Public/Internal Events
Jump to navigation
Jump to search
This page includes methods and tools applicable to the Internal Events chapter (Part 2) of the ASME/ANS Standard.
Note that the definition of a "consensus method" used in this table is as follows: Consensus Method/Model: A method/model that the USNRC has used or accepted for the specific risk-informed application for which it is proposed. (As per PWR Owner's Group comments on draft Regulatory Guide 1.200, documented in a July 2020 letter available at ML20213C660.)
Select Type
Select Technical Element
| ID | Standard Part | Technical Element | Type | Name of method | Brief description of method | Basis for suitability of method | Other basis for validity of method | Reference for method (report, textbook, etc) | Is this method part of a software package, or otherwise associated with a tool? | Associated software | Is this a data set, or associated with a data set? | Name of associated data set |
|---|---|---|---|---|---|---|---|---|---|---|---|---|
| IE-DA-M-1 | Internal Events | DA; | Method | Bayesian Updating Conjugate Distributions | Bayesian updating beta or gamma prior distributions with operating experience assumed to follow a Poison or Binomial likelihood. | Consensus | Standard statistical procedure. | NUREG/CR-6823 | Yes | CAFTA | No | |
| IE-DA-M-2 | Internal Events | DA; | Method | Bayesian Updating Nonconjugate Distributions | Bayesian updating a lognormal distribution with operating experience assumed to follow a lognormal distribution. | Consensus | Standard statistical procedure | NUREG/CR-6823 | Yes | winBUGS | No | |
| IE-DA-M-3 | Internal Events | DA; | Method | Moment Matching | Moment matching is used to transform lognormal distributions into a gamma distribution to perform a Bayesian update with data that is assumed to follow a Poisson distribution. Once the data update is complete, moment matching is re-used to obtain a lognormal distribution. | Consensus | Standard Statistical Practice | NUREG/CR-6823 | No | No | ||
| IE-DA-M-4 | Internal Events | DA; | Method | NUREG/CR-6928 industry Average Parameter Estimates | Component and Initiating Event Data Set | Consensus | Industry recognized source of data. | NUREG/CR-6928 | No | Yes | ||
| IE-DA-M-5 | Internal Events | DA; | Method | INL/EXT-21-62940 CCF Parameter Estimatons | Industry recognized data source for CCF parameters | Consensus | INL/EXT-21-62940 | No | Yes | |||
| IE-DA-M-6 | Internal Events | DA; | Method | NUREG/CR-5497 CCF Parameter Estimates | Industry recognized data set for CCF parameters. | Consensus | NUREG/CR-5497 | No | Yes | |||
| IE-DA-M-7 | Internal Events | DA; | Method | Multiple Greek Letter Common Cause Model | Generating common cause failure probabilities using the multiple greek letter model. | Consensus | NUREG/CR-5485 | Yes | CAFTA | No | ||
| IE-DA-M-8 | Internal Events | DA; | Method | Alpha Factor Common Cause Model | Generating CCF probabilities using the Alpha Factor common cause model. | Consensus | NUREG/CR-5485 | Yes | CAFTA | No | ||
| IE-DA-M-9 | Internal Events | DA; | Method | Distribution fitting using the Simplified Constrained Non-Informative Distribution (SCNID) | Generating component failure rate probabilities using a multiplier on another distribution. | Consensus | NUREG/CR-6823 | No | No | |||
| IE-DA-M-11 | Internal Events | DA; | Method | Loss of Offsite Power Data | Data set for LOOP initiators, recoveries, consequential LOOPs | Consensus | INL/RPT-22-68809 | No | Yes | Analysis of Loss-of-Offsite Power Events - INL/RPT-22-68809 | ||
| IE-DA-M-14 | Internal Events | DA; | Method | Unavailability - Modified F-Distribution | For unavailability distribution calculations, Dominion uses a modified F-Distribution {UA-mean=((t-outage duration/T-exposure)·((n+0.5)/n)·(2n/(2n-2))), n-outages} with a variance of an F-distribution adopted from NUREG/CR-6823 section 6.7.1.2 {Var(UA)=(t-outage duration/T-exposure)^2·((n+0.5)/n)·((?32n?^3-?8n?^2)/((1+2n)(2n-2)^2 (2n-4)))}. While the F-distribution has its own uncertainty parameters, because CAFTA does not have F-distributions as a distribution selection option, it is Dominion assigns these distributions as a gamma distributions since the F-distribution is a ratio of 2 gamma distributions. Per Appendix A.7.11 of NUREG/CR-6823 the mean is defined only if n>1 and the variance if n>2. If the criteria for the mean is not met, then Dominion performs a straight UA calculation using t-outage duration divided by T-exposure and the SCNID method for estimating gamma distribution uncertainty parameters is used. |
Consensus | NUREG/CR-6823 | NUREG/CR-6823 | No | No | ||
| IE-DA-M-16 | Internal Events | DA; | Method | Calculation of mean geometric average given limited population of data points | Given a limited population of data points, assume geometric average as a median and the minimum and maximum values as the 5th and 95th percentile values; mean value is calculated using lognormal distribution | Peer reviewed | Specific application utilizing unique format of failure rate data provided in Advanced Light Water Reactor Requirements Document (ALWR), Volume II, Chapter 1, Appendix A - PRA Key Assumptions and Groundrules, Electric Power Research Institute, Revision 6, December 1993 | n/a | No | No | ||
| IE-DA-M-17 | Internal Events | DA; | Method | Mean Time Between Failure (MTBF) Data Analysis of UPS | Basis for failure rate: Switch - Static Transfer - fails to transfer to back-up power supply | relevant industry data | Industry Data | AMETEK Solidstate Controls - Mean Time Between Failure (MTBF) Data Analysis of UPS | No | Yes | ||
| IE-DA-M-18 | Internal Events | DA; | Method | CEN-403 ESFAS Subgroup Relay Test Interval Extension, Rev 1A | Combustion Engineering, Inc. prepared Topical Report for the C-E Owners Group to provide supporting data for extending the test interval for the ESFAS subgroup relays. This document is used as supporting documentation for the ESFAS subgroup relay failure rate in the parameter data. | NRC approved | Approved by the NRC in a letter from Mr. Bruce A. Boger (USNRC) to Mr. D. F. Pilmer (CEOG) dated February 27, 1996, "Review of CE Owner's Group Topical Report CEN-403, Revision 1. ESFAS Subgroup Relay Test Interval Extension." | n/a | No | Yes | ||
| IE-DA-M-19 | Internal Events | DA; | Method | CEN-327-A, RPS/ESFAS Extended Test Interval Evaluation | Combustion Engineering, Inc. performed Technical Report for the C-E Owners Group to provide supporting data for extending the test interval for the RPS and ESFAS. This document is used as supporting documentation for the RPS/ESFAS subcomponent failure rates in the parameter data | NRC approved | NRC approved via letter from Ashok Thadani (NRC) to Ed Sterling (CEOG) "NRC Evaluation of CEOG Topical Report CEN-327, RPS/ESFAS Extended Test Interval Evaluation", dated November 6, 1989 | n/a | No | Yes | ||
| IE-DA-M-24 | Internal Events | DA; | Method | Non-recovery Probability for Initiating Event LOOP | Given a Station AC Blackout (SBO), if AC power is not recovered within a specific period of time, core damage would be expected to occur. A LOOP Frequency calculation provided the bases for estimating the coping times for avoiding core damage given either successful operation of the turbine driven AFW pump, or failure. The purpose of this calculation is to provide CAFTA recovery factors for recovery of offsite power before core damage given SBO (from a specific type of loss of offsite power), and success or failure of the turbine driven AFW pump | Consensus | Various references and assumptions are used to determine non-recovery probability for I-LOOP in PRA Models, including: 1) Decay Heat Power in Light Water Reactors, ANSI/ANS-5.1-2005 Standard, issued by the American Nuclear Society 2) Treatment of Loss of Offsite Power (LOOP) in Probabilistic Risk Assessments: Technical Basis and Guidelines” Interim EPRI Technical Report, issued September 2009. 3) S.A Eide et al, “Reevaluation of Station Blackout Risk at Nuclear Power Plants”, NUREG/CR-6890, December 2005. 4) “Analysis of Loss of Offsite Power Events – 2008 Update” obtained from the Idaho National Laboratory website maintained for NRC. 5) M. Lloyd, “Treatment of Time Interdependencies in Fault Tree Generated Cutset Results”, EPRI Topical Report 1009187, Issued October 2003 | No | No | |||
| IE-DA-M-25 | Internal Events | DA; | Method | PWROG-18042: FLEX Equipment Failure Data and Analysis | This report provides generic unreliability estimates (failure rates) for standard FLEX equipment in use in the probabilistic risk assessments (PRAs) for the United States (U.S.) nuclear industry. This report also documents the philosophy guiding the effort to update the inputs used for FLEX Equipment in PRA models. | Consensus | PWROG-18042, FLEX Equipment Data Collection and Analysis, Revision 1, August 2021 | No | No | |||
| IE-DA-M-26 | Internal Events | DA; | Method | WCAP-16175-P-A: RCP Seal Failure Model and Data | The purpose of the method to establish a model for estimating the probability of failure of an RCP seal given loss of cooling to the seal. | Consensus | WCAP-16175-P-A, Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants, Revision 0, March 2007 | No | No | |||
| IE-DA-M-27 | Internal Events | DA; | Method | WCAP-16882-NP: Sump Strainer Failure Rates | The purpose of this report is to provide a generic industry model for evaluating the loss of long term core cooling due to debris induced or chemical induced blockages in plant specific Probabilistic Risk Assessments (PRA).The model presented in this report is based on the knowledge and insights gained during the resolution of the Nuclear Regulatory Commission (NRC) Generic Safety Issue (GSI) related to containment sumps, known as GSI-191 (Reference 1) and the subsequent plant modifications to satisfy the NRC’s Generic Letter (GL) 2004-02 (Reference 2). The model takes the form of a stand-alone methodology for determining the probability of a loss of long term core cooling for various initiating events and accident sequences considered in a typical PRA. The result of this methodology is a single basic event probability that can be used in the existing plant PRA model. |
Consensus | WCAP-16882-NP, PRA Modeling of Debris-Induced Failure of Long Term Core Cooling via Recirculation Sumps, Revision 1, November 2009. | No | No | |||
| IE-DA-M-28 | Internal Events | DA; | Method | EPRI TR-1018243: Risk Assessment Method for Extending integrated leak rate test (ILRT) surveillance intervals to 15 years. | This report presents a risk impact assessment for extending integrated leak rate test (ILRT) surveillance intervals to 15 years. The assessment demonstrates that on an industry-wide basis there is small risk associated with the extension, provided that the performance bases and defense-in-depth are maintained. There is an obvious benefit in not performing costly, criticalpath, time-consuming tests that provide a limited benefit from a risk perspective. The first step is to obtain current containment leak rate testing performance information. The data were obtained through an NEI industry surveys, industry failure reports, and previous survey information. The data indicate that there were no failures of a magnitude that approaches that of a large release. This information is used to develop the probability of a pre-existing leak in the containment using the Jeffreys Non-Informative Prior statistical method. |
Peer reviewed | EPRI TR-1018243, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, October 2008, Revision 2-A of 1009325. | No | No | |||
| IE-DA-M-29 | Internal Events | DA; | Method | WCAP-15831-P-A: ATWS Unfavorable Exposure Time (UET) Pressure Relief Intervals | For use as input to the risk-informed PRA model, Unfavorable Exposure Time (UET), must be determined. To calculate UET for a given plant condition and core model, the ANC computer code (ANC: Westinghouse Advanced Nodal Computer Code,” Davidson, S. L., ed., et al.,WCAP-10965-P-A (Westinghouse Proprietary), WCAP-10966-A (Non-Proprietary), September 1986) is used first to determine the critical power as a function of inlet temperature at various cycle burnups. The “critical power” is the power that results in reactor criticality for a given set of conditions (inlet temperature, pressure, etc.). The ANC results are then compared to the Critical Power Trajectory data presented in Section 4.1 corresponding to the ATWS transient conditions that result in a peak RCS pressure of 3200 psig. The time that the ANC calculated critical power is greater than the ATWS Critical Power Trajectory power represents the time of unfavorable reactivity conditions. This time is termed the Unfavorable Exposure Time. | Peer reviewed | WCAP-15831-P-A, WOG Risk-Informed ATWS Assessment and Licensing Implementation Process, Revision2, August 2007 | No | No | |||
| IE-DA-M-30 | Internal Events | DA; | Method | WCAP-16341-P: LERF Event Tree failure probabilities for various equipment and operator responses in the accident progression | The PWROG report provides failure probabilities for various LERF parameter values for various equipment and operator responses in the accident progression events such as: Containment Fails Early (Low RCS Pressure @ VB)), Containment Fails Early due to VB @ High Pressure, Containment Fails Early following a High Pressure Core Uncovery with a subsequent late depressurization prior to VB, Pressure induced SGTR (non SBO) and Thermally induced SGTR (non SBO) | Peer reviewed | WCAP-16341-P, Simplified Level 2 Modeling Guidelines, Revision 0, Novemeber 2005 | No | No | |||
| IE-DA-M-31 | Internal Events | DA; | Method | NUREG/CR-6268: Common-Cause Failure Database and Analysis System: Event Data Collection, Classification, and Coding | This report presents guidance for collecting, classifying, and coding common-cause failure (CCF) events. It updates NUREG/CR-6268, “Common-Cause Failure Database and Analysis System,” published in 1998. The U.S. Nuclear Regulatory Commission’s (NRC’s) Office of Nuclear Regulatory Research (RES) and the Idaho National Laboratory (INL) maintain a CCF database for the U.S. commercial nuclear power industry. The CCF data effort consists of CCF event identification, CCF event coding, and CCF parameter estimation. |
Peer reviewed | NUREG/CR-6268: Common-Cause Failure Database and Analysis System: Event Data Collection, Classification, and Coding, Revision 1, Septemeber 2007 | Yes | CAFTA CCFTool Feature: Common cause failure basic events are calculated using CCFTool. | Yes | Idaho National Laboratory (INL) CCF database | |
| IE-DA-M-32 | Internal Events | DA; | Method | NUREG/CR-5485: CCF Methodology | CCF guidelines are provided in NUREG/CR-5485 to help probabilistic risk assessment (PRA) analysts in modeling common cause failure (COF) events in commercial nuclear power plants. The aim is to enable the analyst to identify important common cause vulnerabilities, incorporate their impact into system reliability models, perform data analysis, and quantify system unavailability in the presence of CCFs. Some of the material in this volume has been presented in previous reports, NUREG/CR-4780 and NUREG/CR-5801. The purpose of this document is to bring together the key aspects of these procedural guidelines supplemented by additional insights gained from their application, enhanced by the capabilities of the CCF software and its data analysis capabilities, recently developed by the United States Nuclear Regulatory Commission (NRC). | Peer reviewed | NUREG/CR-5485. Guidelines on Modeling Common-Cause Failures in Probabilistic Risk Assessment, November 1998. | No | No | |||
| IE-DA-M-34 | Internal Events | DA; | Method | Method to identify risk significant failure modes and basic events that require a Bayesian Update to comply with DA-D1... | Use of the SYSIMP Program to determine RAW/F-V to indentify risk significance failure modes and basic events. | Peer reviewed | Data Notebook (Overview of Data Analysis Process) & PRA Calculation that documents SYSIMP cases. | Yes | SYSIMP: Risk Model Importance Analysis Tool, Version 4.0, EPRI Product ID: 3002010660, Software Manual. 2017. | No | ||
| IE-DA-T-20 | Internal Events | DA; | Tool | Thermohydraulic Analysis | Thermohydraulic analysis using a computer code such a MAAP, MELCOR, RELAP to determine sequence events such as core uncovery, Pzr Relief valve lift, operator timing, etc. Must use a rigorous code that has been benchmarked against events and/or other accepted code. Some may consider the code the method. |
Consensus | Thermohydraulic codes have been used since the inception of PRA. | EPRI MAAP User's Group | No | Several acceptable codes are available - MAAP, MELCOR, RELAP, RETRAN | No | |
| IE-DA-T-21 | Internal Events | DA; | Tool | Containment Performance Analysis | Using a suitable computer code, determine the performance of key containment parameters such as temperature and pressure. Several computer codes are available such as MAAP, CONTAIN, etc. |
Consensus | Containment performance analysis has been used since the inception of PRA | EPRI MAAP User's Group | No | As mentioned above, several computer codes are avaiable. | No | |
| IE-DA-T-22 | Internal Events | DA; | Tool | Fission Product Release Analysis | Analysis of fission product releases from the core to the environment. Typically use a computer code such as MAAP. |
Consensus | MAAP User's Group | Yes | MAAP - Other tools are available that could be used. | No | ||
| IE-DA-T-23 | Internal Events | DA; | Tool | Room Heat-Up Analysis | Using computer codes and/or hand calculations, estimate the room heat up for impact on temperature sensitive equipment. Gothic is a suitable code. |
Consensus | Gothic user's manual will help with the analysis. Various reports on temperature impact on equipment. | No | Gothic is probably the most common tool. | No | ||
| IE-HR-M-2 | Internal Events | HR | Method | Integrated Human Event Analysis System for Event and Condition Analysis (IDHEAS-ECA) | The method is known as the Integrated Human Event Analysis System for Event and Condition Assessment (IDHEAS-ECA). It is published as NUREG-2256. It is based on NUREG 2198, “The General Methodology of an Integrated Human Event Analysis System.” IDHEAS-ECA supports risk informed decisionmaking by providing an HRA method to be used in probabilistic risk assessment (PRA) applications. The NRC staff uses PRA in the review of risk-informed license amendment requests and evaluations of notices of enforcement discretion, operational events (e.g., Management Directive 8.3, “NRC Incident Investigation Program,” and the accident sequence precursor program), and inspection findings (i.e., the significance determination process). IDHEAS-ECA was developed because, in recent years, the scope of application of HRA has expanded into situations beyond the scope of existing HRA methods. IDHEAS-ECA is intended to apply to the same situations modeled by existing HRA methods (e.g., nuclear power plant internal events while at-power) and beyond (e.g., external events, low power and shutdown events, and events for which flexible and coping strategies (FLEX) equipment is used). The IDHEAS ECA method provides step-by-step guidance for analyzing a human action and its context. It models a human action by using five macrocognitive functions: Detection, Understanding, Decisionmaking, Action execution, and Interteam coordination. The failure of a human action is modeled with a set of cognitive failure modes and performance influencing factors, which are then used to calculate the human error probability (HEP). The IDHEAS ECA method includes a software tool that facilitates the documentation of the analysis of a human action and uses the results of the analysis as input to calculate the HEP. IDHEAS-ECA uses human error data documented in an NRC HRA database (referred to as IDHEAS-DATA) as the basis for HEP calculation. |
Peer reviewed, developed by consensus body, tested, evalauted, evaluated against and met ASME/ANS PRA Standard high level requirements | ||||||
| IE-HR-M-8 | Internal Events | HR | Method | Cause Based Decision Tree Method (CBDTM) | Cause Based Decision Tree Method (CBDTM) is used to determine Cognitive Error probability. The CBDTM is used to assess HEPs for procedure-directed actions. The CBDTM methodology assesses HEPs by evaluating separate decision trees that evaluate each of the cognitive failure mechanisms. There are two basic failure mechanisms; failure of the operator-information interface and failure of the operator-procedure interface. Each basic failure mechanism consists of four failure mechanisms. See EPRI TR-100259 and HRAC. | peer reviewed, developed by consensus body, extensive use | ||||||
| IE-HR-M-9 | Internal Events | HR | Method | Human Cognitive Reliability Correlation / Operator Reliability Experiments (HCR/ORE) | Human Cognitive Reliability Correlation / Operator Reliability Experiments (HCR/ORE) is used to determine cognitive error probability. The HCR/ORE correlation represents the cognitive error probability as a function of normalized time (the normalized time is a dimensionless unit which reflects the ratio of time available to crew median response time). Normalized time to be limited to time windows on which observations were made. Extrapolation is not valid. Applicability limited to control room EOP-based actions in response to an initaiting event. See EPRI TR-100259 and HRAC. | peer reviewed, developed by consensus body, extensive use | ||||||
| IE-HR-M-18 | Internal Events | HR | Method | Operator Interviews and Simulator Data | Methodology included performing operator interviews and obtaining simulator data for use in the HRA timings or to substantiate estimated HRA timing analysis where integrated and dynamic operator response information is required to support a timing basis. For example, say VCT depletion time is affected by how operations control let down flow. View a simulator run can aid in establish the basis for flow rate changes that affect when the VCT will be depleted. This method also confirmed response models and scenario development | Use of simulator observations or talk-throughs with operators to confirm the response models for scenarios modeled is a CAT II/CAT III requirement in PRA Standard. (HR-E4). Also PRA Standard requires basing required time to complete actions on actual time measurements in either walkthroughs or talk-throughs of the procedures or simulator observations. (HR-G5) | ||||||
| IE-HR-M-32 | Internal Events | HR | Method | Cognitive Quantification Methodology using ASEP | The approach used for most of the fleet PRAs is very similar to the "EPRI Methodology", but some exceptions in how different aspects of the foundation methodologies are used. The most notable exception is that the ASEP time reliability correlation (NUREG/CR-4772) is used rather than HCR/ORE (EPRI TR-100259) for some sites, and the results are added to those of the Cause-Based Decision Tree Methodology (CBDTM, EPRI TR-100259) rather than using the maximum of the two methodologies. The use of the CBDTM was originally developed to serve as a "floor" value for longer term actions in which the HCR/ORE results are negligible, however, the Constellation approach treats the ASEP curve more like a contribution of a "time stress" failure to the cognitive HEP rather than as a separate, comprehensive representation of a diagnosis failure. This treatment is conservative relative to the defined scope of these methodologies. A high level summary of the quantification process is as follows: - Execution error is determined via THERP (NUREG/CR-1278). - Cognitive error is determined by CBDTM and is supplemented with ASEP for actions with short timelines, though sites in the fleet use the maximum of HCR/ORE and CBDTM. - For FLEX and MCRA (Main Control Room Abandonment) actions, the above methods are supplemented with IDHEAS (NUREG-2199 and RIL-2020-2) based on available industry guidance (e.g., EPRI 3002013018, NUREG-1921 and its supplements). For SDPs, additional work is often performed to quantify actions using SPAR-H (NUREG/CR-6693) to obtain insights into how the NRC may model the operator actions. |
All fleet PRAs have been peer reviewed and these different approaches have been accepted. | ||||||
| IE-HR-M-33 | Internal Events | HR | Method | Recovery Action Modeling | Fleet PRAs used to include operator actions to restore the functionality of previously failed systems using screening level actions; however, unless the action was based on specific procedure steps to address a relevant condition, the "recovery" actions were removed from the PRAs (i.e., the action "Operator Fails to Restore Service Water" should no longer be included in the PRAs). Skill-of-the-craft actions are potential exceptions in that they may not be directly proceduralized, but are actions that are supported by operator training and basic knowledge of operations, such as locally operating a motor operated valve that does not open when it is operated from the main control room. Describing the modeling of "recovery actions" is complicated by the different definitions that are associated with this term, but actions that bypass failures, such as manually initiating a system that as failed to start automatically, are quantified using the same process as other post-initiator actions. Feasibility must be demonstrated for any credited action, though it should be noted that some actions, such as skill of the craft actions, may not require procedures to be available to support non-1.0 HEPs. |
All fleet PRAs have been peer reviewed and these different approaches have been accepted. | ||||||
| IE-HR-M-34 | Internal Events | HR | Method | Dependency Analysis | Fleet PRAs identify operator actions appearing in the same CDF or LERF cutset, assess the dependence between the actions, and apply a Joint Human Error Probability (JHEP) for the operator action combinations. The process is largely based on EPRI 3002003150. The EPRI HRA Calculator dependency module is used to facilitate the consideration of time margin, simultaneity, and other factors that influence dependence, such as common cue instruments or procedure direction. The five-level approach developed in THERP (NUREG/CR-1278) is adopted for the treatment of human action dependence (zero, low, moderate, high, or complete dependence). A JHEP floor value of 1.0E-06 (or 5.0E-7, if any action within the combination has a system window that approaches an Operations shift length, typically 12 hours) is applied to any combination with a calculated JHEP value below the applicable floor. The combinations with Fussell Vesley (F/V) values equal to or greater than 5.0E-03 are manually evaluated in full. This approach is considered to satisfy ASME/ANS PRA Standard SR QU-C1. HFE combinations below the F/V threshold are not reviewed except on an incidental basis per the analysis executed for the combinations above the F/V threshold. |
All fleet PRAs have been peer reviewed and these different approaches have been accepted. | ||||||
| IE-HR-M-36 | Internal Events | HR | Method | ASEP - Accident Sequence Evaluation Program Human Reliability Analysis Procedure | HRA Method. See NUREG/CR-4772 and EPRI HRA Calculator | Peer reviewed, extensive use. | ||||||
| IE-HR-M-37 | Internal Events | HR | Method | NUREG/CR-2300 PRA Procedures Guide | NUREG/CR-2300 PRA Procedures Guide: A Guide to the Performance of PRA at NPP, 1983. [NRC ADAMS ML063560439, ML063560440] | |||||||
| IE-HR-M-38 | Internal Events | HR | Method | NUREG-1792 Good Practices for Implementing Human Reliability Analysis | NUREG-1792 Good Practices for Implementing Human Reliability Analysis, 2005. [NRC ADAMS ML051160213] | |||||||
| IE-HR-T-11 | Internal Events | HR | Tool | HRA Calculator Software | The HRA Calculator is used to identify dependent combinations and assess dependencies. | peer reviewed, developed by consensus body, extensive use | ||||||
| IE-HR-T-12 | Internal Events | HR | Tool | MAAP cases (PWR version) | Revision 4.0.6 of the MAAP4 computer code [EPRI, 2006] using plant specific parameter file was used to determine many HRA Timing Analysis inputs for various accident scenarios. | Peer reviewed. Developed by consensus body and extensively used by industry. PRA Standard requires basing the time available to complete actions on plant-specific thermal/hydraulic analysis or simullations | ||||||
| IE-HR-T-13 | Internal Events | HR | Tool | Battery Depletion Calculations using DCSDM (DC System Database Module) for HRA Timing Analysis. (Includes FLEX and non-FLEX depletion calculations) | The DCSDM is a Microsoft Windows based client server database that is composed of a combination of files that contain data on each battery device and associated cables. The DCSDM also provides a method capable of performing voltage drop calculations to each device. The DCSDM is capable of generating reports that will document the results of the analysis. The DCSDM provides the analysis tools required to store and manipulate the data for each calculation. The DCSDM performs iterative load flow/voltage drop calculations and short circuit calculations for up to 200 nodes. The DCSDM can also perform battery cell sizing calculations and determines battery terminal voltage for each minute during discharge. In cases where a one-minute interval consists of multiple time steps, the time step generating the maximum current will be used to determine the battery voltage for the entire minute. These calculations are based on IEEE Standard 485-1997 and industry accepted practice. |
Developed by Consensus Body and is an industry accepted practice. | ||||||
| IE-HR-T-14 | Internal Events | HR | Tool | Simulator benchmarking | Using feedback from simulator to benchmark and verify operator timing | Simulator standards | ||||||
| IE-HR-T-15 | Internal Events | HR | Tool | Internal Flood HRA Timing Analysis Calculations | Methodology used simple math calculations based supported by various engineering calculations including plant-specfic GOTHIC calculations, flow and differential height calculations, flood elevations for critical equipment, door analysis, etc. Hand calculations were performed to determine system time window for specific flood areas and pipe break sizes for modeled internal flood operator actions. |
Peer Reviewed Internal Flood Model GOTHIC Code used for design basis calculations at site. GOTHIC was modified to support plant specific analysis. GOTHIC Developed by consensus body and extensively used by industry. | ||||||
| IE-HR-T-16 | Internal Events | HR | Tool | Job Performance Measure (JPM) Information for HRA Timing Analysis | Methodology involved using job performance measure timing requirements for use in the HRA timings or to substantiate estimated HRA timing analysis. JPM are developed by the Training Department and are used to test operators and document that operator actions are correctly and within a certain time period. | Use of simulator observations or talk-throughs with operators to confirm the response models for scenarios modeled is a CAT II/CAT III requirement in PRA Standard. (HR-E4). Also PRA Standard requires basing required time to complete actions on actual time measurements in either walkthroughs or talk-throughs of the procedures or simulator observation | ||||||
| IE-HR-T-19 | Internal Events | HR | Tool | Fuel Oil Storage Tank Depletion Calculation: Used for HRA Timing Analysis (which standard) - Jayne to look up standard | Methodology used simple math calculations based on plant-specific parameters for Fuel Oil Storage Tank Volume, FOTP flow capacity, Diesel usage at 2500KW loading and low level pump setpoint and other system information to determine depletion time. | Extensive Use : Basic math calculations using plant specific parameters. PRA Standard requires basing the time available to complete actions on plant-specific thermal/hydraulic analysis or simulations. (HR-G4, CAT III). | ||||||
| IE-HR-T-24 | Internal Events | HR | Tool | WCAP-11993 (Joint Westinghouse Owners Group/Westinghouse Program: assessment of Compliance with ATWS Rule Basis for Westinghouse PWRs): Used for System Time Window | This report presents the results of a joint Westinghouse Owners Group (WOG) and Westinghouse program to quantify the frequency of core damage resulting from Anticipated Transients Without Scram (ATWS) for Westinghouse pressurized water reactors (PWRs). The objectives of this program were to: - document the important factors invol ved in assessing ATWS core damage frequency, - provide a method that can be used in evatuati ng the impact of changes on A TWS core damage frequency, - document the application of the ATWS Rule (10CFR50.62) and basis (SECY-83-293) for Westinghouse PWRs, - assure ATWS core damage criteria and approach compatibility with the Severe Accident Policy. This report is intended to be the vehicle to address ATWS Rule and basis compliance for Westinghouse PWRs. |
Peer Reviewed. Developed by consensus body and extensively used by industry. | ||||||
| IE-HR-T-25 | Internal Events | HR | Tool | GOTHIC Room Heatup Analysis use for HRA Timing Analysis | The thermal hydraulic analysis for the Steam Exclusion areas was evaluated using the GOTHIC (Generation of Thermal Hydraulic Information for Containments) software package. The primary purpose of this analysis is to develop best-estimate environmental (i.e. temperature, pressure, humidity) profiles for the Turbine Building after High Energy Line Break events using the GOTHIC 8.1 software package. There are several systems that could adversely impact the environment in Turbine Building and are evaluated in either PRA Internal Flooding analysis and/or the Design Basis HELB analysis. The design basis GOTHIC model will be modified to support the analysis required for this calculation. Changes to the model include more detailed modeling of the physical building layout, changes to existing heat loads and trip setpoints, and addition of non-safety related loads that could be energized in PRA scenarios. Modifications to the model will be documented and justified in the calculation. |
GOTHIC Code used for design basis calculations at site. GOTHIC was modified to suuport plant specific analysis.GOTHIC Developed by consensus body and extensively used by industry. PRA Standard requires basing the time available to complete actions on plant-specific thermal/hydraulic analysis or simulations. (HR-G4, CAT III) | ||||||
| IE-HR-T-26 | Internal Events | HR | Tool | Technical Manual Used for HRA Timing Analysis | Use information from Technical Manual as input into HRA Timing Analysis. | Peer Reviewed. | ||||||
| IE-HR-T-27 | Internal Events | HR | Tool | Plant Procedures: Steady State Battery Depletion | Using operating information provided in plant procedures to determine timing for steady-state battery depletion. | Considered part of plant-specific operating procedures. | ||||||
| IE-HR-T-28 | Internal Events | HR | Tool | Consumption Rate of Air Supply to Control valves for Control Room Chiller Operation: Used for HRA Timing Analysis | Engineering Calculation: Since the available pressure is known and the required time is known, the methodology for this calculation will be to use simple math to determine the maximum allowable pressure loss per hour. This will be determined by using the capacity of the bottles minus the air pressure that will be left in each bottle (available pressure) divided by the number of hours the bottles are designed to provide a pneumatic source.This is true because of the ideal gas law: PV = mRT. The mass of air consumed by the chiller is then ?m = ?(PV/RT). For this system V(bottle), R, and T are all constant so the change in the mass of air is directly proportional to the change in pressure in the bottles. The calculations were updated to take into account realistic data from the safeguards chiller preventative maintenance procedure. |
Based on Design Basis Calculation. PRA Standard requires basing the time available to complete actions on plant-specfic thermal/hydraulic analysis or simulations . (HR-G4, CAT III) | 3 | |||||
| IE-QU-M-1 | Internal Events | QU | Method | PRAQuant/FRANX/SYSIMP (Architect) | Model quantification software | Extensive use | Software Quality Assurance program | Software Quality Assurance program | Yes | No | ||
| IE-QU-M-5 | Internal Events | QU | Method | Importance Metrics | Use CAFTA for FV and Birnbaum, Use SYSIMP for RAW | Extensive use | Software Quality Assurance program | Software Quality Assurance program | Yes | No | ||
| IE-QU-M-6 | Internal Events | QU | Method | Uncert | Uncertainty analysis | Extensive use | Software Quality Assurance program | Software Quality Assurance program | No | No | ||
| IE-QU-M-7 | Internal Events | QU | Method | Circular Logic Breaking | Circular logic breaking | Extensive use | ASME/ANS PRA Standard | ASME/ANS PRA Standard | Yes | CAFTA | No | |
| IE-QU-M-8 | Internal Events | QU | Method | Mutually exclusive rule development | Use CAFTA modeling and/or a .MUTX file to apply mutually exclusive rules | extensive use | Software Quality Assurance program | Software Quality Assurance program | Yes | CAFTA, PRAQuant | No | |
| IE-QU-M-10 | Internal Events | QU | Method | RG 1.174 | "RG 1.174 is used to characterize generic sources of uncertainty in the Quantification Notebook. EPRI has developed a recommended standard set of sensitivity cases to perform that envelop selected potential sources of uncertainty at a relatively high level. Evaluating this standard set of sensitivity cases instead of attempting to identify and characterize all potential sources of uncertainty associated with these issues has the potential benefit of highlighting the potential impact of these specific issues prior to performing applications. A standard set of four sensitivity cases were evaluated and CDF/LERF values were compared to base case values. The delta risk values can be compared to the acceptance guidelines (RG 1.174) to obtain insights into the sensitivity of the base PRA model results to these generic high level sources of modeling uncertainty. These guidelines state that if risk-informed applications clearly result in a decrease in CDF and LERF, the change satisfies the relevant principle of risk-informed regulation. When an application results in an increase in CDF between 1E-06/rcy to 1E-05/rcy, the application will be considered only if it can be reasonably shown that the total CDF is less than 1E-04/rcy and LERF is less than 1E-05/rcy. The results of these sensitivity cases provide added assurance of model understanding prior to performing applications. |
Consensus | RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3 | No | No | |||
| IE-QU-M-11 | Internal Events | QU | Method | NUREG-1855 | NUREG-1855 describes the characteristics of a risk model and, in particular, a PRA. Modeling uncertainties must be considered in both the base PRA and in specific risk-informed applications. Uncertainties associated with scope and level of detail should be documented in the PRA, but would be only considered for their impact on a specific application. This report assists QU Supporting Requirements for identification of each model uncertainty and how the PRA model is affected and documentation of the assessment in the QU Notebook. |
Consensus | NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk Informed Decision Making, Revision 1 | No | No | |||
| IE-QU-M-12 | Internal Events | QU | Method | EPRI 1016737 | EPRI 1016737 compiled an initial listing of generic sources of uncertainty to be considered for each PRA technical area that complements the guidance in NUREG-1855. This report assists QU Supporting Requirements for identification of each model uncertainty and how the PRA model is affected and documentation of the assessment. |
Consensus | EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, December 2008 | No | No | |||
| IE-QU-M-13 | Internal Events | QU | Method | NUREG/CR-2728 | NUREG/CR-2728 provides guidance for breaking circular logic loops in the model. This report assists QU Supporting Requirements for breaking circular logic loops in fault tree linking. |
Consensus | NUREG/CR-2728, Interim Reliability Evaluation Procedure Guide, January 1983 | No | No | |||
| IE-QU-M-14 | Internal Events | QU | Method | NUREG/CR-6595 | Report is used to define SERF (Small Early Release Frequency) and is referenced in QU notebook to document that no attempt is made in the PRA to quantify smaller early releases with is consistent with the requirements of the ASME risk standard. | Consensus | NUREG/CR-6595, An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Sandia National Laboratory, Revision 1, October 2004 | No | No | |||
| IE-QU-M-15 | Internal Events | QU | Method | Large fault tree / small event tree | Each event tree sequence that leads to either core damage or large early release is individually modeled in the at-power internal events fault tree, as the model follows the large fault tree/small event tree modeling paradigm. Because each core damage and large early release sequence is fully modeled within the fault tree, there are no “transfers” between event trees that could result in “successes” that are not transferred between event trees. The FTREX solution engine uses a delete term approximation method to produce its results. | Extensive use | ASME/ANS PRA Standard | ASME/ANS PRA Standard | No | No | ||
| IE-QU-T-2 | Internal Events | QU | Tool | FTREX | Quantifier | Extensive Use | Software Quality Assurance program | Software Quality Assurance program | Yes | No | ||
| IE-QU-T-3 | Internal Events | QU | Tool | Qrecover/Recovery Rules | post-processing | extensive use | Software Quality Assurance program | Software Quality Assurance program | Yes | No | ||
| IE-QU-T-4 | Internal Events | QU | Tool | Cutset Editor / Reviewing Results | Using CAFTA to edit and view/verify cutset results | extensive use | Software Quality Assurance program | Software Quality Assurance program | Yes | No | ||
| IE-QU-T-9 | Internal Events | QU | Tool | FLAG files | Flag files are generally used for one of two main purposes 1) applying HFE seed values 2) applying data changes BE events for sensitivities or applications | Extensive use | Software Quality Assurance program | Software Quality Assurance program | Yes | PRAQuant, Sysimp | No |