5.3.4.15 Postings. Nuclear criticality safety limits shall be conspicuously posted.53 Postings at the entrance and inside fissionable material storage areas, as applicable, should be considered. 5.3.4.16 Instructions. Signs or other devices should be utilized as appropriate at strategic locations in or near fissionable material storage locations to provide instructions regarding interpretations of, and required responses to, alarms, evacuation routes, and fire fighting. 50ANSIIANS-8.7-1975,R87, section 4.1.2. 51ANSIIANS-8.7-1975,R87, section 4.2.2. 52ANSI/ANS-8.7-1975,R87, section 4.2.2. 53ANSIIANS-8.7-1975,R87, section 4.1.2. DOE G 421.1-1 8-25-99 51 5.3.4.17 Emergency planning. In conjunction with site emergency planning, a fire fighting plan should be developed for fissionable material storage areas and incorporated into the overall facility and site plans. Periodic training drills/exercises should be conducted appropriate to the level of fire hazard associated with the area. 5.3.4.18 Exclusion from storage requirements. Excess fissionable material should not be construed to be "in process" to circumvent the fissionable material storage requirements of this section.

5.3.4.19 Use of shipping containers. Fissionable material may be stored in shipping containers for the purpose of enhancing safety in storage, but not for the purpose of negating the requirements of this section. 5.3.4.20 Material constraints. Fissionable material should be stored in racks or equivalent fixtures capable of securely locating stored material in order to prevent displacement, to ensure spacing control, and to meet designs for criticality safety under normal operational and credible accident conditions. Floor storage within a fissionable material storage facility should only be permitted where control of location or other safety requirements (equivalent to the safety provided by storage racks) are inherently provided by the original containers and their restraints if required for nuclear criticality safety. 5.3.4.21 Pyrophoric materials. All fissionable materials that are determined to be pyrophoric should be put in a safe form (i.e., non-pyrophoric) prior to storage or be stored in approved containers or inert atmospheres that will not permit spontaneous ignition or dispersal. 5.3.4.22 Heat removal. Provisions should be made in a plutonium storage facility to ensure necessary and adequate heat removal for plutonium storage containers as established by facility safety assessments. 5.3.5 Fissionable Material Transportation. The transportation of fissionable materials onsite and offsite shall be governed by written procedures that comply with DOE Order 420.1, Section 4.3; DOE 0 460.1A, formerly DOE Order 5480.3; 49 CFR; 10 CFR 71; and other applicable federal requirements. 5.3.5.1 Onsite transfers. The design and use of onsite shipping containers shall provide criticality safety protection of fissionable material consistent with that protection provided by DOE, NRC, or DOT packages used in interstate transport. Considerations should be given to onsite resources and conditions of material transport that eliminate or mitigate interstate transport hazards (e.g., resources of prompt fire fighting, speed limits of transport, traffic control, method of transport, compensation for weather conditions, lifting height restraints, and others). 5.3.5.1.1 Onsite transport safety analysis. The packaging requirements for onsite transfer of fissionable material are contained in DOE 0 460.1A, formerly DOE Order 5480.3. Safety analysis for onsite transfers shall be in accordance with requirements in this DOE Order. The safety analysis, computational evaluations, and the documentation of the package safety analysis shall be performed in accordance with DOE Order 420.1, Section 4.3. 5.3.5.1.2 Operating procedures. Approved operating procedures applicable to an onsite transfer or shipment of fissionable materials shall be posted or readily available within the loading, unloading, or storage areas for such materials. 52 DOE G 421 .1-1 8-25-99 5.3.5.2 Off site transfers. All transfers of fissionable materials offsite shall be performed in DOE, NRC, or DOT approved fissionable material packages. All required administrative controls and procedures specified for the package use shall be performed. Such DOE, NRC, or DOT approved packages do not require additional criticality safety review for receipt or shipment. 5.3.6 Posting and Labeling. Positive identification of fissionable material is essential to criticality safety. Adequate labeling of fissionable material and clear posting of work and storage areas in which fissionable materials are present are important in avoiding the accumulation of unsafe quantities of such materials. Detailed guidance for posting and labeling follow.

5.3.6.1 Posting of Fissionable Material Handling, Storage, and Work Areas. Posting refers to the placement of signs to indicate the presence of fissionable material, to summarize key criticality safety requirements and limits, to designate work and storage areas, or to provide instruction or warning to personnel. 5.3.6.1.1 Postings for presence of fissionable materials. The presence of significant quantities of fissionable material should be posted at the entrance to work and storage areas such as benches, hoods, glove boxes, cabinets, rooms, zones, and modules where fissionable material is handled, processed, or stored. Posting to identify the presence of significant quantities of fissionable materials may be at the entrance to work areas or storage areas, at room entrances, or entrances to buildings, as appropriate. Such areas shall be periodically reviewed to eliminate extraneous postings. For example, criticality safety requirements should be posted only for those glove boxes or rooms currently containing, or that are intended to contain, significa