Text

NRC070 - Research Activities Fy 2018-2020, NUREG-1925, Rev. 4 (Mar. 2018)
	ML19206A039
Person / Time
Site:	Seabrook
Issue date:	07/24/2019
From:	; NRC/OGC
To:	; Atomic Safety and Licensing Board Panel
	SECY RAS
References
	50-443-LA-2, ASLBP 17-953-02-LA-BD01, RAS 55108
	Download: ML19206A039 (182)
	v • d • e

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of NEXTERA ENERGY SEABROOK, LLC

(Seabrook Station, Unit 1)

Docket No. 50-443-LA-2 ASLBP No. 17-953-02-LA-BD01

Hearing Exhibit

Exhibit Number:

Exhibit Title:

Re sea rch A cti vit ies Offi ce of Nu clea r R egu lato ry Res earc h FY 2 01 8-20 20 U.S.N R C e P r o t e c t i n g P o p l e a n d t h e E n v i r o n m e n t

Research Activities FY 2018-2020

v Abstract.......................................................................................................................................................iii Foreword.....................................................................................................................................................iv

Figures........................................................................................................................................................ix

Abbreviations and Acronyms.................................................................................................................xiii Chapter 1: Agency Programs Support......................................................................................................1 Chapter 2: Thermal Hydraulic Research.................................................................................................13 Chapter 3: Fuel and Core Research.........................................................................................................21 Chapter 4: Severe Accidents and Accident Consequences Research.................................................27 Regulatory Guide Program....................................................................................................................................

NRC Codes and Standards Program.....................................................................................................................

Generic Issues Program........................................................................................................................................

Feasibility Studies..................................................................................................................................................

Report to Congress on Abnormal Occurrences.....................................................................................................

Operating Experience Program.............................................................................................................................

Accident Sequence Precursor Program................................................................................................................

Knowledge Management in the Office of Nuclear Regulatory Research...............................................................

NRC Non-Light Water Reactor (Non-LWR) Research..........................................................................................

NRC Non-Light Water Reactor(on-LWR) Strategy 2 - Acquire/ evelop sufficient computer codes and tools to perform non-LWR regulatory reviews.....................................................................................

TRAC/RELAP Advanced Computational Engine (TRACE) Thermal - Hydraulics Code......................................

Symbolic Nuclear Analysis Package (SNAP) Computer Code Applications.........................................................

Thermal-Hydraulic Simulations of Operating Reactors, New Reactors, and

Integral Pressuized-Water Reactors.....................................................................................................................

Simulation of Anticipated Transients Without SCRAM with Core Instability for Maximum

Extended Load Line Limit Analysis Plus for Boiling-Water Reactors....................................................................

Computational Fluid Dynamics in Regulatory Applications...................................................................................

Code Application and Maintenance Program (CAMP).........................................................................................

Thermal-Hydraulic Cooperative Programs...........................................................................................................

Nuclear Analysis and the SCALE Code................................................................................................................

High-Burnup Light-Water Reactor Fuel................................................................................................................

Fuel Rod Thermal and Mechanical Modeling and Analyses.................................................................................

Spent Nuclear Fuel Burnup Credit........................................................................................................................

Fuel Cooperative Research..................................................................................................................................

Severe Accidents and the MELCOR Code...........................................................................................................

MACCS (MELCOR Accident Consequence Code System)...............................................................................

MELCOR Accident Simulation Using SNAP (MASS)...........................................................................................Source Term Analysis...........................................................................................................................................

Containment Iodine Behavior Research...............................................................................................................

Severe Accident Progression in New and Advanced Nuclear Reactors...............................................................

Technical Basis for the Containment Protection and Release Reduction (CPRR) Rulemaking for Boiling-Water Reactors with Mark I and Mark II Containments.......................................................................

Fukushima Dai-ichi Accident Study with MELCOR 2.2.........................................................................................Sequoyah State-of-the-Art Reactor Consequence Analyses (SOARCA)............................................................. State-of-the-Art Reactor Consequence Analyses: Uncertainty Analyses.............................................................

Research to Support Regulatory and Cost-Benefit Guidance Updates................................................................

Enhancing Guidance for Evacuation Time Estimate Studies................................................................................

T ABLE OF CONTENTS 2 3 4 5 6 7 8 9 10 11 14 15 16 17 18 19 20 22 23 24 25 26 28 29 30 31 32 33 34 35 36 37 38 39 vi Offsite Response Organization Capabilities and Practices for Protective Actions in the Intermediate Phase of Emergency Response to a Nuclear Power Plant Accident..............................................Research to Support Severe Accident Mitigation Alternatives Analyses for Reactor License Renewal....................................................................................................................................

Modeling of Radionuclide Transport in Freshwater Systems Associated with Nuclear Power Plants...........................................................................................................................................

Cooperative Severe Accident Research Program (CSARP)................................................................................

Severe Accident Cooperative Research..............................................................................................................

Fukushima Cooperative Research.......................................................................................................................NRC Standards for Protection Against Ionizing Radiation and ALARA for Radioactive Material in LWR Effluen...................................................................................................................................

Research on Patient Release, Post-Radioisotope Therapy.................................................................................The Million Person Study.....................................................................................................................................

Radiation Exposure Information and Reporting System (REIRS)........................................................................

Radiation Protection Computer Code Analysis and Maintenance Program (RAMP)...........................................Radiological Assessment System for Consequence AnaLysis (RASCAL) Code.................................................RADionuclide Transport, Removal, And Dose Estimation (RADTRAD) Code.....................................................VARSKIN: Computer Code for Skin Contamination Dosimetry............................................................................

Phantom with Moving Arms and Legs (PIMAL)....................................................................................................

Uranium Milling and Decommissioning Computer Codes....................................................................................

Ground-Water Monitoring and Remediation at Nuclear Facilities........................................................................

Effectiveness of Surface Covers for Controlling Fluxes of Water and Radon at DisposalFacilities for Uranium Mill Tailings........................................................................................................................

Radiation Protection Cooperative Research........................................................................................................

Full-Scope Site Level 3 Probabilistic Risk Assessment Project...........................................................................

Probabilistic Risk Assessment Use and S

.............................................................................................Treatment of PRA Uncertainties in Risk-Informed Decisionmaking.....................................................................

SPAR Model Development Program....................................................................................................................

SAPHIRE PRA Software Development Program.................................................................................................

Thermal-Hydraulic Level 1 Probabilistic Risk Assessment (PRA) Success Criteria Activities.............................

Consequential Steam Generator Tube Rupture Program....................................................................................

Risk Analysis Cooperative Research...................................................................................................................Human Reliability Analysis Data Repository........................................................................................................

Human Reliability Analysis Methods....................................................................................................................

Human Performance for New and Advanced Control Room Designs.................................................................Human Performance Test Facility Research........................................................................................................

Human Factors in Nondestructive Examination...................................................................................................

Fitness for Duty....................................................................................................................................................

Safety Culture......................................................................................................................................................

Human Factors Cooperative Research................................................................................................................

Fire Probabilistic Risk Assessment Methodology for Nuclear Power Facilities....................................................

Fire Human Reliability Analysis Methods Development.......................................................................................

Fire Modeling Activities........................................................................................................................................

Cable Heat Release, Ignition, and Spread in Tray Ins lations during Fire......................................................

Fire Effects on Electrical Cables and Impact on Nuclear Power Plant System Performance..............................

Evaluation of Very Early Warning Fire Detection System Performance..............................................................

Chapter 5: Radiation and Environmental Protection Research...........................................................47 Chapter 6: Risk Analysis Research........................................................................................................63 Chapter 7: Human Reliability Research.................................................................................................73 Chapter 8: Human Factors Research.....................................................................................................77 Chapter 9: Fire Safety Research.............................................................................................................85 40 41 42 43 44 45 48 49 50 51 52 53 54 55 56 57 58 59 60 64 65 66 67 68 69 70 71 74 75 78 79 80 81 82 83 86 87 88 89 90 91 viiInternational Testing Program for High Energy Arcing Faults (HEAF) Phase 2....................................................

Eva uate the Effects of Fire on Electrical Cables Coated with Fire Retardant Coatings.....

............................

PRISME 3 - Fire Propagation in Elementary, Multi-Room Scenarios..................................................................Training Programs for Fire Probabilistic Risk Assessment, Human Reliability Analysis, and

Advanced Fire Modeling.......................................................................................................................................

Fire Research and Regulation Knowledge Management.....................................................................................

Fire Safety Cooperative Research.......................................................................................................................

Important Advances in Seismic Hazard for the Central and Eastern United

..........................

Local Effects on Ground Motion Estimation........................................................................................................

Seismic-Induced Ground Failure and Deformations...........................................................................................

Seismic Soil-Structure Interaction.......................................................................................................................

Development of Flood Hazard Information Digest for Operating NPP sites.......................................................

Applications of At-Site Peak-Streamflow Frequency Analyses for Very Low Annual Exceedance Probabilities....................................................................................................................................

Technical Basis for Extending Frequency Analysis Beyond Current Consensus Limits.....................................

Stratigraphic Records of Past Floods for Improved FloodFrequency Analysison the Tennessee River......................................................................................................................................

Technical Basis for Probabilistic Flood Hazard Assessment - Riverine Flooding..............................................

Probabilistic Flood Hazard Assessment Framework Development....................................................................

Structured Hazard Assessment Committee Process for Flooding.....................................................................

Numerical Modeling of Local Intense Precipitation.............................................................................................

Quantification of Uncertainty in Probabilistic Storm Surge Models.....................................................................

Flood Penetration Seal Performance at Nuclear Power Plants..........................................................................

Effect of Environmental Factors on Manual Actions for Flood Protection and Mitigation at Nuclear Power Plants.....................................................................................................................A Simulation-Based Dynamic Analysis Approach for Modeling Plant Response to Flooding Events.............................................................................................................................

Potential Impacts of Accelerated Climate Change..............................................................................................

Erosion Processes in Embankment Dams.........................................................................................................

Cooperative Research on External Flooding......................................................................................................

Cooperative Research on External Events-Seismic...........................................................................................Steam Generator Tube Integrity and Inspection Research.................................................................................Reactor Pressure Vessel Integrity.......................................................................................................................

Irradiation-Assisted Degradation of Reactor Pressure Vessel Internals.............................................................Primary Water Stress Corrosion Cracking Growth Rate Testing.........................................................................

Primary Water Stress Corrosion Cracking Initiation............................................................................................Weld Residual Stress Validation.........................................................................................................................

Leak-Before-Break Analysis...............................................................................................................................

Probabilistic Fracture Mechanics........................................................................................................................

Nondestructive Examination...............................................................................................................................

Research to Support the Review of Subsequent License Renewal Applications...............................................

Degradation of Neutron Absorbers in Spent Fuel Pools.....................................................................................

Aging Management for Dry Storage and Transpor tion of Spent Nuclear Fuel..............................................

Component Integrity Cooperative Research.......................................................................................................

Environmental Degradation Cooperative Research............................................................................................

Concrete Irradiation Effects on Structural Performance......................................................................................

Chemical Degradation of Concrete and Structural Effects..................................................................................

Aging of Prestressed Concrete Containment Vessels........................................................................................

Structural Analysis..............................................................................................................................................

Steel Plate and Concrete Composite Modular Construction...............................................................................

Risk-Informed, Performance-Based Approach to Seismic Safety.......................................................................

Chapter 10: External Events Research...................................................................................................99 Chapter 11: Materials Performance Research......................................................................................

121 Chapter 12: Structural Performance Research....................................................................................137 92 93 94 95 96 97 100 101 102 103 104 105 106 107 108 109 110 111 112 113 114 115 116 117 118 119 122 123 124 125 126 127 128 129 130 131 132 133 134 135 138 139 140 141 142 143 viii Chapter 13: Digital Instrumentation and Controls Research..............................................................145 Chapter 14: Electrical Research............................................................................................................149 Chapter 15: International and Domestic Cooperative Research........................................................153 Digital Instrumentation and Control (DI&C) - Modernization of the Instrumentation & Controls Regulatory Infrastructure........................................................................................Infrastructure Protection: Space Weather and Cyber Security............................................................................

Digital Instrumentation and Control Cooperative Research................................................................................

Electrical Cable Qualification and Condition Monitoring.....................................................................................

BatteryT esting Program....................................

............................................................................................

Electrical Cooperative Research.........................................................................................................................

Halden Reactor Project.......................................................................................................................................

International Operating Experience Database....................................................................................................

146 147 148 150 151 152 154 155

Figure 1.1 Status of the Regulatory Guide Update Project as of June 2017.

................................................................

Figure 1.2 Generic Issues Process Overview...............................................................................................................

Figure 1.3 Risk-Informed Applications Using NRC Operational Data. ...........................................................................Figure 1.4 Historical ASP Program Results. ..................................................................................................................

Figure 1.5 NUREG / KM front covers.

........................................................................................................................

.... Figure 1.6 NRC Non-LWR Mission Readiness Roadmap...................................................................................

.......Figure 2.1 Simplified plant model nodalization......

........................................................................................................Figure 2.2 TRACE, an advanced, best-estimate reactor system code used to model the thermal-hydraulic performance of nuclear power plants. ...........................................................................................................Figure 2.3 Creating input models using SNAP..............................................................................................................Figure 2.4 Animating analysis results using SNAP. ......................................................................................................Figure 2.5 Steady-state conditions in a boiling-water reactor.....

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Figure 2.6 Key primary coolant T/H components including reactor vessel, pumps, and steam generator for a two-loop pressurized-water reactor. ......................................................................................................

Figure 2.8 TRACE/PARCS coupled methodology........................................................................................................ Figure 2.7 Operating state evolutions during ATWS for different operating domains ...................................................Figure 2.9 Power oscillation visualization during simulated ATWS-I....

........................................................................Figure 2.10 Temperature contours of a ventilated dry cask...........................................................

............................... 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.

....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..............

Figure 3.1 NRC nuclear analysis codes for reactor physics........................................................................................ Figure 3.2 The CIRFT device at ORNL. A sister device is installed in a hot-cell to allow for testing of irradiated materials.

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.......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............................................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. .......................................................................................................................Figure 3.5Fuel temperature prediction of a power spike.

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.. Figure 3.6 Comparison of typical reactivity decrements associated with burnup credit allowance.......................

.....Figure 3.7 Integral LOCA test device.

Test segments up to 12 inches long can be tested in this device...................

Figure 4.1 Severe accident experimental programs and MELCOR regulatory applications.

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...Figure 4.2 Example source release timeline for multiple releases at a single site with multiple units.

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...Figure 4.3 Example of the expansion of a keyhole evacuation area as a result of a wind shift.............................

.....Figure 4.4 MASS user interface for AP1000...........................................................................................................

.....Figure 4.5 Accident progression for a PWR.

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...Figure 4.6 Use of source term and relation to other factors in dose calculations.

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.....Figure 4.7 Hypothesized mechanism for gaseous iodine source in the Phébus-FP tests. ..........................................

. Figure 4.8 MELCOR-predicted reactor and containment pressures compared to TEPCO data (Unit 3).

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...Figure 4.9 Example: Sequoyah SOARCA Results. ...................................................................................................

...Figure 4.10 Example: Sequoyah SOARCA UA Results. ...............................................................................................

Figure 4.12 Small, Medium, and Large Population Roadway Networks.

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....Figure 4.11 Microscopic Traffic Simulation...............................................................................................................

...Figure 4.13 Concentric rings 10-50 miles from the Fermi 2 site showing both U.S.

and Canada within 50 miles....

..Figure 5.1 Biokinetic model. ...............................................................................................................................

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...Figure 5.2 I-131 Radiation Treatment of the thyroid.....

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........Figure 5.3 Radiation worker taking measurements...............................................................................................

.......Figure 5.4 Annual Occupational Radiation Dose for PWR/BWR/LWR Reactors..................................................

.....Figure 5.5 RAMP Logo. ...............................................................................................................................

.

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Figure 5.6 RASCAL v4.3.2 Source Term Event Types.

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.........Figure 5.7 RASCAL v4.3.2 Welcome Screen.....

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.................Figure 5.8 Creating RADTRAD input model using SNAP GUI..............................................................................

.......Figure 5.9 SNAP/RADTRAD Control Room Dose Using AptPlot. ................................................................................2 Figure Figure 5.11 VARSKIN Cover Model........................................................................................................................Figure 5.10 VARSKIN ver. 5.3 Welcome...................................................................................................................Figure 5.12 PIMAL Phantoms.....................................................................................................................................

Figure 5.13 MILDOS-AREA Exposure Pathway Calculation.....................................................................................Figure 5.14 Radon flux measurement at Falls City. ..................................................................................................... Figure 6.1 Factors affecting the scope of PRAs for operating NPPs. ............................................................................ Figure 6.2 The PRA Acceptability Concept. ................................................................................................................. Figure 6.3 Example of loss of offsite power SPAR model event tree display with SAPHIRE.

.....................................Figure 6.4 A graphical representation of a simple fault tree.....

...................................................................................... Figure 7.1 One conceptualization of an advanced control room design. ......................................................................Figure 8.1 Human-System Interface in the Control Room. .........................................................................................Figure 8.2 NRC simulation facility at the University of Central Florida................................................

.................Figure 8.3 Examination of Pipe Weld Using Ultrasonic Testing.....

...............................................................................Figure 9.1 Fire Testing of Electrical Components.

........................................................................................................Figure 9.2 Simplified fire PRA event tree representing different sets of fire damage and plant response.

..................Figure 9.3 Reactor Operators in a nuclear power plant main control room ..................................................................Figure 9.4 Horizontal electrical cable circuit integrity test.........................................................................................

.. Figure 9.5 Fire test room configuration.....

....................................................................................................................Figure 9.6 High Energy Arc Fault Testing of Electrical Components.

.......................................................................... Figure 9.7 IEEE-383/1202 modified test with single cable layer tray coated with fire retardant coating.

.....................Figure 9.8 NUREG/KM-0003 cover. ............................................................................................................................. Figure 9.9 High Energy Arc Fault International Test Program Thermal Camera Imaging.

........................................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. ................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......

......................Figure 10.3 tu observations of pore pressure and acceleration at Wildlife Liquefaction Array (WLA)during the August 2012 Brawley earthquake swarm are shown here......................................................

Figure 10.4 A schematic of seismic soil-structure interaction aspects is shown here................................................Figure 10.5 Flood Hazard Information Digest. ...........................................................................................................Figure 10.6 A layered paleoflood deposit sample and typical locations of flood deposits under rock overhangs and in caves along the river.......

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Figure 10.

7 Aleatory hazard curve (in red) from constant at mean values and ..............

.............................Figure 10.8 Reporting Data from NPPs.....................................................................................................................Figure 10.9 Internal Erosion (piping) failure in dam. ..................................................................................................Figure 10.10 Dam breach test facility at Reclamation.....

...........................................................................................Figure 11.1 Steam Generator Tubing.......................................................................................................................Figure 11.2 Tube Integrity research schematic.

..........................................................................................................Figure 11.3 Finite element calculated stress contours around a semi-elliptical surface flaw in the stainless steel cladding of a RPV. ............................................................................................................... Figure 11.4 Cracking of a baffle bolt in a pressurized-water reactor. .........................................................................Figure 11.5 Leakage from PWSCC cracks in a steam generator hot leg nozzle. .......................................................Figure 11.6 PWSCC initiation testing rig and 1.2-inch-tall specimen developed by PNNL. .......................................Figure 11.7 Example Weld Geometry.

........................................................................................................................Figure 11.8 xLPR Version 2.0 Module Structure.....

....................................................................................................Figure 11.9 Probabilistic fracture mechanics calculations distributions.

.....................................................................Figure 11.10 Focal law formation (left) and beam simulation (right) for a phased array ultrasonic probe......

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...Figure 11.11 Fouling from tubercles in service water system (NRC presentation at NRC/NEI public meeting, Dec 4, 2014, ML14338A376).....

..................................... Figure 11.12 Blistering on the aluminum cladding of a Boral© neutron absorber Figure 11.13 Schematics of vertical (top) and horizontal (bottom) DCSS............................................................Figure 12.1 Finite element model of a prestressed concrete reactor containment and contours of maximum principal strain in the liner under accident conditions.....

............................................................................ Figure 12.2 Degradation on concrete compressive strength with neutron fluence (source: Yann Le Pape et al.)....

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Figure 5.1.

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.....Figure 6.

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......... Figure 6.

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Figure 12.5 ASR gel in crecks. ..................................................................................................................................Figure 12.6 Delamination of Crystal River Unit 3 PCCV. ........................................................................................... Figure 12.7 Full-scale finite element model of a SNFT cask and simulation acceleration versus test data.

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Figure 12.8 Full- and 40-percent scale finite element model of a SNFT cask and internal components modeled... Figure 12.9 Illustration of SC constructio..............................................

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Figure 12.10 Evolution of risk-informed performance-based approach to seismic safety. ......................................

... -scale PCCV. testing mock up ...................................................................

...Figure 13.1 Highly Integrated Control Room. ..............................................................................................................

Figure 13.2 Halden Reactor Project. ..............................................................................................

............................Figure 14.1 Electrical switchgear.....................................................................................................................Figure 14.2 BNLbattery facility....

.....................................................................................................................Figure 14.3 EPRI headquarters....

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Figure 15.1 Halden Boiling Water Reactor.

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Figure 12.3 Evolution of ASR...................................................................................................................................Figure 12.4 ASR concrete block sampl..................................................................................................................