Safety Issues for High Temperature Gas Reactors Andrew C. Kadak - - PowerPoint PPT Presentation

Safety Issues for High Temperature Gas Reactors

Andrew C. Kadak Professor of the Practice

Major Questions That Need Good Technical Answers

• Fuel Performance

– Normal operational performance – Transient performance

• Ejected Rod (maximum energy insertion capability)
• Reactivity insertions (seismic, water)

– Accident Performance – Weak fuel issues – Mechanistic source term for high burn-up fuel – Fuel fabrication quality assurance

• Risk Dominant Accident Sequences

– Establish risk informed design to identify risk dominant accident sequences to be analyzed. – Use either IAEA1 or NRC2 risk informed approach to establish safety requirements of plant. – Use of safety goal as a design guide – Application of risk informed “Defense in Depth” – Scope of risk analysis may be easier due to inherent robustness of basic design.

• 1. “Development of Technology Neutral Safety Requirements for Innovative Reactors”, IAEA

TECDOC Draft Dec. 2004 2. “Regulatory Structure for New Plant Licensing, Part 1: Technology Neutral Framework, Dec. 2004, Draft, US NRC.

Consequences F

Initiating Event/Y

• Lev. 1
• Lev. 2
• Lev. 3
• Lev. 4
• Lev. 5

10-2 10-6 10-7

Failure of Level 1 Initiating Event

Minimal emergency actions beyond defined distance from the plant No off-site actions beyond defined distance from the plant Off-site Actions NOTE 2: Doses are derived from IAEA-SS No 115 5 mSv/a (For 1 year period following the accident) 5 mSv/a (For 1 year period following the accident) 5 mSv/a 1 mSv/a (average 5 y) 1 mSv/a (10 µ Sv/a –target) Doses to the public NOTE 1: Doses for NO, AOO, AC are derived from IAEA-SS No 115 500 mSv (limit) (This value derived from Finnish regulation) 50mSv/a (Could be exceeded for rear recovery events) 50 mSv/a 20 mSv/a (average 5 y) (5 mSv/a target) 50 mSv/a ALARA (5 mSv/a target) Doses to Operators Severe plant conditions* (SSPC) Accident conditions (AC) Anticipated Operational Occurrences (AOO) Normal Operations (NO) Plant conditions Consequences

AOO AC SPC Challenges DESIGN BASIS * Severe challenge to the Fission Products Confinement Function

Risk Informed Safety Profile

LEVELS OF DEFENCE IN DEPTH (From INSAG-10)

Control, limiting and protection systems Level 2 Control of deviations from normal

• peration and detection of failures

and other surveillance features Objective Levels of defence Essential means Level 1 Prevention of deviations from normal

• peration and failures

Conservative design and high quality in construction and operation Level 3 Control of accident conditions within the design basis Engineered safety features and accident procedures Level 4 Control of severe plant conditions Complementary measures and accident management Level 5 Mitigation of radiological consequences

• f significant releases of radioactive

materials Off-site emergency response Acceptable failures of the Level of Defence

(events/year) < 10-2 10-2- 10-6 10-6- 10-7 < 10-7

• Expected Significant Accident Sequences

– Air Ingress – Water Ingress (reactivity insertion) – Seismic Events (reactivity insertion) – Loss of Load – Rod Ejection (more significant in block reactors) – Failure of reactor cavity cooling system – Recuperator By-pass events (overcooling) – Graphite dust, plate-out, lift off – Impact of Terrorism – Identification of “cliff edge” effects

Knowledge Required

• Improved understanding of core behavior
• Improved understanding of heat transfer in core and vessel
• pebble and block - bypass flows
• Materials behavior at high temperature in helium (plus

contaminants) including radiation effects and chemical attack on graphite

• Blow down loads and timing of accident event sequences.
• Behavior of fuel, fission product release behavior in

reactor building and structures under accident conditions.

• Development and validation of computer codes used in the

analysis

• Validation of passive performance of safety systems -

natural circulation - heat conduction and convection.

Issues

• Fuel Temperature limits (1600 C ?)
• Regulatory Credit for Basic Design

Strengths

• Need new risk informed licensing process

to allow credit for innovative systems.

Containment

• Based on design and accident analysis of source

term and sequences - a containment of radioactive materials strategy is developed to assure that safety goals are met. – Full pressure containment – Confinement - low pressure - not pressure tight – Dynamic containment/confinement (time dependent) – Performance is quite different than water reactors.

Classification of Safety “Systems”

• Ideally safety system classification should

be done on importance to safety function in a risk informed manner.

• Some “systems” are not components but

parameters in analysis for passive performance (ex. emissivity of reactor vessel).

Expectations

• Water Ingress - generally understood and can be

limited by amount of water ingress - some German experience at AVR

• Seismic - reactivity simulations can assess reactivity

impact.

• Rod ejection - more significant for block reactors but

fuel energy limits like for LWRs can be established for rod worths.

• Testing on heat transfer and flow can be verified by

South African tests and Chinese pebble bed reactor including reactor cavity cooling systems.

• Fuel behavior data to be provided by past German and

focused South African and US testing programs

Challenges

• Verification of high temperature material

behavior (fuel, graphite, metals, carbides)

• Validation of analysis tools
• Air ingress

– Most visible concern among the public – Most significant in terms of potential offsite consequences – Can not be eliminated by “design”

Air Ingress Status

• Most “eliminate” connecting “vessel”

failure as too low a probability event (10-8).

• Break sizes limited to largest connecting

“pipe”.

• Two breaks (top and bottom) considered

unlikely but are analyzed (chimney effect)

• Graphite corrosion behavior not well

modeled in existing codes.

• CFD analysis and confirmatory experiments

needed.

Air Ingress Tests

• Japanese series on prismatic configuration

– Diffusion – Natural Circulation – Corrosion (multi-component)

• German NACOK tests - pebble bed

– Natural circulation – Corrosion

• MIT CFD (Fluent Methodology Development)

Experimental Apparatus - Japanese

C4 2 7

Nitrogen Helium Valves

C3 C1 C2 H4 H3 H2 H1

Figure 16: Apparatus for Isothermal and Non-Isothermal experiments Figure 17: Structured mesh

Isothermal Experiment

0.00 0.20 0.40 0.60 0.80 50 100 150 200 250 300 Time (min) Mole fraction

H-1 & C-1(Calculation) H-2 & C2 (Calculation) H-3 & C3 (Calculation) H-4 & C4 (Calculation) H-1 & C-1(Experiment) H-2 & C2 (Experiment) H-3 & C3 (Experiment) H-4 & C4 (Experiment)

Figure 18: Mole fraction of N2 for the isothermal experiment

Thermal Experiment

Figure 19: The contour of the temperature bound4ary condition

Pure Helium in top pipe, pure Nitrogen in the bottom tank N2 Mole fractions are monitored in 8 points

• Hot leg heated
• Diffusion Coefficients as a

function of temperature

Thermal Experiment

0.2 0.4 0.6 0.8 1 50 100 150 200 Time (min) Mole fraction of N2

H-1(FLUENT) C-1(FLUENT) H-1(Experiment) C-1(Experiment)

Figure 20: Comparison of mole fraction of N2 at Positions H-1 and C-1

0.2 0.4 0.6 0.8 1 50 100 150 200 Time(min) Mole Fraction

H2(Experiment) C2(Experiment) H-2(FLUENT) C-2(FLUENT)

Figure 21: Comparison of mole fraction

• f N2 at Positions H-2 and C-2

Thermal Experiment (Cont.)

0.2 0.4 0.6 0.8 1

50 100 150 200 250

Time(min) Mole Fraction of N2 H4(Exp) C4(Exp) H-4(Calc) C-4(Calc)

Figure 22: Comparison of mole fraction

• f N2 at Positions H-1 and C-1
• 0.15
• 0.10
• 0.05

0.00 0.05 0.10 0.15 0.20 0.25 2 4 6 Time (Second) Velocity (m/second)

Figure 23: The vibration after the

• pening of the valves.

Multi-Component Experiment

2 1 3 4

Heated Graphite Air Helium

Graphite Inserted Multiple gases: O2, CO, CO2, N2, He, H2O Mole fraction at 3 points are measured Much higher calculation requirements Diffusion Coefficients

Figure 34: Apparatus for multi- Component experiment of JAERI

Multi-Component Experiment(Cont.)

0.00 0.03 0.06 0.09 0.12 0.15 0.18 0.21 20 40 60 80 100 120 140 Time(min) Mole Fraction O2(Experiment) O2(Calculation) CO(Experiment) CO(Calculation) CO2(Experiment) CO2(Calculation) Figure 36: Mole Fraction at Point-1 (80% Diffusion Coff.)

Multi-Component Experiment(Cont.)

Figure 37: Mole Fraction at Point-3

0.00 0.04 0.08 0.12 0.16 0.20 0.24 20 40 60 80 100 120 140 Time(min) Mole Fraction

O2(Experiment) O2(Calculation) CO(Experiment) CO(Calculation) CO2(Experiment) CO2(Calculation)

Multi-Component Experiment(Cont.)

Figure 38: Mole Fraction at Point-4

0.00 0.05 0.10 0.15 0.20 0.25 20 40 60 80 100 120 140 Time (min) Mole Fraction O2(Experiment) O2(Calculation) CO(Experiment) CO(Calculation) CO2(Experiment) CO2(Calculation)

NACOK Natural Convection Experiments

Figure 39: NACOK Experiment

Boundary Conditions

Figure 41: Temperature Profile for one experiment

The Mass Flow Rates

Figure 42: Mass Flow Rates for the NACOK Experiment 0.0E+00 1.0E-03 2.0E-03 3.0E-03 4.0E-03 100 300 500 700 900 1100 Temperature of the Pebble Bed (C) Mass Flow Rate (kg/s 5.0E-03 )

T_R=200 DC(Exp.) T_R=400 DC(Exp.) T_R=600 DC(Exp.) T_R=800 DC(Exp.) T_R=200 DC(FLUENT) T_R=400 DC(FLUENT) T_R=600 DC(FLUENT) T_R=800 DC(FLUENT)

Future NACOK Tests

• Blind Benchmark using MIT methodology

to reproduce recent tests.

• Update models
• Expectation to have a validated model to be

used with system codes such as RELAP and INL Melcor.

Air Ingress Mitigation

• Air ingress mitigation strategies need to be

developed

– Realistic understanding of failures and repairs – Must be integrated with “containment” strategy to limit air ingress – Short and long term solution needed

Overall Safety Performance Demonstration and Validation

• China’s HTR-10 provides an excellent test bed for

validation of fundamentals of reactor performance and safety.

• Japan’s HTTR provides a similar platform for

block reactors.

• Germany’s NACOK facility vital for

understanding of air ingress events for both types.

• PBMR’s Helium Test Facility, Heat Transfer Test

Facility, Fuel Irradiation Tests, PCU Test Model.

• Needed - open sharing of important technical

details to allow for validation and common understanding.

Chinese HTR-10 Safety Demonstration

• Loss of flow test

– Shut off circulator – Restrict Control Rods from Shutting down reactor – Isolate Steam Generator - no direct core heat removal only but vessel conduction to reactor cavity

Video of Similar Test

Loss of Cooling Test

Power

Loss of Cooling Test

Power

Summary

• Safety advantages of High Temperature

Reactors are a significant advantage.

• Air ingress most challenging to address
• Fuel performance needs to be demonstrated in
• perational, transient and accident conditions.
• Validation of analysis codes is important
• Materials issues may limit maximum operating

temperatures and lifetimes of some components.

• International cooperation is essential on key

safety issues.