1. Safety aspects of Indian advanced reactors
K.K. Vaze,
Director
Reactor Design and Development Group
Bhabha Atomic Research Centre,
Trombay, Mumbai India 1

2. Post Fukushima Scenario

3. Fukushima Accident
On March 11th, 2011, a gigantic earthquake with a magnitude 9 on the Richter scale shook Japan. The earthquake triggered a tsunami, which was exceptionally high, reached the Fukushima coast about one hour after the earthquake.
All reactors in operation at Fukushima shut down automatically. While the offsite external power source was lost due to the earthquake, emergency diesel generators (EDG) started up properly
Even though the earthquake was of a magnitude far greater than anticipated, there is today no evidence that it produced mechanical or structural damage which would have, in the absence of the tsunami, caused a severe accident. The seismic response analysis and the visual investigations conducted so far did not seem to show major damage to safety-related equipment.

4. Fukushima Accident - contd
The majority of the damage was caused by the tsunami. At Fukushima Daiichi it caused complete loss of AC power, loss of ultimate heat sink and serious degradation of DC power sources. This led to the loss of decay heat removal at three NPP units, to severe reactor core damage, to the loss of containment integrity and to significant radioactive releases to the environment. In addition, the upper part of the fourth unit reactor building was destroyed by hydrogen explosion and the spent fuel pool structures of that unit suffered mechanical damages.

5. Some reassuring thoughts as far India is concerned
Huge earthquakes and huge tsunamis are not commonplace

6. Comparative Seismic Hazard

7. Status of Seismicity – Indian NPPs
Criteria - No Active fault within 5 km
Site Seismic Zone
Narora IV
Rawatbhata II
Kakrapar III
Tarapur III
Jaitapur III
Kaiga III
Kalpakkam II
Kudankulam II

8. Tsunamigenic locations for Indian coast
TARAPUR
KALPAKKAM
ONLY FAR FIELD SOURCES
KUDANKULAM
TECTONIC PLATE BOUNDARIES
18 March 2011

9. How does this benefit us?
Fukushima
Earthquake knocked out Class 4 supply
Tsunami knocked out other supplies
India
EQ and tsunami don’t occur together
Ground motion due to an earthquake causing tsunami is negligible
Earthquakes causing significant ground motion do not cause tsunami
We get warning (~ 2 hrs)

10. Fukushima Accident Lessons Learnt
The key criterion of success:
- recovery of power supply
- water feed for the decay heat removal
As prompt as possible!
Availability of undamageable portable engineering means for power and water supply in the conditions of NPP isolation
Accident prevention and accident mitigation:
- implementation of design fundamental;
- emergency preparation;
- Severe Accident management.
Source: Prevention and Mitigation — Equal Priorities Prof. Vladimir Asmolov, WANO President

11. ACCIDENT MANAGEMENT GOAL
ACCIDENT MANAGEMENT MEASURES
To prevent the core melting
(To keep the integrity of the Ist and IInd physical barriers – Fuel & Clad)
The recovery of the core cooling
To retain melt inside the RPV
(To keep the integrity of the IIIrd physical barrier - RPV)
In-vessel cooling
Ex-vessel cooling
To prevent the containment failure
(To keep the integrity of the IVth physical barrier - Containment)
Core catcher
Hydrogen management
Filtered venting system
Source: Prevention and Mitigation — Equal Priorities Prof. Vladimir Asmolov, WANO President

12. Genesis for development for advanced reactors

13. Securing energy for India’s future is a major challenge
World OECD Non-OECD India India
(developing world) of our dream
Population
(billion)
(stabilised)
Annual av. per ~ ~ ~ ~
capita Electricity (kWh)
Annual
Electricity
Generation (trillion kWh)
Carbon-di-oxide
Emission ?
(billion tons/yr)
India alone would need around 40% of present global electricity generation to be added to reach average 5000 kWh per capita electricity generation
Dr. Kakodkar “Atoms for Prosperity

14. Global climate change is an immediate threat
Just ten years from now, greenhouse emissions from developing nations will equal the emissions from the countries we now call developed. After that, emissions from the developing world will be the major driver of global climate change.
While energy conservation, windmills, and solar panels may help, we cannot hope to rely on such measures alone to meet our world’s expanding appetite for more energy.
John Ritch, Director General of the World Nuclear Association, 15th Pacific Basin Nuclear Conference, Sydney, Oct. 2006
Comparison of sea-ice from 1979 and 2003.
1979
2003
Source:

15. Safety Goals for Advanced Reactors

16. CNS Extraordinary Meeting Summary Report
The displacement of people and the land contamination after the Fukushima Daiichi accident calls for all national regulators to identify provisions to prevent and mitigate the potential for severe accidents with off-site consequences.
Nuclear power plants should be designed, constructed and operated with the objectives of preventing accidents and, should an accident occur, mitigating its effects and avoiding off-site contamination.
The Contracting Parties also noted that regulatory authorities should ensure that these objectives are applied in order to identify and implement appropriate safety improvements at existing plants.

17. Dr. Kakodkar
An essential goal for nuclear safety is “Never Again” should there be any significant off site emergency
Dual level design basis
Design Basis
Risk Lowered to an acceptable level
No impact in public domain
Extreme Event
Maximum potential
No significant off-site emergency
Extra margin between design and ultimate load capacity should be sufficient to cope with this

18. Can the nuclear community set for itself an ambitious goal to meet the challenge of the numbers?
“Four decades from now, in any country of the world, it should be possible to start replacing fossil fuelled power plants, at the same urban or semi-urban site where these are located, with advanced NPPs that would, more economically, deliver at least twice the power that was being produced by the replaced plants”
R.K. Sinha, “The IAEA’s Contribution to the Peaceful Use of Nuclear Power”, Nuclear Power Newsletter, Vol. 3, No. 3, Special Issue, Sept. 2006

19. Number of reactors in operation
Level of safety goals increases with multi-fold increase in deployment of nuclear reactors
Special Siting Criteria, Risk approach
Special Siting Criteria (may/may not); CDF, LERF
Siting criteria
Dose Criteria
Safety
Goals
Advanced future Reactor Systems
Advanced reactors under construction
Reactors under operation (existing technology)
Number of reactors in operation 19

20. Monitored Process Parameter
Achievement of safety goals through enhanced levels of Defence-In-Depth
Strategy for safety measures and features of nuclear installations is two-fold:
To prevent accidents
Preventing the degradation of plant status and performance
If prevention fails, limit their potential consequences and prevent any evolution to further serious conditions
Monitored Process Parameter

21. Passive and Inherent Safety Features are Instrumental in Meeting New Safety Criteria
The conventional reactors or so called “Traditional ones” have seen an extensive use of “active” engineering safety systems for reactor control and protection in the past.
These systems have certain potential concerning termination of events or accidents that are effectively coped with by a protective system limited by the reliability of the active safety systems or prompt operator actions.
Since the reliability of active systems can not be improved above a threshold and that of the operator’s action is debatable, there is growing concern about the safety of such plants due to the large uncertainty involved in Probabilistic Safety Analysis (PSA) particularly in analyzing human faults.
In view of this, a desirable goal for the safety characteristics of an innovative reactor is that its primary defence against any serious accidents is achieved through its design features preventing the occurrence of such accidents without depending either on the operator’s action or the active systems.
That means, the plant can be designed with adequate passive and inherent safety features to provide protection for any event that may lead to a serious accident.
Such robustness in design contributes to a significant reduction in the conditional probability of severe accident scenarios arising out of initiating events of internal and external origin.

22. Example of Applications Passive Systems and Inherent Safety Features in Defence-In-Depth in AHWR

23. The Indian Advanced Heavy Water Reactor (AHWR-Pu)
AHWR is a 300 MWe vertical pressure tube type, boiling light water cooled and heavy water moderated reactor using 233U-Th MOX and Pu-Th MOX fuel.
Bottom Tie Plate
Top Tie Plate
Water Tube
Displacer
Rod
Fuel Pin
Major design objectives
65% of power from Th
Several passive features
7 days grace period
No radiological impact
Passive shutdown system to address insider threat scenarios.
Design life of 100 years.
Easily replaceable coolant channels.
Design validation through extensive experimental programme.
Pre-licensing safety appraisal by AERB
Site selection in progress.
Detailed engineering consultancy in progress
AHWR-Pu is a Technology demonstrator for the closed thorium fuel cycle
AHWR-LEU extends the AHWR technologies with LEU-Th MOX Fuel for the global market
AHWR Fuel assembly 23

24. No unacceptable radiological impact outside the plant boundary with
AHWR incorporates several technolological solutions to a higher level of safety and security against both internal and external threats
Control room and auxiliary systems
Pneumatic supply
Instrumentation & control signals
Electrical power
(Class 1 to 4)
Turbine
Pump
Condenser
Control and S/D systems
Core
External events
Malevolent act
Ultimate heat sink (Cooling tower or sea)
No unacceptable radiological impact outside the plant boundary with
Failure of all active systems, and
Failure of external infrastructure to provide coolant, power and other services, and
Malevolent acts by an insider, one of the consequences of which is the failure of instrumentation signal initiated shutdown actions, and
Inability of plant operators to manage the events and their consequences, for a significantly long time. 24

25. Some important passive safety features of AHWR –1/4
 Heat removal from core under both normal full power operating condition as well as shutdown condition is by natural circulation of coolant.

26. Some important passive safety features of AHWR –2/4
Passive Containment Cooling
(Th-Pu) MOX Fuel pins
(Th-233U) MOX Fuel pins
Central Tube for ECCS water
AHWR FUEL CLUSTER
Passive Containment isolation
Passive injection of cooling water, initially from accumulator and later from the overhead GDWP, directly into fuel cluster.

27. Some important passive safety features of AHWR –3/4
Passive Poison Injection in moderator during overpressure transient
Passive Poison Injection System actuates during very low probability event of failure of wired shutdown systems (SDS#1 & SDS#2) and non-availability of Main condenser

28. Some important passive safety features of AHWR –4/4
Use of moderator as heat sink
Water in calandria vault
Flooding of reactor cavity following LOCA

29. Fukushima and AHWR
AHWR has been assessed for TMI as well as Chernobyl type of accidents
Critics comments: It is easy to become wise after the event (TMI, Chernobyl)
Fukushima type event (Extended SBO) was anticipated even before it happened
Practically no change required in AHWR design to meet Fukushima event
GDWP and passive systems adequate to cater to the extended SBO
No impact in public domain, No need of evacuation
No need of exclusion zone, sterilized zone

30. Prolonged Station Black Out in AHWR Decay heat removal by Isolation Condensers
A strong earthquake with/without Tsunami causing prolonged SBO for several days. Reactor tripped on seismic signal.
Gravity Driven Water Pool is intact.
Heat is removed by Isolation Condensers
GDWP water removes decay heat for ~110 days with periodic containment venting allowed after 10 days.

31. Passive Systems in Defense-In-Depth of AHWR
Level 1 DID:
Elimination of the hazard of loss of coolant flow:
Heat removal from the core under both normal full power operating condition as well as shutdown condition is by natural circulation of coolant.
Reduction of the extent of overpower transient:
Slightly negative void co-efficient of reactivity.
Low core power density.
Negative fuel temperature coefficient of reactivity.
Low excess reactivity

32. Passive Systems in Defense-In-Depth of AHWR (Contd.)
Level 2: Control of abnormal operation and detection of failure
An increased reliability of the control system achieved with the use of high reliability digital control using advanced information technology.
Increased operator reliability achieved with the use of advanced displays and diagnostics using artificial intelligence and expert systems.
Large coolant inventory in the main coolant system.
Level 3: Control of accidents within the design basis
Increased reliability of the ECC system, achieved through passive injection of cooling water directly into a fuel cluster through four independent parallel trains.
Increased reliability of a shutdown, achieved by providing two independent shutdown systems. Further enhanced reliability of the shutdown, achieved by providing a passive shutdown device
Increased reliability of decay heat removal, achieved through a passive decay heat removal system, which transfers the decay heat to GDWP by natural circulation.
Large inventory of water inside the containment (about 8000 m3 of water in the GDWP) provides a prolonged core cooling meeting the requirement of grace period.

33. Passive Systems in Defense-In-Depth of AHWR (Contd.)
Level 4: Control of severe plant conditions, including prevention of accident progression and mitigation of consequences of severe accidents
Use of moderator as heat sink.
Presence of water in the calandria vault
Flooding of reactor cavity following a LOCA.
Level 5: Mitigation of radiological consequences of significant release of radioactive materials
The following features help in passively bringing down the containment pressure and eliminates any releases from the containment following a large break LOCA:
Double containment;
Passive containment isolation
Core catcher
Filtered vent

34. Peak Clad Temp v/s frequency of occurrence – a quantitative probabilistic safety criteria

35. Core Damage Frequency Per Year
Ref: Lecture on Near Term Advanced Nuclear Reactors and Related MIT Research, by Prof. Jacopo Buongiorno, MIT, USA, June 16, 2006.
AHWR
~ 1x10-8

36. Severe Accident Management

37. Incorporation of Hard vent
To Stack
From Containment
Hard Vent system is designed to prevent the over pressurization of the containment beyond design pressure occurring due to failure of multiple safety systems because of an extreme event such as prolonged SBO with non-availability of GDWP water or large seismic event causing cracks in GDWP along with LOCA.
Also retains the radio-activity in the scrubber and minimize activity release beyond the containment boundary.
Scrubber tank contains water + NaOH solution (ph = 8.5).
NaOH combines with Iodine whereas Cs which is in form of CsI, CsOH, CsO2, Cs2CO3 is soluble in water.
A 4 inch Dia pipe is provided at the top of primary containment for venting, which will be connected to scrubber tank.
3 I NaOH = 3 H2O +5 NaI + NaIO3 37

38. Passive Autocatalytic ReCombiner System (PARCS)
Postulated Accidents
DBA : Single failure (LB LOCA): No hydrogen generation
BDBA : Multiple failure (LBLOCA and non-availability of Wired Shutdown System) ~ 30 kg in 300 s.
Prolonged SBO + non-availability of GDWP ~ 450 Kg in 2 hr starting after 40hrs of transient (~5000 m3 at ambient)
Peak H2 generation rate ~ 0.3 kg/s
The released hydrogen will be combined by Passive Autocatalytic Recombiners (PARCS) located at several locations in the containment designed in such a way to reduce the hydrogen concentration in the containment below the flammability limits.
Experiments are being carried out for demonstration of hydrogen removal using PARCS
Recombination rate ~ kg/hr/m2 (for 2 - 4% H2 conc.)
Overall box size : x 400 x (L X B X H)
(8.29 m2 of Catalyst Deposited area)
Estimated Conversion rate : 0.83 kg/hr
No. of Recombiners for one Plant ~ (Total Conversion Rate = 83 kg/hr)

39. Design objective of the core catcher
Design of Core Catcher
Sacrificial Concrete
(300 mm depth)
High porosity concrete
Water pool
(500 mm depth)
Riser Tubes
( 100mm)
Structure of core catcher
7.4 m
Water from GDWP
Sacrificial concrete layer mixes with the melt, reduces its temperature, solidus temperature (typically from 2800oC to 1500oC) and helps in spreading the melt over large surface area
Poison added in sacrificial concrete prevents recriticality
High porosity concrete layer below the sacrificial concrete helps in flooding water from below
Riser tubes inject water within the melt-concrete mixture
The downcomers supply water to the water pool from GDWP passively
Sacrificial concrete
composition
Design objective of the core catcher
Retention of the melt in the cavity
Quenching it within 30 minutes
Stabilize it for substantial period of time (several days) 39

40. Indian High Temperature Reactor Programme

41. Indian High Temperature Reactor Programme
Compact High Temperature Reactor (CHTR)- Technology Demonstrator
100 kWth, 1000 °C, TRISO coated particle fuel
Several passive systems for reactor heat removal
Prolonged operation without refuelling
Status: Design of most of the systems worked out. Fuel and materials under development. Experimental facilities for thermal hydraulics setup. Facilities for design validation are under design.
Status: Optimisation of reactor physics and thermal hydraulics design, selection of salt and structural materials in progress. Experimental facilities for molten salt based thermal hydraulics and material compatibility studies set-up.
Innovative High Temperature Reactor for Hydrogen Production (IHTR)
600 MWth , 1000 °C, TRISO coated particle fuel
Small power version for demonstration of technologies
Active & passive systems for control & cooling
On-line refuelling
Indian Molten Salt Breeder Reactor (MSBR)
Large power, moderate temperature, and based on 233U-Th fuel cycle
Small power version for demonstration of technologies
Emphasis on passive systems for reactor heat removal under all scenarios and reactor conditions
Status: Initial studies being carried out for conceptual design 41

42. Technology for fuel kernel by sol-gel technique is well established – Focus is on technologies for TRISO coating and fuel compact
Initial trials with zirconia kernels completed
Fabrication trials of TRISO fuel using natural UO2 kernel carried out
Fuel compact prepared by two different techniques
High packing density (45-50%) achieved
OPyC
SiC
Zirconia
IPyC
Buffer PyC
X-ray radiographic image of TRISO particle with Zirconia kernel
Radiograph and tomograph of fuel compact made by different technique
SEM images of particle with Nat. UO2 kernel
Fuel Compacts

43. Fabrication of C/C composite tubes and coating with SiC
High Temperature Fluidized bed Coater
(Inset shows fluidized bed
distributor assembly)
Cooling tower
Induction heating system
Sample with graphite fixtures and graphite susceptor
Fluidized Bed Distributor
Heated graphite being dipped in fluidized bed
Ar rotameter
Fluidised bed based SiC coating method developed
High density C-C composite fuel tube samples fabricated in collaboration with National Physical Laboratory, New Delhi
Pre-form was made using high strength carbon fibers
Pre-form subjected to multiple cycles of resin impregnation and hot iso-static pressing with intermediate machining cycles
Machining trials of graphite components (AFD)

44. Thermal hydraulic studies for liquid metal (Pb-Bi)
Liquid Metal Loop (2009)
Major areas of development
Analytical studies and development of computer codes
Liquid metal loop for experimental studies
Loop at 550 °C in operation since 2009
Loop at 1000 °C under commissioning
Steady state and transient experiments carried out
In-house developed code validated
Experimental and analytical studies for freezing and de-freezing of coolant
Test bed for development of instrumentation –level probes, oxygen sensor, EM pump and flowmeters
YSZ based oxygen sensor
Comparison of steady state correlation [Vijayan, 2002] with experimental data 44

45. Sufficient time margin before shutdown or passive alternate heat removal system needs to act
Case-1
250% step increase in power
LOCA
No heat sink
Case-2
Similar to case-1, but with a 300% “spike” in power before stabilizing at 250%
~40 min
~58 min
Sufficient time available to activate primary and/or secondary shutdown system, or passive gas-gap filling system

46. Negligible rise in peak temperatures after shutdown due to decay heat
Minimum temperatures well above freezing point of coolant even after 1 hour

47. Innovative High Temperature Reactor (IHTR) for commercial hydrogen production
600 MWth, 1000 °C, TRISO coated particle fuel
Pebble bed reactor concept with molten salt coolant
Natural circulation of coolant for reactor heat removal under normal operation
Current focus on development:
Reactor physics and thermal hydraulic designs – Optimisation
Thermal and stress analysis
Code development for simulating pebble motion
Experimental set-up for tracing path of pebbles using radio-tracer technology
Pebble feeding and removal systems
TRISO coated particle fuel
Pebble
Hydrogen: 80,000 Nm3 /hr
Electricity: 18 MWe, Water: 375 m3/hr
No. of pebbles in the annular core ~150000
Packing fraction of pebbles ~60%
Packing fraction of TRISO particles ~ 8.6 %
233U Requirement 7.3 %

48. Molten salt corrosion test facility
Thermal hydraulic studies and material compatibility studies for molten salt coolant
Molten salt loop
Major areas of development
Analytical studies and development of computer codes
Molten salt natural circulation loop for experimental studies
Molten fluoride salt corrosion facility using FLiNaK
Experiments being carried out upto 750 °C mainly on Inconel materials
Molten salt corrosion test facility 48

49. Design features of Indian HTRs leading to inherent safety
TRISO coated fuel particles: Retention of fission products up to 1600 °C
High thermal inertia of ceramic core and low power density
Sufficient margin between reactor operation and boiling point of the coolant
Negative temperature coefficient of the core and coolant
Natural circulation of liquid metal / molten salt coolant in single phase
Low pressure of the system
Passive removal of heat under normal operation and postulated accident scenarios
High temperature heat pipe for CHTR
Chemical inertness of the lead based coolant with air/water

50. Molten Salt Breeder Reactor (MSBR)
This concept is attractive to India because of large thorium reserves and possibility of breeding 233U in thermal spectrum – For the third stage of Indian Nuclear Power Programme

51. Schematic of Indian MSBR
Design guidelines
Heat removal by natural circulation of molten salts
Avoid moderator to reduce solid high level waste generation
Ability to tolerate outage of reprocessing plant
Enhanced safety as compared to current reactors for possible deployment near population centres
Turbine
IHX
Condenser
Redox control
(Fertile Salt)
Helium bubbling and Redox control
(Fuel Salt)
Pump
Fuel Salt
Selection of salts, materials and conceptual design in progress
Fertile Salt
Fertile salt drain tank
Coolant salt drain tank
Fissile salt drain tanks

52. Inherent safety features of MSBR (1/2)
Continuous addition of fuel to maintain criticality
Less initial reactivity
Fission products, including xenon and krypton, are continuously taken out of the system,
No excess reactivity reaquired for xenon override
No danger of their release under accident condition
Entire fuel salt inventory can be dumped into smaller subcritical dump tanks, through freeze valves,
Reducing the chances of any untoward incidents.
The molten salt has a high boiling point (~1400°C), hence there is a very low vapor pressure
Normal operating temperatures ~ 700 to 800 C 52

53. Inherent safety features of MSBR (2/2)
The density of fuel salts decreases with increase in temperature,
With increase in temperature fuel salt is pushed out of the core leading to reduction of reactivity
No scenario for ‘fuel melt down’
Modification of existing safety codes required for defining CDF
Molten fluorides are simple ionic liquids
Stable to the irradiation
Do not undergo any violent chemical reactions with air or water
Fuel has no burnup limits
Life is dictate by life of moderator and structural materials 53

54. Accelerator Driven Systems

55. Accelerator-driven Sub-critical reactor system
BARC is developing technologies for Accelerator Driven System (ADS) mainly for Thorium utilization and waste transmutation
Major Role:
Accelerator-driven Sub-critical reactor system
High conversion sub-critical blanket with thorium for producing 233U
Incineration of minor actinides and some fission products
Sub-critical reactor core
Steam generator plant
Turbo-electrical generation plant
Accelerator-driven Sub-critical reactor system

56. Generation of fissile materials from thorium by spallation reaction using high energy proton accelerators
Inherently safe, flexible fuel cycle
Higher burn-up
Reduced doubling time for ADS-breeders
Intense, low-energy-cost neutron source
Fissile factory for U-233 from Th-232
Suitability for transmutation & burning nuclear waste

57. ADS for Transmutation & with Th-fueled reactor

58. Summary
In the Indian context, large scale deployment of nuclear reactors is required, with possible deployment near population centres
Enhanced level of safety is one of the primary goals for advanced reactors under design in BARC
Defence-in-depth
Passive safety devices
PSA studies
Margin assessment
Advanced materials
Advanced Reactor concepts

59. Thank You

60. Modification to Strengthen “Severe Accident Prevention Features”
Improving availability of onsite power supply
- Providing back up emergency DG (air cooled) at a higher location
- Providing a smaller/mobile DG to power essential loads and charge station batteries
Improving steam generator heat sink
- Securing FFW diesel engines pumps from external flood and margins w.r. t earthquake evaluated
Additional diesel engine operated pumps to transfer deaerator storage tank inventory to steam generator
Provision of hook up connections outside reactor building, qualified for maximum anticipated earthquake and flood
Provision for Passive Decay Heat Removal (PDHR) system for 700 MWe
Improving onsite water storage for one month SBO period
Augmentation of water inventory
Sources of water near stations are identified for fire tenders
Hook upto Primary Heat Transport System /ECCS
- Injection into PHT system for making up leakage during SBO
- Injection into PHT for unsuccessful long term ECCS operation

61. Other measures
Introduction of Seismic Trip (already exists in NAPS & KAPS)
Strengthening provision for monitoring of critical parameters under prolonged loss of power
Creation of an emergency response facility capable of withstanding severe flood, cyclones & earthquake
Provision for Tsunami early warning system