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NUREG-1925, Rev. 4, Research Activities Fy 2018-2020.
	ML18071A139
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Issue date:	03/31/2018
From:	Gibson K; Office of Nuclear Regulatory Research
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	Meyd, Donald
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	NUREG-1925 R4
	Download: ML18071A139 (181)
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U.S.NRC Protecting People and the Environment Research Activities FY 2018-2020 Office of Nuclear Regulatory Research

Research Activities FY 2018-2020 NUREG-1925, Rev. 4 Manuscript Completed: September 2017 Date Published: March 2018 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, D.C. 20555-0001 www.nrc.gov

ABSTRACT The Office of Nuclear Regulatory Research (RES) supports the mission of the U.S. Nuclear Regulatory Commission (NRC) by providing technical advice, tools, and information to identify potential safety and security issues and resolve them as appropriate, assessing risk and other nuclear safety and security issues, and developing and coordinating regulatory guidance. This includes conducting confirmatory experiments and analyses, developing technical bases that support the NRCs safety decisions, and preparing the agency for the future by evaluating the safety aspects of new technologies and designs for nuclear reactors, materials, waste, and security. The RES staff uses its own expertise and collaborates with partner offices at the NRC, commercial entities, national laboratories, other Federal agencies, universities, and international organizations.

This NUREG describes research being conducted by RES across a wide variety of disciplines ranging from fuel behavior under accident conditions to seismology to health physics. This is the fourth issuance of NUREG-1925, revised to capture new research and to update ongoing research projects. RES has organized this collection of information sheets by topical areas that summarize projects currently in progress. Each sheet provides the name(s) of the RES technical staff who can be contacted for additional information.

iii

FOREWORD A Message from the Director The Office of Nuclear Regulatory Research (RES) of the U.S. Nuclear Regulatory Commission (NRC) plans, recommends, and implements nuclear regulatory research, confirmatory analyses, standards development, and resolution of potential generic safety issues for nuclear power plants and other facilities and materials regulated by the NRC. The office was established by the U.S. Congress in 1974 to accomplish the NRC mission of protecting the public health and safety, promoting the common defense and security, and protecting the environment. RES partners with other NRC offices, Federal agencies, industry research organizations, international organizations, and universities. This NUREG identifies and describes our key research projects.

Some of the highlighted FY 2018-2020 projects include Level 3 probabilistic risk assessment (Chapter 6),

human reliability analysis activities (Chapter 7), seismic and flooding research (Chapter 10), material aging and degradation analysis (Chapter 11), and international and domestic cooperative research (Chapter 15).

The NRC also continues to focus on other issues such as new and advanced reactor designs, enhancing the reality of probabilistic risk assessments, ground-water monitoring and remediation at operating and decommissioning sites, and emerging radiation protection topics.

We conduct research both in-house and with the use of contractors and interagency agreements. The offices annual budget for contracted work is typically several tens of millions of dollars (direct contract funds do not include staff or full cost allocations), which is broken down as follows:

User needs from the NRCs regulatory offices drive over three-fourths of RES activities.

The Commission drives about 10 percent of RES activities through agency-mandated programs.

A small amount is used to conduct feasibility studies of emerging technologies that may have regulatory applications in the future.

RES has made steady progress enhancing the planning, prioritizing, tracking, and reporting on research projects, which provides greater transparency on the use of its resources and improves stakeholder confidence. We work daily to enhance our effectiveness, efficiency, and agility.

Currently, RES has about 200 staff members. This staff reflects diversity in academic degrees, demographics, and technical disciplines. About 34 percent have PhDs, and another 36 percent have Masters degrees. The wide range of engineering and scientific disciplines includes expertise in thermal-hydraulics, materials science and engineering, instrumentation and controls, severe accident progression, human factors and human reliability, health physics, fire protection, seismology, and probabilistic risk assessment.

We appreciate your interest in and support for nuclear safety and security research. If you have any additional questions or comments on our research projects, please contact the technical staff or the division noted on each specific project summary sheet in this report.

Michael F. Weber Director of Nuclear Regulatory Research iv

TABLE OF CONTENTS Abstract.......................................................................................................................................................iii Foreword.....................................................................................................................................................iv Figures........................................................................................................................................................ix Abbreviations and Acronyms.................................................................................................................xiii Chapter 1: Agency Programs Support......................................................................................................1 Regulatory Guide Program.................................................................................................................................... 2 NRC Codes and Standards Program..................................................................................................................... 3 Generic Issues Program........................................................................................................................................ 4 Feasibility Studies.................................................................................................................................................. 5 Report to Congress on Abnormal Occurrences..................................................................................................... 6 Operating Experience Program............................................................................................................................. 7 Accident Sequence Precursor Program................................................................................................................ 8 Knowledge Management in the Office of Nuclear Regulatory Research............................................................... 9 NRC Non-Light Water Reactor (Non-LWR) Research.......................................................................................... 10 NRC Non-Light Water Reactor (Non-LWR) Strategy 2 - Acquire/develop sufficient computer codes and tools to perform non-LWR regulatory reviews..................................................................................... 11 Chapter 2: Thermal Hydraulic Research.................................................................................................13 TRAC/RELAP Advanced Computational Engine (TRACE) Thermal - Hydraulics Code...................................... 14 Symbolic Nuclear Analysis Package (SNAP) Computer Code Applications.........................................................15 Thermal-Hydraulic Simulations of Operating Reactors, New Reactors, and Integral Pressuized-Water Reactors..................................................................................................................... 16 Simulation of Anticipated Transients Without SCRAM with Core Instability for Maximum Extended Load Line Limit Analysis Plus for Boiling-Water Reactors.................................................................... 17 Computational Fluid Dynamics in Regulatory Applications...................................................................................18 Code Application and Maintenance Program (CAMP)......................................................................................... 19 Thermal-Hydraulic Cooperative Programs........................................................................................................... 20 Chapter 3: Fuel and Core Research.........................................................................................................21 Nuclear Analysis and the SCALE Code................................................................................................................ 22 High-Burnup Light-Water Reactor Fuel................................................................................................................ 23 Fuel Rod Thermal and Mechanical Modeling and Analyses................................................................................. 24 Spent Nuclear Fuel Burnup Credit........................................................................................................................ 25 Fuel Cooperative Research.................................................................................................................................. 26 Chapter 4: Severe Accidents and Accident Consequences Research.................................................27 Severe Accidents and the MELCOR Code........................................................................................................... 28 MACCS (MELCOR Accident Consequence Code System)................................................................................. 29 MELCOR Accident Simulation Using SNAP (MASS)........................................................................................... 30 Source Term Analysis........................................................................................................................................... 31 Containment Iodine Behavior Research............................................................................................................... 32 Severe Accident Progression in New and Advanced Nuclear Reactors............................................................... 33 Technical Basis for the Containment Protection and Release Reduction (CPRR) Rulemaking for Boiling-Water Reactors with Mark I and Mark II Containments....................................................................... 34 Fukushima Dai-ichi Accident Study with MELCOR 2.2.........................................................................................35 Sequoyah State-of-the-Art Reactor Consequence Analyses (SOARCA)............................................................. 36 State-of-the-Art Reactor Consequence Analyses: Uncertainty Analyses............................................................. 37 Research to Support Regulatory and Cost-Benefit Guidance Updates................................................................ 38 Enhancing Guidance for Evacuation Time Estimate Studies................................................................................39 v

Offsite Response Organization Capabilities and Practices for Protective Actions in the Intermediate Phase of Emergency Response to a Nuclear Power Plant Accident.............................................. 40 Research to Support Severe Accident Mitigation Alternatives Analyses for Reactor License Renewal.................................................................................................................................... 41 Modeling of Radionuclide Transport in Freshwater Systems Associated with Nuclear Power Plants...........................................................................................................................................42 Cooperative Severe Accident Research Program (CSARP)................................................................................43 Severe Accident Cooperative Research.............................................................................................................. 44 Fukushima Cooperative Research....................................................................................................................... 45 Chapter 5: Radiation and Environmental Protection Research...........................................................47 NRC Standards for Protection Against Ionizing Radiation and ALARA for Radioactive Material in LWR Effluents.................................................................................................................................... 48 Research on Patient Release, Post-Radioisotope Therapy................................................................................. 49 The Million Person Study..................................................................................................................................... 50 Radiation Exposure Information and Reporting System (REIRS)........................................................................ 51 Radiation Protection Computer Code Analysis and Maintenance Program (RAMP)........................................... 52 Radiological Assessment System for Consequence AnaLysis (RASCAL) Code................................................. 53 RADionuclide Transport, Removal, And Dose Estimation (RADTRAD) Code..................................................... 54 VARSKIN: Computer Code for Skin Contamination Dosimetry............................................................................ 55 Phantom with Moving Arms and Legs (PIMAL)....................................................................................................56 Uranium Milling and Decommissioning Computer Codes.................................................................................... 57 Ground-Water Monitoring and Remediation at Nuclear Facilities........................................................................ 58 Effectiveness of Surface Covers for Controlling Fluxes of Water and Radon at Disposal Facilities for Uranium Mill Tailings........................................................................................................................ 59 Radiation Protection Cooperative Research........................................................................................................ 60 Chapter 6: Risk Analysis Research........................................................................................................63 Full-Scope Site Level 3 Probabilistic Risk Assessment Project........................................................................... 64 Probabilistic Risk Assessment Use and Standards............................................................................................. 65 Treatment of PRA Uncertainties in Risk-Informed Decisionmaking..................................................................... 66 SPAR Model Development Program....................................................................................................................67 SAPHIRE PRA Software Development Program.................................................................................................68 Thermal-Hydraulic Level 1 Probabilistic Risk Assessment (PRA) Success Criteria Activities............................. 69 Consequential Steam Generator Tube Rupture Program....................................................................................70 Risk Analysis Cooperative Research................................................................................................................... 71 Chapter 7: Human Reliability Research.................................................................................................73 Human Reliability Analysis Data Repository........................................................................................................ 74 Human Reliability Analysis Methods.................................................................................................................... 75 Chapter 8: Human Factors Research.....................................................................................................77 Human Performance for New and Advanced Control Room Designs................................................................. 78 Human Performance Test Facility Research........................................................................................................79 Human Factors in Nondestructive Examination...................................................................................................80 Fitness for Duty....................................................................................................................................................81 Safety Culture...................................................................................................................................................... 82 Human Factors Cooperative Research................................................................................................................83 Chapter 9: Fire Safety Research.............................................................................................................85 Fire Probabilistic Risk Assessment Methodology for Nuclear Power Facilities....................................................86 Fire Human Reliability Analysis Methods Development.......................................................................................87 Fire Modeling Activities........................................................................................................................................ 88 Cable Heat Release, Ignition, and Spread in Tray Installations during Fire........................................................ 89 Fire Effects on Electrical Cables and Impact on Nuclear Power Plant System Performance.............................. 90 Evaluation of Very Early Warning Fire Detection System Performance.............................................................. 91 vi

International Testing Program for High Energy Arcing Faults (HEAF) Phase 2.................................................... 92 Evaluate the Effects of Fire on Electrical Cables Coated with Fire Retardant Coatings....................................... 93 PRISME 3 - Fire Propagation in Elementary, Multi-Room Scenarios.................................................................. 94 Training Programs for Fire Probabilistic Risk Assessment, Human Reliability Analysis, and Advanced Fire Modeling....................................................................................................................................... 95 Fire Research and Regulation Knowledge Management..................................................................................... 96 Fire Safety Cooperative Research....................................................................................................................... 97 Chapter 10: External Events Research...................................................................................................99 Important Advances in Seismic Hazard Evaluation for the Central and Eastern United States...........................100 Local Effects on Ground Motion Estimation........................................................................................................101 Seismic-Induced Ground Failure and Deformations........................................................................................... 102 Seismic Soil-Structure Interaction.......................................................................................................................103 Development of Flood Hazard Information Digest for Operating NPP sites....................................................... 104 Applications of At-Site Peak-Streamflow Frequency Analyses for Very Low Annual Exceedance Probabilities....................................................................................................................................105 Technical Basis for Extending Frequency Analysis Beyond Current Consensus Limits..................................... 106 Stratigraphic Records of Past Floods for Improved Flood Frequency Analysis on the Tennessee River...................................................................................................................................... 107 Technical Basis for Probabilistic Flood Hazard Assessment - Riverine Flooding.............................................. 108 Probabilistic Flood Hazard Assessment Framework Development.................................................................... 109 Structured Hazard Assessment Committee Process for Flooding..................................................................... 110 Numerical Modeling of Local Intense Precipitation............................................................................................. 111 Quantification of Uncertainty in Probabilistic Storm Surge Models..................................................................... 112 Flood Penetration Seal Performance at Nuclear Power Plants.......................................................................... 113 Effect of Environmental Factors on Manual Actions for Flood Protection and Mitigation at Nuclear Power Plants..................................................................................................................... 114 A Simulation-Based Dynamic Analysis Approach for Modeling Plant Response to Flooding Events............................................................................................................................. 115 Potential Impacts of Accelerated Climate Change..............................................................................................116 Erosion Processes in Embankment Dams......................................................................................................... 117 Cooperative Research on External Flooding...................................................................................................... 118 Cooperative Research on External Events-Seismic........................................................................................... 119 Chapter 11: Materials Performance Research......................................................................................121 Steam Generator Tube Integrity and Inspection Research................................................................................. 122 Reactor Pressure Vessel Integrity.......................................................................................................................123 Irradiation-Assisted Degradation of Reactor Pressure Vessel Internals............................................................. 124 Primary Water Stress Corrosion Cracking Growth Rate Testing.........................................................................125 Primary Water Stress Corrosion Cracking Initiation............................................................................................126 Weld Residual Stress Validation......................................................................................................................... 127 Leak-Before-Break Analysis............................................................................................................................... 128 Probabilistic Fracture Mechanics........................................................................................................................ 129 Nondestructive Examination............................................................................................................................... 130 Research to Support the Review of Subsequent License Renewal Applications............................................... 131 Degradation of Neutron Absorbers in Spent Fuel Pools..................................................................................... 132 Aging Management for Dry Storage and Transportation of Spent Nuclear Fuel.............................................. 133 Component Integrity Cooperative Research.......................................................................................................134 Environmental Degradation Cooperative Research............................................................................................135 Chapter 12: Structural Performance Research....................................................................................137 Concrete Irradiation Effects on Structural Performance......................................................................................138 Chemical Degradation of Concrete and Structural Effects..................................................................................139 Aging of Prestressed Concrete Containment Vessels........................................................................................140 Structural Analysis.............................................................................................................................................. 141 Steel Plate and Concrete Composite Modular Construction...............................................................................142 Risk-Informed, Performance-Based Approach to Seismic Safety.......................................................................143 vii

Chapter 13: Digital Instrumentation and Controls Research..............................................................145 Digital Instrumentation and Control (DI&C) - Modernization of the Instrumentation & Controls Regulatory Infrastructure........................................................................................146 Infrastructure Protection: Space Weather and Cyber Security............................................................................147 Digital Instrumentation and Control Cooperative Research................................................................................148 Chapter 14: Electrical Research............................................................................................................149 Electrical Cable Qualification and Condition Monitoring..................................................................................... 150 Battery-Testing Program..................................................................................................................................... 151 Electrical Cooperative Research.........................................................................................................................152 Chapter 15: International and Domestic Cooperative Research........................................................153 Halden Reactor Project.......................................................................................................................................154 International Operating Experience Database.................................................................................................... 155 viii

FIGURES Figure 1.1 Status of the Regulatory Guide Update Project as of June 2017. ...................................................................... 2 Figure 1.2 Generic Issues Process Overview...................................................................................................................... 4 Figure 1.3 Risk-Informed Applications Using NRC Operational Data. ................................................................................ 7 Figure 1.4 Historical ASP Program Results. ....................................................................................................................... 8 Figure 1.5 NUREG / KM front covers. ................................................................................................................................. 9 Figure 1.6 NRC Non-LWR Mission Readiness Roadmap................................................................................................... 10 Figure 2.1 Simplified plant model nodalization.................................................................................................................... 13 Figure 2.2 TRACE, an advanced, best-estimate reactor system code used to model the thermal-hydraulic performance of nuclear power plants. ............................................................................................................... 14 Figure 2.3 Creating input models using SNAP................................................................................................................... 15 Figure 2.4 Animating analysis results using SNAP. ........................................................................................................... 15 Figure 2.5 Steady-state conditions in a boiling-water reactor............................................................................................. 16 Figure 2.6 Key primary coolant T/H components including reactor vessel, pumps, and steam generator for a two-loop pressurized- water reactor. ......................................................................................................... 16 Figure 2.7 Operating state evolutions during ATWS for different operating domains........................................................ 17 Figure 2.8 TRACE/PARCS coupled methodology ............................................................................................................ 17 Figure 2.9 Power oscillation visualization during simulated ATWS-I.................................................................................. 17 Figure 2.10 Temperature contours of a ventilated dry cask.................................................................................................. 18 Figure 2.11 During a particular severe accident scenario, hot gases from the core circulate through the hot legs and steam generator in a counter-current flow pattern. The risk of induced failures is considered. ........ 18 Figure 2.12 The advanced accumulator (b) is a water-storage tank with a flow damper in it that switches the flow rate of cooling water injected into a reactor vessel from a large (a) to small (c) flow rate................... 18 Figure 3.1 NRC nuclear analysis codes for reactor physics .............................................................................................. 22 Figure 3.2 The CIRFT device at ORNL. A sister device is installed in a hot-cell to allow for testing of irradiated materials. .......................................................................................................................................... 23 Figure 3.3 Moment-curvature measurements in static tests showing loading and unloading response.

The corresponding stress/strain is displayed on right/top axes, respectively................................................... 23 Figure 3.4 Maxima of absolute strain as a function of number of cycles to failure with curve-fitting extended to include the no-failure data. ........................................................................................................................... 23 Figure 3.5 Fuel temperature prediction of a power spike. .................................................................................................. 24 Figure 3.6 Comparison of typical reactivity decrements associated with burnup credit allowance.................................... 25 Figure 3.7 Integral LOCA test device. Test segments up to 12 inches long can be tested in this device......................... 26 Figure 4.1 Severe accident experimental programs and MELCOR regulatory applications. ............................................ 28 Figure 4.2 Example source release timeline for multiple releases at a single site with multiple units. .............................. 29 Figure 4.3 Example of the expansion of a keyhole evacuation area as a result of a wind shift......................................... 29 Figure 4.4 MASS user interface for AP1000...................................................................................................................... 30 Figure 4.5 Accident progression for a PWR. ...................................................................................................................... 30 Figure 4.6 Use of source term and relation to other factors in dose calculations. ............................................................. 31 Figure 4.7 Hypothesized mechanism for gaseous iodine source in the Phébus-FP tests................................................. 32 Figure 4.8 MELCOR-predicted reactor and containment pressures compared to TEPCO data (Unit 3). ......................... 35 Figure 4.9 Example: Sequoyah SOARCA Results. ............................................................................................................ 36 Figure 4.10 Example: Sequoyah SOARCA UA Results. ...................................................................................................... 37 Figure 4.11 Microscopic Traffic Simulation........................................................................................................................... 39 Figure 4.12 Small, Medium, and Large Population Roadway Networks. ............................................................................. 39 Figure 4.13 Concentric rings 10-50 miles from the Fermi 2 site showing both U.S. and Canada within 50 miles............... 41 Figure 4.14 CSARP participation and technical meetings.................................................................................................. 43 Figure 5.1 Biokinetic model. ............................................................................................................................................... 48 Figure 5.2 I-131 Radiation Treatment of the thyroid........................................................................................................... 49 Figure 5.3 Radiation worker taking measurements............................................................................................................. 50 Figure 5.4 Annual Occupational Radiation Dose for PWR/BWR/LWR Reactors............................................................... 51 Figure 5.5 RAMP Logo. ...................................................................................................................................................... 52 Figure 5.6 RASCAL v4.3.2 Source Term Event Types. ..................................................................................................... 53 Figure 5.7 RASCAL v4.3.2 Welcome Screen..................................................................................................................... 53 Figure 5.8 Creating RADTRAD input model using SNAP GUI........................................................................................... 54 Figure 5.9 SNAP/RADTRAD Control Room Dose Using AptPlot. ..................................................................................... 54 ix

Figure 5.10 VARSKIN ver. 5.3 Welcome.......................................................................................................................... 55 Figure 5.11 VARSKIN Cover Model ................................................................................................................................ 55 Figure 5.12 PIMAL Phantoms........................................................................................................................................... 56 Figure 5.13 MILDOS-AREA Exposure Pathway Calculation............................................................................................ 57 Figure 5.14 Radon flux measurement at Falls City. .......................................................................................................... 59 Figure 5.15 Test Pit at the Bluewater Site.......................................................................................................................... 59 Figure 6.1 Factors affecting the scope of PRAs for operating NPPs. ................................................................................ 63 Figure 6.2 The PRA Acceptability Concept. .................................................................................................................... 65 Figure 6.3 Example of loss of offsite power SPAR model event tree display with SAPHIRE. ........................................ 67 Figure 6.4 A graphical representation of a simple fault tree............................................................................................. 68 Figure 6.5 Level 1 Success Criteria Analysis.................................................................................................................. 69 Figure 6.6 Combustion Engineering Steam Generator................................................................................................... 70 Figure 7.1 One conceptualization of an advanced control room design. ........................................................................ 73 Figure 8.1 Human-System Interface in the Control Room. ............................................................................................. 77 Figure 8.2 NRC simulation facility at the University of Central Florida............................................................................ 79 Figure 8.3 Examination of Pipe Weld Using Ultrasonic Testing...................................................................................... 80 Figure 9.1 Fire Testing of Electrical Components. .......................................................................................................... 85 Figure 9.2 Simplified fire PRA event tree representing different sets of fire damage and plant response. .................... 86 Figure 9.3 Reactor Operators in a nuclear power plant main control room ................................................................... 87 Figure 9.4 Horizontal electrical cable circuit integrity test................................................................................................ .. 90 Figure 9.5 Fire test room configuration............................................................................................................................ 91 Figure 9.6 High Energy Arc Fault Testing of Electrical Components. ............................................................................. 92 Figure 9.7 IEEE-383/1202 modified test with single cable layer tray coated with fire retardant coating. ....................... 93 Figure 9.8 NUREG/KM-0003 cover. ................................................................................................................................ 96 Figure 9.9 High Energy Arc Fault International Test Program - Thermal Camera Imaging. .......................................... 97 Figure 10.1 Shows a comparison of the variability of predicted ground motions for a magnitude 7.5 earthquake as a function of distance for currently available GMPEs at a frequency of 100 Hz. .................. 100 Figure 10.2 Example of damping values used for rock materials in site response analysis. Both weathered and unweathered shales were sampled at similar depth and within range of EPRI rock............................. 101 Figure 10.3 Insitu observations of pore pressure and acceleration at Wildlife Liquefaction Array (WLA) during the August 2012 Brawley earthquake swarm are shown here.......................................................... 102 Figure 10.4 A schematic of seismic soil-structure interaction aspects is shown here...................................................... 103 Figure 10.5 Flood Hazard Information Digest. ................................................................................................................. 104 Figure 10.6 A layered paleoflood deposit sample and typical locations of flood deposits under rock overhangs and in caves along the river......................................................................................................... 107 Figure 10.7 Aleatory hazard curve (in red) from constant at mean values and = ................................................... 108 Figure 10.8 Reporting Data from NPPs............................................................................................................................ 113 Figure 10.9 Internal Erosion (piping) failure in dam. ........................................................................................................ 117 Figure 10.10 Dam breach test facility at Reclamation........................................................................................................ 117 Figure 11.1 Steam Generator Tubing.............................................................................................................................. 122 Figure 11.2 Tube Integrity research schematic. ............................................................................................................... 122 Figure 11.3 Finite element calculated stress contours around a semi-elliptical surface flaw in the stainless steel cladding of a RPV. ................................................................................................................. 123 Figure 11.4 Cracking of a baffle bolt in a pressurized-water reactor. .............................................................................. 124 Figure 11.5 Leakage from PWSCC cracks in a steam generator hot leg nozzle. ............................................................ 125 Figure 11.6 PWSCC initiation testing rig and 1.2-inch-tall specimen developed by PNNL. ............................................ 126 Figure 11.7 Example Weld Geometry. ............................................................................................................................. 127 Figure 11.8 xLPR Version 2.0 Module Structure.............................................................................................................. 128 Figure 11.9 Probabilistic fracture mechanics calculations distributions. .......................................................................... 129 Figure 11.10 Focal law formation (left) and beam simulation (right) for a phased array ultrasonic probe......................... 130 Figure 11.11 Fouling from tubercles in service water system (NRC presentation at NRC/NEI public meeting, Dec 4, 2014, ML14338A376)............................................ 131 Figure 11.12 Blistering on the aluminum cladding of a Boral© neutron absorber (Used with permission.)...................... 132 Figure 11.13 Schematics of vertical (top) and horizontal (bottom) DCSS......................................................................... 133 Figure 12.1 Finite element model of a prestressed concrete reactor containment and contours of maximum principal strain in the liner under accident conditions................................................................................... 137 Figure 12.2 Degradation on concrete compressive strength with neutron fluence (source: Yann Le Pape et al.).......... 138 x

Figure 12.3 Evolution of ASR.......................................................................................................................................... 139 Figure 12.4 ASR concrete block sample......................................................................................................................... 139 Figure 12.5 ASR gel in crecks. ........................................................................................................................................ 139 Figure 12.6 Delamination of Crystal River Unit 3 PCCV. ................................................................................................. 140 Figure 12.7 Full-scale finite element model of a SNFT cask and simulation acceleration versus test data. .................. 141 Figure 12.8 Full- and 40-percent scale finite element model of a SNFT cask and internal components modeled.......... 141 Figure 12.9 Illustration of SC wall construction................................................................................................................ 142 Figure 12.10 Evolution of risk-informed performance-based approach to seismic safety. ............................................... 143 Figure 12.11 Construction of EDFs 1/3-scale PCCV. testing mock up ............................................................................ 144 Figure 13.1 Highly Integrated Control Room................................................................................................................... 145 Figure 13.2 Halden Reactor Project. ............................................................................................................................... 148 Figure 14.1 Electrical switchgear .................................................................................................................................... 149 Figure 14.2 BNL battery facility....................................................................................................................................... 151 Figure 14.3 EPRI headquarters....................................................................................................................................... 152 Figure 15.1 Halden Boiling Water Reactor. .................................................................................................................... 154 xi

U.S.NRC Protecting People and the Environment Research Activities FY 2018-2020 Office of Nuclear Regulatory Research

Research Activities FY 2018-2020 NUREG-1925, Rev. 4 Manuscript Completed: September 2017 Date Published: March 2018 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, D.C. 20555-0001 www.nrc.gov

ABSTRACT The Office of Nuclear Regulatory Research (RES) supports the mission of the U.S. Nuclear Regulatory Commission (NRC) by providing technical advice, tools, and information to identify potential safety and security issues and resolve them as appropriate, assessing risk and other nuclear safety and security issues, and developing and coordinating regulatory guidance. This includes conducting confirmatory experiments and analyses, developing technical bases that support the NRCs safety decisions, and preparing the agency for the future by evaluating the safety aspects of new technologies and designs for nuclear reactors, materials, waste, and security. The RES staff uses its own expertise and collaborates with partner offices at the NRC, commercial entities, national laboratories, other Federal agencies, universities, and international organizations.

This NUREG describes research being conducted by RES across a wide variety of disciplines ranging from fuel behavior under accident conditions to seismology to health physics. This is the fourth issuance of NUREG-1925, revised to capture new research and to update ongoing research projects. RES has organized this collection of information sheets by topical areas that summarize projects currently in progress. Each sheet provides the name(s) of the RES technical staff who can be contacted for additional information.

iii

FOREWORD A Message from the Director The Office of Nuclear Regulatory Research (RES) of the U.S. Nuclear Regulatory Commission (NRC) plans, recommends, and implements nuclear regulatory research, confirmatory analyses, standards development, and resolution of potential generic safety issues for nuclear power plants and other facilities and materials regulated by the NRC. The office was established by the U.S. Congress in 1974 to accomplish the NRC mission of protecting the public health and safety, promoting the common defense and security, and protecting the environment. RES partners with other NRC offices, Federal agencies, industry research organizations, international organizations, and universities. This NUREG identifies and describes our key research projects.

Some of the highlighted FY 2018-2020 projects include Level 3 probabilistic risk assessment (Chapter 6),

human reliability analysis activities (Chapter 7), seismic and flooding research (Chapter 10), material aging and degradation analysis (Chapter 11), and international and domestic cooperative research (Chapter 15).

The NRC also continues to focus on other issues such as new and advanced reactor designs, enhancing the reality of probabilistic risk assessments, ground-water monitoring and remediation at operating and decommissioning sites, and emerging radiation protection topics.

We conduct research both in-house and with the use of contractors and interagency agreements. The offices annual budget for contracted work is typically several tens of millions of dollars (direct contract funds do not include staff or full cost allocations), which is broken down as follows:

User needs from the NRCs regulatory offices drive over three-fourths of RES activities.

The Commission drives about 10 percent of RES activities through agency-mandated programs.

A small amount is used to conduct feasibility studies of emerging technologies that may have regulatory applications in the future.