scholarly journals Carbon-14 in Waste Packages for Spent Fuel in a Tuff Repository

1986 ◽  
Vol 84 ◽  
Author(s):  
R. A. Van Konynenburg ◽  
C. F. Smith ◽  
H. W. Culham ◽  
H. D. Smith

AbstractCarbon-14 is produced naturally by cosmic ray neutrons in the upper atmosphere. It is also produced in nuclear reactors, in amounts much smaller than the global inventory. About one-third of this is released directly to the atmosphere, and the other two-thirds remains in the spent fuel. Both the Environmental Protection Agency and the Nuclear Regulatory Commission have established limits on release of the 14C in spent fuel. This is of particular concern for the proposed repository in tuff, because of the unsaturated conditions and the consequent possibility of gaseous tansport of 14C as CO2Existing measurements and calculations of the 14C inventory in spent fuel are reviewed. The physical distribution and Wemical forms of the 14C are discussed. Available data on the release of 14C from spent fuel in aqueous solutions and in gaseous environments of air, nitrogen, and helium are reviewed. Projected N behavior in a tuff repository is described. It is concluded that 14C release measurements from spent fuel into moist air at temperatures both above and below the in situ boiling point of water as well as detailed transport calculations for the tuff geological environment will be needed to determine whether the IOCFR60 and 40CFR191 requirements can be met.

1982 ◽  
Vol 15 ◽  
Author(s):  
Margaret S. Chu ◽  
James E. Campbell ◽  
Stephen E. Stuckwisch ◽  
Krishan K. Wahi ◽  
Malcolm D. Siegel

The Environmental Protection Agency (EPA) has issued a draft standard (40CFR191) [1] which specifies permissible limits for radionuclide releases from a high-level waste repository to the accessible environment. The U. S. Nuclear Regulatory Commission (NRC) has published a proposed rule (10CFR60) [2] which contains technical criteria for geologic disposal of high-level waste to facilitate compliance with the EPA Draft Standard. The three main numerical criteria specified in 10CFR60 are:


1998 ◽  
Vol 4 (S2) ◽  
pp. 480-481
Author(s):  
R. L. Sabatini ◽  
Toshi Sugama ◽  
Leonidas Petrakis

A BNL-Grace process has been developed to chemically convert in-situ, the chrysotile fibers of sprayed-on fireproofing products to an unregulated glassy material. The effectiveness of this process has been convincingly demonstrated using Analytical Electron Microscopy and X-Ray Diffraction techniques.Asbestos minerals were used in fireproofing materials because of their excellent physical properties including fire resistance, high tensile strength, heat and electrical insulation, and resistance to acids and alkali. But in 1975 the Environmental Protection Agency began regulating materials containing > 1% asbestos.The new in-situ BNL-Grace process, which uses a foamy solution sprayed directly onto asbestos-containing fireproofing chemically digests essentially all the asbestos fibers, transforming them into harmless materials. After treatment, the fireproofing is no longer a regulated material. The process produces essentially no waste.Our problem was to demonstrate that all of the asbestos was converted and that the remaining materials were no longer regulated. Typical analysis methods use conventional optical and Polarized Light Microscopy (PLM) to measure and observe fibers.


Author(s):  
Robert Eby ◽  
Lark Lundberg ◽  
Steve Marske ◽  
Nolan Hertel ◽  
Rod Ice

Abstract The Georgia Tech Research Reactor (GTRR) is a 5-megawatt (MW) heavy-water-cooled nuclear reactor located on the Georgia Institute of Technology (Georgia Tech) campus in downtown Atlanta, Georgia. On July 1, 1997, Georgia Tech administration notified the U.S. Nuclear Regulatory Commission (NRC) of their intent to decommission the GTRR. In the summer of 1999, the NRC issued a license amendment to decommission the GTRR in accordance with NRC’s Regulatory Guide 1.86. In the spring of 1999, Georgia Tech and the State of Georgia contracted CH2M HILL to serve as the Executive Engineer to manage the decommissioning project. Later in the summer of 1999, the IT Corporation was selected as the Decommissioning Contractor. The Decommissioning Contractor began the dismantlement process at the Georgia Tech site in November, 1999. By February, 2000, reactor support systems such as the primary and secondary cooling water systems, and the bismuth cooling system were removed and packaged for off-site disposal. Reactor internals were removed in April, 2000. Removal of the bioshield occurred from May through November, 2000. Throughout January, 2001, various concrete structures, including the Spent Fuel Storage Hole, were decontaminated. Dismantlement and decontamination activities were completed by April, 2001. The Final Survey Report to the NRC is planned to be submitted to the NRC December, 2001, 2001. Final license termination by the NRC is anticipated in the spring of 2002.


Author(s):  
Christopher S. Bajwa ◽  
Ian F. Spivack

The US Nuclear Regulatory Commission (NRC) is responsible for licensing spent fuel storage casks under Title 10 of the Code of Federal Regulations Part 72 (10 CFR Part 72). Under these regulations, storage casks must be evaluated to verify that they meet various criteria, including acceptable thermal performance requirements. The purpose of the evaluation described in this paper is to establish the effectiveness of a medium-effort modeling approach and associated simplifying assumptions in closely approximating spent fuel cask component temperature distributions. This predictive evaluation is performed with the ANSYS® code, and is applicable to externally cooled cask designs. The results are compared against experimental measurements and predictions of the COBRA-SFS finite-difference code developed at Pacific Northwest National Laboratory.


Author(s):  
Joseph S. Miller

The United States utilities started preparing for external events that could lead to a loss of all ac power in the 1980’s, when the Station Blackout (SBO) rulemaking was first introduced by the United States Nuclear Regulatory Commission (USNRC). Following the events at the Fukushima Dai-ichi nuclear power plant on March 11, 2011, the USNRC established a senior-level agency task force referred to as the Near-Term Task Force (NTTF). The NTTF was tasked with conducting a systematic, methodical review of Nuclear Regulatory Commission (NRC) regulations and processes to determine if the agency should make additional improvements to these programs in light of the events at Fukushima Dai-ichi. As a result of this review, the NTTF developed a comprehensive set of recommendations, documented in SECY-11-0093, “Near-Term Report and Recommendations for Agency Actions Following the Events in Japan,” dated July 12, 2011. Documentation of the staff’s efforts is contained in SECY-11-0124, “Recommended Actions to be Taken without Delay from the Near-Term Task Force Report,” dated September 9, 2011, and SECY-11-0137, “Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned,” dated October 3, 2011. To satisfy some of the NRC’s recommendations, the industry described its proposal for a Diverse and Flexible Mitigation Capability (FLEX), as documented in Nuclear Energy Institute’s (NEI) letter, dated December 16, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML11353A008). FLEX was proposed as a strategy to fulfill the key safety functions of core cooling, containment integrity, and spent fuel cooling. The events at Fukushima Dai-ichi highlight the possibility that extreme natural phenomena could challenge the prevention, mitigation and emergency preparedness defense-in-depth layers. At Fukushima, limitations in time and unpredictable conditions associated with the accident significantly challenged attempts by the responders to preclude core damage and containment failure. During the events in Fukushima, the challenges faced by the operators were beyond any faced previously at a commercial nuclear reactor. NRC Order 12-049 (Ref. 1) and NRC Interim Staff Guidance JLD-ISG-2012-01 (Ref. 6) provided additional requirements to mitigate beyond-design-basis external events. These additional requirements impose guidance and strategies to be available if the loss of power, motive force and normal access to the ultimate heat sink to prevent fuel damage in the reactor and spent fuel pool affected all units at a site simultaneously. The NEI submitted document NEI 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation Guide” in August 2012 (ADAMS Accession No. ML12242A378) to provide specifications for the nuclear power industry in the development, implementation, and maintenance of guidance and strategies in response to NRC Order EA-12-049. The US utilities are currently proposing modifications to their plants that will follow specifications provided in NEI 12-06. This paper presents some of the NEI 12-06 requirements and some of the proposed modifications proposed by the US utilities.


1998 ◽  
Vol 37 (6-7) ◽  
pp. 315-321 ◽  
Author(s):  
Michael R. Palermo

In-Situ Capping (ISC) is defined as the placement of a subaqueous covering or cap of clean or suitable isolating material over an in-situ deposit of contaminated sediment. ISC is a potentially economical and effective approach for remediation of contaminated sediment. A number of sites have been remediated by in-situ capping operations worldwide. The U.S. Army Corps of Engineers (USACE) has developed detailed guidelines for planning, designing, constructing, and monitoring in-situ capping projects for the United States Environmental Protection Agency (USEPA). This paper briefly describes the major aspects of in-situ capping as an option and provides a summary of recent case studies.


1984 ◽  
Vol 44 ◽  
Author(s):  
M. J. Steindler ◽  
W. B. Seefeldt

Some nuclear waste is destined for disposal in deep geological formations. The disposal system for wastes from commercial nuclear activities, and perhaps also for high-level wastes from defense-related activities, is to be designed and operated by the Department of Energy (DOE) and licensed by the Nuclear Regulatory Commission (NRC). The Nuclear Waste Policy Act [1] outlines some of the procedures and schedules that are to be followed by DOE in carrying out its assignment in the disposal of high-level nuclear waste (HLW). The regulations of the NRC that deal with HLW [2] are only partly in place, and amendments (e.g., related to the unsaturated zone) are yet to be approved and issued. The Environmental Protection Agency (EPA) has issued only draft versions of the regulations pertaining to HLW disposal [3], but key features of these drafts are at present in adequate agreement with NRC documents. On the basis of the trends that have become evident in the last few years, the DOE will be required to substantiate performance predictions for all pertinent aspects of a repository, especially the performance of the engineered waste package. The basis for demonstrating that the waste package performance in the repository will be in concert with the requirements is data on the waste package materials. These key materials data must clearly be highly reliable, and DOE will be required to assure this reliability. This paper addresses the organization and functions that have been assembled to aid in establishing the quality of materials data that are important in the licensing of a waste repository.


Author(s):  
Gunup Kwon ◽  
Khaled Ata

Abstract Nuclear power plant spent fuels are initially stored in the spent fuel pool. Then, the water cooled fuels are transferred in a concrete or steel cask and transported outside of the Fuel Handling Building (FHB) or the Reactor Building (RB) for long term on site storage. The spent fuel casks are typically stored on a slab-on-grade pad. The slab-on-grade pad is designed according to the U.S. Nuclear Regulatory Commission NUREG-1536 and NUREG-1567. The two Standard Review Plans provide guidance to the regulators for the review of cask storage system license application. The ISFSI pad analysis and design have to consider various loading conditions, such as earthquake and tornado loadings as well as normal operating loading conditions. Seismic analysis of the ISFSI pad requires considering interaction between the pad and the supporting soil. Various cask loading configurations on the pad also have to be considered. Due to the lack of specific guidelines, many ISFSI pad designs show overly conservative reinforcement. This study provides guidelines and procedure for the design of the ISFSI pad that are typically used in the nuclear industry. It is considered that the guidelines and practices described in this study help design engineers understand general guidance provided in the NRC Standard Review Plans.


1995 ◽  
Vol 412 ◽  
Author(s):  
P. A. Laplante ◽  
S. J. Maheras ◽  
M. S. Jarzemba

AbstractTo develop capabilities for compliance determination, the Nuclear Regulatory Commission (NRC) conducts total system performance assessment (TSPA) for the proposed repository at Yucca Mountain (YM) in an iterative manner. Because the new Environmental Protection Agency (EPA) standard for YM may set a dose or risk limit, an auxiliary study was conducted to develop estimates of site-specific dose assessment parameters for future TSPAs. YM site-relevant data was obtained for irrigation, agriculture, resuspension, crop interception, and soil. A Monte Carlo based importance analysis was used to identify predominant parameters for the groundwater pathway. In this analysis, the GENII-S code generated individual annual total effective dose equivalents (TEDEs) for 20 nuclides and 43 sampled parameters based upon unit groundwater concentrations. Scatter plots and correlation results indicate the crop interception fraction, food transfer factors, consumption rates, and irrigation rate are correlated with TEDEs for specific nuclides. Influential parameter groups correspond to expected pathway behavior of specific nuclides. Results for nuclides that transfer more readily to plants, such as 99Tc, indicate crop ingestion pathway parameters are most highly correlated with the TEDE, and those that transfer to milk (59Ni) or beef (79Se, 129I, 135Cs, 137Cs) show predominant correlations with animal product ingestion pathway parameters. Such relationships provide useful insight to important parameters and exposure pathways applicable to doses from specific nuclides.


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