1. Safety of Nuclear Reactors
Professor H. MOCHIZUKI
Research Institute of Nuclear Engineering, University of Fukui

2. Accident (1/2) Design Basis Accident: DBA
Assumption of simultaneous double ended break
Installation of Engineered Safety Features Emergency Core Cooling System: ECCS Accumulated Pressurized Coolant Injection System: APCI Low Pressure Coolant Injection System: LPCI High Pressure Coolant Injection System: HPCI

3. Accident (2/2)
Computer codes are used to evaluate temperature behavior of fuel bundle.
Computer codes should be validated.
Blow-down and ECC injection tests have been conducted using mock-ups.
RELAP5/mod3 and TRAC code are developed and validated.

4. ECCS Main System Diagram of Fugen Turbine Control Rod Drive
Relief valve
Containment Air Cooling System
Sea water
Feed Water System
Residual Heat Removal System (RHR)
Dump valve
Shield Cooling System
High Pressure Coolant Injection System (HPCI)
Low Pressure Coolant Injection System (LPCI)
(APCI)
Bypass valve
Heavy Water Cooling
System
Containment Spray
System
Condensate Tank
Reactor Auxiliary
Component Cooling
Water System
Reactor Core Isolation Cooling System
(RCIC)
Reactor Auxiliary
Component Cooling
Sea-Water System
Main System Diagram of Fugen

5. Blow-down experiment

6. ６MW ATR Safety Experimental Facility

7. Water level behavior after a main steam pipe break

8. Simulated fuel bundle
Local peaking is high for the outer rods due to the neutronic characteristics
Location of maximum axial peaking

9. Thermocouple positions

10. Cladding temperature measured in a same cross section of heater bundle

11. Calculation model of pipe break experiment

12. Comparison between experimental result and simulation

13. Improvement of blow-down analysis by applying statistical method
Downcomer 100 mm break
Scram
ECCS operation

14. Downcomer 150 mm break
Temperature (℃)
Time (sec)

15. Severe accident

16. Heat transfer of melted fuel to material

17. Heat transfer between melted jet and materials

18. Fuel melt experiment using BTF in Canada

19. Fuel melt experiment using CABRI

20. Source term analysis codes
General codes
NRC codes
ORIGEN-2, MARCH-2, MERGE, CORSOR, TRAP-MELT, CORCON, VANESA, NAUA-4, SPARC, ICEDF
IDCOR codes
MAAP, FPRAT, RETAIN
NRC code (2nd Gen.)
MELCOR
Precise analysis codes
Core melt
SCDAP, ELOCA, MELPROG, SIMMER
Debris-concrete reaction
CORCON
Hydrogen burning
HECTOR, CSQ Sandia, HMS BURN
FP discharge
FASTGRASS, VICTORIA
FP behavior in heat transport system
TRAP-MELT
FP discharge during debris-concrete reaction
VANESA
FP behavior in containment
CONTAIN, NAUA, QUICK, MAROS, CORRAL-II

21. In case of containment bypass
CONATIN code
(13)
Air
Containment spray
In case of containment bypass
Containment
recirculation
system
Air
(14)
(11)
(12)
Annulus
Stack
(10)
(9)
(7)
(8)
Filter Blower
(6)
Water flow
(5)
Gas flow
(4)
(3)
(2)
(1)
Steam release pool

22. Fluid- structure interaction analysis during hydrogen detonation

23. Analysis of Chernobyl Accident - Investigation of Root Cause -

24. Schematic of Chernobyl NPP
1. Core
2. Fuel channels
3. Outlet pipes
4. Drum separator
5. Steam header
6. Downcomers
7. MCP
8. Distribution group headers
9. Inlet pipes
10. Fuel failure detection equipment
11. Top shield
12. Side shield
13. Bottom shield
14. Spent fuel storage
15. Fuel reload machine
16. Crane
Electrical power ,000 MW
Thermal power ,200 MW
Coolant flow rate ,500 t/h
Steam flow rate ,400 t/h (Turbine)
Steam flow rate t/h (Reheater) Pressure in DS MPa
Inlet coolant temp C
Outlet coolant temp C
Fuel %UO2
Number of fuel channels ,693

25. Elevation Plan

26. Above the Core of Ignarina NPP

27. Core and Re-fueling Machine

28. Control Room

29. Configuration of inlet valve

30. Drum Separator

31. Configuration of Fuel Channnel

32. Heat Removal by Moderation
Pressure tube
Graphite ring
Maximum graphite temperature is 720℃ at rated power
φ91mm
Heat generated in graphite blocks is removed by coolant
φ114mm
φ88mm
φ111mm
Graphite blocks
Coolant
Gap of 1.5mm

33. RBMK & VVER
Finland
Russia
Lithuania
Germany
Ukraine

34. Objective of the Experiment
Power generation after the reactor scram for several tens of seconds in order to supply power to main components.
There is enough amount of vapor in drum separators to generate electricity.
But they closed the isolation valve.
They tried to generate power by the inertia of the turbine system.

35. Report in Dec. 1986

36. Trend of the Reactor Power
Power excursion
Thermal Power (MW)
20-30% of rated power
Scheduled power level for experiment
200MW
30MW
sec
min
hour
day

37. Time Chart Presented by USSR

38. Result in the Past Analysis (1/2)
T. Wakabayashi, H. Mochizuki, et al., Analysis of the Chernobyl Reactor Accident (I) Nuclear and Thermal Hydraulic Characteristics and Follow-up Calculation of the Accident, J. Atomic Energy Society of Japan, 28, 12 (1986), pp
T. Wakabayashi, H. Mochizuki, et al., Analysis of the Chernobyl Reactor Accident (I) Nuclear and Thermal Hydraulic Characteristics and Follow-up Calculation of the Accident, Nuclear Engineering and Design, 103, (1987), pp
Requirement from the Nuclear Safety Committee in Japan
Recirculation flow rate
Drum pressure
Water level
Feed water
Neutron flux

39. Result in the Past Analysis (2/2)
Power at 48,000 MW
Timing of peak was different. Why???
Power just before the accident was twice as large as the report. Why???
Power at 200 MW
？？？
Result of FATRAC
code is transferred, and initial steady calculation was conducted.

40. Possible Trigger of the Accident
Positive scram due to flaw of scram rods
Pump cavitation
Pump coast-down

41. Calculation Model by NETFLOW++ Code

42. Trigger of the Accident
• Positive scram
P.S.W. Chan and A.R. Daster
Nuclear Science and Engineering,
103, (1989).
Andriushchenko, N.N. et al., Simulation of reactivity and neutron fields change, Int. Conf. of Nuclear Accident and the Future of Energy, Paris, France, (1991).

43. Trigger of the Accident (cont.)
Scram rod (24rods)
inserted by AZ-5 button
1.0
Negative reactivity
8.0
5.0
Graphite block
1.5
Positive reactivity
Graphite displacer
Fuel 2×3.5m
Water Column

44. Simulation from 1:19:00 to First Peak
Data acquired by SKALA

45. Behavior of Steam Quality
Turbine trip
RCP　trip
Two-phase
Water
Push AZ-5 button

46. Void Characteristic Da 2Void fraction increase

47. Nuclear Characteristics
Doppler
Void

48. Peak Power and its Reactivity

49. Relationship between Peak Power and Peak Positive Reactivity

50. Just after the Accident

51. Control Room and Corium beneath the Core