Concept Review of Canadian Supercritical Water … - OFFICIAL USE ONLY / À USAGE EXCLUSIF - EACL...

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Concept Review of Canadian Supercritical Water

Reactor and Probabilistic Safety Assessment

Thambiayah (Nithy) Nitheanandan

2015 March 24


Outline

Introduction

Canadian Supercritical Water Reactor Design

Features

Physics Design and Analysis

Fuel Channel and Material Selection

Thermalhydraulics Analysis

Safety Philosophy for Canadian SCWR

Summary


What is Supercritical Fluid?


Why Supercritical Water as Coolant? Supercritical water-cooled reactor: Benefit: Significantly Improved (more than 40%) cycle efficiency as compared to current LWRs. No steam separation equipment. No steam generators. Known Technology: Supercritical fossil fuel plant technology has been well-established. Originally developed in the 1950s. More than 400 SC fossil plants are operating world-wide. Challenge: Reactor Core Design for the significantly increased operating temperature (up to 625 ºC) and pressures (~25MPa).

LWR (285ºC), ~33% Canadian SCWR (625 ºC), ~48%

30

35

40

45

50

55

CANDU 6 (5 MPa, 265ºC)

17 MPa, 540ºC

25 MPa, 550ºC

27 MPa, 590ºC

29 MPa, 610ºC

Canadian SCWR (25

MPa, 625ºC)

35 MPa, 710ºC

Cyc

le E

ffic

ien

cy [%

]

Steam Pressure and Temperature

LWR 7MPa 285ºC


Outline

Introduction


Features





Summary


Reference Design Schematic

• Grid Power = ~1200 MWe

• Operating Life = 75 Years

• Direct cycle – No steam generators, no

steam separators

• 336 fuel channels at 25 MPa

•Inlet temperature at 350 C

•Outlet temperature at 625 C

• Fuel channels

•Zirconium pressure tube

•Zirconia insulator

• Inside diameter = 5.5 m

• Low-pressure calandria at ~0.3 MPa

• Thorium-13%Plutonium fuel (Ref),

• Enriched UO2 is possible.


• SCW is single-phase fluid. Hence, no steam separators (as in BWRs) or steam generators (as in PWRs and CANDUs) are needed. As a result, a direct steam cycle is adopted.

• SCW turbine technology is mature. SCW turbines exist that operate at 600ºC.

• Batch fuelling and vertical orientation are adopted. This allows the use of a simpler gantry crane fuelling machine instead of a high-pressure robotic on-line fuelling machine typical in CANDUs,

• Passive safety systems are adopted for all safety cases. These systems have very few or no moving components, and hence are more reliable and simpler to maintain.

Main design features and simplifications


Reference Core Design Concept

Light Water

Coolant Inlet

Inlet Plenum

Outlet Header Head

Outlet Header

Fuel Channels

Calandria Vessel

Fuel Channel

Supports

Light Water

Coolant Outlet

Outflow

Inflow

Liner Tube

Flow Tube

Pressure Tube

Fuel Element

Fuel Element

Flow Tube

Liner Tube

Pressure Tube

Flow Reversal

Inlet Plenum Head

(a) Reactor Core

(b) Cross-Over Piece

(c) Bottom of Fuel Channel

Reactivity Control

Drives


Batch Refuelling Single Fuel Channel Replacement


Low-Pressure

Calandria Vessel Heavy Water at 0.3 MPa,

~80 ºC

Horizontal Control

and Shutoff Rods Shutoff rods are fail-safe,

spring loaded

Heavy water Moderator Also used as coolant in

the passive moderator

cooling system


Inlet plenum and outlet header connections

Inlet Plenum Head

Inlet Plenum

Coolant Inlet Nozzle

Coolant Outlet Nozzle

Tubesheet

Outlet Header Head

Outlet Header

Pressure Tube Extension

Outlet Header Supports


Thermal and Mechanical Characteristics Parameter Value

General

Power MWt/MWe 2540 MWt/1200MWe

Number of Fuel Channels 336

Steam Cycle Direct

Coolant Supercritical Light Water

Inlet Conditions 25.3 MPa, 350ºC

Outlet Conditions 25.0 MPa, 625ºC

Inlet Plenum

Inside Diameter 6.25 m

Outside Diameter 7.15 m (7.4 m at flanged connection to lid)

Tubesheet Thickness 1.0 m

Candidate Material ASME SA-508 Grade 3 Class 1

Inlet and Outlet Piping

Inside Diameter of Inlet Piping 31 cm

Thickness of Inlet Piping 3.5 cm

Inside Diameter of Outlet Piping 43 cm*

Thickness of Outlet Piping 5.0 cm*

* This is for Alloy 625. Will be updated for the recommended alloy 690


Fuel Channel Extensions

(a) Fuel Assembly Support

and Locking Mechanism

(b) Locked

(c) Unlocked

Fuel Assembly

Outlet

Locking Mechanism

Spring

Locking Balls

Metallic ‘E’ Seal

Fuel Channel

Extension

Inner Sleeve

Equipped with

ramp

Locking Balls

(Retracted Position)

Outlet

Header

Locking

Mechanism

Fuel Assembly

Support and

Seal

Inlet Plenum

Tubesheet

Fuel Assembly


Outlet Header Thermal Sleeve

Outlet Header

Thermal Isolation Sleeve

Liner Sleeve

Outlet Piping

Expansion Bellows

Guide Bushing


Tubesheet Connection

Fuel Channel Extension

Fuel Assembly

Pressure Tube

Inlet Plenum Tubesheet

Seal Weld


Fuel Assembly

Pressure Tube


Tubesheet


Fuel Channel Bottom Details


Fuel Assembly with Wire Wraps


Computational Fluid Dynamics (CFD)

Analysis of the Inlet Plenum

7 m/s

30 m/s


Flow Distribution in the Inlet Plenum

Outlet plenum (red) and pressure tube extensions

(blue) are inside the inlet plenum.

CFD analysis of the inlet plenum showing velocity

vectors




vectors


Temperature and flow distribution




vectors

Temperature


Structural Analysis of the Inlet Plenum

(Carleton University)


Inlet cross-over piece thermal stress

Stresses are high ~200 MPa. Fatigue may be an issue.

Separation of inlet (blue pipes) and outlet (red pipes) streams reduce stress


Outline

Introduction


Features





Summary


Analytical Tools

•Toolset for conventional HWR modeling:

• Nuclear data libraries: “in-house” ENDF/B-VII.0-based libraries

• Lattice physics: WIMS-AECL

• Post processing: WIMS-UTILITIES

• Core physics: RFSP


WIMS-AECL Lattice Level Modeling

• 2D deterministic code, 89-group cross sections

• Representation of cross section of fuel bundle, channel and surrounding moderator

• Calculates energy and space dependent neutron flux distribution

• Calculates fuel depletion versus time based on flux or power history


WIMS-UTILITIES Post Processing Code Suite

• 2D homogenization / condensation of WIMS-AECL cross sections into 2-group cross sections for RFSP

• Other applications include perturbation based reactor cross section tables, and calculation of kinetics parameters for use in RFSP


RFSP Core Level Modeling • 3D 2-group neutron

diffusion code

• Solves for k-eff and 3d flux / power distribution

• Core follow mode

• Interpolation of reactor tables gives burnup dependent cross sections

• Transient analysis capable when coupled with T/H code, e.g. CATHENA


Lattice Physics – Generic Behaviour

• Balance between maximizing exit burnup and minimizing CVR

• Similar challenge with power shaping / reactivity suppression and exit burnup


Evolution of the Fuel Bundle

• Increased burnup

• Improved fissile utilization

• Reduced CVR

• Reduced LER

• Fuel options include: LEU, LEU-Th, Pu-Th, TRU-Th


Central flow tube as moderator

• Blue = thermal flux

• Orange = fission power


Impact of Fuel Type on Spectrum and

Kinetics

• Harder neutron spectrum with Pu-Th fuel (but still thermal spectrum)

• Requires higher initial fissile content compared to enriched uranium to reach same burnup

• Lower delayed neutron fraction from Pu-Th results in faster kinetics


Reactivity Suppression

• The SCWR is a batch fuelled reactor

• Have to use corresponding fuel management strategy

• High initial reactivity needs to be suppressed

• Depletable neutron absorber (gadolinia) incorporated in fuel

• Gadolinia depletion characteristics rely on spatial self shielding


Adjuster Rod Configuration

•5 horizontal banks of 7 rods each on opposite faces of the core

TOP VIEW SIDE VIEW


Core Channel Power Distribution

•Quarter core channel power map (adjusters in)


Emergency Liquid Injection Shutdown System

keff decrease by more than 100 mk in less than two seconds


Outline

Introduction


Features





Summary


Core Physics Parameters • Average initial fissile = 8.6 wt% heavy element

• Core average exit burnup = 42 MWd/kg

• Cycle length ~ 290 days

(can be improved via optimization of absorbers)

• CVR* ~ -30 to -45 mk

• Fuel temperature coefficient* ~ -0.05 mk/K

• Moderator temperature coefficient* ~ -0.1 mk/K

• Relative Peak Channel Power ~ 1.2

• Relative Axial Peak Power ~ 1.2 – 1.3

• Maximum linear element rating ~ 35 kW/m

• Liquid injection shutdown system depth ~ 100 mk in 1 second

* Calculation based on core without reactivity suppression


• In-core materials should be neutron “transparent”.

• Targeted maximum operating temperatures up to 650oC.

• High strength to retain coolant (pressure tube at 25 MPa).

• Resistance to irradiation induced deformation.

• 75 year life-time operation.

• Separation (seal) between the high-temperature coolant and low-temperature moderator.

• Allowance for single channel replacement.

Requirements of the Canadian SCWR Fuel

Channel


Canadian SCWR Fuel Channel

Fuel Channel Bellows Alloy 625

Welded to the bottom plate of the outlet header which facilitates differential movement between the outlet header and the channel

Pressure Tube Extension Alloy 718

Which connects the pressure tube to the outlet header.

Seal weld between the pressure-tube extension and the inlet plenum which allows coolant to enter into the fuel channel and subsequently to the fuel assembly.

Pressure Tube Excel


Canadian SCWR Fuel Channel

Bellows Pressure tube extension Seal Weld To allow seal welding of the pressure tube to the inlet plenum, the pressure tube transitions from Alloy 718 to zirconium at a point 0.5m above the fuelled region. Excel pressure tube

moderator

tubesheet

Inlet plenum

Outlet header


• In co-extrusion a billet comprising of two materials in the correct proportions and geometry is forced through an extrusion press die under a controlled atmosphere.

• In the extrusion process, the two materials are bonded at the molecular level, providing a joint whose mechanical properties are similar to the parent materials.

• The technical concept and feasibility of a co-extruded Zr-2.5Nb/stainless steel end fitting joint for CANDU reactors was proven.

Co-Extruded Pressure Tube


Dimensions Fuel Channels

Fuel Channel Configuration

Square Array

Pitch 250 mm

Pressure Tube OD and Thickness

181 mm,

12 mm

Pressure Tube Length 6.58 m

Pressure Tube Extension Length

1.0 m Pressure Tube


Tubesheet


Pressure Tube Strength as a Function of

Temperature

• Zirconium alloys are good for

pressure tubes as they are a low neutron absorbing material.

• Mechanical strength decreases with increasing temperature.

• Option is to insulate the pressure tube so that it operates at a lower temperature.

• Low temperature decreases the creep, corrosion/hydrogen pickup rates, and crack propagation rates.

Maximum coolant temperature

SCWR moderator temperature


The Canadian SCWR fuel channel incorporates concepts from the Re-Entrant Channel (REC) design.

Evolution of the Canadian SCWR Fuel

Channel Concept

REC

Light Water

Coolant Inlet

Inlet Plenum

Outlet Header Head

Outlet Header

Fuel Channels

Calandria Vessel

Fuel Channel

Supports

Light Water

Coolant Outlet

Outflow

Inflow

Liner Tube

Flow Tube

Pressure Tube

Fuel Element

Fuel Element

Flow Tube

Liner Tube

Pressure Tube

Flow Reversal

Inlet Plenum Head

(a) Reactor Core

(b) Cross-Over Piece

(c) Bottom of Fuel Channel

Reactivity Control

Drives

The pressure tube is different from the REC in that an insulator rather than the coolant keeps the pressure tube at a low temperature .


• High-strength, creep-resistant zirconium alloy Excel

(Zr - 3.5%Sn - 0.8%Nb - 0.8%Mo – 1130 part-per-million (ppm) O).

• Solution strengthening effect of Sn.

• 75 year life-time operation without re-tube.

• Each pressure tube is in direct contact with the moderator at 120oC and 150oC.

• The pressure tube is thermally insulated from the hot coolant by an insulator.

• The coolant pressure of 25 MPa is transmitted to the pressure tube.

SCWR Pressure Tube


Transverse Strains from NRU Loop Tube

Assemblies (Operating Temperature ~285oC)

The annealed condition of Excel minimizes irradiation creep compared to Zr-2.5Nb.


The two necessary conditions for delayed hydride cracking initiation and propagation:

• Exceeding the Terminal Solid Solubility (TSS) for hydrogen,

• Having high tensile stresses (at flaw tips or due to residual stresses).

Best defence against delayed hydride cracking is for the pressure tubes to remain “hydride-free” during its lifetime.

Hydrogen pickup is a by-product of oxidation.

Delayed Hydride Cracking


Delayed Hydride Cracking Velocity

• DHCV is a strong function of temperature.

• DHCV is greater in annealed Excel than in Zr-2.5Nb.

• The high DHCV would cause rapid crack growth if initiated.

• It is expected that irradiation hardening of Excel would increase DHCV.

25010-6 150200

1000/Temperature (1/K)

HydrogenConcentration

29 mgH/kg29 mgH/kg34 mgH/kg

100 mgH/kg

50100

Temperature (C)

1.8 2.2 2.4 2.82.0 3.02.6

Scatter bandfor Zr-2.5Nb

10-8

10-9

10-10

10-7

oC


Single Channel Replacement

• The bellows and fuel channel extension are removed by cutting the weld above the pressure tube, and removing this through the outlet header.

• The pressure tube maintains the boundary between the moderator and coolant.


Single Channel Replacement

• A sleeve is inserted through the opening in the outlet header to the top of the tubesheet. The sleeve extends upwards to a point above the liquid level.

• A seal is used at the tubesheet that isolates the pressure tube that will be replaced

• Water is pumped out of the channel

• Moderator water level is lowered

• PT is cut and removed, new PT is inserted and welded


Outline

Introduction


Features





Summary


Evolution of Fuel Bundle/Assembly


• Subchannel code (ASSERT) to establish the cladding temperature distributions at normal operating conditions Fully qualified for bundle analyses at sub-critical pressures

Supercritical water properties and heat-transfer correlations implemented

• Computational fluid dynamic (CFD) tool (Star-CCM+) to quantify the effect of wire-wrapped spacers Supercritical water properties implemented Sensitivity of turbulence models and mesh generation assessed

Analytical Tools and Models


Description

Fluid Water at supercritical conditions

Gravity term Include gravity terms

Wall friction Colebrook-White formula for turbulent friction.

Inter-subchannel turbulent mixing

The Rogers and Tahir correlation multiplied by 2 for tight-lattice rod arrays.

Element-to-coolant heat transfer

The Jackson correlation for supercritical pressure conditions.

Appendages/Spacers Bare bundle, no appendages or wire wrap modelled.

Key ASSERT Subchannel Code Models


Assessment of ASSERT Subchannel Code

Against Supercritical Water Bundle Test Data

2 °C underpredictions at sub-critical temp.

7 °C overpredictions at supercritical temp.

398

397

396

395

394

45

400

39390 135 180 225 270 315 3600

Angle (o)

401

399

0

#2

0

#1

P = 25 MPaG = 1000 kg/m2sq = 400 kW/m2

tb = 383.79 oC

#1 Sliding Thermocouples#2 Fixed ThermocouplesASSERT Predictions

484

482

480

478

476

474

472

45

488

47090 135 180 225 270 315 3600

Angle (o)

490

4860

#2

0

#1

P = 25 MPaG = 1000 kg/m2sq = 400 kW/m2

tb = 430.72 oC

#1 Sliding Thermocouples#2 Fixed ThermocouplesASSERT Predictions


Component Dimension

Centre Flow Tube 4.7 cm outer radius, 0.1 cm thickness

Inner Pins (32) 0.475 cm outer radius, 5.40 cm pitch circle radius

Outer Pins (32) 0.5 cm outer radius, 6.57 cm pitch circle radius

Cladding 0.06 cm thick

Liner Tube 7.20 cm inner radius, 0.05 cm thick

Insulator 7.25 cm inner radius, 0.55 cm thick

Outer Liner 7.80 cm inner radius, 0.05 cm thick

Pressure Tube 7.85 cm inner radius, 1.2 cm thick

Fuel Bundle Active Length 500 cm

Fuel Bundle Geometry


Parameters Beginning of Cycle

(BOC) End of Cycle

(EOC)

Outlet Pressure (MPa) 25 25

Mass Flow Rate (kg/s) 5.13 4.73

Inlet Temperature (°C) 350 350

Power (MW) 9.97 9.21

Flow Conditions and Powers


• Bundle geometry is maintained (no distortion or variation)

• No change in cladding surface conditions • No wire-wrapped spacers and grid spacers at two ends • Negligible heat transfer to the downward inlet flow through the centre tube and to the moderator through the insulator and pressure tube

• Negligible conduction heat transfer through the fuel cladding between subchannels

• Uniform flow conditions at the entrance of the fuel bundle

• Deterioration heat transfer regime is not present in bundle with wire-wrapped spacers.

Key Modelling Assumptions


Axial and Radial Power Profiles Applied in the

Assessment


Axial and Radial Heat Flux Profiles


Coolant Temperature Distributions at Channel

Outlet

(Beginning Of Cycle) (End Of Cycle)

Outer Subchannel Middle Subchannel Inner Subchannel


Coolant-Temperature Variations in

Subchannels


Variations of Mass Flow Rate in Subchannels


Cladding-Temperature Variations in

Subchannels

808 °C


• Subchannel analysis focused on bare bundle only (i.e., no spacers)

• Wire-wrapped spacers have been incorporated to maintain gap sizes between rods and minimize vibration

• Previous studies demonstrated enhancement effect of wire-wrapped spacers on heat transfer leading to cladding temperature reduction in tubes, annuli and bundle subassembly Current subchannel analysis would

overpredict the cladding temperature (hence conservative)

• Confirmatory analysis for SCWR fuel assembly using a CFD tool

Effect of Wire-Wrapped Spacers


•Low Reynolds y+ k-ω turbulence model

The first node point set to a corresponding wall y+ value of ~1

•Variable fluid properties

Star-CCM+ CFD Tool Models


Wire-Wrapped Spacer Effect for SCWR Fuel

Pressure: 25 MPa Inlet fluid velocity: 3.08 m/s Inlet fluid temperature: 384.53 °C Heat flux on inner, outer rings: 0.96, 0.91 MW/m2


• Power profiles for a device free core without burnable neutron absorber (BNA) were applied in the analysis

• Power profiles for a core with BNA and adjuster rods are different Local powers at the peak

temperature locations are lower Peak cladding temperatures are

anticipated to be lower Current predicted peak cladding

temperature is conservative

Changes in Axial Power Profiles

Core with BNA and Adjuster Rods (BOC)Core with BNA and Adjuster Rods (EOC)


Outline

Introduction


Features





Summary



Safety advances are made possible through:

Adoption of ambitious safety objectives to drive research

Continued emphasis on principle of “defense-in-depth”

Systematic application of an integrated safety approach,

both deterministic and probabilistic, to ensure that safety

is “built in” rather than “added on”


Integrated Safety Assessment Methodology

Five practical and flexible tools to identify

vulnerabilities and relative contributions to risk

commensurate with design maturity


Completed QSR , PIRT, and PSA to assess the

Risk & Safety of CSCWR – the three most appropriate tools

applicable to a Conceptual Design

Integrated Safety Assessment Methodology


Qualitative Safety Features Review

A relatively new tool to shape the work of the designers ensuring that safety is “built-in”, not “added on to”

Five levels of defence-in-depth provisions assessed The assessment of Level 1 (i.e., prevention) features resulted in:

97 “Favorable” 41 “Neutral” 8 “irrelevant” 2 “unfavourable”, and 101 “to be assessed” when the design matures.


A Sample QSR Assessment Table

Index Content of Recommendation QSR Team Comments Qualitative

Assessment

1.1

Work out and set up a design

for the plant (i.e., the reactor

core, primary circuits and

balance-of-plant (BOP), that will

allow simple procedures for

the reactor operations and

maintenance during normal

conditions (i.e., minimize

process complexity and avoid

inherent instability; systematic

consideration of human factors

and the human–machine

interface for operation and shut

down).

Majority of the reactor core transport system and feedwater

circuit calculation for concept design is complete (Matt). Favourable

Plant energy balance has been completed (Matt) Favourable

Feedwater pump selection completed (Qingwu) Favourable

BOP equipment similar to coal-fired plants (Canadian

SCWR concept uses a live steam pressure and

temperature slightly higher than the pressure and

temperature currently used in conventional coal-fired power

plants).

Favourable

Human-Machine Interface (HMI) is assumed to be similar to

EC6.

Favourable

Vijay C. to provide stability map. Further CATHENA

analyses will be done in the next phase.

Favourable

Through documented monthly review meetings process

complexity has been minimized where possible. (Metin

paper –evolution of design)

Favourable

Technical Recommendations and Foreseen Characteristics and Features Applicable to

Canadian SCWR Concept (DiD Level 1: Prevention)


Phenomena Identification & Ranking Tables

Applied initially in pre-conceptual design phase,

and iteratively thereafter

Used as an early “screen” to identify,

categorize, and characterize phenomena that

are potentially important to risk and safety

Relied heavily on expert elicitation

Formed input to PSA and identified areas in

which additional research is needed


Phenomena Identification & Ranking Tables

A PIRT review of LBLOCA in Canadian SCWR

Concept was completed in 2014

Nine national and international experts

participated

Two Figures-of-Merit were used:

Fuel Cladding Temperature, and

Containment pressure

428 phenomena in the SCWR systems and

components were assessed


Knowledge Gap Identification Using PIRT

Status of

Knowledge

Rank of Importance

H M L I

4 3 2 1 10

3 67 37 4 242

2 25

(gap)

1

(gap) 8 21

1 4

(gap)

0

(gap)

0

(gap) 3


A Sample of Identified Knowledge Gaps

Four phenomena that have high importance

and very limited knowledge where uncertainty

cannot be characterized are:

Counter Current Flooding Limit in the central flow tube

of the fuel assembly

Material degradation of ceramic insulator up to cladding

failure

Material degradation of ceramic insulator after cladding

failure to containment failure

Cracking/embrittlement of ceramic insulator


Probabilistic Safety Assessment of Canadian

SCWR Concept

A preliminary, simplified, Probabilistic Safety

Assessment (PSA) was performed on the safety

systems using Computer Aided Fault Tree

Analysis tool

Three analysis scenarios:

Small-break loss-of-coolant accident

Large-break loss-of-coolant accident

Loss-of-Class-IV power event


Large-Break LOCA Fault Tree Analysis


Conclusions from PSA of Canadian SCWR

The probability of core damage for these three postulated

accident scenarios is at least one order of magnitude

lower compared to other reactor systems

Outcome

Postulated Accident Scenario

Small-Break LOCA

Large-Break LOCA

Loss of Class-IV Power

No Core Damage 1.00 x 10-2 1.00 x 10-4 1.00 x 10-2

Limited Core Damage 1.00 x 10-6 2.10 x 10-8

Core Damage 4.06 x 10-9 4.06 x 10-11 1.34 x 10-10


Summary

The safety of Canadian Super Critical Water

Reactor concept has been assessed using

three tools from the GIF Integrated Safety

Assessment Methodology. The tools used

are: Qualitative Safety Features Review,

Phenomena Identification and Ranking Table, and

Probabilistic Safety Assessment.


Summary (Cont’d)

The assessments performed using Qualitative Safety

Features and Phenomena Identification and Ranking

Tables indicated that except for a small percent of features

and phenomena, the Canadian Super Critical Water

Reactor Concept has acquired adequate knowledge to

progress towards the next stage of design

The Probabilistic Safety Assessment completed using three

postulated accident sequences concluded that the

probability of core damage is at least an order of magnitude

lower than other reactor systems


Thank you

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