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DOE OSTI2017

Ensuring operator reliability in digital control rooms - 102

The U.S. fleet of nuclear power plants is aging, and efforts are being made to modernize instrumentation and control systems including integration of digital human-machine interfaces (HMIs) into the control room. Automation and control technologies have advanced since original systems were installed in the plants. This presents unique opportunities and challenges in ensuring the replacements are a benefit to the operators and the overall safe and reliable operation of the plant. We discuss trends we have observed from assisting utilities in their modernization efforts. Automation and evolving control rooms come with their own host of problems that existing human factors and engineering methods fail to capture. Traditional approaches are analytical approaches to ensuring reliability. Here we discuss some of the shortfalls of these traditional approaches as plants become increasingly more complex. We suggest that ensuring reliability for modernized and next generation control room will require expanding the scope of what was traditionally considered human factors as well as exploring novel approaches and methodologies for dealing with complexity and rapidly advancing technology. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Recent results from a testbed for the reliability of the digital instrumentation and control systems and cybersecurity in nuclear power plants - 392

Transition from analog to digital instrumentation and control systems requires a systematic study of the modifications, and assessment of their reliability. Such studies are often carried out on a test bed. The test bed developed earlier includes a commercially available triple modular redundant (TMR) programmable logic controller (PLC) for testing not only the reliability of the digital I and C but also the cyber-security of the system during plant operation. This paper reports results of our preliminary analyses carried out using the test bed following the injection of faults mimicking faulty sensors. The fault injection tool generates faulty signals in single or multiple sensors located in the primary loop. These scenarios studied under steady-state operation conditions simulate a bad sensor, a cut wire, noise, etc. The faults injected can lead to information being displayed in the main control room (MCR) to not match the conditions in the plant. In addition to a fault injection module, a result analyzer module is also a part of this testbed. Operator actions can be based on information displayed in the MCR, or they can be informed by the result analyzer. Potential operator actions following these single- or multiple faulty signals are assessed. It is concluded that the result analyzer in some cases can be helpful in guiding operators to make the right decisions. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Identification of Significant NPP Cyber-Attack Scenarios based on Risk Information

For the last decades, analog instrumentation and control (I and C) systems have been replaced with digital systems in nuclear power plants (NPPs). While there are various advantages of the digital technology for the plant safety, new threats have been introduced such as cyber security. In fact, there have been cyber-attacks on nuclear facilities adopting digital systems. The cyber-attack on the Iran nuclear facility is a typical example which shows the possibility of a cyber-attack on safety critical infrastructures with intra-network. Cyber security has been introduced as one of the significant issues for NPP safety and recently its importance has been more highlighted, but the research in this field is not mature yet. Moreover, there is not much information about cyber security of NPPs. In this work, as a fundamental research for risk evaluation of cyber-attacks, a method to identify significant cyber-attack scenarios on NPPs was proposed based on the probabilistic risk assessment (PRA) method which is widely used for evaluating the safety of NPPs. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

A Bayesian and HRA-Aided Method for the Novel Reliability Analysis of Software

Technological advancements and nuclear power plant modernization has inspired considerable research in the areas of safety and reliability, yet there remains a lack of consensus for the reliability assessment of digital instrumentation and control (I&C) systems. Motivated by the lack of consensus for reliability analysis methods, this work employs a novel framework that incorporates Bayesian, human reliability, and common-cause failure (CCF) modeling techniques. The novel framework allows the use of state-of-the-art or classical modeling techniques when accounting for human and CCF effects on system reliability. The Bayesian and HRA-Aided Method for the Reliability Analysis of Software (BAHAMAS) is demonstrated by a case study for the quantification of software hazards found in a previous analysis of a digital reactor trip system. The results demonstrate the ability of BAHAMAS to account for human activities during the software development life cycle and their influence on software reliability. BAHAMAS is a flexible tool for extending the coverage of conventional probabilistic risk assessments to include modernized digital I&C systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1975

Beryllium reflected cavity reactor for UF6 critical experiments

Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diam by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials are available. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

Jarvis, G. A.↗

DOE OSTI2017

A business case for nuclear plant control room modernization - 352

This paper presents a generic business case for implementation of technology that supports Control Room Modernization (CRM) in existing nuclear power plants (NPPs). The business case is based on advanced digital instrumentation and control (I and C), human-system interface (HSI), and work process technologies that allow NPP operators to conduct control room operations with improved human factors and efficiency, and therefore operate with fewer operational events. Labor savings can be harvested in terms of reduced overtime and redistribution of work. Significant savings are also demonstrated as reduced time to bring the plant back on line after completion of outages. The benefits are quantified to a rough order of magnitude, providing directional guidance to NPPs that are interested in developing similar business cases. The technologies employed in CRM include large overview displays, task-based operator displays, computer-based procedures, advanced alarm systems, computerized operator support systems, and wireless networks, among others. The CRM business case is founded on the following categories of cost savings: - Integrated Computer-Based Procedures for Operators; - Reduction in Corrective Action Program Work Due to Operator Error; - Reduction in Outage Critical Path Time; - Operation of Local Control Panels; - Reduction in Control Room Support; - Reduction in Plant Events Resulting in Generation Loss; - Paperless Control Room Processes. In addition to cost savings, these technologies also offer improvements to indirect ways that have impact on key performance indicators of the plant. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Reliability Assessment of Safety-Critical Network Communication in a Digitalized Nuclear Power Plant

In analog instrumentation and control (I and C) systems for nuclear power plants (NPPs), most connections are comprised of hardwired point-to-point types. Since they are based on analog technologies, they are fragile to noise and require analog-to-digital conversion time. Therefore, many NPPs have adopted communication network for the exchange of safety-critical data between programmable logic controllers (PLCs) in digital I and C systems to effectively accommodate a huge number of field controllers. However, the primary issue of digital network communication in a safety-critical system involves potential hazards, such as partial or total loss of communication between PLCs, which may result in a failure of safety critical signal generation when it is needed. Therefore, the probability that a safety-critical system in digitalized NPP becomes unsafe due to a network failure must be evaluated to quantify the risk of digital I and C system. In Korea, the Engineered Safety Feature-Component Control System (ESF-CCS) was developed as one of the digital safety-critical system in Advanced Power Reactor- 1400 (APR-1400). The function of a digital engineered safety feature (ESF) actuation system and field actuator controllers of conventional NPPs are performed by the ESF-CCS. In this system, safety-critical network, High Reliable- Safety Data Network (HR-SDN), was employed for the safety-critical data transmission from group controllers (GCs) to loop controllers (LCs). In this study, a framework for assessing the reliability of a HR-SDN network communication between GCs and LCs in the ESF-CCS in addition to a fault-tree model for ESF components considering network failure between GCs and LCs were developed to assess the plant risk effect of a newly developed digitalized ESF-CCS. Based on the case studies and cut set analysis, the risk effect of network failure on plant risk and the important risk contributors of network failure were quantitatively addressed. A framework for identifying the potential hazardous states of network communication in the ESFCCS was proposed. In addition, a fault-tree model for network communication failure was developed to estimate the risk effects of network failure between the GCs and LCs on ESF-CCS signal failure. The developed fault-tree model was then applied to several case studies. The fault-tree model of ESF-CCS signal failure for ESF components required for various mitigation actions regarding NPP DBAs was designed by considering the identified hazardous states of network failure that would result in a failure to provide input signals to the corresponding LC. As a sensitivity study, three case studies were performed based on the periodic test methods for hardware components in HR-SDN module and their test intervals in this study. The quantitative results for three case studies demonstrated that the probability of overall network communication failure, which was calculated as the sum of the failure probability associated with each failure cause, contributed up to 1.58% on the Small-LOCA CD frequency. The results also revealed the important basic events related to network failure in the case studies which included CCF of hardware and software in HR-SDN module. This study is expected to provide insight into the development of the fault-tree model for the network failure in digital I and C system and into the quantification of the risk effect of network failure for safety-critical data transmission in NPPs. As a further study, the risk effects of network failure in the ESF-CCS on the plant risk can be estimated in a more detailed manner by analyzing the failure modes and causes in a sub-component level and the fault detection coverage of the fault tolerance techniques for network failure implemented in ESF-CCS after the detailed configuration of ESF-CCS is decided. In terms of the failure of manual ESF actuation, the relationship among the human action failures must be investigated to consider multiple human error condition since ESF-CCS provides multiple manual actuation to assure the diversity of the system. Since the conventional HEP method cannot accommodate the multiple conditions in a fault tree, condition-based human reliability assessment (CBHRA) method or the dynamic human reliability modeling approach can be applied to include a complex relationship among the automated safety signal generation and the human operator's manual actuation, avoiding the optimistic estimation of HEP of the human action failure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Application of HFE to the design and V and V of a computerized procedure system: benefits and challenges - 234

The application of Human Factors Engineering (HFE) practices to the Human Machine Interface design and subsequent Verification and Validation (V and V) is a widely applied practice when talking about the design of complete control rooms or important control room modernizations, but not so often applied to the design of computerized operator support systems. In order to overcome the challenges faced on designing a Computerized Operation Support System (COSS) that meets the process efficiency requirements and integrates seemingly with hybrid plant operation concept, the involvement of a HFE team early in the COSS design and implementation process along with operation, Instrumentation and Control (I and C) and cybersecurity experts will guarantee the project success. This paper documents the HFE process applied during the design and evaluation of a Computerized Procedure System (CPS) and how final product was shaped to ensure that the transition from paper to computerized procedures will help in meeting Delivering the Nuclear Promise (DNP) goals, improving plant efficiency. The HFE program included the following activities: 1. A review of existing operating experience 2. An analysis of functions and tasks associated to the use and maintenance of CPS 3. A design verification of the system to meet guidelines prescribed in the NUREG-0700 (Rev.2), requirements of the ISG-05 and cybersecurity criteria (NEI 08-09, IEC-62646, EPRI 1015313, these among others). 4. An integrated system validation in a full scope simulator The application of HFE practices in the design and implementation of COSS helps not only to ensure that efficiency and safety improvement requirements are met, but also to evaluate the potential return of the investment, that such tools should bring. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Preliminary Analysis of Damage to MPFDs Caused by Reactor Pulses

The need for next-generation, accident-tolerant fuels has generated significant interest in the restart of domestic fuels testing at the Transient Reactor Test (TREAT) facility. The TREAT reactor is designed to produce large, transient fluxes that replicate the effects of a wide variety of accident conditions on an irradiated fuel sample. Peak powers generated during short pulses approach 20 GW with a neutron flux of approximately 10{sup 17} n cm{sup -2} s{sup -1}. As part of the restart effort, new vehicle designs are being explored for sample irradiation, including the Static Environment Rodlet Transient Test Apparatus (SERTTA) and variants. As part of the SERTTA designs, micro-pocket fission detectors (MPFDs) are being explored for localized flux measurements before, during, and after transient operations. These detectors are also being considered as part of future, core-wide instrumentation for the TREAT reactor. However, the high fluxes associated with the pulses in the TREAT core present certain thermodynamic concerns given their intensity and the amount of energy they will release within the MPFDs within such a short amount of time. Assuming each fission that occurs in a MPFD results in a fission fragment depositing 85 MeV of energy in the substrate on average and each MPFD has 0.5 μg of natural uranium deposited on it, a 100 ms long pulse in the TREAT reactor at 1017 n cm{sup -2} s{sup -1} flux will produce approximately 4.6 x 10{sup 6} fissions in an MPFD. This pulse deposits 6.3 x 10{sup -5} J in the substrate over a 100 ms duration. That amount of energy is sufficient to melt roughly 33% of the Pt under the UO{sub 2} deposition or a circular spot with a diameter of approximately 860 μm. Of course, the above estimate makes a number of simplifying assumptions and ignores heat transfer, but does serve to illustrate the point that this rapid heating effect could be detrimental to the integrity and life of these in-core MPFDs. To study how these layers are affected by large transient neutron fluxes, five samples were prepared, pre-analyzed, and then subjected to a pulse in the central thimble of the Kansas State University TRIGA Mark II reactor to a peak power of 1.58 GW. After being allowed to decay to safer radiation levels, the sample substrates and coatings were inspected using scanning electron microscopy (SEM) to identify any changes to their surface morphology. Indeed, certain changes were found and are discussed in greater detail in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Steady-State Irradiation of Characterized Instruments for Nuclear Thermal Rockets Using In-Pile Experiment Apparatus

The development of nuclear rocket technology is critical for further exploration of extra-terrestrial bodies. Nuclear rockets can provide various advantages over current chemical rockets. However, the harsh environment provides several challenges regarding instrument performance. Temperature extremes, radiation (reactor and space), and the inability to conduct traditional maintenance activities can lead to severe degradation in instrument performance. To better understand the effects of radiation on instrumentation, an irradiation campaign has been designed to test a set of instruments provided by a collaborator in the aerospace industry. Details concerning instrument performance before and after irradiation are provided to allow for an analysis of the tested instruments. Additionally, the design of the irradiation test is described in detail.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2024

Heat Pipe Cooled Microreactors

Microreactor technologies are required to provide reliable carbon-free power generation in remote applications. The heat pipe cooled microreactor concept in particular offers notable advantages due to the passive operation of heat pipes enabling increased reliability and simplicity in a more compact form factor. There is a significant need for experimental work to aid and expedite the deployment of heat pipe microreactors due to their unique technological characteristics. Thus, there has been increased interest in heat pipe experiments by numerous institutions in order to support these efforts. The present work is a comprehensive review of heat pipe experiments, describing instrumentation, methods, phenomena of interest, and recent developments. In addition, legacy work on the operation of high-temperature heat pipes under irradiation is reviewed and discussed. Furthermore, the verification and validation efforts for the flagship heat pipe simulation code Sockeye are reviewed and requirements for future experiments are outlined. Lastly, future directions are proposed for heat pipe experimentation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1976

Research on plasma core reactors

Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

Jarvis, G. A.↗

DOE OSTI2017

Development of human machine interface (HMI) for digital control rooms in nuclear power plants - 119

The analog control systems of traditional Nuclear Power Plants (NPPs) are being replaced by digital systems due to the fact that analog Instrumentation and Control (I and C) systems are becoming obsolete and such technology is no longer supported by manufacturers. The purpose of this study is to jointly evaluate the performance of an operator with a Human Machine Interface (HMI) in a NPP digital control room ensuring the efficiency and the safety of the NPP. The validity and the accuracy of the operators' actions are the primary goals that have to be preserved through the emergence of the HMI based in digital systems. The transition from analog to digital control systems leads to a more advanced support of the NPP; Video Display Units (VDUs) are used for monitoring the state of the NPP while alarm systems help for directing attention. The operators, in order to monitor information from the different VDUs, have to switch attention between different screens (alarms systems and control systems). Therefore, a well-established HMI is a significant challenge to the overall operation of the NPP, that actually helps the operators to learn from the available information in detecting possible anomalies and making the right decisions, to ensure safety, and in parallel to diminish the human error. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

One dimensional imager of neutrons on the Z machine

We recently developed a one-dimensional imager of neutrons on the Z facility. The instrument is designed for Magnetized Liner Inertial Fusion (MagLIF) experiments, which produce D-D neutrons yields of ~3 × 10 12 . X-ray imaging indicates that the MagLIF stagnation region is a 10-mm long, ~100-μm diameter column. The small radial extents and present yields precluded useful radial resolution, so a one-dimensional imager was developed. The imaging component is a 100-mm thick tungsten slit; a rolled-edge slit limits variations in the acceptance angle along the source. CR39 was chosen as a detector due to its negligible sensitivity to the bright x-ray environment in Z. A layer of high density poly-ethylene is used to enhance the sensitivity of CR39. Finally, we present data from fielding the instrument on Z, demonstrating reliable imaging and track densities consistent with diagnosed yields. For yields ~3 × 10 12 , we obtain resolutions of ~500 μm.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2014

Qualification of Daiichi Units 1, 2, and 3 Data for Severe Accident Evaluations - Process and Illustrative Examples from Prior TMI-2 Evaluations

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This initial review focused on the set of sensors deemed most important by post-TMI-2 instrumentation evaluation programs. Instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. In addition, prior efforts focused on sensors providing data required for subsequent forensic evaluations and accident simulations. To encourage the potential for similar activities to be completed for qualifying data from Daiichi Units 1, 2, and 3, this report provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: primary system pressure; containment building temperature; and containment pressure. As described within this report, sensor evaluations and data qualification required implementation of various processes, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design to instruments easily removed from the TMI-2 plant for evaluations. As documented in this report, results from qualifying data for these parameters led to key insights related to TMI-2 accident progression. Hence, these selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are documented in this report to facilitate implementation of similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Development of Criteria for Hardware Descripted Language Programmed-Devices for Safety Systems in Nuclear Power Plants in the U.S

The U.S. Nuclear Regulatory Commission (NRC) is in the process of developing a regulatory guide (RG) describing a method the staff of the NRC considers acceptable, for use in complying with NRC regulations with respect to the development of hardware description language-programmed device (HPD). Examples include: field programmable gate arrays (FPGAs), systems in the power, instrumentation, and control portions of safety systems for nuclear power plants. The regulatory framework that NRC is establishing for nuclear power plants consists of a number of regulations and supporting guidelines applicable to the development of HPD-based safety systems. The NRC issues RGs to describe to the public, methods that the NRC staff considers acceptable for use in implementing specific parts of the agency's regulations; to explain techniques that the NRC staff uses in evaluating specific problems or postulated accidents; and to provide guidance to applicants. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Utilization of the Halden reactor for supporting development of fuel with enhanced accident tolerance

This paper gives an overview of the facilities and capabilities of the Halden Boiling Water Reactor (HBWR) for supporting the development and validation of nuclear fuels. The pressurized water loops, the auxiliary in-reactor experimental systems, the instrumented test fuel assemblies and the post-irradiation facilities are succinctly described. The HBWR is contributing to the validation of new fuel types proposed as candidates for enhanced accident tolerance fuel (EATF) concepts, including uranium nitride compounds and microcell UO{sub 2}. Another point is the validation of new cladding materials for EATF. A new experiment is to be carried out to study the corrosion and irradiation creep and growth behaviour of several candidate claddings compared against a reference of Zircaloy-4. The candidate claddings include coated and surface modified Zr-based alloys, optimized FeCrAl and a Mo-based cladding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Planned Steady State Irradiation of Characterized Instruments for Nuclear Thermal Rockets Using In-Pile Experiment Apparatus

The development of nuclear rocket technology is critical for the further exploration of extraterrestrial bodies. Nuclear rockets are capable of providing various advantages over current chemical rockets. However, the harsh environment provides a number of challenges regarding instrument performance. Temperature extremes, radiation (reactor and space), and the inability to conduct traditional maintenance activities can lead to a severe degradation in instrument performance. To better understand the effects of radiation on instrumentation, an irradiation campaign has been designed to test a set of instruments provided by a collaborator in the aerospace industry. Details concerning pre-irradiation performance are provided for a future comparison to the irradiated performance. Additionally, the design of the irradiation test is described in detail.

Floyd, Dan↗