sfr core design performance & safety - vasile · pdf fileoutline gen iv specifications...

SFR CORE DESIGN PERFORMANCE

AND SAFETY

19 SEPTEMBER 201213 SEPTEMBRE 2012 | PAGE 1CEA | 10 AVRIL 2012

A. VASILE

European Nuclear Education Network Association – Gen IV - INSTN | Alfredo Vasile

OUTLINE

GEN IV specifications

Overview of the Design Methodology

SFR core design: example of results, key issues

Application to ASTRID

Conclusions

13 SEPTEMBRE 2012 | PAGE 2CEA | 19 SEPTEMBER 2012

GEN IV CRITERIA


GOALS FOR GEN IV FUEL CYCLES


Optimization of the use of natural resources

Waste minimization

Proliferation resistance

Recycling options:All paths should be kept available, they could be used in a sequence.

R T

DepU

FP

U Pu MA

R T

DepU

MA

U Pu

FPR T

DepU

FP MA

U Pu

Homogeneous MAsrecycling (GenIV)

Heterogeneous

MAs recyclingU & Pu

recycling

SFR CORE LAYOUT


SFR CORE SUBASSEMBLIES


Central sub-assembliesFuel sub-assemblies (containing the fissile material),

Blanket sub-assemblies & reflector sub-assemblies, Sub-assemblies supported by the diagrid,

Peripheral sub-assemblies Lateral neutron shielding,

Sub-assemblies on a dummy diagrid,

Control rodsFunction: optimize the neutron balance,

Material: enriched boron with absorbing10B,

Spread out in the fissile zone.

SFR CORE DESIGN REQUIREMENTS


Main core functions:Producing & controlling the power level,Ensuring the fissile material filling density, Guaranteeing & controlling the temperature level, Stabilising and maintaining overall core consistency.

Design issues related to these functions: Neutronics

Power density, safety parameters

Thermal-hydraulicsCooling and monitoring,

Thermomechanics (& materials)Stability and consistency,

Fuel physicochemistryTemperature stability,

InstrumentationMonitoring.

SFR CORE DESIGN PROCESS


Pre-design

Design

Core definition

S/A Design



Neutronics

Pre-design

Neutronics Neutronics

Thermo Hydraulic

Mechanic

Fuel Behavior

Materials

Simplified modeling

Preliminary physical studies

Pre design studies Design studies

Feasibility domain

Severe accidents

Detailed modeling

Way of interest

S/A Pre design

Physical behavior

assessmentsCore detailed

image



1st phase: parametric study

Neutronic assessment

Study of the physical phenomena

Most important parameters to improveSafety (Doppler, Void,…)

Performances (BU,..)

2nd phase: images

Compilation of most promising options from the 1st phase

Core images focusing on:

safety, looking for cores with limited void effect

self sustainability (BG, Pu Inventory)

Economic performances (Cycle length, BU,..)

Neutronic

Elementary physical studies

Neutronic

Pre-design

Pre design studies

SUBASSEMBLY DESIGN PROCESS


NeutronicsSpatial Power Distribution within the S/A

and neutronic performances

Fuel ElementFuel and Clad Temperatures

Pressurization Elementary Design

S/A Geometrical DesignNew Volume Fractions

Thermal hydraulic behavior

criteria S/A candidateYESNO

The S/A design need a multi-discipline process :

Neutronic: irradiated fuel characteristics, neutronic performances (depletion, power distribution…)

Mechanics: pressurization

Thermal hydraulics: fuel and pin temperatures

Iterative design process involving the multi-discipline criteria

New methodologies are being developed using multicriteria optimization approach

NEUTRONIC PRE DESIGN STUDIES


Homogeneous cells calculations without self

shielding

2D RZ core geometry, homogeneous compositions

33 group for flux calculation in diffusion approximation

• Pu content, BG , mass of HN

• Cycle length, BU calculations

• DPA, Max and average power …

• Core safety parameters :

Void effect,

Doppler,

Delayed Neutrons fraction.

Fast implementation and reduced time calculations

Assessment of global parameters

Calculations are performed without uncertainties at this stage

CONCEPTUAL DESIGN


Detailed core study (from a pre-design core definition)

Detailed core description

Implementation of control rods and backup systems

Overall optimization

Transfer data to other codes

Neutronics

Thermal hydraulics

mechanicsFuel behavior

Transients & Severe AccidentsSystem

Neutronics Reference Code (MC)

SFR CORE DESIGN STUDIES


Heterogeneous cells calculations Pin and S/A description with fine self

shielding treatment

Detailed core description (3D Hex Z)

Loading Batch Management

33 group for flux calculation in transport theory

Global performances

Reaction rates distribution per S/A and meshes

Feedback coefficients per S/A and meshes

Individual control rod worth and detailed monitoring of safety criteria (10$,cold and hot shutdown, handling error, ,..)

Most accurate calculations and time ratio

Specific calculation scheme

Detailed characterization

Uncertainties depends on the qualification domain

DESIGN METHODOLOGY APPLIED TO SFR SYSTEMS

| PAGE 15

ICAPP'12, Chicago, USA | 24,28 june 2012

EXEMPLE OF PARAMETRIC STUDIES


Ways to improve the breeding gain Ways to improve the Void/Doppler ratio Geometry modifications

Increase of the fuel volume fraction Decrease of the sodium volume fraction Increase of the height / diameter ratio (0<H/D<1) Decrease of the core volume

Increase of the core volume Decrease of the height / diameter ratio (0<H/D<1)

Increase of the sodium volume fraction Increase of the fuel volume fraction

Increase of the core radius Modification of the core radius

« options » addition Minor actinides addition Sodium plenum addition Sodium plenum addition Addition of CaH2 moderator Annular core Addition of B4C moderator Addition of B4C moderator Addition of an internal fertile axial blanket Addition of an internal fertile axial blanket Annular core

Addition of CaH2 moderator Minor actinides addition

Fuel type Nitride Carbide Carbide Oxide Metallic Nitride

Oxide Metallic

Large improvement Large deterioration

No Effect



Key parameters Safety oriented optimization

Major expected benefits

Loss of reactivity per cycle

Decrease• Decrease Control Rods efficiency so as to reduce

consequences of a sudden withdrawal of rod (s) accident (UTOP)

Coolant void effect Decrease

• To minimize energy release potentially leading to severe accident

• Favorable feedback in ULOF transient• Safety margin in case of gas ingress in the core

DopplerOptimizationDecrease or increase

• Favorable feedback in ULOF• Reactivity insertion accident (sudden rod withdrawal) (reactivity

insertion)

Core pressure drop Decrease• To favor natural convection• Favorable natural behavior in case of ULOF (LIPOSO)

Pu enrichment Decrease• To limit consequences of core meltdown accident ?• Fusion limited to few fuel assemblies (7 being the criteria in the

past)

Detection

+ early detection system and high performance core monitoring

• Fast delayed n detection in case of instantaneous total blockage (IBT)

• Acoustic detection of local boiling (IBT)• Enhanced detection system for Sudden rod (s) withdrawal

accident• Enhanced detection of fuel handling mistakes

SFR / PWR


Reactor PWR SFR

Thermal power (MW) 4250 3600

Electrical power (MW) 1450 1450

Fuel UOX (~4%), MOX (~8%) MOX (~20%)

Coolant water sodium

Primary pressure (b) 155 1

Cladding diameter (mm) 9.4 8.2

Bundle geometry square hexagonal

Pins per S/A 264 331

Fissile hight (cm) 420 100

Number of S/A in the core 241 387

Core volume (m3) 46 12

Uranium, Plutonium mass (t) 130 42

Core inlet temperature (°C) 300 400

Core ∆T (°C) 40 150

Power density (W/cm3) 90 300

Reactivity control Control rods (~80), Soluble Boron Control rods (~30)

SFR S/A

PWR S/A

SFR SUBASSEMBLY


Wrapper tube

Fuel bundle Sodium inlet

Axial blanket bundle

PHENIX, SUPER PHENIX AND EFR


Reactor PHENIX SUPER PHENIX EFR

Thermal power (MW) 563 2990 3600

Pellet diameter (mm) 5.507.14

(central hole 2 mm)6.94

(central hole 2 mm)

Cladding diameter (mm) 5.65 8.5 8.20

Pins per S/A 217 271 331

Fissile hight (cm) 85 100 100

Blanket zone hight (radial/upper/lower) (cm) 52/22/30 60/30/30 40/15/25

S/A width across flats (cm) 12.37 17.3 18.3

S/A pitch (cm) 12.72 17.9 18.8

Equivalent core fissile radius (cm) 68 179 194

Core fissile volume (m3) 1.2 10 12

Number of fuel S/AsZone Enrichment 1 Zone Enrichment 2 Zone Enrichment 3

1045054-

364193191

-

38720710872

Number of control rods 6 21 24

Number of safety control rods 1 6 9

Number of blanket S/As78

(2 rings)237

(3 rings)78

(1 rings)


REACTIVITY EFFECTS - INTRINSIC

.comb

.combDOPDOP T

dTKd α=ρ


REACTIVITY EFFECTS - LOCAL

2) LOCAL EXPANSION

Sodium densityCladding and wrapper tube radialCladding and wrapper tube axialFuel axial (free or linked to the cladding)

Remarks:

No radial fuel expansion effect because fuel radial expansion is limitedby the cladding

When the fuel is linked to the cladding the axial expansion is driven by the cladding axial expansion.

iii dT.kd =ρ


REACTIVITY EFFECTS - GLOBAL


Type of effect Effect Related to FormulationValue (SPX)

pcm / °C

Intrinsec Doppler T fuel (��) �ρ ��

��

. �� -0.08

Local

Sodium density T sodium ��) �ρ � �. �� +0.33

Cladding radial T cladding ��) �ρ � �� . �� +0.08

Cladding axial T cladding ��) �ρ � ��. �� +0.07

Wrapper tube radial T wrapper tube (�� ρ � �� . �� +0.01

Wrapper tube axial T wrapper tube (�� ρ � ��. �� +0.04

Fuel axial (linked fuel) T cladding ��) �ρ � ��. �� -0.31

Fuel axial (free fuel) T fuel (��) �ρ � ��. �� -0.22

Global

Diagrid radial T diagrid (��)�ρ

� ��. ��-1.01

Relative expansion core/vessel/control rods

T core, T control rods,T vessel

�ρ � ��. �� 12mm/°C

REACTIVITY EFFECTS


LOSS OF SECONDARY FLOW TRANSIENT

Primary flow Secondary flow

Core power

Sodium and fuel temperatures


LOSS OF SECONDARY FLOW TRANSIENT

Core inlet temp. ��, ρDiagrid < 0, P ��, ρDOPPLER > 0,ρSodium > 0,ρFuel < 0,ρCore/Vessel/Control rods> 0

Core power could be stabilised but isothermal sodium temperature >1000°C)

Reactivity

Doppler

Sodium

Diagrid

Core/Vessel/CR


ADVANCED SODIUM TECHNOLOGICAL REACTOR FOR INDUSTRIAL DEMONSTRATION

Industrial prototype (step before a First Of A Kind)

Integrating French and international SFRs feedback

A Generation IV reactor

Safety :Level at least equivalent to GENIII systemsWith significant improvements on sodium specific issues

Operability :Validate on the long term an ambitious load factorSignificant improvements concerning ISI&R

Ultimate waste transmutation :Continue experimentation of minor actinides transmutation, up to large scales if so decided

An investment cost under control

Irradiation services and testing of longer term options


MAIN FEATURES

S/A : UPuO2 fuel (annular pellets)Fuel pins with spacerHexagonal wrapper tube

General performances objectives for the coreAverage burn-up > 100 GWd/tBreak-even core : breeding gain ≈ 0 (without fertile radial blanketFuel cycle length : 400- 490 efpdTransmutation capabilities for MA transmutationSafety improvement

PP

DHX

IHX PP

REDAN

CORE

Pool type reactor : 1500 MWthSodium cooled3 Primary pumps, 4 IHX


SAFETY PERFORMANCE TARGET

Favorable natural behavior during loss of Flow transients

Target criteria : no sodium boiling for a ULOSSP transient(Unprotected Loss Of Station Supply Power)

Sodium void effect minimizedTarget criteria : Na void effect < 0

Natural behavior favorable for a complete control rod withdrawal (without detection) Target criteria : no fuel fusion

Improvement of behavior in case of sub-assembly Total Instantaneous BlockageEnsure sub -criticality of 7 melted adjacent sub-assemblies

Elimination of Transient of PowerCompaction or gas flow are " practically eliminated" by design.

CFV CONCEPT

| PAGE 30



CFV CORE LAYOUT


CFV CORE LAYOUT

Fissile zone

Blanket zone

Plenum zone

Absorber zone

Theses Features give a

negative Sodium void worth


CORE DESIGN OPTIONS: SFRV2 AND CFV LAYOUTS

SFRv2 – AIM1 – 1500 MWth CFV – V1 - AIM1 – 1500 MWth

Neutronic and thermal-hydraulic core calculations were performed with CEA’s reference system codes ERANOS and CATHARE.

CORE PERFORMANCES

| PAGE 34



CFV AND V2B PERFORMANCES 1/2

Core design CFV V2B 1500

Thermal Power 1500 MW

Elect. Power 600 MW

Fuel Residence Time 1440 EFPD 1560 EFPD

Fuel Cycle Length 360 EFPD 390 EFPD

∆ρ∆ρ∆ρ∆ρ (cycle) - 1550 pcm - 875 pcm

∆ρ∆ρ∆ρ∆ρ (efpd) - 4,3 pcm - 2,2 pcm

Batch 4

Fissile zone diameter 340 cm 312 cm

S/A Pitch 17,5 cm 16,8 cm

Nb fuel elements C1/C2 177 / 114 144 / 144

Pin diameter 9,7 mm 10,73 mm

Pins/Assembly 217 169

CSD/DSD Nb 12 / 6 18 / 6

High Cycle length, positive

to competitiveness

Weak reactivity loss, favorable to limit CR withdrawal


CFV AND V2B PERFORMANCES 2/2


Pvol (Fiss.+ int. fertile) 226 W/cm3 194 W/cm3

Pvol (Fissile)) 258 W/cm3 194 W/cm3

Fuel burnup C1/C2 105/ 69 GWd/t 76 / 67 GWd/t

Pu enrichment C1/C2 23,5 / 20 % 13,9 / 17,6 %

DPA max 113 108

Void effect EOC -0,5 $ +5,1 $

ββββeff 364 pcm 373 pcm

Breeding Gain -0,02 -0,05

Pu inventory (HN) 4,9 t 5,3 t

Max linear power rate 483 W/cm 407 W/cm

Core pressure drop 2,6 b >3b

Total core flow rate 7990 ks/s

Inlet core temperature 400°C

Outlet core temperature 550°C

Criteria reached for CFV, good for safetyand public acceptance

Break Even Core, durability

Favorable for natural convection


CFV AND V2B FEEDBACK COEFFICIENTS

CFV SFRV2 V0

Axial clad expansion 0,061 0,064 pcm /°C

Radial clad expansion 0,089 0,147 pcm /°C

Sodium expansion 0,09 0,492 pcm /°C

Fuel expansion -0,23 -0,254 pcm /°C

Doppler Constant Fissile zone (KD) -664 -526 pcm

Doppler constant Fertile slab (KD) -355 - pcm

Plate expansion -0,88 -0,797 pcm /°C

Favorable Sodium worth

feedback coefficient for TH

transient

Favorable Doppler

constant, upgrade TOP

behavior

TRANSIENT BEHAVIOR

Thermal hydraulic transients

Control rod withdrawal

| PAGE 38



THERMAL HYDRAULIC TRANSIENTS

To investigate the core capabilities, 3 mains unprotected transients were assessed, in comparison to classical SFR core design.

ULOSSP (Unprotected Loss Of Station Supply Power) : blackout transient without scram and without starting up of ultimate emergency systems or decay heat removal

ULOF (Unprotected Loss of Flow): loss of primary pumps without scram, secondary pumps remaining at nominal flow.

ULOHS (Unprotected Lost Of Heat Sink): Secondary pumps tripped in 100s without scram, the primary pumps still remaining in normal operation

Hypothesis for comparison:Halving time for the Primary Pumps is 20sNo optimization of flow rate between S/A

All results presented are given for the hot S/A


CFV ULOSSP

Power

Na density

Inlet Temp

Outlet Temp

Boiling Temp

Doppler


CFV ULOSSP VS USUAL SFR DESIGN

Inlet Temp

Outlet Temp

Boiling Temp

Na density


CFV VS CLASSICAL SFR DESIGN (ULOHS)


SUMMARY ON THERMAL HYDRAULIC TRANSIENTS


ULOSSP 55°C of marging Na boiling ~100s

ULOF Na boiling ~3500s Na boiling ~100s

ULOHS Temp. of neutronic shutdown700°C

Temp. of neutronic shutdown800°C

These results are given without uncertainties and margin were evaluated for the hot S/A but no for the hot sub channel of the hot S/A


CONTROL ROD WITHDRAWAL

All the CRW that can occur on the CFV core can be detected by two devoted independent systems of core detection which stop the reactor by scram.The first system prompted is the core temperature monitoring; the second is the neutron detectionIn case of a total control rod withdrawal (corresponding to the failure of two strong lines of defense) the outer rods do not comply with the criterion of no melting

( ) [ ][ ])(.1.)(.'1 0 tbtkPtP ilinlin oρρ ∆+∆+=

Plin0, initial linear power of the considered fuel sub-assembly, k’i , relative variation of linear power density per unit of inserted reactivity,∆ρ total reactivity worth of the control rod between its initial position in the core and the parking position at the end of the withdrawal.b0 is the relative variation of the total power of the core per unit of reactivity inserted


CONCLUSION

Sodium cooled Fast Reactors core design is an iterative and integrated process.

It must accommodate different requirements mainly related to safety, fissile materials balance and economic aspects.

It includes neutronics, thermal-hydraulics, mecanics, fuel behaviourfor the core itself but also whole plant behavior under normal and accidental conditions.

Pre-conceptual design studies of ASTRID prototype core shows significant improvements related to previous SFRs designs (SPX, EFR).

Nuclear Energy DirectorateReactor Studies Department

Commissariat à l’énergie atomique et aux énergies alternatives

Centre de Cadarache | 13108 Saint Paul Lez Durance

T. +33 (0)442257000

Etablissement public à caractère industriel et commercial | RCS Paris B 775 685 01913 SEPTEMBRE 2012

| PAGE 46

CEA | 10 AVRIL 2012

Thank you for your attention

sfr core design performance & safety - vasile · pdf fileoutline gen iv specifications...

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