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At least 1,675 records · Page 93
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Validation of MPACT BWR capabilities against critical experiments

Over the years, significant validation work for the neutronics code MPACT has been performed against zero-power critical benchmarks and measured data from operating nuclear power plants. As a part of the Modeling and Analysis of Exelon BWRs for Eigenvalue and Thermal Limits Predictability project, new validation efforts relevant to boiling water reactor (BWR) core applications have been performed. This paper presents the results of the critical experiment portion of the BWR validation efforts for MPACT. The Kritz-4 experiments and IPEN/MB-01 BWR-relevant configurations are modeled with MPACT. The Kritz-4 BWR critical experiments were performed at both cold and hot conditions, which is a unique feature among other critical experiment facilities. The MPACT results with linear source (LS) Method of Characteristics (MOC) have very good agreement with the measured criticality. When using the 60-group BWR library with LS and P2 scattering, the maximum k{sub eff} error is 109 pcm, and the average cold-hot bias is only -0.11 pcm/K. The fission rate distributions of MPACT are also verified with Serpent calculations. For the IPEN facility, two configurations are considered: one with a large central void, and one with a cruciform control rod. The MPACT k{sub eff} errors are less than 50 pcm for all IPEN problems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

VERA BWR progression problems

During the first phase of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program, the Virtual Environment for Reactor Applications (VERA) was developed with a focus on capabilities for high-fidelity, multiphysics simulation of pressurized water reactors (PWRs). During this development effort, a set of progression problems was created ranging from smaller pin cell calculations to larger 3D full-core calculations. These progression problems helped to guide the development of the software and served as benchmarks against which to test VERA. Since 2019, efforts have been made to extend the capabilities of VERA to model boiling water reactors (BWRs). Because BWR simulations come with many unique challenges, a set of BWR progression problems was developed to aid in this new effort. The BWR progression problems range from 2D lattice calculations to 3D mini-core problems, and reference neutronic solutions were computed using continuous-energy Monte Carlo codes. MPACT, one of the neutronics code in VERA, was benchmarked using the BWR progression problems. The code is capable of computing solutions to all problems. The eigenvalues computed by MPACT agree well with the Monte Carlo reference solutions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Hydrogen trapping at neutron irradiation produced defects in recrystallized alpha annealed Zircaloy-4 - 2016-0060

Hydrogen is absorbed into zirconium alloy-clad fuel elements and structural components as a result of water-side corrosion in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments. Increasing hydrogen concentrations in the zirconium alloy affect both its mechanical properties and dimensional stability. It is well known that hydrogen can diffuse relatively rapidly through zirconium base alloys (in relation to the lifetime of the components) in response to solid solution concentration, temperature, and stress gradients. Migration of hydrogen within a component through its life must be considered to reliably predict the performance of the component through life and even post-service life. Many of the physical properties needed to describe this process, such as the solubility of hydrogen in zirconium alloys, diffusion coefficients of hydrogen through zirconium alloys, and the Soret effect within zirconium alloys, already have been published. Many of these measurements have been made outside of the reactor core environment on nonirradiated zirconium alloys. This paper presents in-reactor, isothermal hydrogen diffusion data that show an effect of the in-reactor radiation environment on the final measured hydrogen concentrations. Samples used are similar to the hydrogen diffusion couples used by several investigators in the past to measure the hydride dissolution solvus as a function of temperature. The experimental results are interpreted on the basis of a physical model for hydrogen trapping that hypothesizes that hydrogen in solid solution in the zirconium lattice comes into thermodynamic equilibria with the hydride phase as well as into a steady-state equilibrium with trapping at irradiation-induced defects within the lattice. In agreement with the model, the irradiation effect on final measured concentrations is seen to be a decreasing function of the irradiation temperature between ∼271 deg. C and 354 deg. C, with any effect essentially eliminated by an irradiation temperature of 354 deg. C. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effects of chemistry and microstructure on corrosion performance of Zircaloy-2-based BWR cladding - 2016-0066

In boiling water reactors (BWRs), Zircaloy-2 is typically the material of choice for fuel cladding. With increased focus on fuel performance and reliability, consistent cladding corrosion performance under varying BWR conditions has become an acknowledged requirement for modern BWR fuel. To support optimization of corrosion performance, modifications to alloy chemistry and manufacture conditions can be made, supported by ex-reactor corrosion testing and demonstrated through in-reactor operation. In this work, the combined effects of alloy chemistry and fabrication process were investigated. The effect of alloy chemistry is often complicated by variation within a production ingot. In this work, additional measurements were intentionally taken at the tube-shell stage (before cold pilgering) so that chemistry more representative of the final cladding could be obtained and the effect of iron, nickel, and tin could be examined more closely. Although chemistry range of Zircaloy-2 was the primary focus, the effect of iron above the ASTM limit in a new alloy was also explored. Three types of cladding fabrication processes were considered. Two types included a heat treatment applied after the first cold pilgering and annealing, and the treatment was limited to the outer portion of the cladding and differed in the temperature of the heat treatment; the third type had no special heat treatments after the first cold pilgering and annealing. The weight gain of the three types of cladding with varying compositions under two corrosion test conditions (400 deg. C steam and 410/520 deg. C two-stage steam) are discussed in terms of the initial second-phase particle sizes obtained from transmission electron microscopy examinations. Based on poolside inspections and hot-cell examinations, the in-reactor corrosion performance of the two cladding types up to typical discharge exposures are discussed in the terms of corrosion margin, microstructure evolution, and correlation with ex-reactor corrosion test data. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

The effect of iron on dislocation evolution in model and commercial zirconium alloys - 2016-0068

Although the evolution of irradiation-induced dislocation loops has been well correlated with irradiation-induced growth phenomena, the effect of alloying elements on this evolution remains elusive, especially at low fluences. To develop a more mechanistic understanding of the role iron has on loop formation, we used state-of-the-art techniques to study a proton-irradiated Zr-0.1Fe alloy and proton- and neutron-irradiated Zircaloy-2. The two alloys were irradiated with 2-MeV protons up to 7 dpa at 350 deg. C and Zircaloy-2 up to 14.7 X 10{sup 25} n. m{sup -2}, approximately 24 dpa, in a boiling water reactor at approximately 300 deg. C. Baseline transmission electron microscopy showed that the Zr{sub 3}Fe secondary-phase particles in the binary system were larger and fewer in number than the Zr(Fe,Cr){sub 2} and Zr{sub 2}(Fe,Ni) particles in Zircaloy-2. An analysis of the irradiated binary alloy revealed only limited dissolution of Ze{sub 3}Fe, suggesting little dispersion of iron into the matrix, while at the same time a higher -loop density was observed compared with Zircaloy-2 at equivalent proton dose levels. We also found that the redistribution of iron during irradiation led to the formation of iron nano-clusters. A delay in the onset of -loop nucleation in proton-irradiated Zircaloy-2 compared with the binary alloy was observed. The effect of iron redistributed from secondary-phase particles because of dissolution on the density and morphology of and loops is described. The implication this may have on irradiation-induced growth of zirconium fuel cladding is also discussed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Investigating the effect of zirconium oxide microstructure on corrosion performance: a comparison between neutron, proton, and nonirradiated oxides - 2016-0069

Scanning precession electron diffraction in the transmission electron microscope has been used to simultaneously map the phase, orientation, and grain morphology of oxides formed on Zircaloy-2 after three and six cycles in a boiling water reactor in unprecedented detail. For comparison, a region of a preoxidized autoclave-formed oxide was also proton-irradiated at the Dalton Cumbrian Facility. The proton irradiation was observed to cause additional stabilization of the tetragonal phase that was attributed to the stabilizing effect of irradiation- induced defects in the oxide. In the reactor-formed oxides, no extra stabilization of the tetragonal grains was observed under neutron irradiation, as indicated by the similar tetragonal phase fraction and transformation twin-boundary distributions between the nonirradiated and reactor-formed oxides. It is suggested that the damage rate is too low in the newly formed oxide to cause significant stabilization of the tetragonal phase. This technique also reveals that the oxide formed under reactor conditions has a more heterogeneous microstructure, and the growth of well-oriented columnar monoclinic grains is significantly reduced compared with a nonirradiated oxide. High-angle annular dark-field scanning transmission electron microscopy also revealed the development of extensive networks of intergranular porosity and eventually grain de-cohesion in the reactor-formed oxides. These results suggest that the tetragonal-monoclinic transformation is not responsible for the accelerated corrosion exhibited under reactor conditions. It is proposed that the usual out-of-reactor oxide growth and nucleation processes are significantly modified under reactor conditions, resulting in a more heterogeneous and randomly oriented oxide microstructure with reduced columnar grain growth. It is suggested that this disordered oxide microstructure allows for the formation of extensive intergranular porosity that could lead to accelerated in-reactor corrosion. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

A Provably Accurate Randomized Sampling Algorithm for Logistic Regression

In statistics and machine learning, logistic regression is a widely-used supervised learning technique primarily employed for binary classification tasks. When the number of observations greatly exceeds the number of predictor variables, we present a simple, randomized sampling-based algorithm for logistic regression problem that guarantees high-quality approximations to both the estimated probabilities and the overall discrepancy of the model. Our analysis builds upon two simple structural conditions that boil down to randomized matrix multiplication, a fundamental and well-understood primitive of randomized numerical linear algebra. We analyze the properties of estimated probabilities of logistic regression when leverage scores are used to sample observations, and prove that accurate approximations can be achieved with a sample whose size is much smaller than the total number of observations. To further validate our theoretical findings, we conduct comprehensive empirical evaluations. Overall, our work sheds light on the potential of using randomized sampling approaches to efficiently approximate the estimated probabilities in logistic regression, offering a practical and computationally efficient solution for large-scale datasets.

Chowdhury, Agniva↗
DOE OSTI2025

[LCA UP] LNG Ocean Shipping

This unit process includes operation of a liquefied natural gas (LNG) ocean tanker using steam, DFDE/TFDE, ME-GI or X-DF engines. The tanker is fueled by a combination of boil-off gas (BOG) produced via evaporation of LNG, fuel oil, and diesel.

Group - Transport Process; LCA; LCA Unit Process; ↗
DOE OSTI2025

Distillable amine-based solvents for effective pretreatment of multiple biomass feedstocks

Exploring the potential of advanced distillable solvents as efficient biomass pretreatment agents is critical for biorefineries, enhancing fermentable sugar yields while enabling solvent recovery and recycling without suffering significant losses. Here, we employ distillable amine-based solvents for pretreating a wide range of lignocellulosic feedstocks, aiming to facilitate the industrial release of fermentable sugars from diverse feedstocks through enzymatic hydrolysis. Twenty-two diverse feedstocks, sourced from different geographical regions and representing various biomass categories, were surveyed for chemical (mainly carbohydrates and lignin) and lignin (S, G, and H units) profiles. Several solvents, including ethanolamine, ethanolammonium acetate, butylamine, butylammonium acetate, and triethylamine, were tested for the pretreatment of eight selected biomasses. Among these solvents, butylamine emerged as the most effective due to its favorable sugar release, excellent solvent removal rate, and low boiling point, facilitating solvent recovery and recycling. Extending butylamine pretreatment to all 22 feedstocks demonstrated desirable sugar yields and highly efficient solvent removal in the majority of the biomass sources tested. Agricultural residues and their mixtures showed particularly favorable sugar release. Despite minimal changes in cellulose crystallinity, XRD characterization of sorghum, poplar, and pine before and after butylamine pretreatment showed a decrease in intensity and a slight shift of certain peaks, indicating alterations in cellulose structure. Fourier-transform infrared spectroscopy and thermogravimetric analysis analyses suggested disruption of biomass linkages in hemicellulose and lignin, enhancing enzymatic digestibility. Scale-up experiments of the mixed agricultural feedstocks in a 1 L Parr reactor achieved over 90% glucose liberation and more than 99% butylamine removal, highlighting the scalability of the method. The resulting hydrolysates supported the growth of diverse bacterial and fungal strains, indicating downstream compatibility with commercial fermentation processes. This study presents butylamine as an effective, recoverable pretreatment solvent for a wide range of lignocellulosic feedstocks, offering a promising solution to key biorefinery challenges. The demonstrated scalability and compatibility with various biomass types and blends underscore its potential for industrial application, advancing sustainable biofuel and biochemical production.

biomass composition↗
DOE OSTI2016

Investigation of Thermal Acoustic Effects on SRF Cavities within CM1 at Fermilab

Two TESLA-style 8-cavity cryomodules have been operated at Fermilab Accelerator Science and Technology (FAST), formerly the Superconducting Radio Frequency (SRF) Accelerator Test Facility. Operational instabilities were revealed during Radio Frequency (RF) power studies. These observations were complemented by the characterization of thermal acoustic effects on cavity microphonics manifested by apparent noisy boiling of helium involving vapor bubble and liquid vibration. The thermal acoustic measurements also consider pressure and temperature spikes which drive the phenomenon at low and high frequencies.

43 PARTICLE ACCELERATORS↗
DOE OSTI2013

Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-High Fluence Conditions

Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm 2 , far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing, and fracture toughness master curve issues.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Integrated Test and Evaluation of 3He Replacement Technologies for Safeguards and Nonproliferation Applications – boron-based HLNC-B counter build

A full scale thermal neutron coincidence counter (High Level Neutron Counter – Boron: HLNB) was designed and built during FY12. Following successful functionality demonstration, initial performance evaluation using LANL available neutron and gamma-ray sources was performed during FY13. The results of this activity were shared with international community at the NNSA sponsored workshop on 3 He alternatives organized at LANL in June 2013. Field test of the HLNB is anticipated during FY14 pending funding approval. In addition to HLNB related activities, LANL supported and hosted experimental evaluation of Advanced Boron Uranium Neutron Collar (ABUNCL) designed by GE/RS and evaluated by PNNL. The evaluation involved use of LANL low enriched uranium (LEU) fresh fuel rods in a boiling water reactor (BWR) fuel arrangement. Details of the activities performed during FY13 are described in the following sections.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2013

RELAP-7 Simulation Resolving an SBO Scenario on a Simplified Geometry of a BWR

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). RELAP-7 will become the main reactor systems toolkit for the Risk-Informed Safety Margin Characterization Pathway of the Light Water Reactor Sustainability Program and the next generation tool in the RELAP reactor safety/systems analysis application series (i.e., the replacement for RELAP5). The code is being developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). During Fiscal Year 2013, a number of physical components with two-phase flow capability have been developed to support the simplified boiling water reactor (BWR) station blackout analyses. The case selected for demonstration calculation is built from the specifications documented in an Organization for Economic Cooperation and Development benchmark problem for BWR turbine trip analysis. The reference design for the benchmark problem was from the Peach Bottom-2 nuclear station, which is a General Electric BWR-4 design. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for reactor core isolation cooling (RCIC) and the wet well of a BWR containment. The case was initially run to steady-state with RELAP-7. The station blackout transient simulations were subsequently initiated by using the INL-developed RAVEN code. Two scenarios for the station blackout simulations have been considered. Scenario I represents an extreme station blackout accident with no safety injection functioning. The reactor core could experience dry out fairly quickly with this scenario. Scenario II represents a more probable station blackout accident progression with the RCIC system functioning. In this scenario, the RCIC system is fully coupled with the reactor primary system and the safety injection to provide makeup cooling water to the reactor core from the suppression pool is dynamically simulated. With the RCIC system functioning, the core dry out is significantly postponed when compared to the results from Scenario I. This fully coupled RCIC system simulation capability represents the first-of-a-kind simulation capability. Sensitivity studies have also been carried out to study the effect of RCIC control strategy with varying core makeup cooling water mass flow rates. The next stage of development will be to demonstrate more refined BWR station blackout analyses with more realistic geometries, to be reported in the next demonstration simulation report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Modifications and Optimization of the Organic Rankine Cycle to Improve the Recovery of Waste Heat, Milestone Report #5: Prototype Test

The opportunity for recovery of lost energy in thermal processes is estimated to be approximately 80 GW in the United States. In other words, this could save approximately 450 MM tons of CO 2 from entering the atmosphere. A significant fraction of this energy (about 10.7 GW) is emitted as exhaust (waste) heat from gas turbines at temperatures between 450 and 600°C. Existing waste heat recovery processes (e.g., the OREGEN Organic Rankine Cycle [ORC]) attempt to recover this energy from such high-temperature exhaust streams by employing an intermediate heat transfer oil that protects the ORC working fluid from exposure to the high temperatures in the exhaust stream because the high temperatures (about 300°C) would cause thermal breakdown of the organic working fluid in the ORC. The direct evaporator technology developed as part of this project allows direct contact between the hot exhaust and the ORC working fluid and mitigates the risk of fluid degradation by employing a “protective staging” (shown in Figure 1). The direct evaporator experiment conducted as part of this work aims to validate the premise of the direct evaporator in an ORC cycle: heat transfer from a hot exhaust gas at 500°C to a cyclopentane working fluid at conditions representative of an ORC cycle with boiling and superheating of the cyclopentane without risking thermal degradation. In addition, the following related data were collected for the 300-hour duration test: • Data that show the temperature of the working fluid did not exceed 300°C, • Degradation data from samples of the working fluid collected at various times during the test, and • A measure of fouling in the heat exchanger (HX) tubes (caused by thermal breakdown of the working fluid).

42 ENGINEERING↗
DOE OSTI2013

Investigation of the Performance of D 2 O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

This report presents FY13 activities for the analysis of D 2 O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D 2 O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D 2 O coolant, MOX with different 235 U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233 U over other fissile like 239 Pu and 235 U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239 Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void reactivity until the target burnup of 51 GWd/t. Several other distinctions between U-Pu and Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D 2 O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D 2 O-HCPWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Discoloration of the wetted surface in the 6.1D dissolver

During a camera inspection of a failed coil in the 6.1D dissolver, an orange discoloration was observed on a portion of the dissolver wall and coils. At the request of H-Canyon Engineering, the inspection video of the dissolver was reviewed by SRNL to assess if the observed condition (a non-uniform, orange-colored substance on internal surfaces) was a result of corrosion. Although the dissolver vessel and coil corrode during dissolution operations, the high acid conditions are not consistent with the formation of ferrous oxides (i.e., orange/rust-colored corrosion products). In a subsequent investigation, SRNL performed dissolution experiments to determine if residues from the nylon bags used for Pu containment could have generated the orange discoloration following dissolution. When small pieces of a nylon bag were placed in boiling 8 M nitric acid solutions containing other components representative of the H-Canyon process, complete dissolution occurred almost immediately. No residues were obtained even when a nylon mass to volume ratio greater than 100 times the 6.1D dissolver value was used. Degradation products from the dissolution of nylon bags are not responsible for the discoloration observed in the dissolver.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

System Study: Reactor Core Isolation Cooling 1998–2012

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗