ScribbleLive

@James Ward everything is highly edited so it is hard to see the whole picture of an area

Those cams numbers that jo notes below, for 2&3 suppression chambers, are huge

@James Ward This vid shows where that picture came from www.tepco.co.jp

17 apr plant parameters (see suppression chamber values): www.meti.go.jp

www.tepco.co.jp www.tepco.co.jp

@Nancy - I agree 100% with your comment below: "everything is highly edited so it is hard to see the whole picture of an area". Exactly why the work you're doing is so valuable.

@James Ward Thanks. I just wish they would be more forthcoming with video and images. It would go far to improve their credibility. @Elaine, I have not seen anything of 5 or 6 other than some early roof shots.

@elainekirk if anyone sees new image data on 5 or 6 let me know.

@elainekirk that trial footage could really help at least get a rough idea of some more intact insides.

I think we maybe should consider that these phenomena are players in the situation. Probably not the only factor, but a prime mover.

The vent-to-vent desynchronization effects on LOCA steam condensation loads in BWR pressure suppression pool

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UNIT 1

References and further reading may be available for this article. To view references and further reading you must purchase this article.


Yutaka Kukitaaand Ken Namatameb

aJapan Atomic Energy Research Institute, Tokai-Mura, Ibaraki-Ken, 319-11 Japan

bInstitute of Nuclear Safety, Japan, 1-4-28 Mita, Minatoku, Tokyo, 108 Japan

Received 5 November 1984. Available online 19 February 2003.

Abstract
During a loss-of-coolant accident of a boiling water reactor (BWR), direct-contact steam condensation in the pressure suppression pool will induce undesirable dynamic pressure loads onto pool boundary structures. The magnitudes of the pool boundary loads will be influenced by finite desynchronization among the condensation processes, which take place at the outlets of multiple (about one hundred) vent pipes. This paper investigates the effects of the condensation desynchronization on the pool boundary loads, with regard to test data obtained from large-scale tests which represented seven full-sized vent pipes in a BWR Mark II pressure suppression containment system. The desynchronization effects are studied by analyzing the test data in the time and frequency domains. Attempts are made to extrapolate the experimental results to the actual plant geometry. Mechanisms responsible for the desynchronization are also discussed

www.sciencedirect.com

Pressure suppression pool mixing in passive advanced BWR plants

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References and further reading may be available for this article. To view references and further reading you must purchase this article.


Robert E. Gamble, , a, Thuy T. Nguyena, Bharat S. Shiralkara, Per F. Peterson, b, Ralph Greifc and H. Tabata, d

a GE Nuclear Energy, 175 Curtner Avenue M/C 365, San Jose, CA 95125, USA

b Department of Nuclear Engineering, University of California, Berkeley, CA 94720-1730, USA

c Department of Mechanical Engineering, University of California, Berkeley, CA 94720-1730, USA

d Japan Atomic Power Company, 1-6-1 Otemachi, Chiyoda-ku, Tokyo 100-0004, Japan

Received 5 July 2000; accepted 5 July 2000 Available online 11 April 2001.

Abstract
In the SBWR passive boiling water reactor, the long-term post-accident containment pressure is determined by the combination of noncondensible gas pressure and steam pressure in the wetwell gas space. The suppression pool (SP) surface temperature, which determines the vapor partial pressure, is very important to overall containment performance. Therefore, the thermal stratification of the SP due to blowdown is of primary importance. This work looks at the various phases and phenomena present during the blowdown event and identifies those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests. This is important in determining which of the large body of blowdown to SP data is adequate for application to the stratification problem. The mixing by jets from the main vents is identified as the key phenomena influencing the thermal response of the suppression pool and analytical models are developed to predict the jet influence on thermal stratification. The analytical models are implemented into a system simulation code, TRACG, and used to model thermal stratification behavior in a scaled test facility. The results show good general agreement with the test data.

Article Outline
www.sciencedirect.com

www.osti.gov

www.osti.gov


Unit 2 suppression pool
www.osti.gov
EXPERIMENTAL AND THEORETICAL STUDY OF STEAM
CONDENSATION INDUCED WATER HAMMER PHENOMENA
Imre Ferenc Barna, Gábor Baranyai and György Ézsöl
KFKI Atomic Energy Research Institut(AEKI) of the Hungarian
Academy of Sciences, Thermo-hydraulic Department,
Address P.O. Box 49, H-1525 Budapest, Hungary,EU
Tel:00-36-392-2222/1472 , Fax:00-36-1-395-9293,
Email: barnai@sunserv.kfki.hu
Abstract We investigate steam condensation induced water hammer (waha)
phenomena and present experimental and theoretical results. Some of the
experiments were performed in the PMK-2 facility, which is a full-pressure
thermohydraulic model of the nuclear power plant of VVER-440/312 type and
located in the Atomic Energy Research Institute Budapest, Hungary. Other
experiments were done in the ROSA facility in Japan.
On the theoretical side waha is studied and analyzed with the WAHA3 model
based on two-phase flow six first-order partial differential equations that present
one dimensional, surface averaged mass, momentum and energy balances. A
second order accurate high-resolution shock-capturing numerical scheme was
applied with different kind of limiters in the numerical calculations. The applied
two-fluid model shows some similarities to Relap5 which is widely used in the
nuclear industry to simulate nuclear power plant accidents. Experimentally
measured and theoretically calculated waha pressure peaks are in qualitative
agreement.
I. INTRODUCTION
Safety of nuclear reactors is a fundamental issue.
Nuclear and thermo-hydraulic processes in the active zone
of modern reactors are well known and well-controlled,
explosions are out of question. However, violent unwanted
thermo-hydraulic transients in the primer loop may cause
serious derangement or pipe breakage. Such an unplanned
transient is the steam condensation induced water
hammer(waha). In thermal loops of atomic reactors or in
other pipelines where water steam and cold water can mix,
quick and dangerous transients can occur causing
explosions which mean high financial expenses or even
cost human lives.
In the following we will introduce the ROSA1 and
the PMK-2 facility 2 which is an integral experimental
device and capable to produce waha effects.
On the other side we present the WAHA3 3 model
we use, which is a complex physical model suitable to
simulate various quick transients in single and two-phase
flows.
In the last two decades the nuclear industry
developed a few complex two-phase flow-codes like
Relap5 4 , Trac 5 or Cathare 6 which are feasible to solve
safety analysis of nuclear reactors and model complicated
two-phase flow transients.
The model, WAHA37 is very similar to Relap5. This
means that the conservation equations and all the applied
correlations are essentially the same. The main difference
between the above mentioned models and our WAHA3
code is basically the applied numerical scheme; other
commercial codes have a ratio of spatial and time resolution
Dx / Dt which describes usual flow velocities. This code,
however is capable of capturing shock waves and describes
pressure waves which may propagate quicker than the local
speed of sound. To our knowledge there is no special
model and computer code for numerical simulation of water
hammer in the field of nuclear thermal-hydraulics.
WAHA3 model can successfully reproduce the
experimental data of different one- or two-phase flow
Proceedings of ICAPP ‘09
Tokyo

back to unit 1
www.nks.org
Combined Effects Experiments with
the Condensation Pool Test Facility
Markku Puustinen
Lappeenranta University of Technology, Finland

Abstract
This report summarizes the results of the condensation pool experiments in
spring 2006, where steam and steam/air mixture was blown into the pool through
a DN200 blowdown pipe. Altogether three experiments, each consisting of several
blows, were carried out with a scaled down test facility designed and constructed
at Lappeenranta University of Technology. The main purpose of the experiments
was to study the effects of non-condensable gas present in the discharge
flow. Particularly pressure pulses inside the blowdown pipe and at the
pool bottom caused by chugging were of interest.
The test pool was an open stainless steel tank with a wall thickness of 4 mm
and a bottom thickness of 5 mm containing 15 m3 of water. The nearby PACTEL
test facility was used as a steam source. During the experiments the initial pressure
of the steam source was 0.5 MPa and the pool water bulk temperature
ranged from 40 C to 70 C. The test facility was equipped with high frequency instrumentation
for capturing different aspects of the investigated phenomena. The
data acquisition program recorded data with the frequency of 10 kHz. A digital
high-speed video camera was used for visual observation of the pool interior.
Air, in quantities even less than 1 %, reduced the condensation rate considerably.
The high pressure pulses registered inside the blowdown pipe due to water
hammer propagation during chugging almost disappeared when the combined
discharge period of steam and air started. With non-condensable gas fractions
above 3 % the damping of pressure oscillations inside the blowdown pipe was
practically complete. Air quantities in the vicinity of 2 % started to have an effect
also on the oscillations measured by the pressure sensor at the pool bottom.
Both the amplitude and frequency of the pressure pulses decreased considerably.
The experiments demonstrated that even small quantities of non-condensable
gas can have a strong diminishing effect on pressure oscillations and structural
loads.
Key words
condensation pool, steam/air blowdown, non-condensable gas, pressure oscillations
NKS-148
ISBN 978-87-7893-211-2
Electronic report,

@elainekirk so tepco owns housing units and resorts. Maybe they should use some of those unoccupied units to house people who can't go home due to the reactor disaster

@elainekirk possibly helpful. It shows an offset well. Adds to other info. I guess they are getting that from somewhere.

@Olivier: Articles like this makes me lose faith in mankind

Fukuskima 2003. Nipponese BUSINESS recruit homeless people to clean up nuclear plants www.elmundo.es
by Olivier at 3:59 PM

@Janis Moulton : this is known. tepco hirers homeless people for work in their plants. it is not new.

@mod: in this case, delete me too (2x) thx.

Check out neilhartmannlife.wordpress.com for documentary on Japanese nuclear temp workers' health risks. It is not just TEPCO, or Japan, by the way. The whole industry relies on "nuclear slaves" as a Russian Chernobyl liquidator explained.

@alblee Ah, got it. :-) Sounds like a plan. A running static record of some of this would be a good idea, also adding all those older (pre quake) incidents from the news or press releases like damaged pipes at the reactors.

[Yes, @Nancy. - Alblee]

If anyone finds a larger version of that image let me know. I tried a few URL hacks to see if I could find it with no luck. I think there is a larger version based on their naming structure.