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Nuclear Power | BWR Boiling Water Reactor | Timeline

Introduction

Here are notes on the BWR Boiling Water type reactors used at Fukushima


@@Specifications and facts

GE Mark I Containment

Reactorsref

Initially to be retired in 2011, but relicense for 10 more years. Damaged beyond repair,

Damaged beyond repair

Damaged beyond repair, possible prompt critical explosion of fuel scattering nuclear materials in 1000 ft high plume, sending large parts up to 2 miles away, detectable uranium in US west coast.

Shut down but fuel pool exploded and caught fire, destroying most walls and roof. Possible fission in pool.

Glossary

References


@@Contents


@@links

@@Generation

ALL MODERN US PLANTS ARE 2ND GENERATION
http://news.nationalgeographic.com/news/energy/2011/03/110323-fukushima-japan-new-nuclear-plant-design/ 5 of 6 reactors at Fukushima Daiichi are General Electric BWR-3 with Mark I containment building. 92 plants around the world use the BWR system, and 32 use the Mark I containment. They are considered "second" generation as are all reactors in the US and France. The first generation according to World Nuclear Association were developed in the 50s and 60s and most are gone.

Generation III reactors integrate passive or inherent safety system that operate without intervention or power, and have been around since 1982. Only 15 of 442 are G3 now, with 14 under consruction.


@@List of BWR reactors

see http://en.wikipedia.org/wiki/List_of_boiling_water_reactors

This is a list of operational and decommissioned nuclear reactors of the Boiling Water Reactor type, commonly used for generating electrical power in nuclear power plants.

Commercial Boiling Water Reactors

NPP

RX#

Nation

City

State

Status

MWth

MWe

Gen

NSSS Ver.

Cont. Ver.

Year Licensed

Year Lic Expires

License Renewals

Notes

Browns Ferry Nuclear Power Plant

1

US

Decatur

AL

Operating

3458

1065

GEN2

BWR/4

MK-1

1974

2033

1

Browns Ferry Nuclear Power Plant

2

US

Decatur

AL

Operating

3458

1118

GEN2

BWR/4

MK-1

1974

2034

1

Browns Ferry Nuclear Power Plant

3

US

Decatur

AL

Operating

3458

1114

GEN2

BWR/4

MK-1

1976

2036

1

Brunswick Nuclear Generating Station

1

US

Southport

NC

Operating

2923

938

GEN2

BWR/4

MK-1

1976

2036

1

Brunswick Nuclear Generating Station

2

US

Southport

NC

Operating

2923

900

GEN2

BWR/4

MK-1

1974

2034

1

Clinton Nuclear Generating Station

1

US

Clinton

IL

Operating

3473

1043

GEN2

BWR/6

MK-3

1987

2026

0

Columbia Nuclear Generating Station

2

US

Richland

WA

Operating

3486

1131

GEN2

BWR/5

MK-2

1984

2023

0

Cooper Nuclear Station

1

US

Nebraska City

NE

Operating

2419

760

GEN2

BWR/4

MK-1

1974

2034[1]

1

Dresden Nuclear Power Plant

2

US

Morris

IL

Operating

2957

867

GEN2

BWR/3

MK-1

1971

2029

1

Dresden Nuclear Power Plant

3

US

Morris

IL

Operating

2957

867

GEN2

BWR/3

MK-1

1971

2031

1

Duane Arnold Energy Center

1

US

Cedar Rapids

IA

Operating

1912

581

GEN2

BWR/4

MK-1

1974

2034

1

Edwin I. Hatch Nuclear Generating Station

1

US

Baxley

GA

Operating

2804

876

GEN2

BWR/4

MK-1

1974

2034

1

Edwin I. Hatch Nuclear Generating Station

2

US

Baxley

GA

Operating

2804

883

GEN2

BWR/4

MK-1

1978

2038

1

Enrico Fermi Nuclear Generating Station

2

US

Newport

MI

Operating

3430

1122

GEN2

BWR/4

MK-1

1985

2025

0

Grand Gulf Nuclear Generating Station

1

US

Vicksburg

MS

Operating

3898

1266

GEN2

BWR/6

MK-3

1984

2024

0

Hope Creek Nuclear Generating Station

1

US

Hancocks Bridge

NJ

Operating

3840

1061

GEN2

BWR/4

MK-1

1986

2026

0

James A. Fitzpatrick Nuclear Generating Station

1

US

Oswego

NY

Operating

2536

852

GEN2

BWR/4

MK-1

1974

2014

0

LaSalle County Nuclear Generating Station

1

US

Ottawa

IL

Operating

3489

1118

GEN2

BWR/5

MK-2

1982

2022

0

LaSalle County Nuclear Generating Station

2

US

Ottawa

IL

Operating

3489

1120

GEN2

BWR/5

MK-2

1983

2023

0

Limerick Nuclear Power Plant

1

US

Philadelphia

PA

Operating

3458

1134

GEN2

BWR/4

MK-2

1985

2024

0

Limerick Nuclear Power Plant

2

US

Philadelphia

PA

Operating

3458

1134

GEN2

BWR/4

MK-2

1989

2029

0

Monticello Nuclear Generating Plant

1

US

Monticello

MN

Operating

1775

572

GEN2

BWR/3

MK-1

1971

2030

1

Nine Mile Point Nuclear Generating Station

1

US

Oswego

NY

Operating

1850

621

GEN2

BWR/2

MK-1

1974

2029

1

Nine Mile Point Nuclear Generating Station

2

US

Oswego

NY

Operating

3467

1135

GEN2

BWR/5

MK-2

1987

2046

1

Oyster Creek Nuclear Generating Station

1

US

Toms River

NJ

Operating

1930

619

GEN2

BWR/2

MK-1

1969

2029

1

Peach Bottom Nuclear Generating Station

2

US

Lancaster

PA

Operating

3514

1112

GEN2

BWR/4

MK-1

1973

2033

1

Peach Bottom Nuclear Generating Station

3

US

Lancaster

PA

Operating

3514

1112

GEN2

BWR/4

MK-1

1974

2034

1

Perry Nuclear Generating Station

1

US

North Perry

OH

Operating

3758

1231

GEN2

BWR/6

MK-3

1986

2026

0

Pilgrim Nuclear Generating Station

1

US

Plymouth

MA

Operating

2028

685

GEN2

BWR/3

MK-1

1972

2012

0

Quad Cities Nuclear Generating Station

1

US

Moline

IL

Operating

2957

867

GEN2

BWR/3

MK-1

1972

2032

1

Quad Cities Nuclear Generating Station

2

US

Moline

IL

Operating

2957

867

GEN2

BWR/3

MK-1

1972

2032

1

River Bend Nuclear Generating Station

1

US

Baton Rouge

LA

Operating

3091

967

GEN2

BWR/6

MK-3

1985

2025

0

Susquehanna Steam Electric Station

1

US

Berwick

PA

Operating

3952

1300

GEN2

BWR/4

MK-2

1982

2022

0

Susquehanna Steam Electric Station

2

US

Berwick

PA

Operating

3952

1140

GEN2

BWR/4

MK-2

1984

2024

0

Vermont Yankee Nuclear Power Plant

1

US

Brattleboro

VT

Operating

1912

620

GEN2

BWR/4

MK-1

1973

2012

0

Big Rock Point

1

US

Charleviox

Michigan

Dismantled

240

ND

GEN1

BWR/1

PREMOD

1964

1997

N/A

End of economical operation.

Dresden Nuclear Power Plant

1

US

Morris

IL

Stored

700

ND

GEN1

BWR/1

PREMOD

1959

1978

N/A

End of economical operation.

Elk River Station

1

US

Elk River

MN

Stored

58

ND

GEN1

BWR/1

PREMOD

1962

1968

N/A

Demonstration plant.

Humboldt Bay

3

US

Eureka

CA

Dismantled

200

ND

GEN1

BWR/1

PREMOD

1962

1976

N/A

End of economical operation.

La Crosse Boiling Water Reactor

1

US

Genoa

WI

Stored

165

ND

GEN1

BWR/1

PREMOD

1967

1987

N/A

End of economical operation.

Millstone Nuclear Power Plant

1

US

Waterford

CT

Stored

2011

ND

GEN2

BWR/4

MK-1

1986

1998

N/A

Technical issues.

Pathfinder Nuclear Generating Station

1

US

Sioux Falls

SD

Dismantled

190

ND

GEN1

BWR/1

PREMOD

1964

1967

N/A

Demonstration plant.

Shoreham Nuclear Generating Station

1

US

Wading River

NY

Dismantled

2436

ND

GEN2

BWR/5

MK-2

1989

1989

N/A

Completed but successsfully opposed by anti-nuclear movement before entering commercial operation.

Vallecitos Nuclear Center

1

US

Sunol

CA

Stored

50

ND

GEN1

BWR/1

PREMOD

1957

1963

N/A

Demonstration plant.

GE BONUS

1

PR

Rincón

Puerto Rico

Stored

50

ND

GEN1

VARIANT

PREMOD

1964

1968

N/A

Demonstration plant.

Tokai Nuclear Power Plant

2

JP

Tokai

Ibaraki Prefecture

Operating

?

1100

GEN2

BWR/5

MK-2

1978

N/A

N/A

Fukushima I (Daiichi) Nuclear Power Plant

1

JP

Okuma

Fukushima Prefecture

Damaged

1380[2]

460

GEN2

BWR/3[3]

MK-1

1971

N/A

N/A

Scheduled Shutdown 26 Mar 2011[4]; Damaged during Fukushima I nuclear accidents

Fukushima I (Daiichi) Nuclear Power Plant

2

JP

Okuma

Fukushima Prefecture

Damaged[5]

2381

784

GEN2

BWR/4

MK-1

1974

N/A

N/A

Scheduled Shutdown 18 Jul 2014[4]; Damaged during Fukushima I nuclear accidents

Fukushima I (Daiichi) Nuclear Power Plant

3

JP

Okuma

Fukushima Prefecture

Damaged[6]

2381

784

GEN2

BWR/4

MK-1

1976

N/A

N/A

Scheduled Shutdown 26 Mar 2016[4]; Damaged during Fukushima I nuclear accidents

Fukushima I (Daiichi) Nuclear Power Plant

4

JP

Okuma

Fukushima Prefecture

Maintenance

2381

784

GEN2

BWR/4

MK-1

1978

N/A

N/A

Scheduled Shutdown 12 Oct 2018[4]

Fukushima I (Daiichi) Nuclear Power Plant

5

JP

Okuma

Fukushima Prefecture

Maintenance

2381

784

GEN2

BWR/4

MK-1

1978

N/A

N/A

Scheduled Shutdown 18 Apr 2018[4]

Fukushima I (Daiichi) Nuclear Power Plant

6

JP

Okuma

Fukushima Prefecture

Maintenance

3293

1100

GEN2

BWR/5

MK-2

1979

N/A

N/A

Scheduled Shutdown 24 Oct 2019[4]

Fukushima I (Daiichi) Nuclear Power Plant

7

JP

Okuma

Fukushima Prefecture

Under Construction

?

1380

GEN3

ABWR

ABWR

2013

N/A

N/A

ABWR.

Fukushima I (Daiichi) Nuclear Power Plant

8

JP

Okuma

Fukushima Prefecture

Under Construction

?

1380

GEN3

ABWR

ABWR

2014

N/A

N/A

ABWR.

Fukushima II (Daini) Nuclear Power Plant

1

JP

Naraha & Tomioka

Fukushima Prefecture

Operating

3293

1100

GEN2

BWR/5

MK-2

1982

N/A

N/A

Fukushima II (Daini) Nuclear Power Plant

2

JP

Naraha & Tomioka

Fukushima Prefecture

Operating

3293

1100

GEN2

BWR/5

MK-2

1984

N/A

N/A

Fukushima II (Daini) Nuclear Power Plant

3

JP

Naraha & Tomioka

Fukushima Prefecture

Operating

3293

1100

GEN2

BWR/5

MK-2

1985

N/A

N/A

Fukushima II (Daini) Nuclear Power Plant

4

JP

Naraha & Tomioka

Fukushima Prefecture

Operating

3293

1100

GEN2

BWR/5

MK-2

1987

N/A

N/A

Hamaoka Nuclear Power Plant

1

JP

Omaezaki

Shizuoka Prefecture

Operating

?

540

GEN2

BWR/4

MK-1

1974

N/A

N/A

Hamaoka Nuclear Power Plant

2

JP

Omaezaki

Shizuoka Prefecture

Operating

?

840

GEN2

BWR/5

MK-2

1978

N/A

N/A

Hamaoka Nuclear Power Plant

3

JP

Omaezaki

Shizuoka Prefecture

Operating

?

1100

GEN2

BWR/6

MK-3

1987

N/A

N/A

Hamaoka Nuclear Power Plant

4

JP

Omaezaki

Shizuoka Prefecture

Operating

?

1137

GEN2

BWR/6

MK-3

1993

N/A

N/A

Hamaoka Nuclear Power Plant

5

JP

Omaezaki

Shizuoka Prefecture

Operating

?

1380

GEN3

ABWR

ABWR

2004

N/A

N/A

ABWR.

Higashidori Nuclear Power Plant, A

1

JP

Higashidori

Aomori Prefecture

Operating

?

1100

GEN3

TOSHIBA

TOSHIBA

2005

N/A

N/A

Toshiba BWR variant.

Higashidori Nuclear Power Plant, A

2

JP

Higashidori

Aomori Prefecture

Under Construction

?

1385

GEN3

ABWR

ABWR

2010

N/A

N/A

ABWR.

Higashidori Nuclear Power Plant, B

1

JP

Higashidori

Aomori Prefecture

Under Construction

?

1385

GEN3

ABWR

ABWR

2008

N/A

N/A

ABWR.

Higashidori Nuclear Power Plant, B

2

JP

Higashidori

Aomori Prefecture

Under Construction

?

1385

GEN3

ABWR

ABWR

2011

N/A

N/A

ABWR.

Onagawa Nuclear Power Plant

1

JP

Onagawa

Miyagi Prefecture

Operating

?

524

GEN2

TOSHIBA

TOSHIBA

1984

N/A

N/A

Toshiba BWR variant.

Onagawa Nuclear Power Plant

2

JP

Onagawa

Miyagi Prefecture

Operating

?

825

GEN2+

TOSHIBA

TOSHIBA

1995

N/A

N/A

Toshiba BWR variant.

Onagawa Nuclear Power Plant

3

JP

Onagawa

Miyagi Prefecture

Operating

?

825

GEN3

TOSHIBA

TOSHIBA

2002

N/A

N/A

Toshiba BWR variant.

Shika Nuclear Power Plant

1

JP

Shika

Ishikawa Prefecture

Operating

1593

540

GEN2+

MITSU

MITSU

1993

N/A

N/A

Mitsubishi BWR variant.

Shika Nuclear Power Plant

2

JP

Shika

Ishikawa Prefecture

Operating

3926

1358

GEN3

ABWR

ABWR

2006

N/A

N/A

ABWR.

Shimane Nuclear Power Plant

1

JP

Matsue

Shimane Prefecture

Operating

?

460

GEN2

BWR/4

MK-1

1974

N/A

N/A

Shimane Nuclear Power Plant

2

JP

Matsue

Shimane Prefecture

Operating

?

820

GEN2

BWR/6

MK-3

1989

N/A

N/A

Shimane Nuclear Power Plant

3

JP

Matsue

Shimane Prefecture

Under Construction

?

1373

GEN3

ABWR

ABWR

2011

N/A

N/A

ABWR.

Tsuruga Nuclear Power Plant

1

JP

Tsuruga

Fukui Prefecture

Operating

?

357

GEN2

BWR/4

MK-1

1970

N/A

N/A

Olkiluoto Nuclear Power Plant

1

FI

Eurajoki

Western Finland

Operating

2500

860

GEN2

ABB

BWR 75

1979

N/A

N/A

First Asea Brown Boveri BWR 75.

Olkiluoto Nuclear Power Plant

2

FI

Eurajoki

Western Finland

Operating

2500

860

GEN2

ABB

BWR 75

1982

N/A

N/A

Gundremmingen Nuclear Power Plant

A

DE

Gundremmingen

Bavaria

Decommissioned

?

250

GEN1

BWR/1

PREMOD

1962

N/A

N/A

End of economical operation.

Gundremmingen Nuclear Power Plant

B

DE

Gundremmingen

Bavaria

Operating

?

1344

GEN2

Typ 72

Typ 72

1984

2016

N/A

Gundremmingen Nuclear Power Plant

C

DE

Gundremmingen

Bavaria

Operating

?

1344

GEN2

Typ 72

Typ 72

1984

2017

N/A

Isar Nuclear Power Plant

1

DE

Landshut

Bavaria

Operating

2575

912

GEN2

Typ 69

Typ 69

1979

2011

N/A

Krümmel Nuclear Power Plant

1

DE

Krümmel

Schleswig-Holstein

Operating

3690

1402

GEN2

Typ 69

Typ 69

1984

2016

N/A

Lingen Nuclear Power Plant

1

DE

Lingen

Lower Saxony

Decommissioned

?

268

GEN1

BWR/1

PREMOD

1968

1979

N/A

End of economic operation.

Philippsburg Nuclear Power Plant

1

DE

Philippsburg

Baden-Wuerttemberg

Operating

2575

890

GEN2

Typ 69

Typ 69

1979

2011

N/A

Würgassen Nuclear Power Plant

1

DE

Würgassen

North Rhine-Westphalia

Decommissioned

?

670

GEN2

Typ 69

Typ 69

1971

1994

N/A

Technical difficulties.

Tarapur Atomic Power Station

1

IN

Tarapur

Maharashtra

Operating

670

160

GEN1

BWR/1

PREMOD

1969

2021

1

Overhaul 2006, next modernization 2021.

Tarapur Atomic Power Station

2

IN

Tarapur

Maharashtra

Operating

670

160

GEN1

BWR/1

PREMOD

1969

2021

1

Overhaul 2006, next modernization 2021.

Garigliano Nuclear Power Plant

1

IT

Garigliano

Campania

Decommissioned

?

150

GEN1

BWR/1

PREMOD

1963

1982

N/A

End of economical operation.

Caorso Nuclear Power Plant

1

IT

Caorso

Emilia Romagna

Decommissioned

?

882

GEN2

BWR/?

MK-?

1978

1990

N/A

Shutdown following anti-nuclear machinations.

Dodewaard Nuclear Power Plant

1

NL

Neder-Betuwe

Gelderland

Decommissioned

183

58

GEN1

BWR/1

PREMOD

1969

1997

N/A

End of economical operation.

Laguna Verde Nuclear Power Plant

1

MX

Alto Lucero

Veracruz

Operating

2027

682.5

GEN2

BWR/5

MK-2

1989

N/A

N/A

Laguna Verde Nuclear Power Plant

2

MX

Alto Lucero

Veracruz

Operating

2027

682.5

GEN2

BWR/5

MK-2

1994

N/A

N/A

Cofrentes Nuclear Power Plant

1

ES

Cofrentes

Valencia

Operating

3237

1096

GEN2

BWR/6

MK-3

1984

2034

0

Santa María de Garoña Nuclear Power Plant

1

ES

Santa María de Garoña

Burgos

Operating

1381

466

GEN2

BWR/3

MK-1

1971

2013

1

Licence extended in 2009 by four years.

Barsebäck Nuclear Power Plant

1

SV

Barsebäck

Skane Lan

Decommissioned

1800

600

GEN2

ABB

BWR

1975

1999

N/A

Decommissioned due to political decisions.

Barsebäck Nuclear Power Plant

1

SV

Barsebäck

Skane Lan

Decommissioned

1800

600

GEN2

ABB

BWR

1975

2005

N/A

Decommissioned due to political decisions.

Forsmark Nuclear Power Plant

1

SV

Forsmark

Uppland

Operating

2928

1010

GEN2

ABB

BWR 69

1980

N/A

N/A

Forsmark Nuclear Power Plant

2

SV

Forsmark

Uppland

Operating

2928

1010

GEN2

ABB

BWR 69

1981

N/A

N/A

Forsmark Nuclear Power Plant

3

SV

Forsmark

Uppland

Operating

3300

1190

GEN2

ABB

BWR 75

1985

N/A

N/A

Oskarshamn Nuclear Power Plant

1

SV

Oskarshamn

Småland

Operating

1375

494

GEN2

ABB

BWR

1972

N/A

N/A

Oskarshamn Nuclear Power Plant

2

SV

Oskarshamn

Småland

Operating

1800

661

GEN2

ABB

BWR

1975

N/A

N/A

Oskarshamn Nuclear Power Plant

3

SV

Oskarshamn

Småland

Operating

3900

1450

GEN2

ABB

BWR 75

1985

N/A

N/A

Ringhals Nuclear Power Plant

1

SV

Varberg

Halland

Operating

2500

830

GEN2

ABB

BWR

1976

N/A

N/A

Leibstadt Nuclear Power Plant

1

CH

Leibstadt

Canton of Aargau

Operating

?

1220

GEN2

BWR/6

MK-3

1984

PERM

N/A

In nuclear power referendum of 2005, 68% of Swiss voters were in favor of continued operation at plant.

Mühleberg Nuclear Power Plant

1

CH

Mühleberg

Canton of Berne

Operating

?

385

GEN2

BWR/4

MK-2

1972

2012

4

Popular support for plant demonstrated by canton referendum for relicencing, 64% of voters in favor in 2000, reduced to 51.2% in 2011. Second LWR for site in planning process.

[edit] See also

·         List of nuclear reactors by country

·         List of nuclear power stations


@@Mark I Containment

Summary

The Mark I was the first of 3 containments, and the weakest design. The Mark II replaces a steel torus with a lined concrete compartment. Only the Mark III introduces a strong concrete dome over the reactor like most other reactors, and moves fuel storage into another building. Some faults of Mark I containment:

·         "Secondary containment" is just a steel or concrete thin-walled building kept at slight negative pressure. This was destroyed by hydrogen exposions in Fukushima Daichi units 1, 3, and 4, along with all equipment including overhead crane and refueling platform.

·         Destruction of roof and ducting also destroyed any ability to filter harmful radiation from pool or vented gases. However it did make it possible to drop air into the pool by water canon, helicopter and concrete pump as improvised methods.

·         Hydrogen gases were vented into upper building rather than up the stack. It is unknown, but suspected Japanese government did not require that Fukushima units had installed "hardened" vents designed to vent exposive gases up the stack through harded ducts. The duct on reactor 3 was completely severed after the hydrogen explosion. It is possible power failure deactivated fans and values to properly function.

·         Fuel storage pools require power to cool and circulate. They are placed above the hot reactor which can also heat water on top of warm fuel rods which are producing their own heat. The gates between top of reactor and equipment pool use air-filled seals which may not work without power, or in earthquake and allow leakage. It is certain there was a fire in pool unit 4, less certain what happened in Unit 3 which suffered a spectacular explosion sending a dark debris cloud 1000 ft in the air, and probably being the source of pieces of radioactive debris and plutonium as far as 2 miles away.

Links:

·         2011 NEI Mark I Containment Report after accident

·          

New York times April 6, 2011, 11:54 AM Chemistry 201: Why Is Fukushima So Gassy? By MATTHEW L. WALD

Description: http://graphics8.nytimes.com/images/2011/04/06/business/contain/contain-blog480.jpg
American experts are urging operators of the Fukushima Daiichi plant to fill the empty space in the primary containment – the area around the reactor vessel, shown here in red – with inert nitrogen gas to avert explosions.

http://www.gereports.com/how-it-works-white-paper-on-mark-i-containment/

Mark I Containment Facts and The New York Times

(comment - a whitewash which papers over failings of system which released a one-tenth scale version of Chernobyl in terms of radiactive releases and building destruction)

Posted on March 19, 2011, 11:47 pm, by GEreporter, under Other.

The New York Times published an online story last night and an accompanying graphic about the Mark I boiling water reactor (BWR) containment system used in the Fukushima Daiichi nuclear power plant. The story contains errors and distorts the facts about the technology with misleading comparisons of the BWR design and that of the pressurized water reactor (PWR).

The story claims that the BWR design is a “simpler containment.” The language suggests that simpler means weaker, which is not the case. In fact, there are containment design requirement differences. A PWR operates at over 2,000 pounds per square inch (psi). Conversely, a BWR operates at about half that – around 1,000 psi . In the event of a leak inside the containment system of a PWR, there are higher pressures that cause faster and larger volumes of instantaneous steam to be released. Therefore the containment has to have more free space in order to absorb the larger volumes of steam that are released.

Alternatively, a BWR operates at lower pressure and doesn’t have the same acceleration during a steam loss. Because the pressure loss is not as fast, the containment is not required to have as much free space and can be smaller in size. In either case, both designs are reviewed by the same regulatory processes – the same rules and the same requirements.

The Times also compares Three Mile Island and Fukushima, saying that the PWR reactor at Three Mile Island withstood a hydrogen blast. In fact, the hydrogen blast at Three Mile Island occurred within the primary containment. The hydrogen blast at Fukushima occurred in the reactor building – which is the secondary and not the primary containment. The indirect comparison between the blast at Three Mile Island and the blast at Fukushima is misleading.

Their graphic also confuses primary and secondary containment systems. They write: “Calculating how much heat needed to be disposed of, and with the torus to do that job, GE persuaded regulators that only a modest outer containment building was necessary.”

In fact:

·         The torus, which is a large, rounded suppression pool that sits next to the reactor core, is not the outer containment, it is primary containment.

·         By design, the outer containment is not a pressure containing building.

·         The reactor building (secondary containment) is kept at a slight vacuum to limit radioactivity release during refueling operations and during certain design scenarios.

·         The design pressures (meaning the pressures that each containment system is designed to withstand) of a BWR and a PWR primary containment are similar.

The Times also mischaracterizes the torus in a later description in the second paragraph. In fact, the torus is not kept at partial vacuum. It’s kept at ambient pressure. The secondary containment is kept at slight vacuum for reasons listed above.

The Times cites GE as one of its sources. GE did speak to them but the information they present is wrong.

Without a doubt, these issues are incredibly complex but that makes it all the more important for The New York Times to treat these issues with care.

Description: http://files.gereports.com/wp-content/uploads/2011/03/Mark1Containment.jpg

Click to enlarge.

How it Works: White Paper on Mark I Containment

(comment - this is largely a whitewash of how the GE design performed so "well")

A new report has just been developed on the Mark I containment design that is in use at the Fukushima Daiichi nuclear power plant. It also explores U.S. and regulatory actions that made the design safer over the past several decades, noting “the Mark I pressure suppression containment is a proven technology that has been enhanced with confirmatory testing, enhanced knowledge and advanced analysis over time.”

In addition to explaining the boiling water reactor (BWR) design, the report, which can be downloaded from the Nuclear Energy Institute website, also makes initial observations about the performance of the containment system at Fukushima.

·         “Coincident long-term loss of both on-site and off-site power for an extended period of time is a beyond-design-basis event for the primary containment on any operating nuclear power plant.”

·         “The Mark I containment vessels appeared to have held pressure to well above the design pressure.”

·         “The response of the reactor pressure vessel and reactor in general agree with severe accident management studies performed in the 1980s and early 1990s.”

Other topics in the report include: “Containment Operation During a Loss of Coolant Accident,” “Evolution of the Design,” and “Containment Operation During a Station Blackout.”

Regarding operation during a blackout, the report notes: “In the late 1980s and early 1990s, BWR operators made procedure changes and modifications to cope with events which involved the loss of the normal offsite power and normally available emergency diesel generators…. As a result of the September 11, 2001 terrorist attacks, additional actions and equipment were put in place at certain U.S. plants to allow water makeup to the reactor and the fuel pools should significant damage occur to the reactor buildings. These changes include pre-staged diesel-driven pumps, piping, and procedures that would support water makeup from various water supplies without the need for electrical power.”

* Download the full report from the NEI website.

Learn more in these GE Reports stories:
* Setting the Record Straight on Mark I Containment History
* Mark I Containment Facts and The New York Times
* The Mark I Containment System in BWR Reactors
* An Update on GE Disaster Relief Efforts in Japan
* Facts on the Nuclear Energy Situation in Japan (Update)

The NEI website is also providing updates on the situation in Japan.

The Mark I Containment System in BWR Reactors

While events are still unfolding on the ground at the damaged Fukushima Daiichi Power Plant, GE continues to provide technical assistance to TEPCO through our joint venture partners in Japan and to the U.S. Nuclear Regulatory Commission (NRC), which is in turn providing assistance to the Japanese government. There are also some facts that GE Hitachi Nuclear Energy can attempt to clarify, such as those concerning the Mark I containment system in use at the reactors in the Fukushima Daiichi Power Plant.

The Mark I containment has a proven track record of safety and reliability for over 40 years and there are 32 BWR Mark I reactors operating as designed worldwide.

While the technology was commercialized 40 years ago, it has continued to evolve. Over the last four decades, the Mark I has been modified in the form of retrofits to address technology improvements and changing regulatory requirements.

All of the modifications were made in accordance with regulatory requirements. In the United States, for example, the NRC issued a generic industry requirement in 1980 for the Mark I containment that the industry used to make modifications.

We understand that all of the BWR Mark I containment units at Fukushima Daiichi also addressed these issues and implemented modifications in accordance with Japanese regulatory requirements.

The modifications made to Mark I containments include:

·         Quenchers were installed to distribute the steam bubbles in order to produce rapid condensation and to reduce loads on the unit. In a reactor, exhaust steam is piped into a suppression chamber, which is known as the torus and is a large, rounded suppression pool that sits next to the reactor core. It is used to remove heat when large quantities of steam are released from the reactor. In the torus, the steam bubbles go under water. With the modification to the Mark I, the quenchers, which are also underwater, make steam bubbles smaller by breaking up the larger bubbles. This in turn reduces pressure.

·         Another modification is the installation of deflectors inside the torus. When that steam goes in, the water level rises. The deflectors that were added break up the pressure wave that is produced and help relieve pressure on the torus.

·         A further modification was made to the saddles on which the torus sits — basically the series of leg-like structures that support it. The construction was fortified, as was the steel, to accommodate the loads that are generated.

Description: http://files.gereports.com/wp-content/uploads/2011/03/bwr_2.jpg

A BWR reactor: The schematic above shows the torus at left, which is doughnut-shaped.

In a BWR, the containment strategy is a bit different. A BWR's containment consists of a drywell where the reactor and associated cooling equipment is located and a wetwell. The drywell is much smaller than a PWR containment and plays a larger role. During the theoretical leakage design basis accident the reactor coolant flashes to steam in the drywell, pressurizing it rapidly. Vent pipes or tubes from the drywell direct the steam below the water level maintained in the wetwell (also known as a torus or suppression pool), condensing the steam, limiting the pressure ultimately reached. Both the drywell and the wetwell are enclosed by a secondary containment building, maintained at a slight sub-atmospheric or negative pressure during normal operation and refueling operations. The containment designs are referred to by the names Mark I (oldest; drywell/torus), Mark II, and Mark III (newest). All three types house also use the large body of water in the suppression pools to quench steam released from the reactor system during transients. The Mark I containment was used in those reactors at the Fukushima I Nuclear Power Plant which were involved in the Fukushima I nuclear accidents. Even before the Fukushima incident, Mark I containment had been criticized as being more likely to fail. [4]

In most BWR designs the spent fuel pool is outside of the containment building.

From a distance, the BWR design looks very different from PWR designs because usually a square building is used for containment. Also, because there is only one loop through the turbines and reactor, and the steam going through the turbines is also slightly radioactive, the turbine building has to be considerably shielded as well. This leads to two buildings of similar construction with the taller one housing the reactor and the short long one housing the turbine hall and supporting structures.

Boiling Water Reactor Wikipedia

First series of production BWRs (BWR/1–BWR/6)

The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR: the torus (used to quench steam in the event of a transient requiring the quenching of steam), as well as the drywell, the elimination of the heat exchanger, the steam dryer, the distinctive general layout of the reactor building, and the standardization of reactor control and safety systems. The first, General Electric, series of production BWRs evolved through 6 iterative design phases, each termed BWR/1 through BWR/6. (BWR/4s, BWR/5s, and BWR/6s are the most common types in service today.) The vast majority of BWRs in service throughout the world belong to one of these design phases.

·         1st generation BWR: BWR/1 with Mark I containment.

·         2nd generation BWRs: BWR/2, BWR/3 and some BWR/4 with Mark I containment. Other BWR/4, and BWR/5 with Mark-II containment.

·         3rd generation BWRs: BWR/6 with Mark-III containment.

Containment variants were constructed using either concrete or steel for the Primary Containment, Drywell and Wetwell in various combinations.[5]

Apart from the GE designs there were others by ABB, MITSU, Toshiba and KWU. See List of boiling water reactors.

Disadvantages

·         Complex calculations for managing consumption of nuclear fuel during operation due to "two phase (water and steam) fluid flow" in the upper part of the core. This requires more instrumentation in the reactor core. The innovation of computers, however, makes this less of an issue.

·         Much larger pressure vessel than for a PWR of similar power, with correspondingly higher cost. (However, the overall cost is reduced because a modern BWR has no main steam generators and associated piping.)

·         Contamination of the turbine by short-lived activation products. This means that shielding and access control around the steam turbine are required during normal operations due to the radiation levels arising from the steam entering directly from the reactor core. This is a moderately minor concern, as most of the radiation flux is due to Nitrogen-16, which has a half-life measured in seconds, allowing the turbine chamber to be entered into within minutes of shutdown.

·         Though the present fleet of BWRs are said to be less likely to suffer core damage from the "1 in 100,000 reactor-year" limiting fault than the present fleet of PWRs are (due to increased ECCS robustness and redundancy) there have been concerns raised about the pressure containment ability of the as-built, unmodified Mark I containment – that such may be insufficient to contain pressures generated by a limiting fault combined with complete ECCS failure that results in extremely severe core damage. In this double failure scenario, assumed to be extremely unlikely prior to the Fukushima I nuclear accidents, an unmodified Mark I containment can allow some degree of radioactive release to occur. This is supposed to be mitigated by the modification of the Mark I containment; namely, the addition of an outgas stack system that, if containment pressure exceeds critical setpoints, is supposed to allow the orderly discharge of pressurizing gases after the gases pass through activated carbon filters designed to trap radionuclides.[7]

·         A BWR requires active cooling for a period of several hours to several days following shutdown, depending on its power history. Full insertion of BWRs control rods safely shuts down the primary nuclear reaction. However, radioactive decay of the fission products in the fuel will continue to actively generate decay heat at a gradually decreasing rate, requiring pumping of cooling water for an initial period to prevent overheating of the fuel. If active cooling fails during this post-shutdown period, the reactor can still overheat to a temperature high enough that zirconium in the fuel cladding will react with water and steam, producing hydrogen gas. In this event there is a high danger of hydrogen explosions, threatening structural damage to the reactor and/or associated safety systems and/or the exposure of highly radioactive spent fuel rods that may be stored in the reactor building (approx 15 tons of fuel is replenished each year to maintain normal BWR operation) as happened with the Fukushima I nuclear accidents.

·         Control rods are inserted from below for current BWR designs. There are two available hydraulic power sources that can drive the control rods into the core for a BWR under emergency conditions. There is a dedicated high pressure hydraulic accumulator and also the pressure inside of the reactor pressure vessel available to each control rod. Either the dedicated accumulator (one per rod) or reactor pressure is capable of fully inserting each rod. Most other reactor types use top entry control rods that are held up in the withdrawn position by electromagnets, causing them to fall into the reactor by gravity if power is lost.


@@Mark III Containment

Mark III Containment

http://nuclearstreet.com/nuclear-power-plants/w/nuclear_power_plants/mark-iii-containment.aspx

The Mark III primary containment consists of several major components, many of which can be seen below. The drywell (13) is a cylindrical, reinforced concrete structure with a removable head. The drywell is designed to withstand and confine steam generated during a pipe rupture inside the containment and to channel the released steam into the suppression pool (10) via the weir wall (11) and the horizontal vents (12). The suppression pool contains a large volume of water for rapidly condensing steam directed to it. A leak tight, cylindrical, steel containment vessel (2) surround the drywell and the suppression pool to prevent gaseous and particulate fission products from escaping to the environment following a pipe break inside containment.

 

Description:  From: http://www.nrc.gov/reactors/bwrs.html

Boiling Water Reactors

In a typical commercial boiling-water reactor, (1) the core inside the reactor vessel creates heat, (2) a steam-water mixture is produced when very pure water (reactor coolant) moves upward through the core, absorbing heat, (3) the steam-water mixture leaves the top of the core and enters the two stages of moisture separation where water droplets are removed before the steam is allowed to enter the steam line, and (4) the steam line directs the steam to the main turbine, causing it to turn the turbine generator, which produces electricity. The unused steam is exhausted into the condenser where it is condensed into water. The resulting water is pumped out of the condenser with a series of pumps, reheated and pumped back to the reactor vessel. The reactor's core contains fuel assemblies that are cooled by water circulated using electrically powered pumps. These pumps and other operating systems in the plant receive their power from the electrical grid. If offsite power is lost emergency cooling water is supplied by other pumps, which can be powered by onsite diesel generators. Other safety systems, such as the containment cooling system, also need electric power. Boiling-water reactors contain between 370-800 fuel assemblies. See also our animated diagram.

from http://www.nrc.gov/reactors/bwrs.html

In a typical commercial boiling-water reactor, (1) the core inside the reactor vessel creates heat, (2) a steam-water mixture is produced when very pure water (reactor coolant) moves upward through the core, absorbing heat, (3) the steam-water mixture leaves the top of the core and enters the two stages of moisture separation where water droplets are removed before the steam is allowed to enter the steam line, and (4) the steam line directs the steam to the main turbine, causing it to turn the turbine generator, which produces electricity. The unused steam is exhausted into the condenser where it is condensed into water. The resulting water is pumped out of the condenser with a series of pumps, reheated and pumped back to the reactor vessel. The reactor's core contains fuel assemblies that are cooled by water circulated using electrically powered pumps. These pumps and other operating systems in the plant receive their power from the electrical grid. If offsite power is lost emergency cooling water is supplied by other pumps, which can be powered by onsite diesel generators. Other safety systems, such as the containment cooling system, also need electric power. Boiling-water reactors contain between 370-800 fuel assemblies. See also our animated diagram.

Description: Typical Boiling Water Reactor (BWR)
Description: Boiling Water Reactor (BWR)


@@Perry Nuclear Generating Station

Summary: Wikipedia. The Perry Nuclear Power Plant reactor is a General Electric BWR-6 boiling water reactor design, with a Mark III containment design. The original core power level of 3,579 megawatts thermal was increased to 3,758 megawatts thermal in 2000, making Perry one of the largest BWRs in the United States. Built at a cost of $6 billion, Perry-1 is one of the most expensive power plants ever constructed.[citation needed] Perry was originally designed as a two-unit installation, but construction on Unit 2 was suspended in 1985 and formally cancelled in 1994. At the time of cancellation, all of the major buildings and structures for the second unit were completed, including the 500-foot-tall (150 m) cooling tower.

The Nuclear Regulatory Commission's estimate of the risk each year of an earthquake intense enough to cause core damage to the reactor at Perry was 1 in 47,619, according to an NRC study published in August 2010.[3][4]

http://www.nrc.gov/info-finder/reactor/perr1.html
Location: Perry, OH (35 miles NE of Cleveland, OH) in Region III
Operator: FirstEnergy Nuclear Operating Co.
Operating License: Issued - 11/13/1986, Expires - 03/18/2026
Docket Number: 05000440

Reactor Type: Boiling Water Reactor
Electrical Output:
1261 MWe
Reactor Vendor/Type:
General Electric Type 6
Containment Type:
Wet, Mark III

Safety problems

from:http://www.fox8.com/news/wjw-perry-nuclear-power-plant-concerns-after-japan-earthquake-txt,0,924650.story Local Nuclear Plant Shares Traits with One in Japan
Their fuel tanks are above ground, ours are buried and sealed in a vault," said Todd Schneider, a spokesman for FirstEnergy. "Our containment structures are larger and that protects the reactor better," he added.

from: http://www.fox8.com/news/wjw-fire-nuclear-power-plant-txt,0,2418243.story
2 People Injured in Nuclear Power Plant Fire
Bill Sheil Fox 8 News Anchor 6:35 a.m. EDT, March 29, 2010 two people were transported to the hospital for "heat stress" following a fire at its Perry Nuclear Power Plant. Crews managed to extinguish the fire before 9:45 p.m. Sunday, and there was no contamination at the Center Road facility, according to FirstEnergy. .. a "small fire" started around in a lubrication plant that feeds one of the water pumps that cools the nuclear reactor.... The pump had to be taken offline, so the utility has reduced output of the plant to 75 percent of its capacity. From: Perry Nuclear Power Plant, North Perry, Ohio April 30th, 2011 3:50 pm ET By Richard Zimmerman Perry’s troubles began as early as March of 2010, when a small fire burned for several hours in its water pump lubrication system. In May of that year, plant engineers also had to manually ‘power down’ the reactor, when it became clear that the core’s automatic emergency shutdown was not operating properly. The facility has also been plagued in recent years by a series of safety problems and violations that prompted the Nuclear Regulatory Commission to monitor safety procedures every three months throughout 2005. Perry also had to be shut down for a period due to problems with circulating pumps failing to provide adequate coolant water to temper the core reactor

http://www.journalgazette.net/apps/pbcs.dll/article?AID=/20110427/NEWS11/110429505
April 27, 2011 6:30 p.m. High radiation levels found at [Perry] Ohio nuclear plant MEGHAN BARR | Associated Press CLEVELAND — High radiation levels recorded at a nuclear reactor in northeast Ohio have prompted a special inspection by the U.S. Nuclear Regulatory Commission. Workers immediately evacuated it on April 22 when radiation levels rose while it was shutting down for a refueling outage.. The highest radiation exposure to any of the workers was 98 millirems (= 1 millisievert), which is equivalent to two or three chest X-rays, a spokesman for the plant's owner said. The NRC's limit for radiation exposure in a year is 5,000 (=50 msv) millirems, he said.

from: http://www.fox8.com/news/wjw-perry-power-plant-special-investigation-radiation-levels-txt,0,1576630.story
FirstEnergy, that owns the plant, will conduct their own investigation but a spokesman says workers did not use proper methods for removal of a source range monitor and it appears the increased radiation levels were due to human erro

@@Pictures

Mark I schematic
http://www.beyondnuclear.org/home/2011/3/12/fukushima-dai-ichi-unit-1-reactor-schematic.html

Fukushima Dai-ichi Unit 1 reactor schematic

Description: http://www.beyondnuclear.org/storage/post-images/bwr-mk1i.jpg?__SQUARESPACE_CACHEVERSION=1299944680212

The schematic diagram above shows the GE Mark I Boiling Water Reactor reactor building structure, the Fukushima Dai-ichi Unit 1.

 

The explosion at Fukushima has apparently disintegrated the upper third of  the reactor building. The video and pictures currently available indicate that the "blow out panels" of the  reactor building and roof cover were blown away by an energetic explosion likely due to a hydrogen gas detonation. The reactor core refuelling deck and the surface of the elevated irradiated nuclear fuel pool are now exposed to the atmosphere. Essentially, the photos show the remaining steel I-beam structure for the weather cover that was over the refueling deck and the top of the "spent fuel" pool. These panels are designed to "blow out" at overpressure.

The actual "pressure suppression system" structures credited for containment sit below this structure inside the concrete reactor building, namely the drywell and wetwell or "torus." The drywell is the large inverted lightbulb steel structure which is 100 feet tall and a nominal wall thickness of 1.5 inches.  The reactor vessel sits inside this structure. In the event of a coremelt accident involving high pressure and high temperature, the highly radioactive steam and pressure would be vented into the drywell and then routed through the large diameter pipes to the "wet well" or "torus" which is the large 18 foot diameter hollow doughnut-shaped structure that surrounds the drywell. The torus contains approximately 1 million gallons of water and designed to receive the pressurized radioactive steam where it is supposed to be quenched and contained.

The status of the reactor containment in the reactor building remains unclear, but apparently remains intact. Fuel damage has apparently occurred because elevated levels of radioactive iodine and cesium are being monitored outside of reactor containment.

What is additionally unclear is how much cooling water is left in the fuel storage pools and whether or not there has been damage to irradiated fuel stored in that pool.  There are reports of sea water being brought in to cool this facility.

Radiation levels are reported to have fallen following the explosion. David Lochbaum, Senior Reactor Safety Engineer for the Union Concerned Scientists, has reported that the explosion may acutally have occurred in the turbine hall building adjacent to the reactor building.

An anonymous Japanese government spokesperson has attributed the explosion to the buildup of a combination of hydrogen and oxygen that detonated inside the concrete reactor building but not the crdedited containment structure. Industry reports that the containment structure itself was not compromised or breached by the explosion.

Mark II Reactor Building
http://www.nucleartourist.com/areas/auxbldg.htm
http://www.nucleartourist.com/images/rx-bldg1.jpg
Description: http://www.nucleartourist.com/images/rx-bldg1.jpg


@@Reactor Cap

link Here is the cap (cap of containment vessel) being lifted by the resident crane, but this may not be an actual cap of a GE Mark I design:
Description: http://media.tumblr.com/tumblr_li9d0wQNnc1qbnrqd.jpg

@@Onagawa

From: http://en.wikipedia.org/wiki/Onagawa_Nuclear_Power_Plant

[edit] Reactors on Site

Unit

Type

Start of Operation

Electric Power

Onagawa - 1

BWR

June 1, 1984

524 MW

Onagawa - 2

BWR

July 28, 1995

825 MW

Onagawa - 3

BWR

January 30, 2002

825 MW

[edit] Unit 1

Shut down manually on 25 February 2005 because it was determined that the reactor containment leaked small amounts of nitrogen. The unit was restarted once Nuclear and Industrial Safety Agency was satisfied that the countermeasures taken by the plant operator to prevent a reoccurrence were adequate.[3]

[edit] Unit 2

·         May 2006 it was confirmed that a pipe was leaking due to debris damage.

·         June 7, 2006 Difficulties with pressure control prompted further inspections.

·         July 7, 2006 METI and the Nuclear and Industrial Safety Agency determined that the plant's performance was not satisfactory.[citation needed]

[edit] Unit 3

·         July 7, 2006 Due to pipe integrity concerns the reactor was shut down.

·         November 25, 2006 Following repairs the reactor was restarted.

·         March 11, 2011 2011 Tohoku earthquake damaged the turbines after a fire broke out and was shut down.

[edit] Incidents

[edit] 2001

Small fire in the administrative offices. Did not affect functioning of the plant.

[edit] 2005

The Onagawa Nuclear Power Plant was affected by the 2005 Miyagi earthquake and recorded vibrations above what the plant was designed for. Analysis after the event, however, found no damage to the reactor systems. Some people reported seeing smoke come from the plant after the earthquake and reported it, thinking that it indicated an accident, but the smoke was actually produced by the backup diesel generators.[citation needed]

[edit] 2011

A fire from the turbine section of the plant following the 2011 Tohoku earthquake was reported by Kyodo News.[4]

On March 13, 2011, levels of radiation on site reached 21µSv/hour, a level at which Tohoku Electric Power Company were mandated to declare state of emergency, and they did so at 12:50, declaring the lowest-level such state. Within 10 minutes the level had dropped to 10µSv/hour.[5][6][7] The Japanese authorities assume the temporarily heightened values were due to radiation from the Fukushima I nuclear accidents and not from the plant itself.[8][9] On March 13 20:45 UTC, the IAEA announced that radiation levels at the Onagawa plant had returned to normal background levels.[8]

The three units remain in cold shutdown since the earthquake of 11 March. Two hundred people who lost their homes to the tsunami took refuge in the plant.[10] The April 7th aftershock damaged 2 of the 3 external power lines to the plant but cooling was maintained through the third line.[11]


@@Refueling

link Based on the news story this was part of, this is what things look like during refueling ops of a GE Mark I design (water level is brought up and a 'water' path made over to the fuel storage pool so the rods never 'see air'):

picture: Description: http://www.gb.ilneurone.com/wp-content/plugins/rssposter-pro-0.8.8/cache/2f0ce_article-1367524-0B3B48D800000578-106_472x754.jpg

@@Unit 1

http://sketchup.google.com/3dwarehouse/details?mid=11b70c1651c0b152f0725315050aa28f&prevstart=0

The reactor for unit 1 was supplied by General Electric. All construction was done by Kajima construction company. Unit 1 was a 439 MW boiling water reactor (BWR-3) constructed in July 1967. It commenced commercial electrical production on March 26, 1971, and was initially scheduled for shutdown in early 2011. In February 2011, Japanese regulators granted an extension of ten years for the continued operation of the reactor. At 15:36 JST on 12 March 2011 there was an explosion at Unit 1 following a major earthquake and tsunami.


 

@@Venting Hydrogen

VENTING HYDROGEN INTO REACTOR BUILDING A BAD IDEA
http://www.gereports.com/the-mark-i-containment-system-in-bwr-reactors/
Willem Jan Oosterkamp says: April 5, 2011 at 7:54 am "Venting hydrogen into the reactor building is not a sound technical solution. I assume that the containment has failed, barring better information on the Fukushima plants."