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The transfer of energy produced by the interaction of the intense pulses of short-ranged fusion microexplosion products with materials is one of the most difficult problems in inertially-confined fusion (ICF) reactor design. The short time and deposition distance for the energy results in local peak power densities on the order of 10/sup 18/ watts/m/sup 3/. High local power densities may cause change of state or spall in the reactor materials. This will limit the structure lifetimes for ICF reactors of economic physical sizes, increasing operating costs including structure replacement and radioactive waste management. Four basic first wall protection methods have evolved: a dry-wall, a wet-wall, a magnetically shielded wall, and a fluid wall. These approaches are distinguished by the way the reactor wall interfaces with fusion debris as well as the way the ambient cavity conditions modify the fusion energy forms and spectra at the first wall. Each of these approaches requires different heat transfer considerations.

This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the United States, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general, the agreement between methods was very good, providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

Abstract The long term core and primary loop heatup of an HTGR subsequent to loss of all forced circulation has been analyzed using a modified version of the CORCON code. The results indicate that if the liner cooling system is operating, or can be restarted within about 60 h, safe cooldown can be achieved, but significant core damage will occur. Without functioning liner cooling system the core heatup will lead to PCRV concrete degradation and the resulting concrete gas releases will ultimately cause containment building failure after 6 to 10 days.

Abstract Analysis of Primary Containment Transients (APRICOT) is an ERDA sponsored project in which a variety of reactor safety analysis groups around the world have been invited to participate by performing calculations to verify capabilities of large computer codes used to analyze postulated core disputive accidents of liquid metal fast breeder reactors. Nine groups have performed calculations of the first three problems which were set, using ten computer codes. Two problems were simple test problems for which analytical solutions exist, namely an ideal gas shock tube, and a suddenly pressurized spherical cavity in an infinite elastic medium. The third problem concerns an explosion in a partially water-filled overstrong cylindrical containment vessel for which experimental data exist. A critique of the results of these calculations is given in this paper.

Abstract Next generation commercial reactor designs emphasize enhanced safety by means of improved safety system reliability and performance. These objectives are achieved via safety system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet, the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs will necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing U.S. advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes may require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

Abstract Check valves are used extensively in nuclear plant safety systems and balance-of-plant (BOP) systems. Their failures have resulted in significant maintenance efforts and, on occasion, have resulted in water hammer, overpressurization of low-pressure systems and damage to flow system components. Consequently, in recent years check valves have received considerable attention by the Nuclear Regulatory Commission (NRC) and the nuclear power industry. Oak Ridge National Laboratory (ORNL) is carrying out a comprehensive two phase aging assessment of check valves in support of the Nuclear Plant Aging Research (NPAR) program. As part of the second phase, ORNL is evaluating several developmental and/or commercially available check valve diagnostic monitoring methods; in particular, those based on measurements of acoustic emission, ultrasonics, and magnetic flux. These three methods were found to provide different (and complementary) diagnostic information. The combination of acoustic emission with either ultrasonic or magnetic flux monitoring yields a monitoring system that succeeds in providing sensitivity to detect all major check valve operating conditions. The three check valve monitoring methods described in this paper are still under development and are presently being tested as part of a program directed by the Nuclear Industry Check Valve Group (NIC) in conjunction with the Electric Power Research Institute (EPRI). Phase 1 of this program (water testing) is being carried out at the Utah Water Research Laboratory located on the Utah State University campus.

Los Alamos National Laboratory is a participant in the 2D/3D program. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of facility design, construction, and operation; provision of boundary and initial conditions for test facility operations based on analysis of pressurized water reactors; performance of pretest and posttest predictions and analyses; and use of experimental results to validate and assess the single- and multidimensional nonequilibrium features in the Transient Reactor Analysis Code (TRAC). During Fiscal Year 1986, Los Alamos conducted analytical assessment activities using data from the Cylindrical Core Test Facility and the Slab Core Test Facility. Los Alamos also continued to provide support analysis for the planning of Upper Plenum Test Facility experiments. Finally, Los Alamos either completed or is currently working on three areas of TRAC modeling improvement. In this paper, Los Alamos activities during Fiscal Year 1986 are summarized; several significant accomplishments are described in more detail to illustrate the work activities at Los Alamos.

Abstract Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750°F (1783 K) and an oxide debris melting temperature of 4350°F (2672 K). COINTAIN analyses were performed for the debris/concrete interaction occurring without consideration of the possible existence of an overlying pool of water. Results indicate failure of the drywell head seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed.

Abstract The behavior of LMFBR core debris following a hypothetical core-disruptive accident (HCDA) depends upon a wide range of physical and chemical phenomena. Current understanding of the key phenomena are summarized for core-debris behavior within the reactor vessel, within the reactor cavity, and within the concrete base mat below the reactor cavity. In-vessel behavior was the principal consideration of post-accident heat removal for the FFTF reactor. Several concepts of engineered core-retention systems within the reactor cavity have been considered for other reactors, including the cooled crucible concept, the sacrificial barrier concept, the stable barrier concept, and the catch tray concept. Behavior within the concrete base mat is an important part of a general concept of inherent core retention which depends upon an understanding of the complex interactions of core debris with concrete.