ML22264A169

HEAF Research Overview

ML22264A169
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Issue date:	09/21/2022
From:	Coyne K, Kenneth Hamburger, Nick Melly, Markhenry Salley, Gabe Taylor NRC/RES/DRA
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Hamburger, Kenneth 301 415 2022
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26thInternational Conference on Structural Mechanics in Reactor Technology (SMiRT 26) 17th International Post-Conference Seminar on FIRE SAFETY IN NUCLEAR POWER PLANTS AND INSTALLATIONS Overview of U.S. Nuclear Regulatory Commission Research Activities to Update Probabilistic Risk Treatment of High Energy Arcing Faults Kevin Coyne*, Nicholas Melly, Kenneth Hamburger, Gabriel Taylor, Mark Henry Salley United States Nuclear Regulatory Commission, Washington, DC, 20555-0001, United States of America ABSTRACT International nuclear power plant (NPP) operating experience has demonstrated that High Energy Arcing Faults (HEAFs), which result from electrical system failures, can cause a sus-tained arc that leads to the rapid release of energy in the form of heat, vaporized metal, and mechanical force. Because of their hazard and potential risk significance, the U.S. Nuclear Regulatory Commission (NRC) has implemented a multi-year research plan to develop more realistic models and methods to assess HEAF hazards. This paper summarizes, and provides a status update on, the research objectives and key tasks, including: (1) model development and survey of plant electrical applications and configurations; (2) physical testing needed to inform and validate the HEAF hazard model and assess component fragility; and (3) updates to Probabilistic Risk Assessment data and methods to improve the realism and fidelity of the HEAF hazard model. This work also supports the ongoing international Organisation for Eco-nomic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) HEAF 2 Project.

INTRODUCTION Electrical system failure HEAF events are energetic electrical arcing faults that can lead to the rapid release of energy. This energy release can result in high heat fluxes in the vicinity of the HEAF, vaporization of metal, release of ionized gas and smoke, and mechanical shocks to nearby equipment. The characteristics of HEAF events result in faster development of the fire and associated damage compared to more traditional fire events that include delay time re-lated to the fire ignition and growth stages [1]. As a result, HEAF initiated fire events can rapidly propagate to other equipment and result in unexpected challenges associated with the rapid onset of fire related damage and impacts from smoke and ionized gases.

Because of their potential to induce an initiating event (e.g., loss of feedwater, reactor trip) and lead to failure of adjacent mitigating equipment, HEAF events can have risk-significant impacts.

The NRCs Accident Sequence Precursor (ASP) program systematically evaluates U.S. NPP operating experience to identify potential severe accident precursors [2]. An ASP precursor is defined as an initiating event with a conditional core damage probability (CCDP) greater than or equal to 1 x 10-6 (one in a million) or a degraded plant condition resulting in an increase in core damage probability (CDP) greater than or equal to 1 x 10-6. The ASP program identified nine HEAF related precursor events since 1989. Seven of the nine ASP HEAF precursors were analysed as an initiating event and quantified using CCDP, with results ranging from

~ 2 x 10-6 to ~ 4 x 10-4. For these events, the HEAF events resulted in plant transients, loss of electrical busses, and losses of offsite power. Two of the ASP HEAF events were assessed

as a degraded condition due to the nature of the specific condition and were quantified by CDP, with results ranging from ~ 4 x 10-6 to ~ 4 x 10-4. Key insights derived from these as-sessments are the risk significance associated plant transients, loss of electrical busses, and loss of offsite power for HEAF events. Further, design and quality control deficiencies that lead to high resistance breaker connections increase the potential for initiation of HEAF events.

The insights from the U.S. experience are consistent with international operating experience for HEAF events [3].

Given the potential risk importance of HEAF events, the NRC has undertaken substantial work over the last two decades, in collaboration with the Electric Power Research Institute (EPRI) and the OECD/NEA, to develop methods and models to support the risk assessment of HEAFs.

Specific activities include: (1) model development and survey of plant electrical applications and configurations; (2) physical testing needed to inform and validate the HEAF hazard model and assess component fragility; and (3) updates to Probabilistic Risk Assessment (PRA) data and methods to improve the realism and fidelity of the HEAF hazard model. The results of this program are expected to improve the realism of HEAF models in fire probabilistic risk assess-ment (Fire PRAs).

Regulatory Context The NRCs fire protection requirements are contained in Title 10 to the U.S. Code of Federal Regulations (CFR) Part 50.48, Fire protection. These regulations require a NPP to implement a fire protection plan that meets the requirements of 10 CFR 50, Appendix A, General Design Criterion (GDC) 3, Fire protection. GDC 3 specifies, in part, that NPP systems, structures, and components (SSCs) important to safety be designed and located to minimize the proba-bility and effect of fires and explosions. The NRCs regulations provide two options for meeting this requirement: (1) a deterministic option described under 10 CFR 50.48(b) that references the requirements contained in 10 CFR 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, and generally applicable to NPPs li-censed before 1979 or as specified in license conditions for later plants; or (2) a risk-informed, performance-based option provided under 10 CFR 50.48(c) which allows use of the National Fire Protection Association (NFPA) Standard NFPA 805, 2001 Edition [4], with some excep-tions. The approach described in Appendix R provides deterministic requirements, such as limiting the damage from a fire to ensure that one train of equipment necessary to achieve hot shutdown is free from fire related damage. These criteria generally are supported by prescrip-tive requirements covering specific fire protection features such as the use of 1-hour or 3-hour fire barriers, installed automatic fire detection and suppression, 20 ft (6.1 m) of horizontal sep-aration with no intervening combustibles, or use of alternate shutdown capability. Conversely, NFPA 805 specifies high level goals, performance objectives, and performance criteria for nuclear safety and radioactive release. The licensee may use engineering analysis, probabil-istic risk assessment, or fire modeling calculation to demonstrate that the performance goals, objectives, and criteria are satisfied. As described in the Statements of Consideration for the final rulemaking approving use of NFPA 805 [5], the methodology incorporates a number of attributes consistent with a performance-based approach, notably:

• measurable or calculable parameters exist to monitor the system, including facility performance,

• objective criteria to assess performance are established based on risk insights, deter-ministic analyses, and/or performance history,

• flexibility to determine how to meet established performance criteria in ways that will encourage and reward improved outcomes, and 2

• a framework exists to assess the failure to meet performance criterion and to ensure that such failures, while undesirable, will not constitute or result in an immediate safety concern.

The NRC provides guidance for implementing the NFPA-805 option in Regulatory Guide (RG) 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants [6]. Although a Fire PRA is not mandatory to transition to NFPA 805, RG 1.205 notes that a plant-specific Fire PRA enables a licensee to fully realize the safety and cost benefits of making a transition to NFPA 805, particularly by supporting risk-informed changes to the fire protection program that can be made without prior NRC approval. To the extent that a Fire PRA is used to support NFPA 805 implementation, the NRC must find the PRA ap-proach, methods, and data acceptable. The NRC endorsed NUREG/CR-6850/EPRI 1011989, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Final Report [7] and its Supplement 1 [8], as one acceptable method for conducting a Fire PRA. For the purposes of assessing HEAF events, NUREG/CR-6850 and its supplement define a zone of influence (ZOI) surrounding the initiation point of a HEAF event within which certain equipment is as-sumed to be damaged. For example, the following ZOI guidelines are provided for HEAF events in NUREG/CR-6850:

• For HEAF events within an electrical cabinet, any unprotected cables in the first overhead cable tray within 1.5 m (5 ft) vertical distance of the top of the cabinet and 0.3 m (1 ft) horizontally from the cabinet face will be within the ZOI and assumed to damaged. Addi-tionally, any equipment within 0.9 m (3 ft) horizontal distance from the cabinet front or rear panel and at or below the top of the cabinet will be within the ZOI (NUREG/CR-6850, Volume 2, Appendix M).

• For HEAF events within an iso-phase bus ducts, the ZOI is assumed to be a sphere cen-tred on the fault point and measuring 1.5 m (5 ft) feet in radius (NUREG/CR-6850, Sup-plement 1).

• For HEAF events within non-iso phase bus ducts, the ZOI includes: (1) a downward ex-panding cone from the point of arcing enclosing a total solid angle of 30° to a maximum diameter of 6.0 m (20); (2) a sphere with a radius of 0.45 m (1.5 ft) from the point of arcing (assumed to be the center of the bus duct) (NUREG/CR-6850, Supplement 1).

Motivation for HEAF Research In 2013 the OECD/NEA FIRE (Fire Incidents Records Exchange) Database Project published an analysis of HEAF events that occurred up to mid-2012 [3], which recommended:

to perform experiments for obtaining comprehensive scientific fire data on the HEAF phenomena known to occur in nuclear power plants through carefully designed experi-ments, to be able to develop more realistic models to account for failure modes and con-sequences of HEAF, to advance the state of knowledge and provide better characteriza-tion of HEAF in fire PRA, and, in particular, to answer key questions, which cannot yet be answered form analyzing the HEAF events in the OECD FIRE Database. Such key ques-tions can be how to prevent and to detect HEAF or what would be the best way for limiting pressure phenomena and minimizing fire barrier element failures.

In response to this recommendation, during the period of 2014 through 2016, twenty-six full-scale HEAF experiments were conducted under an OECD/NEA HEAF Project testing program

[9]. These experiments used equipment typically found in NPP applications and controlled for four primary parameters: (1) location of arcing within the electrical enclosure; (2) arcing fault current; (3) voltage; and (4) arc duration. Key findings included the potential for increased severity of a HEAF event when aluminum was present and the increased likelihood of a HEAF 3

event leading to ensuing fires for longer duration arcing events (i.e., greater than 2 seconds).

As a result of the insights developed from this testing campaign, in addition to other relevant U.S. HEAF operating experience, the NRC issued Information Notice (IN) 2017-04, High En-ergy Arcing Faults in Electrical Equipment Containing Aluminum Components [10]. IN 2017-04 noted that the equipment with copper components exhibited similar damage states as those postulated in NUREG/CR-6850; however, results obtained for equipment containing aluminum components exhibited damage states beyond NUREG/CR-6850 predictions. This observation was a key motivation for additional HEAF research activities.

OVERVIEW OF HEAF RESEARCH PROJECT PLAN The NRCs HEAF research project plan [11] is built upon earlier operating and experimental experience, including the results of the OECD/NEA HEAF Project testing program. This test program also provided valuable insights for measuring HEAF phenomena, including the sen-sor design requirements for measurement of heat release rates and total incident energy. In addition, the NRC developed an International Phenomena Identification and Ranking Table (PIRT) expert elicitation in 2017 [12] to prioritize research and regulatory needs.

The project plan consists of five main tasks:

1. Development and validation of a computational fluid dynamics (CFD) model capable of predicting HEAF hazards for a wide variety of electrical (e.g., system voltage, available current, fault duration) and equipment configurations (e.g., geometry and materials). The development of a verified and validated predictive model will reduce the amount of testing required to provide representative HEAF results applicable to a broad range of nuclear power plants. However, as discussed below, the development of the CFD (computational fluid dynamics) model required additional physical testing to assess HEAF phenomena and configuration information regarding operating plant electrical distribution systems.

2. Survey of the U.S. nuclear fleet to ensure that full-scale HEAF experiments are repre-sentative of plant configurations, and to better understand the location, configuration, and material composition of equipment to support PRA method development. The survey pro-vides information on equipment manufacturers, models, voltages, insulation, and the ma-terials (e.g., aluminum, copper, steel).

3. Physical testing to support the development and validation of the CFD model. This includes a limited set of experiments that span the range of critical parameters to ensure that the development and validation of the model provide acceptable results. Three scales of ex-periments are included: small, medium (or open-box), and full-scale tests, with each se-ries designed to investigate aspects of the HEAF phenomena that are best observed at these different length scales. As noted in the research project plan:

• Small-scale experiments characterize the morphology and oxidation states of alumi-num and copper particles generated by a HEAF event,

• Medium-scale open box experiments characterize the electrical arc to support pre-diction of arc energy emitted during a HEAF event and provide information on enclo-sure and electrode mass loss data for both copper and aluminum electrodes, and

• Full-scale experiments provide prototypical nuclear power plant data on enclosure breach, pressure effects, and serve as the representative scenarios for CFD models validated.

4. Fragility testing to assess target damage considering the shorter, higher energy source term associated with HEAF events and support development of a fragility model. The fra-gility model supports the development of the PRA method for assessing HEAF risk.

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5. PRA method development to improve the realism and fidelity of the HEAF hazard model.

This task includes an evaluation of U.S. operating experience, updated fire ignition fre-quencies, updated target fragility thresholds, and updated non-suppression probabilities.

Figure 1 provides an overview of the HEAF research project plan activities. These activities are grouped into three main areas: analytical tasks shown in blue (e.g., tasks 1, 2, and 5),

experimental tasks shown in red (e.g., tasks 3 and 4), and outputs from this work shown in green (tools, methods, and data).

Figure 1 Overview of HEAF research project activities A significant portion of the research project plan is being performed under the auspices of the NRC / EPRI Memorandum of Understanding (MOU). This arrangement allows the research activity to better leverage the knowledge and expertise to support the objectives of this work.

Each of the tasks is discussed in more detail in the following sections.

DEVELOPMENT AND VALIDATION OF A HEAF CFD MODEL This activity focused on the development of a CFD model of HEAF events capable of calcu-lating the incident energy for a variety of equipment configurations and materials. The pre-dicted incident energy at various distances from the modelled electrical enclosure is subse-quently used to determine the ZOIs for specific equipment. The objective of this work was the 5

development of a tool that could leverage experimental data but provide information for con-figurations that were not subject to full-scale testing. This provided a more cost-effective and flexible approach considering the high costs associated with conducting full-scale HEAF ex-periments. The Fire Dynamics Simulator (FDS) was used to support this effort. FDS was ini-tially released in 2000, has undergone extensive verification and validation, and provides sig-nificant capabilities to study fire dynamics and combustion. FDS includes a hydrodynamic model capable of solving the low-speed, thermally driven flow using a large eddy simulation turbulence model [13]. FDS includes a combustion model and a radiative heat transport model, in addition to capabilities for accommodating a wide variety of geometric configurations. A calculation matrix for FDS simulation runs was built on information gathered from surveying the U.S. nuclear fleet (e.g., switchgear manufacturer, bus bar materials, potential fault loca-tions), operating experience, and previous testing. Over 130 simulation runs were identified, covering the following characteristics:

• Low voltage switchgear - bus bar material (aluminum, copper), arc duration, arc lo-cation, arc energy (34 FDS simulations);

• Medium voltage switchgear - bus bar material (aluminum, copper), arc duration, arc location, arc energy (42 FDS simulations);

• Non-segregated bus ducts - duct and bus bar material (aluminum, copper), arc dura-tion, arc location, and arc energy (57 FDS simulations).

Figure 2 FDS calculated thermal plume for a 226 MJ HEAF (medium voltage switchgear cabinet)

An overview of the preliminary results of this work was published for public comment in May 2022 [14].

To provide additional confidence in the ZOIs calculated using FDS, the NRC staff also devel-oped a modified model based on IEEE 1584-2018, IEEE Guide for Performing Arc-Flash Hazard Calculations [15]. This standard was developed to estimate incident energy at various distances from an arc flash event for the purpose of electrical safety. The NRC effort modified the IEEE arc flash model and its application in several ways, including fault current input (arc current rather than bolted fault current), inclusion of an enclosure breaching time, and solving for incident energy to determine a ZOI based on plant specific target fragilities [16]. Results obtained from the modified arc-flash confirmatory calculations are consistent with the more detailed FDS analysis for orientation experience of the more severe exposure conditions, thereby providing additional confidence in the CFD-derived ZOIs.

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SURVEY OF THE U.S. NUCLEAR FLEET EPRI developed and implemented a survey of U.S. nuclear power plants to obtain information about electrical distribution systems, including materials used, fault clearing times, electrical distribution configurations, switchgear manufacturer, and switchgear configuration. The re-sults of this survey are summarized in an EPRI technical report [17].

Physical Testing to Characterize HEAF Phenomena The physical testing portion of the research project plan was focused on obtaining information needed to develop and validate the FDS CFD model. Three types of experiments were per-formed:

• Small-scale experiments were conducted to better characterize aluminum and copper par-ticle size distribution, rates of particle production, and particle morphology produced dur-ing electrical arc faults. Additionally, these tests provided accurate current measurements from which other electrical characteristics (e.g., arc impedance, energy, and power) could be determined. Thirty-six experiments were conducted, using variations in voltage, arc current, arc duration, arc gap size, and bus bar materials. The results from these small-scale experiments inform the energy balance model used in the FDS CFD model to predict additional energy release from particle involvement in the arc fault. The results of these experiments are documented in a Sandia National Laboratories (SNL) report [18].

Figure 3 Small-scale arc fault test system

• Medium-scale open box experiments were similar to the small-scale experiments, but conducted at a larger scale. The open box allowed for more direct observation of the arc, enclosure breach, material loss, and electrical properties. The experiments were con-ducted in steel cubic metal boxes with one open face and a three-phase arcing fault lo-cated in the center of the box. Low- and medium-voltage experiments were run with both aluminum and copper electrodes. The experiments varied arc current and arc duration.

Eleven low voltage (~ 1000 V) and six medium voltage (~ 6900 V) experiments were 7

conducted to obtain mass loss information form the enclosure and electrodes. The results are summarized in an NRC Research Information Letter [19].

Figure 4 Open box experiment OB2 (copper electrode, 1000 V, 15 kA)

• Full-scale experiments were conducted in 2018 for nearly identical medium-voltage

(~ 6900 V) switchgear with aluminum bus bars. These experiments used two different cur-rent and arc durations for a total of four full-scale tests. Insights from the experimental series included timing information related to enclosure breach, event progression, mass loss measurements for electrodes and steel enclosures, peak pressure rise, particle anal-ysis, along with visual and thermal imaging data to better understand and characterize the hazard [20]. These experiments were run by the NRC to evaluate the aluminum HEAF hazard and are identified in the test matrix for the broader OECD/NEA HEAF 2 Project, an international joint project conducted under the auspices of the NEA [21].

Figure 5 Medium voltage switchgear testing, Test 2-19 (6.9 kV, 25.8 kA rms)

Collectively, these experimental campaigns enhanced the state-of-knowledge about HEAF phenomena, enabled more realistic modelling of HEAF events, and provided numerous les-sons learned for experimental and sensor design to capture key HEAF phenomena.

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EQUIPMENT FRAGILITY TESTING AND MODELLING The objective of this portion of the research project plan was to establish the target fragilities for equipment that might be present within the HEAF ZOI. HEAF events generate short dura-tion, extremely high heat fluxes and have different characteristics than traditional combustion fires. Therefore, the existing fragility for equipment exposed to a longer duration thermal fire, particularly cables, may not be applicable to the shorter duration higher heat flux expected from HEAF exposures. The fragility testing was conducted at the Solar Furnace at Sandia National Laboratories [22]. This facility is capable of concentrating sunlight to generate a heat flux of up to 6 MW/m2 within a circle that is approximately 5 cm in diameter. Several different failure modes were evaluated, including ignition, damage as a function of total energy, electri-cal failure of cables, and sub-jacket temperature.

Figure 6 Parabolic dish at the Solar Furnace Figure 7 Solar Furnace Test 1-34 (3 cable bundle, ~ 25 MJ/m2)

Results from this testing, along with the full- and medium-scale data was used to develop a fragility model by the joint NRC/EPRI working group. The fragility model was based on exper-imental data, operating experience, and the current state of HEAF knowledge [23]. The work-ing group focused on electrical cables; the most common target assessed for Fire PRA.

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Damage thresholds for both thermoplastic and thermoset cables were developed as a function of total incident energy. In addition, the working group developed damage criteria for electrical bus ducts and provided guidance for crediting electric raceway fire barrier systems. These damage criteria are used in conjunction with the hazard modelling results to determine the ZOI.

HEAF PRA METHOD DEVELOPMENT The previously described HEAF research activities supported the development of new meth-ods and data to assess HEAF events more realistically within a nuclear plant PRA. In addition to providing updated ZOI limits based on operating experience and experimental testing, the research also led to updated HEAF ignition frequency and non-suppression probabilities [24].

While NUREG/CR-6850 and its supplement provided only one ZOI for each equipment type (switchgear, bus ducts, and isophase bus ducts), the new method provides a more realistic ZOI estimates based mainly on fault clearing time and arc energy. The new ZOIs were deter-mined based on FDS simulations for load centers, switchgear, and non-segregated bus ducts.

Based on the current state of knowledge, the working group could not make any distinction in the ZOIs based on the type of conductor material (copper or aluminum) in the PRA method.

Future testing as part of the OECD/NEA HEAF Phase 2 Project, is expected to provide con-firmatory evidence regarding this modelling conclusion.

Figure 8 Representative changes in ZOI based on fault clearing time (left figure has shorter time)

The final PRA method report is expected to be issued in final form in late calendar year 2022 or early 2023, following the public comment period.

CONCLUSIONS The NRCs fire research program has an overarching objective to improve the realism of fire modelling based on operating experience, experimental data, and analysis. The HEAF re-search project plan has effectively leveraged U.S. and international experience, in addition to 10

a productive collaborative relationship between the USNRC and EPRI, to improve the model-ling of this important phenomenon and guidance on its application.

ACKNOWLEDGEMENTS This work would not have been possible without the collaboration of staff from the Electrical Power Research Institute, Sandia National Laboratories, KEMA Labs, Brandon Stanton Inc.,

and the U.S. National Institute of Standards and Technology. The NRC gratefully acknowl-edges the support from the following individuals:

• U.S. Nuclear Regulatory Commission Working Group members, including Thinh Dinh, J. S, Hyslop, Kenn Miller, Charles Moulton, Reinaldo Rodriguez, David Stroup, Sunil Weerakkody, Jen Whitman, and Antonios Zoulis;

• Electric Power Research Institute: Ashley Lindeman, Marko Randelovic, Tom Short, and Fernando Ferrante. In addition, the following individuals provided substantial sup-port to EPRI during this project: Jason Floyd, Dane Lovelace, Sean Hunt, Ken Fleischer, P. Shannon Lovvorn, and Victor Ontiveros;

• U.S. National Institute of Standards and Technology: Kevin McGrattan, Anthony D.

Putorti, Scott Bareham, Edward Hnetkovsky, Christopher Brown, Wai Cheong Tam, Erik Link, Michael Selepak, Philip

Deardorff,

and Andre Thompson;

• Sandia National Laboratories: Chris LaFluer, Paul Clem, Byron Demosthenous, Aus-tin Glover, Raymond Martinez, Anthony Tambakuchi, Kenneth Armijo, Alvaro Augusto Cruz-Cabrera, James Taylor, Rana Weaver, and Caroline Winters;

• OECD/NEA High Energy Arcing Fault Project participants, including the project Sec-retariat Markus Beilmann and participating countries including Belgium, Canada, France, Germany, Japan, Korea, the Netherlands, Republic of Korea, and Spain.

• KEMA Labs: Frank Cielo and the staff at KEMA Labs;

• Brandon Stanton, Inc.: Rob Taylor and the staff at BSI.

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[22] United States Nuclear Regulatory Commission (U.S. NRC): HEAF Cable Fragility Test-ing at the Solar Furnace at the National Solar Thermal Test Facility, Research Infor-mation Letter (RIL) 2021-09, (SAND2021-11327), Washington, DC, USA, September 2021, https://www.nrc.gov/docs/ML2125/ML21259A256.pdf.

[23] United States Nuclear Regulatory Commission (U.S. NRC) and Electric Power Research Institute (EPRI): Target Fragilities for Equipment Vulnerable to High Energy Arcing Faults, Research Information Letter (RIL) 2022-01, (EPRI 3002023400), Washington, DC, and Palo Alto, CA, USA, May 2022, https://www.nrc.gov/docs/ML2213/ML22131A339.pdf.

[24] United States Nuclear Regulatory Commission (U.S. NRC) and Electric Power Research Institute (EPRI): High Energy Arcing Fault Frequency and Consequence Modeling, Draft Report for Comment, Draft NUREG-2262, Washington, DC, and Palo Alto, CA, USA, July 2022, https://www.nrc.gov/docs/ML2215/ML22158A071.pdf.

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