Accident Risks for High
Temperature Reactors
M. V. Ramana
Nuclear Futures Laboratory and Program on Science and Global Security, Princeton University
Matthias Englert
Öko-Institut e.V., Darmstadt
Friederike Frieß
IANUS, University of Technology, Darmstadt
Presented at 1st International Conference on Nuclear Risks
Vienna, 16-17 April 2015
Overview
•
Brief description of Chinese HTR program
•
Long history of interest - HTGRs in many
countries
•
Operational record
•
Claim about inherent safety
•
Severe accidents
Brief description of
Chinese HTR program
China & High Temperature Reactors
•
1970s - R&D
•
1988: Cooperation contract
between Tsinghua & Jülich
•
1992: government approval
for construction of HTR-10
•
1995-2000: Construction
•
2003: Full power to grid
HTR-PM
•
2001: High-temperature gas cooled reactor pebble-bed
module (HTR-PM) project launched
•
2004: Preliminary decision to set thermal power output at 485
MWt, subsequently 2X250 MWt
•
High priority under “Chinese Science and Technology Plan” for
the period 2006–2020
•
2008: Implementation plan and budget approved
•
2011: Final approval, just two weeks before Fukushima
•
2013: Construction commenced in eastern Shandong province
Plans to Increase Capacity
(6×250
MWt
HTR-PM
modules
+
1×660
MWe
steam
turbine)
Source: http://
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View of Proponents
•
High Temperature Reactors will be built in large
numbers once the HTR-PM is up and running
•
Main selling point is safety (“challenge to the
present regulation”)
Source: https://
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v=3RsiV1wVxW4
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Two Questions
•
What can we expect in terms of small accidents
(“incidents”)?
•
What are the possibilities for severe accidents
with potential for release of radioactive inventory
to the environment?
Breeder Reactor Performance
& Sodium Leaks
Phenix
Date of Grid Connection 13-Dec-73
Cumulative Load factor
40.5
PFR
BN-600
Superphenix
10-Jan-75
08-Apr-80
14-Jan-86
26.9
74.1
7.9
Source: Power Reactor Information System Database, International Atomic Energy Agency, 8 July 2013
Reactor Phenix Superphenix BN600 BN350
PFR
DFR
Number
of leaks
20
7
31
7
27
15
FFTF MONJU KNK II FBTR
1
1
21
Source: Guidez J, Martin L, Chetal SC, et al. (2008) Lesson Learned from sodium-cooled fast reactor
operation and their ramification for the reactors with respect to enhanced safety and reliability. Nuclear
Technology, 164, 207–220.
9
6
“One possible cause is a series of chemical interactions
between the carbon contained in the metallic components
of these reactors and the sodium used to cool reactors;
these interactions can cause a system’s metal parts to
corrode, eventually leading to leaks”
10
What can we expect in terms of
small accidents (“incidents”)?
Learning from History
Dawn of the Nuclear Age
Many Attempts
Source: http://www.jaea.go.jp/jaeri/english/ff/ff43/randd01.html
Dragon (UK,1964-1975)
Problems
•
“Severe and rapid” corrosion in the heat
exchangers
•
Leakage of helium into secondary circuit (luckily
no water leakage into primary circuit)
•
Diffusion of hydrogen produced by corrosion
into primary circuit
Source: Lockett, G.E., and S.B. Hosegood. “Engineering Principles of High Temperature
Reactors.” Jülich, Germany, 1968.
Peach Bottom 1
•
March 1966: Initial criticality
•
May 1966: Plant shut down for steam generator repair work
•
June 1967: Start of commercial operations
•
January 1968: Failed fuel element (increase in radioactivity in primary cycle detected)
•
October 1968: Eleven additional failed fuel elements detected; plant shut down for
maintenance and surveillance
•
January 1969: Restart with fresh fuel; 78 failed fuel elements by October 1969
•
July 1970: Plant restarted with new fuel design (operation for 900 days)
•
October 1974: Decision to shut down - cost of new fuel + meeting NRC requirements
too high compared to benefits
Image Source: https://lasttechage.files.wordpress.com/2011/04/peachbottom-u1-htgr-600x326.jpg
Information Source: Everett III, James L., and Edward J. Kohler. “Peach Bottom Unit No. 1: A High Performance Helium
Cooled Nuclear Power Plant.” Annals of Nuclear Energy 5, no. 8–10 (1978): 321–35. doi:10.1016/0306-4549(78)90017-8.
Fort St. Vrain - 1
•
January 1974: Initial criticality
•
August 1974 and January 1975: Moisture
ingress into primary system
•
July 1976: Helium leak
•
December 1976: Connected to grid
Fort St. Vrain - Core
Temperature Oscillations
126
/
H.G. Olson et al.
The Fort St. Vrain H T G R X
REGIONS
•
Source: Olson,
H. G., H. L.
Brey, and D. W.
Warembourg.
“The Fort St.
Vrain High
Temperature
Gas-Cooled
Reactor: X. Core
Temperature
Fluctuations.”
Nuclear
Engineering and
Design 72, no. 2
(1982): 125–37.
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TIME, MINUTES
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Fort St. Vrain - Performance
800"
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New York Times, December 8, 1988
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What are the possibilities for
severe accidents with potential for
release of radioactive inventory to
the environment?
Claim
•
“The inherent safety features of modular HTGR power plants
guarantees and requires that under all conceivable accident
scenarios the maximum fuel element temperatures will never
surpass its design limit temperature without employing any
dedicated and special emergency systems [e.g. core cooling
systems or special shut-down systems, etc.]. This ensures
that accidents [e.g. similar to LWR core melting] are not
possible so that unacceptable large releases of radioactive
fission products into the environment will never occur”
Source: Zhang, Zuoyi, Zongxin Wu, Dazhong Wang, Yuanhui Xu, Yuliang Sun, Fu Li, and Yujie Dong.
2009. “Current Status and Technical Description of Chinese 2 × 250 MWth HTR-PM Demonstration
Plant.” Nuclear Engineering and Design 239 (7): 1212–1219.
No pressurized leak tight
containment
Safety concept
No emergency cooling system (ECCS)
Decay heat can be removed through complete
nature mechanism, such as heat conduction,
heat radiation, etc. ---Inherent safety
Containment: vented low pressure containment
(VLPC)
Safety goal: cumulative frequency <1.0E-6 for
Source:
Dong, Yujie.
2011. “Status
of Development
and Deployment
Scheme ofeffective
HTR-PM in
BDBA
which
causes
off-site
personal
the People’s Republic of China.” presented at the Interregional Workshop on Advanced
dose >50mSv
Nuclear Reactor Technology for Near Term Deployment, Vienna, Austria, July 4.
Technically, off-site emergency planning
measures can be simplified remarkably
Differences and Similarties
Radioactive release paths not comparable to LWR scenarios (loss of
coolant, core melt down)
Graphite will not melt. BUT graphite burns! Problematic accidents by water
and air ingress in the core with corrosion of fuel elements (water-graphite),
hydrogen, and graphite fire (air-graphite)
Release of radioactivity
- loss of fuel matrix barrier (TRISO, Graphite) at temperatures over 1600 Degree Celsius.
[Promise: “The reactor core is designed and laid out that the fuel element temperature never
exceeds that safety limit [1620 Degree C] under any operation and accidental
condition” [Zheng et al. 2010]
- contamination of the primary circuit (fuel particle failure, broken pebbles; graphite dust with
radioactive particles)
Air Ingress
•
•
Air Ingress may lead to the
burning of graphite
Critical variable is mass flow
Water Ingress
HTR-PM demo plant
•
One of the severe accidents considered by most
safety analyses of the HTR
•
Under-moderated core => positive reactivity
increase when water enters core (reduction in
neutron leakage + increased resonance escape)
•
Graphite corrosion
Water Ingress
Design basis
Double ended break of heat generator steam tube.
Water from Heat exchanger blown into core, 600 kg total
Reactivity increase
Covered by negative temperature coefficient.
Assumption: Blower stops, safety valves close, shutdown
with control rods and absorber spheres, steam generator
draining works
Beyond design basis
Combination of blower does not stop, valve do not close,
no shutdown.
Worst scenarios: Anticipated transient without scram
(ATWS), Large break in steam generator plate
Larger amount of water enters core leading to corrosion.
Pebble Flow
Pebble compaction
- may lead to overheating especially close to reflector but
also locally
- Overheating increases diffusion of Cs137, Sr90 etc.
Pebble destruction
- will release of radioactive particles that often attaches to
graphite dust.
- impede pebble flow
Dust accumulates in tube bends etc.
Might be released during accidents
AVR experience
Problems
No Pressurized Leak Proof Containment
Local hot spots and overheating
Less understanding of accident scenarios as
compared to LWRs
Not much experience
TRISO fabrication and diffusion issues