Research Program of a Super Fast Reactor

Yoshiaki OKA1), Yuki ISHIWATARI1), Jie LIU1), Takayuki TERAI1), Shinya NAGASAKI1), Yusa MUROYA1), Hiroaki ABE1), Hideo MORI2), Masato AKIBA3), Hajime AKIMOTO3), Keisuke OKUMURA3), Naoaki AKASAKA3),

and Shoji GOTO4)

1) Dep. Nuclear Engineering and Management/Nuclear Professional School, University of Tokyo

2) Department of Mechanical Engineering, Kyushu University3) Japan Atomic Energy Agency4) Tokyo Electric Power Company

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Revised from the presentation at ICAPP’06, June 7, 2006, RenoPresent study is the result of “Research and Development of the Super Fast Reactor” entrusted to The University of Tokyo by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).

Outline

• Background and purpose

• Overview of the program

‒ Development of the Super FR concept

‒ Heat transfer and thermal-hydraulic experiments

‒ Material developments

• Summary

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Generation IV Reactor Concepts

SCWR(Super FR / Super LWR)

SFR LFR

VHTR GFR MSR3

Features of Super FR / Super LWR• Simple & compact plant systems‒ No water/steam separation

‒ Low flow rate, high enthalpy coolant

• High temperature & thermal efficiency (500C, ～44%）

• Utilizations of current LWR and Supercritical FPP technologies‒ Major components are used within the temperature range of past experiences

Supercritical FPP（once-through boiler）

Super FR/ Super LWRBWR PWR

Recent Progress for Super LWR

• Concept of once-through direct cycle thermal reactor system with supercritical water has been established

• Excellent safety characteristics have been shown

‒ Effective core cooling by depressurization

‒ Mild behaviors during abnormal transients

‒ Excellent ATWS characteristics

• Major R&D subjects are revealed quantitatively

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Background and Purpose

1. Innovations of FR, cost reduction over thermal reactor2. Management and disposal of spent fuel3. Contributions to energy security and environment protections

• Compatible with tight fuel lattice without stability or pumping power issues• Higher power density than thermal reactor (Super LWR) due to no moderator necessary for FR⇒ Cost reduction

Cost: Super FR < Super LWR < LWR

• To pursue the potential of Super FR by developing the concept and conducting R&D of TH and materials

Subjects

Features of Super FR

Purpose

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Outline of the Research Program

① Development of the Super FR concept

② Heat transfer and thermal-hydraulic experiments

③ Material development

• Research program of Super FR was entrusted by MEXT as one of the Japanese NERI in Dec, 2005. Research period will be until March 2010.

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Development of the Super FR concept • Purpose: To establish Super fast reactor concept

• Design goals‒ Not breeding, but high power core using Pu from spent LWR fuel

‒ High thermal efficiency (High core outlet temperature)

‒ Negative void reactivity

‒ Good plant controllability and safety characteristics

‒ Simple plant systemContents① Fuel and core designs and structural considerations at the high temperature

② Method developments for evaluating nuclear transmutation ability

③ Backend risk analyses④ Preparation of a 3-D CFD code⑤ Reactor characteristics (control, safety) 8

Fuel and Core Designs

Seed FA

Fuel rodCR guide tube

Blanket FA

ZrH2 layer

1/6 Core (example*)

• MOX fuel with SS cladding (Fuel rod analysis)• Core design: 3-D core calculation, subchannel analysis• Development and verification of 3-D CFD analysis code (replacement for large scale experiments)

• Environmental impact analysis by coupling transmutation and backend analyses

9* Cao, L., Oka, Y., Ishiwatari, Y., et al., JNST, 452,(2008),138

Core Structure and Plant Control and SafetyCR guide tube

CR guide tube

Seed

Blanket

InletOutlet

Upper dome

Lower plenum

RPV and the coolant flow

• Large temperature change‒ In-core and RPV structural deigns

• High power density, tight fuel lattice• Control and Safety‒ Control system designs, safety system designs‒ Abnormal transient and accident analyses‒ Applicability of MPS method for condensation phenomenon of supercritical fluids

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HT Correlation for Determining Design Limit• High accuracy is required for predicting max. clad T

⇒ fuel integrity and core outlet temperature• Currently available:‒ heat transfers in smooth circular tubes‒ Large uncertainties (Differences in MCST predictions～44℃)

• Required correlations for design:‒ Fuel bundle geometry including grid spacers‒ Improved accuracy in the high temperature region‒ Upward flow / downward flow

℃ BOC MOC EOC

Watts 637 638 647

Oka-Koshizuka 604 606 603

Bishop 627 629 635

Dittus- Boelter 617 618 616250 300 350 400 450 500 550

5000

10000

15000

20000

25000

30000

35000

40000

45000Hydraulic diameter: 4.17mm

mass flux: 937.9 kg/m2/s

heat flux: 562 kW/m2

Heat transfer coefficient (W/m2 K)

Bulk temperature (℃)

Watts Dittus-Boelter Bishop Oka-Koshizuka

High T region

MCST (Max. clad surface temperature) predictions by different HT correlations

Requirements for exp.and materials

Core Design (Super LWR)

Radial・axialflux factor

Local flux factor

EngineeringUncertainties

Nominal steady state Core average condition（Design target）

Nominal peak steady state condition

Max. peak steady stateCondition (thermal limit)

Statistical thermal design

LIMIT for designtransients

AbnormalTransients

Plant Safety analyses

MarginFailure LIMIT

3-D core calculations

Fuel rod analyses

MCST:740℃

Subchannel analyses

ΔT1=150℃

ΔT２=58℃

ΔT3=32℃

ΔT4=50℃

MCST:850℃

Ave. outlet:500℃

Startup (Super LWR)

8 10 12 14 16 18 20 22 240

100

200

300

400

500

600

700

800

0

10

20

30

40

50

60

70

80

90

100

Inlet flow rate

Core power

Inlet temperature

Average outlet temperature

MCST

Temperature [℃]

Pressure [MPa]

Ratio (%)

BT

• Sliding pressure startup system (nuclear heating starts at subcritical pressure)

• Clad temperature increase in pressurization phase is due to BT• Power / flow region is limited by CHF• CHF may be increased by grid spacers

8 10 12 14 16 18 20 22 240

5

10

15

20

25

30

35

40

Operable region during pressurization phase

Flow rate = 35%

Inlet temperature = 280oC

Max. allowable power Min. required power

Power [％]

Pressure [MPa]

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Heat Transfer and Thermal-Hydraulic Experiments

• To develop basic heat transfer and thermal-hydraulic database (correlations) for design

‒ Effects of fuel bundle geometry and grid spacers on heat transfer

‒ Heat transfers in both steady state and transient conditions (supercritical pressure)

‒ CHF and effects of grid spacers on CHF during startup conditions (subcritical pressure)

‒ Critical flow (in case of LOCA) and condensation (to suppression pool) characteristics

Objective

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Contents of the Experiments• Single tube test

‒ Obtain basic data

‒ Development of heat transfer correlation

• Fuel bundle test

‒ Effects of grid spacers

‒ Steady state and transient tests

‒ Flow directions upward / downward

• CHF test (subcritical pressure)

‒ Obtain CHF data near the critical pressure

‒ Development of HT correlation

• Critical flow and condensation tests

‒ Obtain basic data for supercritical fluids

Single tube

ID:4.4mm

Seven-rod bundle

Heater pin (OD:8mm)×7 pins

Pressure vessel

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Kyushu University Test Loop(before Remodeling)

Pre-heaterTest loopTest section

Circulationpump

• Surrogate fluid: HCFC22‒ Critical pressure: 5 MPa‒ Critical temperature: 96 ℃

• Test conditions‒ Pressure: max. 5.5 MPa‒ Temperature: max. 160 ℃‒ Mass flux: max. 0.15kg/s

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JAEA Test loop

Overview of the system

3.5m

4m

• Working fluid: Supercritical water• Test conditions‒ Pressure: max. 25 MPa‒ Temperature: max. 400 ℃ (to be upgraded to 500 ℃)‒ Max. circulation mass flux: 1.3kg/sec

Circulation pump

Pump head: 70m (400℃)

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Materials R&D

• Development of fuel cladding material• Development of thermal insulator• Development of methodology for measuring dissolution of radioactive materials and corrosion products

Blanket FA

Seed FA

Contents

Long lived fuel rod(Development of cladding)

Maintaining coolant temperatureReduction of thermal stress

(Development of thermal insulator)

ZrH layer (For achieving negative void reactivity) 18

0 10 20 30 40 50 60-80

-60

-40

-20

0

20

40

60

80

Primary membrane stress [MPa]

Fuel rod ave. burnup [GWd/t]

Segment no. 8 Segment no. 9 Segment no. 10

Time to rupture [h]Creep rupture strength [MPa]

750℃

700℃

650℃

600℃

10

102

103

10 102 103 104 105

600℃650℃700℃750℃

PNC1520 PNC316

Need for Developing High Creep Strength Clad• Max. stress on clad at peak T (700-750℃): 70-100MPa

‒ Exceed creep strength of SS for LWR (SUS316L)

‒ Advanced SS for LMFBR (PNC1520) almost satisfies the requirement but SCC susceptibility, corrosion and neutron absorption properties need to be improved

• High creep strength clad needs to be developed for Super FR

Creep rupture strength of advanced SS

Fuel rod analysis results (Super LWR)

700-750℃

Development of Cladding Material

• Primary optimized test material

‒ To be developed based on PNC1520 experience

‒ Irradiation test by HIT of the Univ. of Tokyo

‒ Further equipments will be developed as required to test corrosion & irradiation effects

HIT (High Fluence Irradiation Facility, Univ. of Tokyo) 20

Development of Thermal Insulator• Large ΔT (～250℃) at the temperature boundary

• Thermal insulator is required for:‒ reduction of thermal stress

‒ maintaining moderator temperature

0 100 200 300 400 500 600 700

100

200

300

400

500

σ＞Su

(1/2×Su)＜σ＜Su

σ＜(1/2×Su)

Mid wall temperature [℃]

ΔT (℃)

Thermal stress on the wall

No thermal insulation

Thermally insulated

Max. thermal stress• Low heat conductivity

• Low neutron absorption

• Good thermal shock resistance

• Good dimensional stability

RequirementsInsulated No insulation

Stainless steel

HotCold

T HotCold

T

• Survey and tests of candidate materials (e.g., ZrO2)

Su: tensile strength

Summary

• Japanese research program has been launched to clarify the concept of the once-through direct cycle supercritical-pressure light water cooled fast reactor (Super FR) including the fuel and core designs

• The followings will be achieved‒ Development of the Super FR concept

‒ Establishment of the basic thermal-hydraulic database by experiments

‒ Development of the main in-core materials

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