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DOE OSTI2013

Thermal Modeling Studies for Active Storage Modules in the Calvert Cliffs ISFSI

Temperature measurements obtained for two storage modules in the Calvert Cliffs Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI) as part of the Used Fuel Disposition Campaign of the Department of Energy (DOE) were used to perform validation and sensitivity studies on detailed computational fluid dynamics (CFD) models of the concrete storage modules, including the dry storage canister within the modules. The storage modules in the Calvert Cliffs Nuclear Power Station’s ISFSI are a site-specific version of the standard NUHOMS® HSM. The two modules inspected each contained a 24P DSC loaded with 24 CE 14x14 spent fuel assemblies. The thermal analysis was performed using the STAR-CCM+ package, and the models developed for the specific ISFSI modules yielded temperature predictions in actual storage conditions for the concrete structure, the DSC and its contents, including preliminary estimates of fuel cladding temperatures for the used nuclear fuel. The results of this work demonstrate that existing CFD modeling tools can be used to obtain reasonable and accurate detailed representations of spent fuel storage systems with realistic decay heat loadings when the model omits specific conservatisms and bounding assumptions normally used in design-basis and safety-basis calculations. This paper presents sensitivity studies on modeling detail (for the storage module and the DSC), boundary conditions, and decay heat load, to evaluate the effect of the modeling approach on predicted temperatures and temperature distributions. Because nearly all degradation mechanisms for materials and structures comprising dry storage and transportation systems are dependent on temperature, accurate characterization of local temperatures and temperature gradients that the various components of these systems will experience over the entire storage period has been identified as a primary requirement for evaluation of very long term storage of used nuclear fuel.

long term storage↗

DOE OSTI2015

In-service Inspection of Radioactive Waste Tanks at the Savannah River Site – 15410

Liquid radioactive wastes from the Savannah River Site (SRS) separation process are stored in large underground carbon steel tanks. The high level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks will be required to be in service for times exceeding the initial intended life. Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. A structural integrity program is in place to maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. In-service inspection (ISI) is an essential element of a comprehensive structural integrity program for the waste tanks at the Savannah River Site (SRS). The ISI program was developed to determine the degree of degradation the waste tanks have experienced due to service conditions. As a result of the inspections, an assessment can be made of the effectiveness of corrosion controls for the waste chemistry, which precludes accelerated localized and general corrosion of the waste tanks. Ultrasonic inspections (UT) are performed to detect and quantify the degree of general wall thinning, pitting and cracking as a measure of tank degradation. The results from these inspections through 2013, for the 27 Type III/IIIA tanks, indicate no reportable in-service corrosion degradation in the primary tank (i.e., general, pitting, or cracking). The average wall thickness for all tanks remains above the manufactured nominal thickness minus 0.25 millimeter and the largest pit identified is approximately 1.70 millimeter deep (i.e., less than 10% through-wall). Improvements to the inspection program were recently instituted to provide additional confidence in the degradation rates. Thickness measurements from a single vertical strip along the accessible height of the primary tank have been used as a baseline to compare historical measurements. Changes in wall thickness and pit depths along this vertical strip are utilized to estimate the rate of corrosion degradation. An independent review of the ISI program methodology, results, and path forward was held in August 2009. The review recommended statistical sampling of the tanks to improve the confidence of the single strip inspection program. The statistical sampling plan required that SRS increase the amount of area scanned per tank. Therefore, in addition to the baseline vertical strip that is obtained for historical comparisons, four additional randomly selected vertical strips are inspected. To date, a total of 104 independent vertical strips along the height of the primary tank have been completed. A statistical analysis of the data indicates that at this coverage level there is a 99.5% confidence level that one of the worst 5% of all the vertical strips was inspected. That is, there is a relatively high likelihood that the SRS inspection program has covered one of the most corroded areas of any of the Type III/IIIA waste tanks. These data further support the conclusion that there are no significant indications of wall thinning or pitting. Random sampling will continue to increase the confidence that one of the worst 5% has been inspected. In order to obtain the additional vertical strips, and minimize budget and schedule impacts, data collection speed for the UT system was optimized. Prior to 2009, the system collected data at a rate of 32 square centimeters per minute. The scan rate was increased to 129 - 160 square centimeters per minute by increasing the scanner step and pixel sizes in the data acquisition set-up. Laboratory testing was utilized to optimize the scan index/pixel size such that the requirements for wall thinning and pit detection were still maintained. SRS continues to evaluate improvements to ultrasonic equipment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Digital Sensor Technology

The nuclear industry has been slow to incorporate digital sensor technology into nuclear plant designs due to concerns with digital qualification issues. However, the benefits of digital sensor technology for nuclear plant instrumentation are substantial in terms of accuracy and reliability. This paper, which refers to a final report issued in 2013, demonstrates these benefits in direct comparisons of digital and analog sensor applications. Improved accuracy results from the superior operating characteristics of digital sensors. These include improvements in sensor accuracy and drift and other related parameters which reduce total loop uncertainty and thereby increase safety and operating margins. An example instrument loop uncertainty calculation for a pressure sensor application is presented to illustrate these improvements. This is a side-by-side comparison of the instrument loop uncertainty for both an analog and a digital sensor in the same pressure measurement application. Similarly, improved sensor reliability is illustrated with a sample calculation for determining the probability of failure on demand, an industry standard reliability measure. This looks at equivalent analog and digital temperature sensors to draw the comparison. The results confirm substantial reliability improvement with the digital sensor, due in large part to ability to continuously monitor the health of a digital sensor such that problems can be immediately identified and corrected. This greatly reduces the likelihood of a latent failure condition of the sensor at the time of a design basis event. Notwithstanding the benefits of digital sensors, there are certain qualification issues that are inherent with digital technology and these are described in the report. One major qualification impediment for digital sensor implementation is software common cause failure (SCCF).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2014

Seismic risk management solution for nuclear power plants

Nuclear power plants should safely operate during normal operations and maintain core-cooling capabilities during off-normal events, including external hazards (such as flooding and earthquakes). Management of external hazards to expectable levels of risk is critical to maintaining nuclear facility and nuclear power plant safety. Seismic risk is determined by convolving the seismic hazard with seismic fragilities (capacity of systems, structures, and components). Seismic isolation (SI) is one protective measure showing promise to minimize seismic risk. Current SI designs (used in commercial industry) reduce horizontal earthquake loads and protect critical infrastructure from the potentially destructive effects of large earthquakes. The benefit of SI application in the nuclear industry is being recognized and SI systems have been proposed in American Society of Civil Engineer Standard 4, ASCE-4, to be released in the winter of 2014, for light water reactors facilities using commercially available technology. The intent of ASCE-4 is to provide criteria for seismic analysis of safety related nuclear structures such that the responses to design basis seismic events, computed in accordance with this standard, will have a small likelihood of being exceeded. The U.S. nuclear industry has not implemented SI to date; a seismic isolation gap analysis meeting was convened on August 19, 2014, to determine progress on implementing SI in the U.S. nuclear industry. The meeting focused on the systems and components that could benefit from SI. As a result, this article highlights the gaps identified at this meeting.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2012

Preliminary Public Design Report for the Texas Clean Energy Project: Topical Report - Phase 1, June 2010-July 2011

Summit Texas Clean Energy, LLC (Summit) is developing the Texas Clean Energy Project (TCEP or the project) to be located near Penwell, Texas. The TCEP will include an Integrated Gasification Combined Cycle (IGCC) plant with a nameplate capacity of 400 megawatts electric (MWe), combined with the production of urea fertilizer and the capture, utilization and storage of carbon dioxide (CO 2 ) sold commercially for regional use in enhanced oil recovery (EOR) in the Permian Basin of west Texas. The TCEP will utilize coal gasification technology to convert Powder River Basin sub-bituminous coal delivered by rail from Wyoming into a synthetic gas (syngas) which will be cleaned and further treated so that at least 90 percent of the overall carbon entering the facility will be captured. The clean syngas will then be divided into two high-hydrogen (H 2 ) concentration streams, one of which will be combusted as a fuel in a combined cycle power block for power generation and the other converted into urea fertilizer for commercial sale. The captured CO 2 will be divided into two streams: one will be used in producing the urea fertilizer and the other will be compressed for transport by pipeline for offsite use in EOR. The TCEP was selected by the U.S. Department of Energy (DOE) Office of Fossil Energy (FE) for cost-shared co-funded financial assistance under Round 3 of its Clean Coal Power Initiative (CCPI). A portion of this financial assistance was budgeted and provided for initial development, permitting and design activities. Front-end Engineering and Design (FEED) commenced in June 2010 and was completed in July 2011, setting the design basis for entering into the detailed engineering phase of the project. During Phase 1, TCEP conducted and completed the FEED, applied for and received its air construction permit, provided engineering and other technical information required for development of the draft Environmental Impact Statement, and completed contracts for the sale of all of the urea and most of the CO 2 . Significant progress was made on the contracts for the purchase of coal feedstock from Cloud Peak Energy’s Cordero Rojo mine and the sale of electricity to CPS Energy, as well as a memorandum of understanding with the Union Pacific Railroad (UPRR) for delivery of the coal to the TCEP site.

Mattes, Karl↗

DOE OSTI2016

Transient behavior of a scaled RCCS test facility under postulated fault and accident scenarios

Tests were performed on the Natural convection Shutdown heat removal Test Facility (NSTF) to simulate design basis accident and postulated fault scenarios. Residing at Argonne National Laboratory, the NSTF stands nearly 26-m in total height and reflects a ½ scale reactor cavity cooling system (RCCS) for high temperature gas cooled reactors. The following manuscript details three test conditions performed on the experimental test facility. The first simulated the reactor pressure vessel (RPV) boundary condition during depressurized conduction cool down accident with small primary leak, and was repeated during both winter and summer seasons. The second examined a short-circuit break between the inlet and outlet flow paths, and was performed in three incremental stages of nominal flow area break size. The third and final test case studied system behavior with varying amounts of cooling channel blockages, up to and including 50% flow areas. Nominal component temperatures, heat removal performance, and system stability will be presented to characterize the behavior at these conditions.

decay heat removal↗

DOE OSTI2016

Nondestructive Examination for Nuclear Power Plant Cable Aging Management Programs

cable aging inspection↗

DOE OSTI2017

The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions

This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding was subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings was compared to that of hydrided zirconium-based nuclear fuel cladding alloy. The hoop strain evolution during pressure pulses were collected in situ; the permanent diametral strains of both accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture.

Accident tolerant fuel↗

DOE OSTI2015

Impacts of Vehicle (In)Security

Nuclear and radioactive material is routinely transported worldwide every day. Since 2010, the complexity of the transport vehicle to support such activities has grown exponentially. Many core functions of a vehicle are now handled by small embedded computer modules with more being added each year to enhance the owner’s experience and convenience. With a system as complex as today’s automobile, the potential for cyber security issues is certain. Hackers have begun exploring this new domain with public information increasingly disseminated. Because vehicles are allowed into and around secure nuclear facilities, the potential for using a vehicle as a new cyber entry point or vector into the facility is now plausible and must be mitigated. In addition, compromising such a vehicle could aide in illicit removal of nuclear material, putting sensitive cargo at risk. Because cyber attacks can now be introduced using vehicles, cyber security, needs to be integrated into an organization’s design basis threat document. Essentially, a vehicle now extends the perimeter for which security professionals are responsible.Electronic Control Units (ECU) responsible for handling all core and ancillary vehicle functions are interconnected using the controller area network (CAN) bus. A typical CAN network in a modern automobile contains 50 or more ECUs. The CAN protocol now supports a wide variety of areas, including automotive, road transportation, rail transportation, industrial automation, power generation, maritime, military vehicles, aviation, and medical devices. In many ways, the nuclear industry is employing the CAN bus protocol or other similar broadcast serial networks. This paper will provide an overview of the current state of automobile and CAN Bus security, as well as an overview of what has been publicly disclosed by many research organizations. It will then present several hypotheses of how vehicle security issues may impact nuclear activities. An initial discussion of how a vehicle can be used as a new threat vector to penetrate secure facilities will be presented. This includes how a modern automobile can be used as the exploitation mechanism for nearby devices such as laptops, cell phones, and wireless access points. Additional discussion will highlight how vehicle security might impact transportation of nuclear material through remote exploitation of a moving vehicle. The final discussion will include what possible implications might be relative to the physical protection systems at nuclear facilities. The audience will also be given details regarding the complexity of attack, thus implying the likelihood of successful exploitation, and information on how such attacks may be mitigated. Emerging security products for automobiles will be discussed and other mitigation methods will be detailed (e.g. disabling vehicle cellular modems). As a result, the audience will have a greater understanding of how to add vehicle security as a part of a comprehensive nuclear security policy.Finally, this paper will highlight the similarities between CAN Bus and other broadcast serial bus networks such as Profibus or DeviceNet, helping educate the reader on how susceptible this type of networking is to nefarious attacks and how it might affect components connected to many different nuclear systems, including control systems, safety systems, emergency systems, and support systems.

47 OTHER INSTRUMENTATION↗

DOE OSTI2018

Bison Fuel Performance Modeling on RIA in a PWR

This report is prepared to fulfill a Level 2 milestone for CASL on “deliver and document RIA fuel performance modeling using Bison fuel performance code.” The work performed is focused on modeling the full length commercial fuel rod response under a design basis Reactivity Initiated Accident (RIA), in particular, a hypothetical Control Rod Ejection Accident (CREA) in a Pressurized Water Reactor (PWR). Fuel rods with different burnups, 0 (fresh fuel), 20, 40, and 60 GWd/tU have been modeled with a reactivity worth insertion of $1.24 at Hot Zero Power (HZP) condition. Bison failure modeling method, and a brief summary of validations on selected RIA cases from CABRI and NSRR testing facilities are also described. The work performed has demonstrated the successful application of Bison code in the modeling of thermal and mechanical responses in RIAs for PWRs. In particular, Bison code has a more mature capability in the modeling of the frictional contact between pellet and cladding of the high burnup fuels, which is critical for modeling the Pellet Clad Mechanical Interaction (PCMI). The modeling of RIA using Bison code shows fuels in PWR environment could have different responses than in the simulated RIA tests, and results indicate that a transition from PCMI to Departure from Nucleate Boiling (DNB) could happen at relatively lower enthalpies: a better heat transfer from the fuel to coolant could increase cladding temperature and ductility, but at the same time also reduces the margin to DNB, which could cause high temperature oxidation and ballooning type of failures. The precise determination of the transition into DNB may rely on a more accurate input of the power histories in the simulated RIAs from VERA, the core simulator of CASL. The limiting conditions in PWRs could be quite different than those tests with non-typical testing conditions, and the improvement of the modeling of thermal responses, such as, clad coolant heat transfer, gap heat transfer, clad thermal inertia, and radial power profiles could be important considerations of the RIAs. In modeling the fuel responses in a fast transient, a number of variables could affect the initial burnup states, and some material models, e.g., fuel mechanical models, fission gas release, and corrosion models, and parameters can affect modeling results as well. Those uncertainties and advanced modeling capabilities need to be investigated in the future study of RIAs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants using Non-destructive Evaluation

License renewal of existing nuclear power plants (NPPs) to extend operation from 40 to 60 to 80 years will require effective aging management programs for the 100’s of kilometers of electric power, control and instrumentation cables installed in each plant. Degradation of jacket, insulation and other cable components can reduce the ability of cables to withstand design-basis accidents. The groundwork for a research and development roadmap to address aging cable management in NPPs was laid at a July 2012 workshop1 that brought together subject matter experts from the U.S. Nuclear Regulatory Commission (NRC), the U.S. Department of Energy (DOE) national laboratories, the Electric Power Research Institute (EPRI), universities, and cable manufacturers and inspectors. The most common nondestructive evaluation (NDE) methods for gauging cable status include indentation and visual inspection. These methods are limited to evaluation of the surface easily accessed cable assemblies. Other current methods (such as time and frequency domain analysis, inductance and capacitance measurements, tan d, etc.) provide flaw detection, but are not predictive. The workshop identified the need to advance and develop new NDE methods to enable in-situ cable condition assessment. Key chemical, physical, and electrical indicators of cable aging with exposure to heat, radiation, moisture, chemicals, and mechanical stress need to be determined. Finally, models correlating the key cable indicators with NDE signals need to be developed that can aging cable remaining useful life.

NDE, Nuclear Cables, Polymer Aging↗

DOE OSTI2019

RIA Testing in TREAT

The Transient Reactor Test facility (TREAT) was constructed in the late 1950’s. The reactors flexible design and multi-mission nature provided thousands of transient irradiations for numerous reactor fuels and transient types before being placed in standby in 1994. The reactor resumed operations in 2017 to reclaim its crucial role in nuclear-heated safety research. A principal goal of restarting the reactor was to perform transient testing of Accident Tolerant Fuels (ATF), supporting their safe adoption by the commercial Light Water Reactor (LWR) fleet. In support of the Accident Tolerant Fuels mission, multiple capsules have been designed and are currently being designed to test fuel specimens under varying conditions. The Minimal Activation Retrievable Capsule Holder (MARCH) system was designed to support high throughput testing by using a modular approach where an outer containment structure will act as the primary safety containment. The MARCH system is designed to house various modules for testing different fuel systems in different environments. The near-term tests for the ATF concepts will use the Static Environment Rodlet Transient Test Apparatus (SERTTA), aka MARCH-SERTTA. MARCH-SERTTA is capable of testing a single ATF rodlet 10 cm in length in a static pool of water with initial conditions up to 200°C and 500 psi. The purpose of the MARCH-SERTTA device is to perform integral tests evaluating the Pellet Clad Mechanical Interaction (PCMI) performance of LWR fuel concepts in design basis Reactivity Initiated Accidents (RIA). This will be accomplished by testing a variety of ATF fuel concepts under varying initial thermal-hydraulic conditions and varying prompt energy pulses. The flexibility of TREAT and the MARCH-SERTTA capsule enables testing of a unique, non-fueled, rod specimen under similar conditions as the fueled specimens to perform separate-effect tests. This non-fueled rod will be used to study the thermal-hydraulic behavior (specifically determine the critical heat flux) during these fast prompt transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

ELECTRICAL BREAKDOWN STRENGTH AND AC WITHSTAND IN HARVESTED EPR INSULATIONS FOR NUCLEAR POWER PLANTS

While the operating voltages for some instrument and control (I&C) cables can be greater than 4 kV to operate pumps, controls, and provide auxiliary power, the operating voltage for most I&C cables is near 600 V. Although the electrical stress is not nearly as high in the insulation of these cables, an ac withstand two and one-half times greater than the operating voltage after a simulated design basis exposure (DBE) has been used in environmental qualification (EQ) to determine the ability of an insulation to meet a 40-year lifetime at an operating voltage. While multiple cable lengths greater than 10 feet are needed to replicate this EQ, systematic measurement of the electrical insulation strength in harvested cable insulations can be beneficial to better understand the remaining useful life of the insulation. This understanding can only be realized through systematic comparison to some level of DBE, and electrical insulation strength measurement can provide a baseline for comparison with other mechanical, chemical, and electrical characterizations that have been utilized. Accelerated aging of two types of ethylene propylene rubber (EPR) insulations from harvested cables was carried out, and the electrical breakdown strength and indenter modulus for both insulations were determined. Results showed that degradation of electrical breakdown strength occurred in only one type of BIW insulation and that this degradation did track with that measured by indenter modulus. The methodology for using Weibull statistics for electrical breakdown strength relative to time to failure is discussed.

Duckworth, Robert↗

DOE OSTI2018

Insight into Thermal Aging of Jacket-bonded Ethylene-propylene Rubber Cable Insulation

Electrical cables are integral to nuclear power plant operation and control during power generation, during planned outages, and during design basis events. Polymeric insulation and jacketing used in cable construction does age over time as a function of environmental stresses including heat, moisture, and radiation. Degradation of polymer mechanical properties can lead to breaks in the insulation and consequent electrical shorting. In this work we consider the thermal aging behavior of one of the two most common insulation polymer classes in nuclear power plants, ethylene-propylene rubber (EPR), both as an independent material and within a cable in which the EPR is bonded to a chlorinated polyethylene (CPE) jacket. Understanding the changes in these two materials with aging, both independently and in concert, will allow the establishment of aging lifetime curves and enable the development of effective non-destructive methods for tracking the condition of installed cables composed of similar material systems. Additional and more effective methods for monitoring the condition of installed cables are needed as a technical basis for continued safe use of aging cables in long term operation and to minimize operating costs for cable aging management through better informed retain/repair/replace decisions.

Light Water Reactors, electrical cables, aging, EP↗

DOE OSTI2017

Tracking Aging in Nuclear Electrical Cable Polymers

The electrical cable network is the nerve center of a nuclear plant, providing power, instrumentation and control of critical systems during operation, planned outage, and during design basis, unplanned outage events. It is therefore important to manage cable system health through awareness of local environments adverse to cable longevity, monitoring of cable condition, and predictive understanding of cable degradation [1-3]. Nondestructive methods can be used to assess the condition of the full length of installed cables and of localized cable regions to support cable repair and replacement planning decisions [4-5]. Informed knowledge of the effects of environment on cable performance over time combined with cable state assessment data can minimize costs associated with cable management by both avoiding outages and ramifications from unexpected cable failure and by avoiding premature replacement of functional cables. In this work, the aging behavior of commonly found nuclear cable insulation and jacketing materials is investigated along with established and emerging methods for locating and quantifying cable degradation. Ethylene-propylene rubber (EPR) and chlorinated polyethylene (CPE), examples of prevalent insulation and jacket materials used in nuclear power plants (NPPs), are thermally aged at elevated temperature to induced material changes similar to those experienced by cable material under long term service conditions. Destructive and nondestructive laboratory tests are used to follow changes in the materials with aging toward establishing predictive capabilities to help extend the life of installed nuclear cables and reduce cable management operational costs.

Cable aging, EPR, CPE, NDE, nuclear↗

DOE OSTI2018

Defect Length and Profile Influence on Frequency Domain Reflectometry

Degradation of the cable jacket, electrical insulation, and other cable components of installed cables within nuclear power plants (NPPs) is known to occur as a function of age, temperature, radiation, and other environmental factors. Although system tests verify cable function under normal loads, demonstration of some cable’s ability to perform under exceptional loads associated with design-basis events is essential to assuring plant integrity. The cable’s ability to perform safely over the initial 40 year planned and licensed life has generally been demonstrated and there have been very few age-related cable failures. With greater than 1000 km of power, control, instrumentation, and other cables typically found in an NPP, replacing all the cables would be a severe cost burden. Justification for life extension to 60 and 80 years requires a cable aging management program that includes condition monitoring to justify cable performance under normal operation as well as accident conditions. A variety of tests are available to assess various aspects of electrical and mechanical cable performance, but none are suitable for all cable configurations nor does any single test confirm all features of interest. One particularly promising test that is beginning to be used more and more by utilities is frequency domain reflectometry (FDR). FDR is a nondestructive electrical inspection technique used to detect and localize faults in power and communication system conductors along the length of a cable from a single connection point. FDR detects discontinuities in the electrical impedance that arise due to cable splices or similar changes along the path of the conductor pair. In addition, FDR has the potential to provide sensitivity to insulation degradation by detecting small changes in impedance between the cable conductors being examined. The technique is also sensitive to cable bends, the particular lay of the cable in tray, proximity to other cable, and other factors that bear consideration when interpreting the test results. This paper examines various influences on the FDR approach particularly using FEM modeling methods and focusing on defect length and defect profile.

Cable Inspection, Frequency Domain Reflectometry, ↗

DOE OSTI2021

Modeling FLEX Human Actions Using the EMRALD Dynamic Risk Assessment Tool

Most current efforts have modified and applied existing human reliability analysis (HRA) methods to treat human actions related to beyond-design-basis external events in which diverse and flexible coping strategies (FLEX) equipment would be deployed. However, many questions remain regarding the suitability of legacy HRA methods to address FLEX human actions, sparking the need for a new method of reasonably evaluating them. In this context, Idaho National Laboratory (INL) researched a relatively new approach to treating FLEX human actions via Event Modeling Risk Assessment Using Linked Diagrams (EMRALD) software. EMRALD was developed to support the increasing need for dynamic probabilistic risk assessment (PRA) models that can respond to evolving plant conditions during simulations. A couple benefits were identified when analyzing FLEX human actions through this software. In general, it is especially useful for evaluating a strategy’s feasibility, including FLEX human actions that require a relatively long time to perform. In this paper, we suggest how FLEX human actions can be modeled using the EMRALD software. Two different HRA modeling approaches using EMRALD are introduced: (1) procedure-based modeling and (2) PRA/HRA-based modeling. The former approach was introduced in the authors’ previous paper, whereas this paper mainly discusses how the latter approach works for an extended loss of AC power (ELAP) scenario with relevant procedures and PRA models. A hybrid method combining the two modeling approaches is introduced at the paper’s conclusion.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2012

Development of an object-oriented dynamics simulator for a LFR DEMO

A control-oriented dynamics simulator for a Generation IV Lead-cooled Fast Reactor (LFR) demonstrator (DEMO) has been developed aimed at providing a flexible, simple and fast-running tool allowing to perform design-basis transient and stability analyses, and to lay the foundations for the study of the system control strategy. For such purposes, a model representing a compromise between accuracy and straightforwardness has been necessarily sought, and in this view an object-oriented approach based on the Modelica language has been adopted. The reactor primary and secondary systems have been implemented by assembling both component models already available in a specific thermal-hydraulic library, and ad hoc developed nuclear component models suitably modified according to the specific DEMO configuration. The resulting overall plant simulator, incorporating also the balance of plant, consists in the following essential parts: core, integrated steam generator/primary pump block, cold and hot legs, primary coolant cold pool, turbine, heat sink, secondary coolant pump. Afterwards, the reactor response to typical transient initiators has been investigated: feedwater mass flow rate and temperature enhancement, turbine admission valve coefficient variation, increase of primary coolant mass flow rate, and transient of overpower have been simulated; results have been compared with the outcomes of analogous analyses performed by employing a lumped-parameter DEMO plant model. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗