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Plutonium Immobilization Rack and Magazine Preliminary Design

Description: The purpose of this report is to document our current preliminary design for the Can-in-Canister rack and magazine. Since this is a developmental project with testing still ongoing, these designs will probably change as we become more knowledgeable of the functions, reliability, and cost of these designs.
Date: December 11, 1998
Creator: Stokes, M.W.
Partner: UNT Libraries Government Documents Department

Evaluation of cast carbon steel and aluminum for rack insert in MCO Mark 1A fuel basket

Description: This document evaluates the effects ofusing a cast carbon steel or aluminum instead of 3O4L stainless steel in the construction ofthe fuel rack insert for the Spent Nuclear Fuel MCO Mark IA fuel baskets. The corrosion, structural, and cost effects are examined.
Date: March 21, 1997
Creator: Graves, C. E.
Partner: UNT Libraries Government Documents Department

Remote inspection of the IFSF spent fuel storage rack

Description: The Irradiated Fuel Storage Facility (IFSF) is a dry storage facility for spent nuclear fuels located at the Idaho Chemical Processing Plant; it was constructed in the 1970`s specifically for the Fort Saint Vrain spent reactor fuels. Currently, it is being used for various spent fuels. It was not known if IFSF would met current DOE seismic criteria, so re-analysis was started, with the rack being analyzed first. The rack was inspected to determine the as-built condition. LazrLyne and VideoRuler were used in lieu of using a tape measure with the camera. It was concluded that when a visual inspection shows widely varying weld sizes, the engineer has to use all resources available to determine the most probable specified weld sizes.
Date: May 1, 1996
Creator: Uldrich, E.D.
Partner: UNT Libraries Government Documents Department

324 Facility B-Cell quality process plan

Description: B-Cell is currently being cleaned out (i.e., removal of equipment, fixtures and residual radioactive materials) and deactivated. TPA Milestone M-89-02 dictates that all mixed waste and equipment be removed from B-Cell by 5/31/99. The following sections describe the major activities that remain for completion of the TPA milestone. These include: Size Reduce Tank 119 and Miscellaneous Equipment; Load and Ship Low-Level Waste; Remove and Size Reduce the 1B Rack; Collect Dispersible Material from Cell Floor; Remove and Size Reduce the 2A Rack; Size Reduce the 1A Rack; Load and Ship Mixed Waste to PUREX Tunnels; and Move Spent Fuel to A-Cell;
Date: April 29, 1998
Creator: Carlson, J.L.
Partner: UNT Libraries Government Documents Department

324 Facility B-cell quality process plan

Description: B-Cell is currently being cleaned out (i.e., removal of equipment, fixtures and residual radioactive materials) and deactivated. TPA Milestone M-89-02 dictates that all mixed waste and equipment be removed from B-Cell by 5/31/99. The following sections describe the major activities that remain for completion of the TPA milestone. These include: Size Reduce Tank 119 and Miscellaneous Equipment; Load and Ship Low-Level Waste; Remove and Size Reduce the 1B Rack; Collect Dispersible Material from Cell Floor; Remove and Size Reduce the 2A Rack; Size Reduce the 1A Rack; Load and Ship Mixed Waste to PUREX Tunnels; and Move Spent Fuel to A-Cell;
Date: July 29, 1998
Creator: Carlson, J.L.
Partner: UNT Libraries Government Documents Department

Independent review of design and analysis for Holtec spent fuel storage racks of CPP 666 Pool 1

Description: This document summarizes the analyses and review performed to develop and validate the design of the new fuel storage racks for the Idaho Chemical Processing Plant (ICPP) Fuel Storage Area (FSA). Holtec International is responsible for the design and fabrication of the storage racks. This report describes the issues raised in the review effort and the resolutions to these issues. The conclusion is reached that the review issues for the racks of Pool 1 have been satisfactorily resolved in the final design and analysis for these racks. Section 1 of this report gives a brief description of the project. Section 2 describes the approach that Holtec used in analyzing the racks and results from these analyses. Section 3 describes the independent review process. Section 4 discusses the identification of and resolution to comments on the design analysis. Section 5 describes additional analysis performed to address major concerns with the Holtec design analysis. Section 6 presents a summary of AEC`s independent review, which is based on AEC`s final review report. Finally, Section 7 gives the Lockheed Idaho Technologies Company (LITCO) position on the acceptability of Holtec`s design.
Date: March 1, 1996
Creator: Miller, G.K.
Partner: UNT Libraries Government Documents Department

Results for additional calculations for Task Order 98-009B-01,addendum 3 to: HNF-SD-SNF-CSER-005, Revision 3

Description: Several sets of new calculations were performed to support the Spent Nuclear Fuel project nuclear criticality safety evaluation. These new calculations include partial loading of Mark IA inner elements after the outers have been loaded, a new, more robust design for the central pipe insert for the Mark IA fuel baskets, Single Pass Reactor fuel loading, the lowering of a Mark IV-loaded MCO through the concrete operating deck as-it is inserted into the Canister Storage Building storage array, and one additional scrap basket loading error. None of these calculations exceeded the criticality safety limit.
Date: November 25, 1997
Creator: Schwinkendorf, K.N., Westinghouse Hanford, Richland, WA
Partner: UNT Libraries Government Documents Department

Applying x-ray digital imaging to the verification of cadmium in fuel-storage components

Description: The High Flux Isotope Reactor utilizes large underwater fuel-storage arrays to stage irradiated fuel before it is shipped from the facility. Cadmium is required as a thermal neutron absorber in these fuel-storage arrays to produce an acceptable margin of nuclear subcriticality during both normal and off-normal operating conditions. Due to incomplete documentation from the time of their fabrication, the presence of cadmium within two stainless-steel parts of fuel-storage arrays must be experimentally verified before they are reused in new fuel-storage arrays. A cadmium-verification program has been developed in association with the Waste Examination and Assay Facility located at the Oak Ridge national Laboratory to nondestructively examine these older shroud assemblies. The program includes the following elements (1) x-ray analog imaging, (2) x-ray digital imaging, (3) prompt-gamma-ray spectroscopy measurements, and (4) neutron-transmission measurements. X-ray digital imaging utilizes an analog-to-digital convertor to record attenuated x-ray intensities observed on a fluorescent detector by a video camera. These x-ray intensities are utilized in expressions for cadmium thickness based upon x-ray attenuation theory.
Date: March 1997
Creator: Dabbs, R. D. & Cook, D. H.
Partner: UNT Libraries Government Documents Department

Experimental assessment of the thermal performance of storage canister/holding fixture configurations for the Los Alamos Nuclear Materials Storage Facility

Description: This report presents experimental results on the thermal performance of various nested canister configurations and canister holding fixtures to be used in the Los Alamos Nuclear Materials Storage Facility. The experiment consisted of placing a heated aluminum billet (to represent heat-generating nuclear material) inside curved- and flat-bottom canisters with and without holding plate fixtures and/or extended fin surfaces. Surface temperatures were measured at several locations on the aluminum billet, inner and outer canisters, and the holding plate fixture to assess the effectiveness of the various configurations in removing and distributing the heat from the aluminum billet. Results indicated that the curved-bottom canisters, with or without holding fixtures, were extremely ineffective in extracting heat from the aluminum billet. The larger thermal contact area provided by the flat-bottom canisters compared with the curved-bottom design, greatly enhanced the heat removal process and lowered the temperature of the aluminum billet considerably. The addition of the fixture plates to the flat-bottom canister geometry greatly enhances the heat removal rates and lowers the canister operating temperatures considerably. The addition of the fixture plates to the flat-bottom canister geometry greatly enhances the heat removal rates and lowers the canister operating temperatures considerably. Finally, the addition of extended fin surfaces to the outer flat-bottom canister positioned on a fixture plate, reduced the canister temperatures still further.
Date: November 1, 1997
Creator: Bernardin, J.D.; Naffziger, D.C. & Gregory, W.S.
Partner: UNT Libraries Government Documents Department

Utilization of a finite element model to verify spent nuclear fuel storage rack welds

Description: Elastic and plastic finite element analyses were performed for the inner tie block assembly of a 25 port fuel rack designed for installation at the Idaho National Engineering and Environmental Laboratory (INEEL) Idaho Chemical Processing Plant (ICPP). The model was specifically developed to verify the adequacy of certain welds joining components of the fuel storage rack assembly. The work scope for this task was limited to an investigation of the stress levels in the inner tie welds when the rack was subjected to seismic loads. Structural acceptance criteria used for the elastic calculations performed were as defined by the rack`s designer. Structural acceptance criteria used for the plastic calculations performed as part of this effort were as defined in Subsection NF and Appendix F of Section III of the ASME Boiler and Pressure Vessel Code. The results confirm that the welds joining the inner tie block to the surrounding rack structure meet the acceptance criteria. The analysis results verified that the inner tie block welds should be capable of transferring the expected seismic load without structural failure.
Date: July 1, 1998
Creator: Nitzel, M.E.
Partner: UNT Libraries Government Documents Department

Decontamination of FAST (CPP-666) fuel storage area stainless steel fuel storage racks

Description: The purpose of this report was to identify and evaluate alternatives for the decontamination of the RSM stainless steel that will be removed from the Idaho Chemical Processing plant (ICPP) fuel storage area (FSA) located in the FAST (CPP-666) building, and to recommend decontamination alternatives for treating this material. Upon the completion of a literature search, the review of the pertinent literature, and based on the review of a variety of chemical, mechanical, and compound (both chemical and mechanical) decontamination techniques, the preliminary results of analyses of FSA critically barrier contaminants, and the data collected during the FSA Reracking project, it was concluded that decontamination and beneficial recycle of the FSA stainless steel produced is technically feasible and likely to be cost effective as compared to burying the material at the RWMC. It is recommended that an organic acid, or commercial product containing an organic acid, be used to decontaminate the FSA stainless steel; however, it is also recommended that other surface decontamination methods be tested in the event that this method proves unsuitable. Among the techniques that should be investigated are mechanical techniques (CO{sub 2} pellet blasting and ultra-high pressure water blasting) and chemical techniques that are compatible with present ICPP waste streams.
Date: October 1, 1993
Creator: Kessinger, G.F.
Partner: UNT Libraries Government Documents Department

Effects of assembly local power distribution on storage rack criticality

Description: Fuel storage rack criticality calculations have been performed for a set of 8 x 8 BWR k/sub infinity/-equivalent fuel bundles. The results of these calculations indicate that the storage rack multiplication factor is not determined solely by the k/sub infinity/'s of the individual fuel assemblies, but is also sensitive to the local power, and underlying enrichment, burnable poison, water hole and burnup distributions. Furthermore, in order to insure a true conservative upper bound on the storage rack multiplication factor, reference assemblies with the most centrally peaked power distributions should be selected. 2 refs., 1 fig.
Date: January 1, 1985
Creator: Todosow, M. & Carew, J.F.
Partner: UNT Libraries Government Documents Department

Fail-safe storage rack for fuel rod assemblies

Description: This report discusses a fail-safe storage rack which is provided for interim storage of spent but radioactive nuclear fuel rod assemblies. The rack consists of a checkerboard array of substantially square, elongate receiving tubes fully enclosed by a double walled container, the outer wall of which is imperforate for liquid containment and the inner wall of which is provided with perforations for admitting moderator liquid flow to the elongate receiving tubes, the liquid serving to take up waste heat from the stored nuclear assemblies and dissipate same to the ambient liquid reservoir. A perforated cover sealing the rack facilitates cooling liquid entry and dissipation.
Date: December 31, 1991
Creator: Lewis, D. R.
Partner: UNT Libraries Government Documents Department

Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report, October 1-December 31, 1978

Description: Experimental measurements are being taken on critical configurations of clusters of fuel rods mocking up LWR-type fuel elements in close proximity water storage. The results will serve to benchmark the computer codes used in designing nuclear power reactor fuel storage racks.
Date: March 1, 1979
Creator: Baldwin, M.N.; Hoovler, G.S.; Eng, R.L. & Welfare, F.G.
Partner: UNT Libraries Government Documents Department

Validation of the RESRAD-RECYCLE computer code.

Description: The RESRAD-RECYCLE computer code was developed by Argonne National Laboratory under the sponsorship of the U.S. Department of Energy. It was designed to analyze potential radiation exposures resulting from the reuse and recycling of radioactively contaminated scrap metal and equipment. It was one of two codes selected in an international model validation study concerning recycling of radioactively contaminated metals. In the validation study, dose measurements at various stages of melting a spent nuclear fuel rack at Studsvik RadWaste AB, Sweden, were collected and compared with modeling results. The comparison shows that the RESRAD-RECYCLE results agree fairly well with the measurement data. Among the scenarios considered, dose results and measurement data agree within a factor of 6. Discrepancies may be explained by the geometrical limitation of the RESRAD-RECYCLE's external exposure model, the dynamic nature of the recycling activities, and inaccuracy in the input parameter values used in dose calculations.
Date: February 1, 2002
Creator: Cheng, J.-J.; Yu, C.; Williams, W. A. & Murphie, W.
Partner: UNT Libraries Government Documents Department

Achieving increased spent fuel storage capacity at the High Flux Isotope Reactor (HFIR)

Description: The HFIR facility was originally designed to store approximately 25 spent cores, sufficient to allow for operational contingencies and for cooling prior to off-site shipment for reprocessing. The original capacity has now been increased to 60 positions, of which 53 are currently filled (September 1994). Additional spent cores are produced at a rate of about 10 or 11 per year. Continued HFIR operation, therefore, depends on a significant near-term expansion of the pool storage capacity, as well as on a future capability of reprocessing or other storage alternatives once the practical capacity of the pool is reached. To store the much larger inventory of spent fuel that may remain on-site under various future scenarios, the pool capacity is being increased in a phased manner through installation of a new multi-tier spent fuel rack design for higher density storage. A total of 143 positions was used for this paper as the maximum practical pool capacity without impacting operations; however, greater ultimate capacities were addressed in the supporting analyses and approval documents. This paper addresses issues related to the pool storage expansion including (1) seismic effects on the three-tier storage arrays, (2) thermal performance of the new arrays, (3) spent fuel cladding corrosion concerns related to the longer period of pool storage, and (4) impacts of increased spent fuel inventory on the pool water quality, water treatment systems, and LLLW volume.
Date: December 31, 1994
Creator: Cook, D.H.; Chang, S.J.; Dabs, R.D.; Freels, J.D.; Morgan, K.A.; Rothrock, R.B. et al.
Partner: UNT Libraries Government Documents Department

Buckling analysis of spent fuel basket

Description: The basket for a spent fuel shipping cask is subjected to compressive stresses that may cause global instability of the basket assemblies or local buckling of the individual members. Adopting the common buckling design practice in which the stability capacity of the entire structure is based on the performance of the individual members of the assemblies, the typical spent fuel basket, which is composed of plates and tubular structural members, can be idealized as an assemblage of columns, beam-columns and plates. This report presents the flexural buckling formulas for five load cases that are common in the basket buckling analysis: column under axial loads, column under axial and bending loads, plate under uniaxial loads, plate under biaxial loadings, and plate under biaxial loads and lateral pressure. The acceptance criteria from the ASME Boiler and Pressure Vessel Code are used to determine the adequacy of the basket components. Special acceptance criteria are proposed to address the unique material characteristics of austenitic stainless steel, a material which is frequently used in the basket assemblies.
Date: May 1, 1995
Creator: Lee, A.S. & Bumpas, S.E.
Partner: UNT Libraries Government Documents Department

OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

Description: In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. ...
Date: June 1, 1996
Creator: DeHart, M.D.; Parks, C.V. & Brady, M.C.
Partner: UNT Libraries Government Documents Department

Seismic analysis of submerged spent fuel storage structure

Description: The purpose of this calculation is to provide structural integrity analysis for the loaded new spent fuel rack arrays against possible seismic excitation. The seismic design calculation is based on the UCRL-15910 spectrum with peak ground acceleration of 0.32g and 5% damping. This spectrum may be considered as an upper bound of the newly developed Oak Ridge site-specific spectrum with 0.29g peak ground acceleration and 5% damping. Both are more conservative than the current design basis seismic acceleration of 0.15g for HFIR. The calculation is carried out by using ABAQUS version 5.2 and the response spectrum option. Since the new racks are to be submerged in HFIR pool, the pool water induced virtual mass has been conservatively taken into consideration. The result shows that if the silo buckling is regarded as failure than, with 95% confidence, the 5% probability of failure ground acceleration is as much as 2.334g. As compared with the design basis of 0.32g, the structure is very safe against earthquake.
Date: December 31, 1994
Creator: Chang, S.J.
Partner: UNT Libraries Government Documents Department

Analysis of ICPP fuel storage rack inner tie and corner tie substructures

Description: Finite element models were developed and analyses performed for the tie plate, inner tie block assembly, and corner tie block assembly of a 25 port fuel rack assembly designed for installation in Pool 1 of Building 666 at the Idaho Chemical Processing Plant. These models were specifically developed to investigate the adequacy of certain welds joining components of the fuel storage rack assembly. The work scope for the task was limited to an investigation of the stress levels in the subject subassemblies when subjected to seismic loads. Structural acceptance criteria used for the elastic calculations performed were as found in the overall rack design report as issued by the rack`s designer, Holtec International. Structural acceptance criteria used for the plastic calculations performed as part of this effort were as defined in Subsection NF and Appendix F of the ASME Boiler & Pressure Vessel Code. The results of the analyses will also apply to the 30 port fuel storage rack design that is also scheduled for installation in Pool 1 of ICPP 666. The results obtained from the analyses performed for this task indicate that the welds joining the inner tie block and corner tie block to the surrounding rack structure meet the acceptance criteria. Further, the structural members (plates and blocks) were also found to be within the allowable stress limits established by the acceptance criteria. The separate analysis performed on the inner tie plate confirmed the structural adequacy for both the inner tie plate, corner tie plate, and tie block bolts. The analysis results verified that the inner tie and corner tie block should be capable of transferring the expected seismic load without structural failure.
Date: January 1, 1996
Creator: Nitzel, M.E. & Rahl, R.G.
Partner: UNT Libraries Government Documents Department

AFR Spent Fuel Storage Program. Technical progress report, April 1980-June 1980

Description: Work on this project is focused on developing design and licensing information for the model facility. The three major subcontracts for soils and structural design, rack design, and the security system design are progressing satisfactorily. Design modification work at AGNS is near completion. Licensing documentation is approximately 50% complete and progressing at a satisfactory pace to meet scheduled projections.
Date: July 25, 1980
Partner: UNT Libraries Government Documents Department

Evaluation of copper for divider subassembly in MCO Mark IA and Mark IV scrap fuel baskets

Description: The K Basin Spent Nuclear Fuel (SNF) Project Multi-Canister Overpack (MCO) subprojection eludes the design and fabrication of a canister that will be used to confine, contain, and maintain fuel in a critically safe array to enable its removal from the K Basins, vacuum drying, transport, staging, hot conditioning, and interim storage (Goldinann 1997). Each MCO consists of a shell, shield plug, fuel baskets (Mark IA or Mark IV), and other incidental equipment. The Mark IA intact and scrap fuel baskets are a safety class item for criticality control and components necessary for criticality control will be constructed from 304L stainless steel. It is proposed that a copper divider subassembly be used in both Mark IA and Mark IV scrap baskets to increase the safety basis margin during cold vacuum drying. The use of copper would increase the heat conducted away from hot areas in the baskets out to the wall of the MCO by both radiative and conductive heat transfer means. Thus copper subassembly will likely be a safety significant component of the scrap fuel baskets. This report examines the structural, cost and corrosion consequences associated with using a copper subassembly in the stainless steel MCO scrap fuel baskets.
Date: September 29, 1997
Creator: Graves, C. E.
Partner: UNT Libraries Government Documents Department

CSER 79-028, Addendum 2: Security bar addition to pedestal storage racks in Room 3 in 2736-Z Building

Description: The Plutonium Finishing Plant (PFP) is installing security bars on plutonium storage racks in Room 3 in 2736-Z Building to meet International Atomic Energy Agency (IAEA) material control requirements. Figures show the existing arrangement and design of the security bars. The security bars are to be fabricated of aluminum or carbon steel. The detailed fabrication sketches are reproduced in Appendix C. The security bars are to be installed close to the chains of plutonium so a determination of their effect on criticality safety needs to be made. The addition of security bars to the storage array of 2.5 kg plutonium buttons in Room 3 can effect reactivity by reflecting neutrons back into the plutonium in the storage cans, by absorbing neutrons, and by moderating neutrons between stored plutonium buttons. The small amount of metal added by the storage bars in comparison to the amount of concrete in the walls and aluminum in the shelf monitors already in place would not significantly increase the k{sub eff} of the storage array. Several computer calculations in previous analyses show that the security bars will have a negligible affect on reactivity.
Date: November 18, 1994
Creator: Miller, E. M.
Partner: UNT Libraries Government Documents Department

AFR/Design and Licensing Information/BNFP As a Model. Technical progress report, October 1979-December 1979

Description: Work on the AFR spent fuel storage program is focused on developing design and licensing information for the model facility. To date, the design effort is proceeding on schedule. A subcontract for soils and structural design of the fuel transfers canal is ready for DOE review. Proposals for high-density rack design to bring the model facility up to a nominal 1750 MTU capacity have been received and are being reviewed internally. This subcontract should be available for DOE review in early February. A security subcontract has been awarded to International Energy Associates Limited and the preliminary design effort is underway. Licensing activities are progressing satisfactorily. Review of the governing local, state, and federal regulations has been completed. Format and schedule for the safety analysis report and the environmental report have been established.
Date: January 25, 1980
Partner: UNT Libraries Government Documents Department