11 research outputs found
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Effects of condensation modeling on transient behavior of pressurized water reactors
In simulating pressurized water reactor (PWR) transients with large-scale systems codes such as TRAC and RELAP, the effect of condensation has been recognized as a controlling mechanism in the prediction of plant response. For transients involving contraction of or loss of primary coolant, the rate of condensation (primarily in the pressurizer) controls the system refill characteristics. Several separate but interacting phenomena occur during the process of pressurizer refill: steam compression, system heat losses, thermal stratification or mixing of liquid, and condensation. The relative importance of each of these processes and the degree of interaction between them during different transients is very complex. The existing condensation models do not adequately describe the interplay between these effects and this leads to uncertainties in the predicted system response. Further experimental data and code assessment are required to provide data necessary for improving condensation models. Three examples of transients involving uncertainties introduced by condensation modeling are (1) pressurized thermal shock (PTS) transients, (2) small break loss-of-coolant accidents (SBLOCA), and (3) steam generator tube ruptures (SGTR)
The influence of localized states charging on 1/f^{\alpha} tunneling current noise spectrum
We report the results of theoretical investigations of low frequency
tunneling current noise spectra component (1/f^{\alpha}). Localized states of
individual impurity atoms play the key role in low frequency tunneling current
noise formation. It is found that switching "on" and "off" of Coulomb
interaction of conduction electrons with one or two charged localized states
results in power law singularity of low-frequency tunneling current noise
spectrum 1/f^{\alpha}. Power law exponent in different low frequency ranges
depends on the relative values of Coulomb interaction of conduction electrons
with different charged impurities.Comment: 7 pages, 5 figure
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Risk-based analysis for prioritization and processing in the Los Alamos National Laboratory 94-1 program
A previous report, {open_quotes}Analysis of LANL Options for Processing Plutonium Legacy Materials,{close_quotes} LA-UR-95-4301, summarized the development of a risk-based prioritization methodology for the Los Alamos National Laboratory (LANL) Plutonium Facility at Technical Area-55 (TA-55). The methodology described in that report was developed not only to assist processing personnel in prioritizing the remediation of legacy materials but also to evaluate the risk impacts of schedule modifications and changes. Several key activities were undertaken in the development of that methodology. The most notable was that the risk assessments were based on statistically developed data from sampling containers in the vault and evaluating their condition; the data from the vault sampling programs were used as the basis for risk estimates. Also, the time-dependent behavior of the legacy materials was explicitly modeled and included in the risk analysis. The results indicated that significant reductions in program risk can be achieved by proper prioritization of the materials for processing
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Steam-generator-tube-rupture transients for pressurized-water reactors
Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures
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Small-break LOCA recovery in B and W plants. [PWR]
A break of approximately 0.0012 m/sup 2/ in the cold leg of a B and W plant results in an interruption of natural circulation when steam accumulates in the hot-leg U-bend. A small-break loss-of-coolant accident of this size was simulated by TRAC-PF1 to evaluate strategies for recovery and for re-establishing natural circulation. In the absence of operator action, core cooling occurs when water supplied by the high-pressure-injection system boils, then is discharged through the break. Raising the steam-generator secondary level, venting steam from the steam-generator secondary, venting steam from the hog-leg U-bend, bumping the reactor-coolant pumps, and injecting a portion of the high-pressure-injection system into the hot-leg U-bend aided in cooling and depressurizing the primary system but were ineffective in re-establishing natural-circulation flows in the primary-coolant loops
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Systems analysis and simulation of fissile materials disposition alternatives
A detailed process flow model has been developed for use in the Fissile Materials Disposition program. The model calculates fissile material flows and inventories among the various processing and storage facilities over the life of the disposition program. Given existing inventories and schedules for processing, we can estimate the required size of processing and storage facilities, including equipment requirements, plant floorspace, approximate costs, and surge capacities. The model was designed to allow rapid prototyping, parallel and team development of facility and sub-facility models, consistent levels of detail and the use of a library of generic objects representing unit process operations
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Special small-break applications with TRAC. [PWR]
Input models for the Transient Reactor Analysis Code (TRAC) are described and applications of these models to reactor transients involving small breaks in the primary coolant pressure boundary are demonstrated. The operation of the primary overpressure protection system (relief and safety valves) and the thermal-hydraulic response of the reactor to these transients are obtained from numerical simulations. Also, the effects of steam generator recirculation, steam generator tube rupture, Emergency Core Cooling (ECC) injection and reactivity feedback on the course and consequences of these transients are investigated. These models allow reliable predictions of accident signatures that can help determine the adequacy of equipment and procedures at nuclear power plants to prevent and to control severe accidents
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Irradiation performance of helium-bonded uranium--plutonium carbide fuel elements. [LMFBR]
The current irradiation program of helium-bonded uranium--plutonium carbide elements is achieving its original goals. By August 1978, 15 of the original 171 helium-bonded elements had reached their goal burnups including one that had reached the highest burnup of any uranium--plutonium carbide element in the U.S.--12.4 at.%. A total of 66 elements had attained burnups over 8 at.%. Only one cladding breach had been identified at that time. In addition, the systematic and coordinated approach to the current steady-state irradiation tests is yielding much needed information on the behavior of helium-bonded carbide fuel elements that was not available from the screening tests (1965 to 1974). The use of hyperstoichiometric (U,Pu)C containing approx. 10 vol% (U,Pu)/sub 2/C/sub 3/ appears to combine lower swelling with only a slightly greater tendency to carburize the cladding than single-phase (U,Pu)C. The selected designs are providing data on the relationship between the experimental parameters of fuel density, fuel-cladding gap size, and cladding type and various fuel-cladding mechanical interaction mechanisms
Roles for process monitoring in support of nuclear materials accounting
The Next Generation Safeguards Initiative includes a process monitoring (PM) project with near-term and long-term objectives. One role for PM is to support nuclear materials accounting (NMA), particularly in cases where NMA cannot meet loss detection goals, such as at large throughput facilities. At large facilities, interim inventory verifications are frequent, perhaps every 10 days, and there is limited ability to reduce in-process inventory, which is often measured with relatively high uncertainty. Therefore, inprocess inventory can have a relatively large impact on NMA performance as measured by the standard error of the material balance. In the context of safeguarding a large aqueous reprocessing facility to recover plutonium from spent nuclear fuel, this paper describes four roles for solution monitoring (SM) as a type of PM in support of NMA. First, SM helps understand facility status at the time of interim inventory. Second, SM can provide a by-difference estimate and associated uncertainty of material holdup in process equipment that is not directly measureable but is bracketed by measurement points. Third, SM can assess the adequacy of measurement error models such as those used to quantify the uncertainty in solution volume measurements. Fourth, SM together with models of unit operations can provide an inferred or estimated book value for waste or other low-Pu-mass streams that allows tighter control limits than are possible with NMA alone
A dissolver diversion scenario illustrating the value of process monitoring
In large throughput spent nuclear fuel reprocessing plants such as the Rokkasho Reprocessing Plant (RRP) there is a low detection probability for material losses of interest to the IAEA (8 kg of Pu) using even the most optimistic near-real-time accounting (NTRA) methods currently employed. A particularly low detection probability is seen in the head end where “input” shipper declarations (via reactor burnup calculations) having relatively large uncertainties (5-10%) are compared to “output” measurements consisting of waste (leached hulls) measurements plus accountability tank measurements. Currently, a dissolver monitoring system applied by the IAEA utilizes semi-quantitative neutron assay of hull batches to detect changes in the neutron count rate that could indicate excess Pu in the leached hulls. The goal of the exercise reported in this paper is to provide an alternative dissolver process monitoring concept. The approach is to infer the completeness of spent fuel dissolution from easily- monitored process parameters. To provide a framework, a scenario was developed and evaluated where fuel and its contained Pu is purposely left undissolved, resulting in excess Pu in the hulls. The magnitude of the scenario was calculated based on the loss of 8 kg of Pu over the course of 90 working days. Based on the chemical models and material balance calculations presented here, relatively large changes in temperature, acid concentration or reaction time are needed for the stated material loss. Further, these process changes would be easily observable using current process monitoring technologies, but further work is needed to evaluate authentication strategies and performance under plant and long term conditions. Total uncertainties will depend upon the errors associated with model calculations and measurement errors. Estimation of these uncertainties is the next logical step for understanding the value of process monitoring in this scenario