[en] Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

[en] ²³⁸PuO₂ process operations are housed in a complex of 76 gloveboxes and introductory hoods connected by means of an overhead trolley housed in a tunnel. Because a significant number of the gloveboxes used for ²³⁸PuO₂ processing were installed before the original startup of the facility in 1978, they have been in service for nearly 20 years. During a recent heat source production campaign, numerous contamination releases in the ²³⁸PuO₂ processing area were traced to degraded elastomer gaskets used for glovebox connections, and attachment of feed-throughs, service panels, and windows. Evaluation of the degraded gaskets revealed that a combination of radiolytic degradation related to the high specific activity of ²³⁸Pu, and extended service at high altitude in a low (to extremely low) humidity environment had resulted in accelerated gasket aging. However, it was also apparent that gasket design was the most important factor in actual contamination release. All of the contamination releases that were traced to degraded gaskets occurred in variations of a design that used a spline to expand an elastomeric gasket into the space between a connecting flange, window, or service panel, and a glovebox opening. No contamination releases were traced to the gasket design that employed bolted clamps to compress the gasket between a connecting flange, window, or panel, and the exterior surface of a glovebox opening. As a result of these findings, the Actinide Ceramics group at LANL (NMT-9) has initiated a routine replacement and upgrade program to replace aging gloveboxes. All of the new gloveboxes will utilize the preferred gasket design, which is expected to reduce the number and frequency of contamination releases

[en] In order to detect the presence of fuel rods, it was selected a reflection optical transducer, which provides a measurable electrical signal when the beam at a certain distance is interrupted then there is a reflection causing a excitation to the sensor that provides a change of state at the output of transducer. This step is coupled through an operational amplifier which drives the opto coupler circuit isolating this step of the interface and a personal computer. This work presents the description of components, designs, signal coupler and opto isolater circuit, interface circuit and tutorial assemble program. (Author)

[en] With the closure of many experimental facilities, the nuclear safety analyst has to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. Although deterministic methods often do not provide exact models of a system, a substantial amount of reliable information on nuclear systems can be obtained using these methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico (UNM) in cooperation with the Radiation Transport Group at Los Alamos National Laboratory (LANL) has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. DANTSYS is the new name of the group of codes formerly known as: ONEDANT, TWODANT, TWOHEX, TWOGQ, and THREEDANT. The primer is designed to teach bu example, with each example illustrating two or three DANTSYS features useful in criticality analyses. Starting with a Quickstart chapter, the primer gives an overview of the basic requirements for DANTSYS input and allows the user to quickly run a simple criticality problem with DANTSYS. Each chapter has a list of basic objectives at the beginning identifying the goal of the chapter and the individual DANTSYS features covered in detail in the chapter example problems. On completion of the primer, it is expected that the user will be comfortable doing criticality calculations with DANTSYS and can handle 60--80% of the situations that normally arise in a facility. The primary provides a set of input files that can be selective modified by the user to fit each particular problem

[en] This paper addresses the use of real-time software to assist handlers of fissionable nuclear material. We focus specifically on the issue of workstation mass limits, and the need for handlers to be aware of, and check against, those mass limits during material transfers. Here ''mass limits'' generally refer to criticality safety mass limits; however, in some instances, workstation mass limits for some materials may be governed by considerations other than criticality, e.g., fire or release consequence limitation. As a case study, we provide a simplified reliability comparison of the use of a manual two handler system with a software-assisted two handler system. We identify the interface points between software and handlers that are relevant to criticality safety

[en] The report reviews a method for modeling and controlling two serial link manipulators which mutually lift and transport a rigid body object in a three dimensional workspace. A new vector variable is introduced which parameterizes the internal contact force controlled degrees of freedom. A technique for dynamically distributing the payload between the manipulators is suggested which yields a family of solutions for the contact forces and torques the manipulators impart to the object. A set of rigid body kinematic constraints which restricts the values of the joint velocities of both manipulators is derived. A rigid body dynamical model for the closed chain system is first developed in the joint space. The model is obtained by generalizing the previous methods for deriving the model. The joint velocity and acceleration variables in the model are expressed in terms of independent pseudovariables. The pseudospace model is transformed to obtain reduced order equations of motion and a separate set of equations governing the internal components of the contact forces and torques. A theoretic control architecture is suggested which explicitly decouples the two sets of equations comprising the model. The controller enables the designer to develop independent, non-interacting control laws for the position control and internal force control of the system

[en] In 1995, at a conference on criticality safety, a special session was devoted to the Monte Carlo open-quotes eigenvalue of the worldclose quotes problem. Argonne presented a paper, at that session, in which the anomalies originally observed in that problem were reproduced in a much simplified model-problem configuration, and removed by a version of stratified source-sampling. The original test-problem was treated by a special code designed specifically for that purpose. Recently ANL started work on a method for dealing with more realistic eigenvalue of the world configurations, and has been incorporating this method into VIM. The original method has been modified to take into account real-world statistical noise sources not included in the model problem. This paper constitutes a status report on work still in progress

[en] A concern has been identified regarding the viability of redesigning and requalifying existing glovebox lines for use as glovebox lines integral to future mission activities in the 773-A laboratory building at the Savannah River Site (SRS). The Bechtel Savannah River Inc. (BSRI) design engineering team has been requested to perform an evaluation which would investigate the reuse of these existing gloveboxes versus the procurement of completely new glovebox systems. The existing glovebox lines were manufactured for the Plutonium (Pu) Metallograph Facility, Project 3253, located in building 235-F at SRS. These gloveboxes were designed as independent, fully functional Pu 'metal' and Pu 'oxide' processing glovebox systems for this facility. These gloveboxes, although fully installed, have never processed radioactive material. The proposed use for these gloveboxes are: (1) to utilize the Pu 'metal' glovebox system for the primary containment associated with the Pre-Processing/Re-Processing Laboratory for obtaining radioactive glass compound viscometer analysis and (2) to utilize the Pu 'oxide' glovebox system for primary containment associated with the Pu 'Can in Can' Demonstration for proof of principle testing specific to long term Pu immobilization and storage technology. This report presents objective evidence that supports the engineering judgment indicating the existing gloveboxes can be requalified for the proposed uses indicated above. SRS has the ability to duplicate the test parameters, with site forces, that will meet or exceed the identical acceptance criteria established to qualify the existing gloveboxes. The qualification effort will be a documented procedure using the leak test criteria characteristic of the original glovebox purchase. Two equivalent tests will be performed, one for post modification leak test acceptance and one for post installation leak test acceptance. (Abstract Truncated)

[en] Subcriticality measurement by Mihalczo method was investigated varying Cf source intensity and neutron background level. It is found that Cf source intensity must be chosen carefully when neutron background level is too high to be ignored compared to Cf source intensity. Under current measurement accuracy, the statistic error becomes double when the ratio of neutron background occupies 90% of all neutron sources in the system. On the other hand, under low neutron background level, Cf source intensity does not have any effect on the accuracy and error of measured keffs. Also investigated was the detector position dependence of measured keffs. The detector fields of view for locally positioned detector was utilized to correct the dependence. When corrected, the measured keffs showed little dependence on detector position, which verified the validity of the correction method. (author)

[en] A general methodology has been developed to evaluate the criticality potential of the wide range of waste forms planned for geologic disposal. The range of waste forms include commercial spent fuel, high level waste, DOE spent fuel (including highly enriched), MOX using weapons grade plutonium, and immobilized plutonium. The disposal of these waste forms will be in a container with sufficiently thick corrosion resistant barriers to prevent water penetration for up to 10,000 years. The criticality control for DOE spent fuel is primarily provided by neutron absorber material incorporated into the basket holding the individual assemblies. For the immobilized plutonium, the neutron absorber material is incorporated into the waste form itself. The disposal criticality analysis methodology includes the analysis of geochemical and physical processes that can breach the waste package and affect the waste forms within. The basic purpose of the methodology is to guide the criticality control features of the waste package design, and to demonstrate that the final design meets the criticality control licensing requirements. The methodology can also be extended to the analysis of criticality consequences (primarily increased radionuclide inventory), which will support the total performance assessment for the respository