a fundamental study on safety analysis technology

Abstract—In the field of nuclear engineering, very high

temperature reactor (VHTR) is drawing attention as the next generation nuclear plant (NGNP). The VHTR is appropriate to produce the huge amounts of hydrogen and electricity. For the optimum design and safety of VHTR system, it is required to develop the safety analysis technology of VHTR. In this study, a fundamental research was performed to establish the VHTR safety analysis technology. As part of an effort to understand the current status of VHTR safety analysis technology, this study investigated the VHTR phenomena identification and ranking table (PIRT), event categorization and acceptance criteria developed at home and abroad. This article presents the review outcome of current status and prospect of the safety analysis technology of VHTR.

Keywords—Nuclear engineering, PIRT, Safety analysis, VHTR

I. INTRODUCTION HE very high temperature reactor (VHTR) is defined as a helium-cooled, graphite moderated reactor with a core

outlet temperature in excess of 900°C and a long-term goal of achieving an outlet temperature of 1000°C. Due to the very high temperature in core outlet, the VHTR is suited for the production of huge amounts of hydrogen in addition to electricity. From this characteristic, the VHTR is drawing attention as the next generation nuclear plant (NGNP) in nuclear engineering field. In the VHTR, there are two candidates depending on the core type; one is prismatic core option and the other is pebble bed core option. This study focused on the prismatic modular reactor (PMR) as shown in Fig. 1.

The construction of demonstration reactor is needed prior to the construction of commercial reactor because the design

Seongsu Jeon (corresponding author) is with the FNC Technology Co. Ltd., 135-308, Seoul National University, Daehak-Dong, Gwanak-Gu, Seoul, 151-742, S. Korea (e-mail: [email protected]).

Suhyun Hwang is with the FNC Technology Co. Ltd., 135-308, Seoul National University, Daehak-Dong, Gwanak-Gu, Seoul, 151-742, S. Korea (e-mail: [email protected]).

Yeonjun Choo is with the FNC Technology Co. Ltd., 135-308, Seoul National University, Daehak-Dong, Gwanak-Gu, Seoul, 151-742, S. Korea (e-mail: [email protected]).

Changwook Huh is with Korea Institute of Nuclear Safety, 19 Guseong, Yuseong, Daejeon, 305-338, S. Korea (e-mail: [email protected]).

Changyong Jin is with Korea Institute of Nuclear Safety, 19 Guseong, Yuseong, Daejeon, 305-338, S. Korea (e-mail: [email protected]).

Kyuntae Kim is Korea Institute of Nuclear Safety, 19 Guseong, Yuseong, Daejeon, 305-338, S. Korea (e-mail: [email protected]).

concept of VHTR is different from PWR. South Korea plans to obtain the operating license for demonstration reactor of VHTR by 2022. For the optimum design and safety of VHTR system, it is required to develop the safety analysis technology of VHTR. Since the current safety analysis technologies for nuclear facilities are based on the light water reactors (LWRs), it is required to develop the safety analysis technologies considering the characteristics specific to VHTR. For the development of the safety analysis technology, it is essential to develop the phenomena identification and ranking table (PIRT) and to establish the event classification and acceptance criteria for VHTR.

The objective of this research is to investigate VHTR PIRT, event categorization and acceptance criteria developed at home and abroad prior to the development of the fundamental technology of safety analysis. This article presents the review outcome of current status and prospect of the safety analysis technology development of VHTR in South Korea.

Fig. 1 PMR plant configuration

A Fundamental Study on Safety Analysis Technology Development of VHTR

Seongsu Jeon, Suhyun Hwang, Yeonjun Choo, Changwook Huh, Changyong Jin, and Kyuntae Kim

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International Conference on Trends in Mechanical and Industrial Engineering (ICTMIE'2011) Bangkok Dec., 2011

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mailto:[email protected]

II. CURRENT STATUS OF SAFETY ANALYSIS METHODOLOGY DEVELOPMENT

Safety analysis for the licensing of current LWRs has been primarily based on deterministic safety analysis methodology. Thus, the current safety analysis for reactor design and licensing does not utilize the risk-informed methodology, but only depends on deterministic methodology with conservative assumptions to compensate for uncertainties. However, a number of new comprehensive safety analysis methods, sometimes based on PSA, have been developed and are increasingly used to improve almost all facets of plant operation and maintenance. Therefore, it is technically possible to develop a methodology for integrated design-safety analysis of VHTR taking advantage of the PSA technique.

Korea Institute of Nuclear Safety (KINS) reviewed the conventional deterministic safety assessment methodologies and the newly suggested risk-informed safety assessment methodologies that had been suggested for selecting the licensing basis events (LBEs) or design basis events (DBEs) which play an utmost important role in assuring nuclear safety. The risk-informed safety assessment methodologies include the technology-neutral framework (TNF) approach developed by the USNRC. In order to identify desirable characteristics of the safety assessment methodology to be used to ensure safety of domestic future reactors, the strengths and weaknesses of the conventional deterministic and risk-informed methods were carefully reviewed. For the effective harmonious use of deterministic and risk-informed methodologies, it has been determined that the PSA technique, providing a systematic means of identifying accident sequences after accounting for system interactions, common cause failures, and human errors, must be somehow utilized along with the deterministic techniques. In this study, technology-neutral safety assessment methodology has been developed and can be applicable to all types of future reactors. The design safety can be analyzed by selecting LBEs based on the methodology. The nine step approach (Table I) developed herein could be used to evaluate the design safety of future reactors including VHTR under development in Korea [1]. Although the risk-informed technology-neutral safety assessment approach can be effectively used for LBE selection and safety analysis, the existing safety philosophy must be also considered in the integrated safety assessment of future reactors.

TABLE I

NINE STEP APPROACH METHODOLOGY Step 1 Determination of multiple barrier to be protected Step 2 Systematic understanding for initiating events Step 3 Quantification of occurrence frequency of initiating events

Step 4 Development of event sequences for each initiating event using PSA event tree

Step 5 Determination of category according to occurrence frequency of event sequences / classification according to frequency category

Step 6 Determination of type according to the effects of event sequences for plants and classification according to type

Step 7 Selection of limiting event sequences belong to each occurrence frequency category and type

Step 8 Detailed quantitative analysis for each limiting event sequences Step 9 Check for satisfaction of acceptance criteria and safety margin

III. VHTR PIRT

The PIRT provides a structured means of identifying and analyzing a wide variety of off-normal sequences that potentially challenge the viability of complex technological systems. As applied to VHTR, the PIRT is used to identify a spectrum of safety-related sequences or phenomena that could affect those systems, and to rank order those sequences on the basis of their frequencies, their potential consequences, and state of knowledge related to associate phenomena. It is to be used as an early screening tool to identify, categorize, and characterize phenomena and issues that are potentially important to risk and safety of VHTR.

A. PIRT Procedures The PIRT specific to VHTR is being developed based on the

approach, which consists of nine distinct steps [2] as follows. • Step 1: Define the issue that is driving the need for a PIRT. • Step 2: Define the specific objectives for the PIRT. • Step 3: Define the hardware and the scenario for the PIRT. • Step 4: Define the evaluation criterion. • Step 5: Identify and review the current knowledge base. • Step 6: Identify plausible phenomena; PIRT elements. • Step 7: Develop importance ranking for phenomena. • Step 8: Assess knowledge level (KL) for phenomena. • Step 9: Document PIRT results. KNGR LBLOCA PIRT [3] and SMART-P PIRT [4] use 15

Steps as shown in Fig. 2. Though there are some differences between VHTR and KNGR/SMART-P PIRT, the philosophy and procedures are very similar.

Fig. 2 KNGR and SMART-P PIRT procedures

B. Review on the Characteristics of NRC, KAERI, and ANL PIRTS

Since a specific design has not yet been selected for the choice of the US VHTR (NGNP), it was decided early on to focus on a generic plant and reactor design with broadly typical features. Both a generic Pebble Bed Reactor (PBR) design and a generic Prismatic Modular Reactor (PMR) design were selected as the reference plant for KAERI [5] and ANL PIRT [2]. The generic PBR design selected is a version of the 400 MWt South African PBMR design. The generic PMR design selected is a version of the 600 MWt GT-MHR.

The reference plant of NRC PIRT [6] is assumed to be a modular high temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GT-MHR) version (a


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prismatic-core modular reactor-PMR) or a pebble bed modular reactor (PBMR) version (a pebble bed reactor-PBR) design, with either a direct- or indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production.

The difference of VHTR PIRT characteristics among NRC, KAERI and ANL will be reviewed here. The typical shape for each PIRT is shown in Fig. 3.

(a) NRC PIRT Example

(b) KAERI PIRT Example

(c) ANL PIRT Example

Fig. 3 Examples of NRC, KAERI and ANL PIRTs The event sequences of KAERI PIRT consist of High

Pressure Conduction Cooldown (HPCC), Low Pressure Conduction Cooldown (LPCC), and Load Change (LC). The event sequences of ANL PIRT consist of Water Ingress, Rod Withdrawal ATWS (Anticipated Transient Without Scram), and Hydrogen Plant Upset. On the other hand, the event sequences of NRC PIRT consist of Normal operation, General LOFC (Loss Of Forced Circulation), Pressurized LOFC, Depressurized LOFC, Air ingress LOFC, Reactivity (ATWS), and IHX failure (molten salt). Although there are some differences in the event sequences between NRC and KAERI/ANL, the event sequences except for water ingress can be matched with each other.

For the event sequence of water ingress, steam in-leakage from a high-pressure SG would be a dominant risk factor. Otherwise, primary water-cooled heat exchanger secondary systems (in Brayton cycle designs) would run at lower operating pressures and present minimal risks of any substantial water-steam ingress. And NGNP design options with a high-pressure steam generator (Rankine cycle) in the primary loop are not considered in NRC PIRT. Hence, water ingress was eliminated from current NRC PIRT.

Conventionally, the PIRT for each event sequence is divided by system/component categories. Table II shows system/ component classification of KAERI/ANL PIRT. The structure of NRC PIRT is different with KAERI/ANL PIRT as shown in Fig. 3. The column “Issue” of NRC PIRT contains the phenomena and system/component information together and doesn’t distinguish the system/component separately.

KAERI/ANL PIRT evaluates the importance level by dividing sequence phase as shown in Fig. 3. The safety criteria, which supply the basis for the determination of importance level, are shown in Table III. But in NRC PIRT, sequence phase is not divided for each event sequence. And NRC PIRT supplies more detailed safety criteria compared with KAERI/ANL PIRT as shown in Table IV.

Knowledge level and its rationale are supplied in NRC PIRT, but that aren’t in KAERI/ANL PIRT. It is described in KAERI report that knowledge level ranking will be carried out later, when the existing experimental and analytical databases will be surveyed and evaluated.

TABLE II

SYSTEM/COMPONENT CLASSIFICATION System Component

Reactor Vessel

Inlet Plenum Riser

Top Plenum and Components Core & Reflectors

Outlet Plenum and Components Lower Head

Pressure Boundary

Reactor Coolant

Loop

Hot/Cold Pipe Compressor (Direct) or Circulator (Indirect) Intermediate Heat Exchange and Circulator

Mixing Junction (US VHTR) Shutdown

Cooling System Heat Exchanger and Pump

RCCS Reactor Cavity (Confinement)

RCCS Tube (Air Duct) RCCS Piping, Air Cooler and Chimney

TABLE III

KAERI/ANL SEQUENCE PHASE AND SAFETY CRITERIA Event Phase Safety Criteria

High Pressure Conduction Cooldown

1: Coastdown Fuel & Vessel Temperature 2: Conduction Cooldown

Low Pressure Conduction Cooldown

1: Blowdown Fuel & Vessel Temperature 2: Air Ingress

3: Natural Convection

Load Change 1: Flow Reduction Local Hot Spot 2: New Steady State

Water ingress 1: Pre Turbomachine-Trip Peak Fuel Temperature Peak Vessel Temperature

Confinement Radiation Limits 2: Post Turbomachine-Trip

Rod Withdrawal

ATWS

1: Pre Turbomachine-Trip Peak Fuel Temperature

Vessel Wall Temperature 2: Coastdown 3: Post Turbomachine-Trip

Equilibrium

Hydrogen Plant Upset

1: Pre Protection-System Trip Core Outlet Temperature

Vessel Pressure 2: Post Protection-System Trip


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TABLE IV NRC SAFETY CRITERIA

PIRT FOM (Figure of Merit)

Normal operation

- fuel time at temperature - fuel failure fraction - core support structures - time at temperature - vessel and vessel supports temperature - RCCS cavity temperature - primary system boundary integrity - dose to worker/public

General LOFC

- fuel failure fraction - limit vessel temperature - vessel and vessel support integrity - maintain coolable geometry - vessel support temperatures - concrete temperature

Pressurized LOFC

- vessel support temperatures - concrete temperature - fuel temperature - upper vessel support, vessel - damage to SCS HX - pressure boundary failure

Depressurized LOFC

- dose - peak fuel temperature - structural integrity of RCCS - failure of additional pipes

Air ingress LOFC

- fuel temperature - fuel and structural damage - fuel failure fraction - core, concrete integrity - core, core support structure - reactor vessel support - dose to public - Cavity temperature and pressure - RCCS integrity - vessel support, vessel temperature

Reactivity (ATWS)

- fuel failure fraction - corrosion of core supports - dose to public - core support

IHX failure (molten salt)

- public and worker dose - vessel, vessel support, and core support temperatures

IV. EVENT CLASSIFICATION

It is important to establish the event classification and acceptance criteria for the development of safety analysis technology. The current event classification applied to LWRs according to event consequence is as follows;

- Increase in heat removal by the secondary system - Decrease in heat removal by the secondary system - Decrease in reactor coolant system flow rate - Reactivity and power distribution anomalies - Increase in reactor coolant inventory - Decrease in reactor coolant inventory - Radioactive release from a subsystem or component - Anticipated transient without scram (ATWS) The acceptance criteria consist of three categories as follows; - Reactor coolant pressure boundary - Dose limit - Fuel integrity The event classification and acceptance criteria for VHTR

are being developed based on the current technology system as well as VHTR specific characteristics.

V. CONCLUSIONS

The VHTR design characteristics are very different from existing LWRs. For the optimum design and safety of VHTR system, it is required to develop the safety analysis technology of VHTR. In this study, a fundamental research was performed to establish the VHTR safety analysis technology. The nine step approach methodology for the evaluation of the design safety of VHTR and previous VHTR PIRTs were reviewed. From the review of previous VHTR PIRTs, it is found that 1) NRC PIRT doesn’t include the detailed phase of event sequence and system/component; 2) KAERI/ANL PIRT doesn’t evaluate knowledge level. Therefore, new PIRT will be developed to obtain the completed PIRT shape by complementing the weakness of the PIRT of NRC, KAERI, and ANL. Furthermore, event classification and acceptance criteria are being developed to supply the basis for VHTR licensing technology.

REFERENCES [1] KINS/HR-1011, “Methodology Development for Establishment of

Licensing Basis Events and Safety Analysis of GEN-IV Reactors”, 2010. [2] NUREG/CR-6944, “Next Generation Nuclear Plant Phenomena

Identification and Ranking Tables (PIRTs)”, 2008. [3] INEEL, “Phenomena Identification and Ranking Tabulation, Korean

Next Generation Reactor, Large Break Loss of Coolant Accident”, February 2001.

[4] KAERI/TR-2546/2003, “Development of a Preliminary Phenomena Identification and Ranking Table of Thermal hydraulic Phenomena for SMART-P”, July 2003.

[5] KAERI/TR-3050/2005, “Generation of a Preliminary PIRT for Very High Temperature Gas-Cooled Reactors”, September 2005.

[6] ANL-GenIV-071, “Prioritization of VHTR System Modeling Needs Based on Phenomena Identification, Ranking and Sensitivity Studies”, April 2006.


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a fundamental study on safety analysis technology

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