Anthony J. Baratta

Bridging the Information Gap at Three Mile Island: Radiation Monitoring by Citizens:

This paper describes a unique effort to provide citizens with information about radiation exposure and to rebuild public confidence in information supplied by government agencies. The Citizen Radiation Monitoring Program (CRMP) trained lay citizens to monitor, interpret, and publicize radiation levels in 12 communities surrounding Three Mile Island (TMI) during the purge of radioactive Krypton from the reactor a year after the accident. One outgrowth of the accident was an erosion in the communitys confidence in information about risks associated with the clean-up. Traditional mechanisms for information exchange between government officials and the public had a deviation-amplifying effect on this information gap. The CRMP sought to reverse this effect by (1) involving the communities in the programs design and operation, (2) trusting local citizens to measure and publicize radiation levels, and (3) addressing emotional as well as technical issues associated with the purge. Several implications for disseminating public information are offered.

Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes

Abstract The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway.

Cost and Schedule Reduction of Nuclear Power Plant Construction Using 4D CAD and Immersive Display Technologies

A strategic goal of the Nuclear Engineering Institute as defined in Vision 2020 includes the addition of 50,000 MW of safe and reliable nuclear plant capability in the United States by 2020. To meet this goal, the new plants will need to be developed in an efficient and safe manner. New visualization technologies including 4D CAD and immersive display environments offer promising avenues to improve design, construction, and operation processes of these complex facilities. This paper presents our progress to evaluate the application of 4D CAD and immersive display tools to reduce the time to and costs of constructing new nuclear plants. The results from the initial phase of this research show that the use of a virtual immersion environment to display the 3D plant design is beneficial to the designer and contractor. Our current focus is on the development of tools to allow construction experts to better visualize the construction sequence using 4D CAD so they can provide critical design feedback prior to starting construction. This will ultimately reduce the number of construction interferences in the field and help minimize the quantity of changes during construction.

Development of Parallel Coupling System between Three-Dimensional Nodal Kinetic Code ENTREE and Two-Fluid Plant Simulator TRAC/BF1

The high-speed three-dimensional neutron kinetic code ENTRÉE was developed based on the polynomial and semi-analytical nonlinear iterative nodal methods (PNLM and SANLM) with also introducing the discontinuity factor. In order to enhance the efficiency of transient calculation, the nonlinear correction-coupling coefficients are intermittently updated based on the changing rate of core state variables. By giving the analytical form for two-node problem matrix elements, the additional computing time in SANLM was minimized. A fast algorithm was developed for the multi table macro-cross section rebuilding process. The reactivity component model was implemented based on the variation of the neutron production and destruction terms. The code was coupled with the two-fluid thermal hydraulic plant simulator TRAC/BF1 through PVM or MPI protocols. Two codes are executed in parallel with exchanging the feedback parameters explicitly. Based on the LMW PWR transient benchmark, it was shown that both PNLM and SANLM spend less than 20% excess computing time in comparison with the coarse mesh finite difference method (CFDM). The implementation of the discontinuity factor was verified based on the DVP problem. Adequacy and parallel efficiency of the coupling system TRAC/BF1-ENTREE was demonstrated based on the BWR cold water injection transient proposed by NEA/CRP.

Features and performance of a coupled three-dimensional thermal-hydraulic/kinetics TRAC-PF1/NEM pressurized water reactor (PWR) analysis code

Abstract This paper summarizes the current status of the Pennsylvania State University (PSU) version of the coupled three-dimensional (3-D) thermal-hydraulic/kinetics TRAC-PF1/NEM code for pressurized water reactor (PWR) transient and accident analysis and describes applications to reactivity insertion accident (RIA) simulations as well as recent developments. The TRAC-PF1/NEM methodology utilizes closely coupled 3-D thermal-hydraulics and 3-D core neutronics transient models to simulate the vessel and a 1-D simulation of the primary system. An efficient and flexible cross-section generation procedure was developed and implemented into TRAC-PF1/NEM. These features make the coupled code capable of modeling PWR reactivity transients, including boron dilution transients, in a reasonable amount of computer time. Three-dimensional studies on hot zero power (HZP) rod ejection and main steam line break (MSLB) transients in a PWR, as well as a large break loss-of-coolant-accident (LBLOCA) and boron dilution transients, were accomplished using TRAC-PF1/NEM. The results obtained demonstrate that this code is appropriate for analysis of the space-dependent neutronics and thermal-hydraulic coupled phenomena related to most current safety issues.

Fracture mechanism of borated stainless steel

The mechanical properties and fracture mechanism of irradiated and unirradiated boron containing Type 304 stainless steel are studied. Four different batches with different boron weight percentages are used. One of these batches was manufactured by a conventional wrought technique, while the others were manufactured by a powder metallurgy technique. The irradiated specimens were subjected to a fluence level of 5 × 1019 or 1 × 1021 n/m2. The mechanical and fracture tests were performed at temperatures of 233, 298, and 533 K. No significant effects on the mechanical properties or fracture behavior were observed as a result of neutron irradiation and/or temperature. The ductility and toughness of the borated steel were found to decrease with increasing boron content. The effect of boride on void nucleation and linkage was found to play an important role in the fracture behavior of borated steel.

Determination of the end state of the Three Mile Island Unit 2 accident using neutron transport analysis

Since the March 1979 accident, the source range monitors (SRMs) at Three Mile Island Unit 2 (TMI-2) have been reading several orders of magnitude higher than would be expected in a normal shutdown core. A study in which these anomalous SRM readings are analyzed and the cause determined is reported. Here, the DOT 4.3 two-dimensional transport code was used to simulate the SRM response and the response of an axial string of solid-state track recorders by modeling the neutronics of the damaged TMI-2 core. This modeling has indicated the presence of -- 10 tonnes of fuel material in the lower vessel plenum, a condition that was subsequently verified by direct observation. The computational model, the method of cross-section preparation, and an analysis of the various core neutron sources are described, as well as the results obtained from this effort.

Analysis of the source range monitor during the first four hours of the Three Mile Island Unit 2 accident

The source range monitor (SRM) data recorded during the first 4 h of the Three Mile Island Unit 2 (TMI-2) accident following reactor shutdown were analyzed. An effort to simulate the actual SRM response was made by performing a series of neutron transport calculations. Primary emphasis was placed on simulating the changes in SRM response to various system events during the accident so as to obtain useful information about core conditions at the various stages. Based on the known end-state reactor conditions, the major system events and the actual SRM readings, self-consistent estimates were made of core liquid level, void fraction in the coolant, and locations of core materials. This analysis expands the possible interpretation of the SRM data relative to core damage progression. The results appear to be consistent with other studies of the TMI-2 Accident Evaluation Program, and provide information useful for the development and determination of the TMI-2 accident scenario.

Monitoring Krypton-85 During Three Mile Island Unit 2 Purging

The Penn State noble gas monitor played an important role in measuring environmental levels of /sup 85/Kr during the purging of the Three Mile Island Unit 2 primary containment. It filled a gap in the U.S. Environmental Protection Agency monitoring program, which existed between their real time monitors and their cryogenic gas chromatographic separation technique. During the 15-day purging period, the system analyzed a total of 124 samples, of which 37 were quantified to contain /sup 85/Kr in concentrations ranging from 3 X 10/sup 4/ to 1.5 X 10/sup 6/ pCi/m/sup 3/. The maximum whole body beta dose rate was found to be 0.28 mrem/h.

Multidimensional TMI-1 Main-Steam-Line-Break Analysis Methodology Using TRAC-PF/NEM

Abstract A comparison of a point-kinetics calculation and a full three-dimensional thermal-hydraulic/kinetics calculation using TRAC-PF1/NEM is presented. The coupled TRAC-PF1/NEM methodology uses version 5.4 of the TRAC-PF1/MOD2 code, developed by the Los Alamos National Laboratory, and a special kinetics module, developed at The Pennsylvania State University and based on the nodal expansion method. Cross sections are obtained from two-dimensional tables generated using CASMO-3. The results of the analysis show that the point-kinetics calculation is conservative and predicts a return to power. The three-dimensional analysis shows no return to power despite an extended overfeeding of the affected generator with feedwater. The difference is believed to be caused by the inability of the standard point-kinetics method to properly account for the moderator density feedback, local effects, and flux redistribution, which occur during the transient.