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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 261
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Book Description
This document provides the documented safety analysis (DSA) and Central Plateau Remediation Project (CP) requirements that apply to surveillance and maintenance (S & M) activities at the 221-B Canyon Building and ancillary support structures (B Plant). The document replaces BHI-010582, Documented Safety Analysis for the B-Plant Facility. The B Plant is non-operational, deactivated and undergoing long term S & M prior to decontamination and decommissioning (D & D). This DSA is compliant with 10 CFR 830, Nuclear Safety Management, Subpart B, ''Safety Basis Requirements.'' The DSA was developed in accordance with U.S. Department of Energy (DOE) standard DOE-STD-1120-98, Integration of Environment, Safety, and Health into Facility Disposition Activities (DOE 1998) per Table 2 of 10 CFR 830 Appendix A, DOE Richland Operation Office (RL) direction (02-ABD-0053, Fluor Hanford Nuclear Safety Basis Strategy and Criteria) for facilities in long term S & M, and RL Direction (02-ABD-0091, ''FHI Nuclear Safety Expectations for Nuclear Facilities in Surveillance and Maintenance''). A crosswalk was prepared to identify potential inconsistencies between the previous B Plant safety analysis and DOE-STD-1120-98 guidance. In general, the safety analysis met the criteria of DOE-STD-1120-98. Some format and content changes have been made, including incorporating recent facility modifications and updating the evaluation guidelines and control selection criteria in accordance with RL direction (02-ABD-0053). The facility fire hazard analysis (FHA) and Technical Safety Requirements (TSR) are appended to this DSA as an aid to the users, to minimize editorial redundancy, and to provide an efficient basis for update.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 261
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Book Description
This document provides the documented safety analysis (DSA) and Central Plateau Remediation Project (CP) requirements that apply to surveillance and maintenance (S & M) activities at the 221-B Canyon Building and ancillary support structures (B Plant). The document replaces BHI-010582, Documented Safety Analysis for the B-Plant Facility. The B Plant is non-operational, deactivated and undergoing long term S & M prior to decontamination and decommissioning (D & D). This DSA is compliant with 10 CFR 830, Nuclear Safety Management, Subpart B, ''Safety Basis Requirements.'' The DSA was developed in accordance with U.S. Department of Energy (DOE) standard DOE-STD-1120-98, Integration of Environment, Safety, and Health into Facility Disposition Activities (DOE 1998) per Table 2 of 10 CFR 830 Appendix A, DOE Richland Operation Office (RL) direction (02-ABD-0053, Fluor Hanford Nuclear Safety Basis Strategy and Criteria) for facilities in long term S & M, and RL Direction (02-ABD-0091, ''FHI Nuclear Safety Expectations for Nuclear Facilities in Surveillance and Maintenance''). A crosswalk was prepared to identify potential inconsistencies between the previous B Plant safety analysis and DOE-STD-1120-98 guidance. In general, the safety analysis met the criteria of DOE-STD-1120-98. Some format and content changes have been made, including incorporating recent facility modifications and updating the evaluation guidelines and control selection criteria in accordance with RL direction (02-ABD-0053). The facility fire hazard analysis (FHA) and Technical Safety Requirements (TSR) are appended to this DSA as an aid to the users, to minimize editorial redundancy, and to provide an efficient basis for update.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 16
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Book Description
This document identifies administrative controls that are established for the operating contractor of the B Plant facility during the B Plant surveillance and maintenance phase. When approved by the DOE-RL, this document is part of the B Plant safety authorization basis. Look for the associated B Plant Surveillance and Maintenance Phase Safety Analysis Report, HNF-3358.
Author: T. Brown
Publisher:
ISBN:
Category :
Languages : en
Pages : 5
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Book Description
The objective of the Department of Energy's (DOE) Fernald Closure Project (FCP), in suburban Cincinnati, Ohio, is to safely complete the environmental restoration of the Fernald site by 2006. Over 200 out of 220 total structures, at this DOE plant site which processed uranium ore concentrates into high-purity uranium metal products, have been safely demolished, including eight of the nine major production plants. Documented Safety Analyses (DSAs) for these facilities have gone through a process of simplification, from individual operating Safety Analysis Reports (SARs) to a single site-wide Authorization Basis containing nuclear facility Bases for Interim Operations (BIOs) to individual project Auditable Safety Records (ASRs). The final stage in DSA simplification consists of project-specific Integrated Health and Safety Plans (I-HASPs) and Nuclear Health and Safety Plans (N-HASPs) that address all aspects of safety, from the worker in the field to the safety basis requirements preserving the facility/activity hazard categorization. This paper addresses the evolution of Safety Basis Documentation (SBD), as DSAs, from production through site closure.
Author: International Atomic Energy Agency
Publisher:
ISBN: 9789201215109
Category : Technology & Engineering
Languages : en
Pages : 0
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Book Description
On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
Author: Noel R. Kerr
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
Author: International Nuclear Safety Advisory Group
Publisher:
ISBN:
Category : Business & Economics
Languages : en
Pages : 118
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Book Description
The present report is a revision of Safety Series No. 75-INSAG-3 (1988), updating the statements made on the objectives and principles of safe design and operation for electricity generating nuclear power plants. It includes the improvements made in the safety of operating nuclear power plants and identifies the principles underlying the best current safety policies to be applied in future plants. It presents INSAG's understanding of the principles underlying the best current safety policies and practices of the nuclear power industry.
Author:
Publisher:
ISBN:
Category : Meat
Languages : en
Pages : 74
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Book Description
Author: R. T. McCracken
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, "Safety Basis Requirements," requires acontractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.110 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants"2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology thatwas developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 72
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Book Description
This interim safety basis (ISB-008) replaces the B Plant Safety Analysis Report, WHC-SD-WM-SAR-013, Rev. 2 (WHC 1993a). ISB-008 uses existing accident analyses, modified existing accident analyses, and new accident analyses to prove that B Plant remains within the safety envelope for transition, deactivation, standby, and shutdown activities. The analyses in ISB-008 are in accordance with the most current requirements for analytical approach, risk determination, and configuration management. This document and supporting accident analyses replace previous design-basis documents.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
With the implementation of DOE Order 420.1B, Facility Safety, and DOE-STD-3007-2007, 'Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities', a new requirement was imposed that all criticality safety controls be evaluated for inclusion in the facility Documented Safety Analysis (DSA) and that the evaluation process be documented in the site Criticality Safety Program Description Document (CSPDD). At the Hanford site in Washington State the CSPDD, HNF-31695, 'General Description of the FH Criticality Safety Program', requires each facility develop a linking document called a Criticality Control Review (CCR) to document performance of these evaluations. Chapter 5, Appendix 5B of HNF-7098, Criticality Safety Program, provided an example of a format for a CCR that could be used in lieu of each facility developing its own CCR. Since the Plutonium Finishing Plant (PFP) is presently undergoing Deactivation and Decommissioning (D & D), new procedures are being developed for cleanout of equipment and systems that have not been operated in years. Existing Criticality Safety Evaluations (CSE) are revised, or new ones written, to develop the controls required to support D & D activities. Other Hanford facilities, including PFP, had difficulty using the basic CCR out of HNF-7098 when first implemented. Interpretation of the new guidelines indicated that many of the controls needed to be elevated to TSR level controls. Criterion 2 of the standard, requiring that the consequence of a criticality be examined for establishing the classification of a control, was not addressed. Upon in-depth review by PFP Criticality Safety staff, it was not clear that the programmatic interpretation of criterion 8C could be applied at PFP. Therefore, the PFP Criticality Safety staff decided to write their own CCR. The PFP CCR provides additional guidance for the evaluation team to use by clarifying the evaluation criteria in DOE-STD-3007-2007. In reviewing documents used in classifying controls for Nuclear Safety, it was noted that DOE-HDBK-1188, 'Glossary of Environment, Health, and Safety Terms', defines an Administrative Control (AC) in terms that are different than typically used in Criticality Safety. As part of this CCR, a new term, Criticality Administrative Control (CAC) was defined to clarify the difference between an AC used for criticality safety and an AC used for nuclear safety. In Nuclear Safety terms, an AC is a provision relating to organization and management, procedures, recordkeeping, assessment, and reporting necessary to ensure safe operation of a facility. A CAC was defined as an administrative control derived in a criticality safety analysis that is implemented to ensure double contingency. According to criterion 2 of Section IV, 'Linkage to the Documented Safety Analysis', of DOESTD-3007-2007, the consequence of a criticality should be examined for the purposes of classifying the significance of a control or component. HNF-PRO-700, 'Safety Basis Development', provides control selection criteria based on consequence and risk that may be used in the development of a Criticality Safety Evaluation (CSE) to establish the classification of a component as a design feature, as safety class or safety significant, i.e., an Engineered Safety Feature (ESF), or as equipment important to safety; or merely provides defense-in-depth. Similar logic is applied to the CACs. Criterion 8C of DOE-STD-3007-2007, as written, added to the confusion of using the basic CCR from HNF-7098. The PFP CCR attempts to clarify this criterion by revising it to say 'Programmatic commitments or general references to control philosophy (e.g., mass control or spacing control or concentration control as an overall control strategy for the process without specific quantification of individual limits) is included in the PFP DSA'. Table 1 shows the PFP methodology for evaluating CACs. This evaluation process has been in use since February of 2008 and has proven to be simple and effective. Each control identified in the applicable new/revised CSE is evaluated via the table. The results of this evaluation are documented in tables attached to the CCR as an appendix, for each CSE, to the base document.
