The United States Nuclear Regulatory Commission (NRC) typically begins its narrative on the “lessons learned” from the Fukushima Daiichi nuclear catastrophe with Japan’s March 11, 2011accident. Not surprisingly, the agency has avoided addressing the most critical lesson recognized in the accident’s official investigative report by Japan’s National Diet. In their finding, the unfolding radiological catastrophe is “manmade” and the result of “willful negligence” of government, regulator and industry colluding to protect Tokyo Electric Power Company’s financial interests. Likewise, here in the US, addressing identical reactor vulnerabilities remain subject to a convoluted corporate-government strategy of “keep away” with public safety as the “monkey in the middle” going back more than four decades and, for now, three nuclear meltdowns later.
In the latest development, by a 3-1 vote issued on August 19, 2015, the majority of the four sitting Commissioners with NRC ruled not to proceed with their own proposed rulemaking and bar public comment and independent expert analyses on the installation of “enhanced” hardened containment vents on 30 U.S. reactors. In the event of a severe nuclear accident, roughly one-third of U.S. atomic power plants currently rely upon a flawed radiation protection barrier system at General Electric (GE) Mark I and Mark II boiling water reactors that are essentially identical to the destroyed and permanently closed units at Fukushima Daiichi. The nuclear catastrophe has resulted in widespread radioactive contamination, massive population relocation, severe economic dislocation and mounting costs projected into the hundreds of billions of dollars.
Fundamentally at fault, the GE Mark I and Mark II boiling water reactor “pressure suppression containment system” designed for internalizing such a nuclear accident is roughly one-sixth the volumetric size of pressurized water reactor containment designs like Three Mile Island. Under accident conditions, the reactor pressure vessel and the operation of the emergency core cooling system is depressurized into the “drywell” containment component which in turn routes steam, heat, combustible gases and radioactivity into the “wetwell” component where it is supposed to be quenched and scrubbed in a million gallons of water. The GE design was first identified as too small to contain potential accident conditions in 1972 by Atomic Energy Commission memos. The internal communications would eventually be released years later under the Freedom of Information Act after more GE reactors were granted operating licenses. The memos revealed that the undersized containment system is highly vulnerable to catastrophic failure from over-pressurization in the event of a severe accident. This long recognized chink in GE’s “defense-in-depth” armor was graphically confirmed with the global broadcast of the Fukushima explosions.
Fukushima further demonstrated that “voluntary” GE containment modifications requested by NRC in the early 1990’s are not reliable under real accident conditions. Most U.S. Mark I operators voluntarily installed a hardened vent on the “wetwell” or “torus” containment component. The same modification was installed in Japanese reactors including Fukushima Daiichi. The voluntary containment modifications in the U.S. were carried out under a NRC regulation (10 CFR 50.59) that avoids licensee disclosures in the public hearing process, claiming that the design changes did not raise significant safety issues. Other than the paper trail, even the NRC inspectors were not aware of the final as-built containment modifications.
Following a review by the NRC Fukushima Lessons Learned Task Force commenced in 2011, the Commission ultimately issued a revised two-phase Order (EA-13-109) on June 6, 2013 for the unreliable containment systems. The Order requires all U.S. GE Mark I and Mark II boiling water reactor operators in Phase 1 to re-install a more “reliable” severe accident capable hardened containment vent on the “wetwell” component by June 2018 to assist in preventing core damage and remain functional during a severe accident. In Phase 2, operators are to install a severe accident capable hardened vent on the “drywell” component by June 2019, or alternately, provide a severe accident management strategy that precludes the need to vent an accident from the drywell. However, the Commission Order also rejected the collective research and technical recommendation of its lessons learned task force advanced on November 26, 2012. The staff recommendation was to require severe accident capable hardened containment vents be equipped with high-efficiency external filters as added defense-in-depth to trap much of the radiation while releasing to the atmosphere the extreme heat, high pressure and non-compressible explosive gases generated by the nuclear accident. The nuclear industry’s lobby group, the Nuclear Energy Institute (NEI) vehemently opposed the installation of external engineered filters arguing it was unnecessary to add any more assurances other than the original design calculations for a radiation scrubbing effect underwater in the “wetwell” component.
The Commission’s Order, while rejecting the staff’s recommendation to add the engineered filters in vent lines, instructed staff to pursue researching a proposed rulemaking to gather more expert stakeholder input on how radiation filtration strategies would improve public safety and reduce severe accident consequences and costs.
The nuclear industry, under the direction of NEI, began developing the alternate strategy to the Phase 2 installation of a “drywell” containment vent which they describe as Severe Accident Water Addition (SAWA) and Severe Accident Water Management (SAWM). Under the industry SAWA plan, in the event of a severe accident, water would be introduced into the reactor pressure vessel and/or the drywell containment control temperatures and cool core debris. If the drywell hardened containment is not installed, flooding up the “drywell” containment requires constant severe accident water management to prevent the simultaneous over-flooding of the “wetwell” that would preclude the use of “wetwell” hardened vent for containment protection as well as crediting the large volume of water’s scrubbing effect for radiation release reduction. Both SAWA and SAWM will require an unspecified number of operator manual actions performed in containment that require a range of apparatus (tools, respirators, keys, etc) as well as shielding from potentially high radiation fields.
The proposed NRC rulemaking to develop regulatory guidance for severe accidents and reconsider filtration strategies was renamed “Containment Protection and Release Reduction” (CPRR) and published as a draft informational document in May 2015. The proposed CPRR rulemaking introduced the staff analyses of four alternatives for consideration by the Commission, the public and industry stakeholders and the NRC’s own independent Advisory Committee on Reactor Safeguards (ACRS).
Alternative 1 adopted the status quo, namely, making the established Order EA-13-109 generically applicable, drop the rulemaking activity and take no further regulatory action. It adopts the Order’s Phase I requirement to install a severe accident capable vent on the “wetwell” containment component without an external engineered filter and instead relying upon the containment water to scrub out radioactivity before a release to the atmosphere. It adopts the NEI approach under Phase 2 to develop alternative capability for severe accident water addition and management to cool reactor core debris should it burn through the reactor vessel and end up on the bottom of the drywell;
Alternative 2 would pursue the rulemaking to make Order EA-13-109 generically applicable for protection of BWR Mark I and II containments against over-pressurization;
Alternative 3 would proceed with a public rulemaking while making Order EA-13-109 generically applicable for improved protection of Mark I and Mark II containment systems with severe accident water addition to the drywell but dropped any further consideration, analysis and public comment for external engineered filters for severe accident capable hardened containment vents;
Alternative 4 would pursue the rulemaking to protect containment against multiple failure modes and release reduction measures with controlled releases through hardened containment venting systems. This alternative would include making Order EA-13-109 generically applicable and require external water addition in the reactor pressure vessel or the drywell containment component. In addition, licensees would be required to further reduce fission product releases by maximizing wetwell scrubbing capability or the installation of an engineered filter in severe accident capable containment vent pathway.
The NRC staff’s draft information paper recommended the adoption of Alternative 3 to pursue a rulemaking to make the Order generically applicable without external engineered filters.
What happened when the information paper for draft guidance was sent up to the Commission is alarming. NRC Commissioner Svinicki took the NRC staff information paper and turned it into a notation vote paper for the Commission to decide, leaving no further consultation, no independent expert input even from the NRC’s Advisory Committee on Reactor Safeguards, the public and their independent experts. Commission Chair Stephen Burns, Commissioner Kristine Svinicki and Commissioner William Ostendorf voted to approve Alternative 1.
Commissioner Jeff Barran’s comments submitted with his minority vote to go forward with the rulemaking are illuminating.
“Consistent with the Commission’s direction, the staff prepared a draft regulatory basis for a containment protection and release reduction (CPRR) rulemaking. The draft regulatory basis was provided to the Commission as an information paper. Prior to the conversion of the information paper to a notation vote paper, the staff planned to issue a Federal Register notice requesting public comment on the draft regulatory basis, ‘hold a public meeting to provide members of the public an opportunity to ask questions and have discussions about the draft CPRR regulatory basis,’ and present the draft regulatory basis to ACRS in order to obtain independent expert feedback on the document. After this public comment period and ACRS review, the staff was scheduled to provide the final regulatory basis to the Commission by September 19, 2015, the proposed rule by September 19, 2016, and the final rule by December 19, 2017.”
“In my view, it is premature for the Commission to consider the draft regulatory basis at this time without the benefit of public comment or ACRS review. I approve the staff’s established plan, based on clear Commission direction, to seek public comment and ACRS review of the draft regulatory basis prior to its submission to the Commission for a notation vote. “
“Furthermore, there is no reason for the Commission to vote on the draft regulatory basis before the ACRS (Advisory Committee on Reactor Safeguards) has reviewed and provided recommendations on the document. Under the staff’s original schedule, the ACRS planned to hold a subcommittee meeting and provide a letter to the Commission after the staff received and addressed public comments on the draft regulatory basis. The staff should resume this course. Though the staff previously presented the draft results of the rulemaking analysis to the ACRS, this will be the first time the ACRS will examine the draft regulatory basis as a whole and share its thoughts with the Commission. We should wait for the ACRS letter before making substantive decisions about the draft regulatory basis.”
“This is an important post-Fukushima rulemaking. A wide range of stakeholders will have a variety of perspectives on the four alternatives presented in the draft regulatory basis. We should hear their views and critiques of these alternatives and the staff’s regulatory analysis before taking any alternatives off of the table. Therefore, consistent with existing Commission direction, the staff should carry out its plan to seek public comment and the ACRS review of the draft regulatory basis prior to submission to the Commission in the next few months for a notation vote.”
The Commission’s August 19th majority vote is effectively a gag order on the American public’s opportunity for formal input to fortify the continued operation of GE Mark I and Mark II reactors against the next nuclear catastrophe. Ironically, the international nuclear industry is simultaneously cashing in on the effort to restart Japan’s nuclear power plants where their Nuclear Regulation Authority has ordered state-of-the-art engineered external filters on severe accident capable hardened containment vents as a prerequisite to resume operation. On August 17, 2015, AREVA issued a press release announcing that it had just delivered it fourteenth filtered containment vent system to the Hamaoka Unit 4 reactor operated by Chubu Electric Power Company where 70% of the Japanese public no longer trust the industry and its regulator and remain opposed to any further nuclear power operations.
August 19, 2015 Commission Notation Vote Sheet
AEC memo September 25, 1972
November 26, 2012, SECY 2012-0157
AREVA press release]]>
A commissioner of the Nuclear Regulation Authority of Japan told reporters this week that the nation’s boiling water reactors will not be allowed to restart until they have installed filtered ventilation systems to reduce the amount of radioactive materials released from containment systems during nuclear emergencies.
“Without this (filtering systems), reactors will not reach the level” to be allowed to operate, said NRA Commissioner Toyoshi Fuketa.
The ruling affects 26 reactors, over half of Japan’s nuclear force, some of which already have ventilation systems installed without filters. Installing these systems will likely take more than three years and cost several billion yen.
In 2012, Tomoko Murakami, a Tokyo-based nuclear researcher at Institute of Energy Economics, told Bloomberg reporters his opinion of potentially requiring plants to install filtered vents, “We all know there is no such thing as perfect safety. The point is what criteria should be used to decide the restart of a reactor? I don’t think installing a filtered venting would be one of the criteria.”
Just last week, the United States Nuclear Regulatory Commission announced its recommendation to require those nuclear reactors to install filtered ventilation systems. NRC staff and industry experts estimated the cost of these upgrades to be between $15 million and $30 million.
In the United States, there are 31 BWR Mark I and Mark II units in operation, many Mark I units already have hardened vent systems installed and will only need to add the filtration systems, but 8 Mark II reactors will need to install both the hardened vents and filtration systems.
No U.S. reactors have filtered vents, Scott Burnell, a spokesman at the NRC told Reuters.
Other countries in Europe and around the world already require filtered containment systems, or are considering including them in their own response to the Fukushima nuclear disaster. A contract to recently install a filtered venting system at a plant in Romania was valued at about $48 million.
The nuclear industry in the United States has been pushing back against the costly upgrades. The Nuclear Energy Institute, the industry’s main lobbying group in the United States, estimated the value of such an investment was $0.
Source: NEI Guidance Document
Source: Kyodo News
The operator of the plant Nuclenor said, that the additional 153 million euros of taxes that the BWR nuclear reactor at Garona would incur in 2013 would “increase current economic losses to the point of sending Nuclenor into bankruptcy.” The reactor pressure vessel was constructed in 1966, and the plant was brought online in 1971.
Nuclenor said in order to keep the plant online, it would need to invest around 120 million euros, while also facing higher taxes.
The Nine Mile Point Nuclear Station is a two-unit nuclear power plant located in the Town of Scriba, approximately five miles northeast of Oswego, New York, on the shore of Lake Ontario. While both units onsite are General Electric BWRs, the Unit 1 reactor is one of the two oldest nuclear reactors still in service in the United States; New Jersey’s Oyster Creek Nuclear Generating Station is the other.
Thursday night, during normal operation, a leakage of nitrogen to maintain Primary Containment pressure within specification was noted at Unit 1 (General Electric BWR Mark 1) in excess of that allowed per Technical Specification 3.3.3.a (An overall integrated leakage rate of less than 1.5% by weight of the containment air per day (La), at 35 psig (Pac).).
This event required the containment system to be declared inoperable, and nuclear reactor to be manually scrammed and shutdown until the utility is able to ensure the safety of the containment system to ensure control of the release of radioactive material.
The primary containment is normally pressurized during reactor operation, and nitrogen used for inerting could leak out of the containment. An investigation of the containment leakage is in progress. Decay heat is being removed via shutdown cooling (SDC).
Nine Mile Point Unit 1 Primary Containment System
The Primary Containment Structure (PCS) consists of a drywell, a suppression chamber in the shape of a torus, and a connecting vent system between the drywell and the suppression chamber. It also includes valves and piping associated with the vacuum breaker system and the structural portions of primary containment penetrations.
The containment systems are designed to control and monitor the primary containment environment. The containment systems consist of the combustible gas control system, primary containment area cooling system, containment atmospheric monitoring system, torus temperature monitoring system, torus drain system, and the integrated leak rate monitoring system.
Containment Inerting System
The containment inerting system is used to inert and deinert primary containment and to makeup nitrogen as required to maintain low oxygen concentration and containment pressure. The containment atmosphere dilution system is designed to monitor and maintain the oxygen concentration of the primary containment atmosphere to less than four percent during a LOCA. Additionally, the system is an alternative system for venting the primary containment to the atmosphere, if necessary.
Previous containment degradation problems at Oyster Creek Nuclear Power Plant
The NRC has previously noticed that Mark 1 containments are vulnerable to degradation like that which occurred at Oyster Creek Generating Station as a result of water intrusion in the air gap from leakage past the refueling seal and subsequent wetting of the sand cushion at the bottom of the air gap.
Source: NRC Event Notifications]]>
Q385, Rev. 4, “Justification for recommended actions considering TEPCO desire to limit containment venting based on EPRI and BWROG Severe Accident Management Guideline Basis,” is attached. This version addresses:
– Clarifications provided by EPRI
– Clarifications from NRC’s Protective Measures Team (PMT)
– Sourcing information on H2/02 graph
– Comments from the 4-2-11 0300 teleconference with the site team and the industry
Please provide formal concurrence of your organizations equivalent to the level indicated on the RST assessment concurrence officials attachment. Target for completion is 4/2/11, 10:00
Pages From ML12102A207 – April 2nd, 2011 – Justification for Recommended Actions Considering TEPCO Desire t…]]>
The report is based on 600 interviews, inspections, and other data, and concludes that the tsunami was much larger than TEPCO had expected.
The company denied that it was taking too long to disclose data, “We did not mean to hide information, but there was a lack of enough explanation.”
The report notes that after the Unit 1 explosion, conditions on the site prevented workers from continuing to conduct emergency operations at Units 2 and 3, leading to the subsequent meltdowns at all three reactors. Prior to the Fukushima Daiichi nuclear disaster, the risk of a severe accident at one unit affecting emergency operations at another unit had not been considered.
Largely the report from the utility appeared content to attempt to justify its actions, still clinging to the defense that the utility “could not predict an occurrence of the event which was far beyond our expectation,” despite the fact that the utility and regulators knew the tsunami threat was serious, yet purposefully delayed any meaningful investigation or action.
TEPCO says there is still a need for studying what actions should be taken in the event that a nuclear reactor has lost all of its functions, and I would agree, it is not uncommon to find ourselves in situations where we not adequately prepared for the unexpected.
European regulators group Ensreg also says that operators must plan for the possibility that their chosen design may be flawed.
“The important lesson from Fukushima is that you might be wrong on your design basis,” Philippe Jamet, chairman of the peer review board for the EU’s nuclear safety stress tests said in a Platts article. “So you need an extra layer [of safety precautions] for when you are wrong on how your design basis copes with unpredictable external events,” he said.
The Fukushima disaster was not ‘unexpected’ as much as it was ignored, tsunamis of that magnitude had occurred in the past, TEPCO knew that a tsunami of only 10 meters would cause a potential loss of power, and they should have reasonably foreseen it to happen again.
The location of the plant was on a bluff which was originally 35-meters above sea level. During construction, however, TEPCO lowered the height of the bluff by 25-meters. The lowered height would keep the running costs of the seawater pumps low.
TEPCO’s analysis of the tsunami risk when planning the site’s construction determined that the lower elevation was safe because the sea wall would provide adequate protection for the maximum tsunami assumed by the design basis. However, the lower site elevation did increase the vulnerability for a tsunami larger than anticipated in design.
“Why are the pumps at water level?” one might ask. Because it is much easier and cheaper to push water from the base of the building then to suck it up from the top, thus pumps at water level.
An in-house study in 2008 pointed out that there was an immediate need to improve the protection of the power station from flooding, as a 10.2 meter wave could inflict damage to the generators. Officials of the department at the company’s headquarters insisted that such a risk was unrealistic and did not take the prediction seriously.
Three years later the report was sent to NISA, where it arrived on the 7 March 2011, just 4 days before the plant was hit by the tsunami.
I don’t know if I personally would be able to claim that I was unprepared if I knowingly designed a plant to withstand a magnitude 7 earthquake in a part of the world known for generating up to a magnitude 9, or if I had then continued to operate with a 6 meter tsunami protection wall, despite knowing that a 10 meter tsunami could cause a loss of power event, and also knowing there had been three 10 meter tsunamis in the last 100 years.
The requirements for shutting down a nuclear reactor after a earthquake or tsunami are well-known, the fact that equipment was not in place, provisions for alternate back up power, and pumping equipment in case of emergencies was not in place is attributable to politics and profit-based decisions.
Still, it never ceases to amaze me how inadequate the backup systems are for nuclear power plants. Look at the hydrogen explosions as an example, they demonstrate either fundamental lack of understanding for the capabilities for the generation of hydrogen, or a severe lack of hydrogen control, the devil is in the details.
In general, the scientific community is fairly knowledgeable when it comes to hydrogen and its potential for combustion. There has been a lot of research done not only within the nuclear industry, but within the scientific community as a whole, so it is hard to believe that it was due to a lack of understanding that if hydrogen levels accumulated there would be a potential for an explosion.
An inherent potential problem of the BWR is the fact that if the cooling system fails for whatever reason, the water in the fuel rod tank starts to over-heat, it gets even worse when the zirc-reaction moves from creating steam to reaching a critical state of separation of hydrogen and oxygen gases from the super-heated water once the temperature reaches 2000 degrees Celsius or 3632 degrees Fahrenheit.
This then creates an extremely hazardous condition when the hydrogen gas accumulates at the top, and the oxygen gathers at the bottom of the reactor. This can easily lead to gases being forced from inside the reactor, one scenario involves the reactor head lifting due to a release from the upper drywell head, others only require migration through any of the penetrations or equipment hatches.
Unless the hydrogen can be released into the atmosphere immediately there will be an explosion; ie Fukushima. The Three Mile Island workers were extremely lucky in that they were able to release the radioactive hydrogen gas into the atmosphere before it ignited thus avoiding the unthinkable.
So we have a good understanding of when the hydrogen is generated, where it may proliferate to (everywhere), how much is generated, and what happens next.
Around the world, industry workers and officials are haunted by the visuals of the Fukushima reactor buildings exploding like they did, for most, this was the first time that many of them had even considered the possibility. Prior to Fukushima, regulators and licensees had relied on hardened vents to prevent an explosive mixture from building up in the reactor building.
New data from Fukushima is confirming what some in the industry had already long feared, that venting might not be the best option. Some experts have long claimed that if workers wait to vent after the containment is full of hydrogen, due to the tendency of hydrogen to leak through various connections in the reactor building itself, and mix with air, it is likely that venting might not be able to eliminate the risk the hydrogen poses.
Hydrogen easily moves through places where we think it won’t, it is almost impossible to create a valid recreation through research alone without experimentation. With hydrogen you don’t even need an ignition source, even static electricity could set off an explosion.
This is a situation that regulators around the world didn’t really evaluate fully, and at this point, the information from Fukushima is not yielding any easy answers. There is a lot of information that has been brought up, leaking seals, core-concrete interactions, and yet no one in the regulatory or industry camps has any idea how to completely mitigate the hydrogen problem away.
With the money they make, all nuclear plants around the world should be forced to put in back-up systems for back-up systems in case of complete failure to prevent the release of radioactive materials.
The industry claims to rely on defense-in-depth, but even that falls woefully short if every possibility is not explored, if every ramification not identified. Ideally, every safety precaution should be taken and enforced to prevent the loss of life.
The Ensreg report likewise, recommended an EU-level assessment of national hazard margins, safety reviews at least every 10 years, containment integrity and “prevention of accidents while limiting all possible consequences.”
But simply put, there is no foolproof way of preventing nuclear disaster, there is no such thing as “leak-tight”. The industry spin-masters will always find a way to blame such things on “human error,” but the truth is, nuclear power is by its nature an unstable compound that is just waiting for the inevitable sequence of events that will reopen ‘Pandora’s Box’ time and again.
The one thing that we can all agree on is, it is going to be a very long time before we have any real comprehension of the Fukushima disaster, it will be immensely difficult to arrange the collection of critical information and consensus on mitigative actions, yet the industry does not see that in conflict in any way with current expansion projects.
A lot of this will be affected upon which opinions are voiced, which organizations are involved, who is paying who to do what, and ultimately where the interests lie and the incentives are.
Source: JiJi Press
Related Articles on Page 2…
A New Strategy for the Fukushima Daiichi Disaster
The Domino Theory (the Plausible Hypothetical Worst Case) is Not Any Longer Credibly Possible – Pages From ML12122A221
IAEA Activities in Response to TheFukushima Accident Report by the Director General
Science of Nuclear Safety Post-Fukushima
Nuclear Power in aPost-Fukushima World
Conditions of Fukush’Iima Dai-Ichi Nuclear Power Station Unit 1(as o6f 7 00 April 22nd, 2011) – Pages From ML12122A221-6
COMGBJ-11-0002 – NRC Actions Following the Events in Japan – Pages From ML12122A221-7
Evacuation of Workers Due to 100 R-Hr Fields at What Appears to Be the Unit 2 Turbine Hall – Pages From ML12122A221-2
Fukushima Exceeds Authority of Stafford Act – FEMA Not in Lead – Pages From ML12122A221-3
Meeting Agenda on Small Modular Reactors – Pages From ML12122A221-4
HHS State-Territory 2011 Pacific Basin Earthquake-Tsunami Conference Call – Pages From ML12122A221-5
White House Briefing Call – Pages From ML12122A221-6
Agenda for Industry Consortium Daily Call (NOTE TIME CHANGE to 2000 HRS DAILY) – Pages From ML12122A221-7
Japan Distribution List – Pages From ML12122A221-8
Meeting Minutes Form 4-1-11 2000 Industry Consortium Call – Pages From ML12122A221-9
FYI – You Might Have Seen Those High Resolution Images Already – Pages From ML12122A221-10
Commercial RADIACS From INPO – Pages From ML12122A221-11
Typical Reverse Osmosis Systems and Their Ability to Filter Outradionuclides – Pages From ML12122A221-3
REVISED – In Preparation for 2000 Hour NRC Consortium Call for Japan – Pages From ML12122A221-4
Cooling Fukushima Daiichi Reactors Through the Steel Head of the Drywell – Pages From ML12122A221-6
NRC Staff Speculates That Part of the Unit 2 Core May Be Out of the Reactor Pressure Vessel – Pages From ML12122A221-7
ML111940514 – Fukushima and the NRC’sResponseJuly July 13, 2011
Perspectives of A Commissioner in Times of A Historical EventNRC Commissioner William C. OstendorffApril 13, 2011
FYI – Information From Video Conference on 4 5 11 That May Be of Interest to the Consortium – Pages From ML12122A221
4-6-11 0300 EDT Call With Site Team – Pages From ML12122A221-2
We Need to Get a Better Understanding of the AMS Data From DOE – Pages From ML12122A221-3
TEPCO Daily Update Thursday April 7 – Pages From ML12122A221-4
ML12143A336 – Filtered Containment Venting Systems Advisory Committee on Reactor Advisory Committee on Reactor Safeguards Safeguards May 22, 2012
Fukushima Daiichi Reactor Diagrams
Ml12144a058 – Resilient Control for Critical Infrastructures And Systems
Reliability Estimation for a Digital Instrument And Control System
Ml12144a045 – Regulatory Commitment Issues Licensing Action Task Force May 10, 2012
English Translation of Investigations on Long Term Behaviour of Lead and Lead Alloys
From “Shelter” to New Safe Confinement
Davis-Besse Nuclear Power Plant
Ml12052a179 – Summary of Telephone Conference Call Held on July 19, 2011,, Concerning Requests for Additional Information Pertaining to The Davis-Besse Nuclear Power Statio
San Onofre Nuclear Generating Station
LER 2012-002-00, Unit 3 Steam Generator Tube Degradation Indicated ByFailed in-Situ Pressure Testing San Onofre Nuclear Generating Station (SONGS), Unit 3
LER 2012-002-00, Emergency Diesel Generator Vibration Trip Not By Passed For Non-Accident Conditions San Onofre Nuclear Generating Station (SONGS), Units 2 and 3
2011 Annual Radio Logical Envi Ronmental Operating Repor tSan Onofre Nuclear Generating Station Uni Ts 1, 2 and 3 And Independent Spent Fuel Storage Faci l i Ty
Related Articles on Page 2…
Subject: FW: JA Low Level Radioactive Waste review
Location: Telecon: 703-695-4042, Passcode: 869823
Start: Thu 4/7/2011 4:00 PM
End: Thu 4/7/2011 5:00 PM
Show Time As: Tentative
Meeting Status: Not yet responded
Organizer: Idar, Deanne J CIV OSD POLICY
For information only.
NRC will participate in a interagency call to discuss Low Level Radioactive Waste that may return to US from
—- Original Appointment—–
From: Idar, Deanne J CIV OSD POLICY
Sent: Thursday, April 07, 2011 1:54 PM
To: Idar, Deanne I CIV OSD POLICY; ‘[email protected]’; ‘Bentz, Julie A.’; ‘[email protected]’;
‘RMTPACTSUELNRC’; ‘RMTPACTSU_HHS’; ‘RMTPACTSUMLO’; ‘RMTPACTSU_SRO’; ‘[email protected]’; ‘NITOPS’; ‘Connery, Joyce’; PMT03 Hoc; ‘David Bowman’; ‘Mustin, Tracy’; ‘Szymanski, John’; ‘Zerr, Thomas J.’; ‘Regan, Sean P.’; ‘Bahar, Michael’; Komp, Greg R Mr CIV USA HQDA ASO; ‘[email protected]’; ‘Munning, Gregory A Capt Code 07, 07’; F (,• 6ý Aponte, Manuel COL OSD POLICY; Lane, Aikojean CIV OSD POLICY; Gross, Laura, CIV, 05D-POLICY; Malone, Stephen C CTR JCS J3; Owens, Janice; LIA03 Hoc; LIA02 Hoc; ‘Tilden, Jay’; Hoc, PMTI2; f, “–(: _’RMTPACTSU [email protected]‘; Farrand, David E SEA04 04N; ‘Steele, Jeffrey M CIV SEA 08 NR’; ‘[email protected]’; ‘[email protected]’; Curry, Michael R
Cc: ‘Smith-Kevern, Rebecca’; ‘McCaughey, Bill’; ‘McGinnis, Edward’; ‘Phillip J Finck’
Subject: IA Low Level Radioactive Waste review
When: Thursday, April 07, 2011 4:00 PM-5:00 PM (GMT-05:00) Eastern Time (US & Canada).
Where: Telecon: 703-695-1042, Passcode: 869823
Thanks to all that have responded. We have all 11 lines identified as follows.
EPA, DOS, and DOT – would you confirm whether or not someone could participate at 1600?
Updated Participant list follows:
I. DOE: Mr. Doug lonkay, LLW/MLLW team for Christine Gelles, DOE/EM HQ Office of Disposal Operations Office
2. DOE: Mr. Edward McGinnis, Nuclear Energy (NE)
3. DOE: Mr. Phillip Finck, INL
4. DOS: Ms. lanet Gorn (unconfirmed)
5. DOT: (Pending)
6. EPA: Mr. Dan Schultheisz (unconfirmed)
7. NRC: Ms. Janice Owens (unconfirmed)
8. NSS: Julie Bentz or Charles Miller
•9. DoD: Mr. Greg Komp, USA – HQDA ASO, DoD LLRW Disposition Advisory Committee Chair (unconfirmed)
10. DoD: OSD(P)/CBRN – Deanne 1. Idar’ and Laura Gross
II. DOD: Naval Reactors: Jeff Steele and Dave Farrand for Steve Trautman, Deputy Director at Naval Reactors
OSTP: John Szymanski
Thank you for your follow-up with POC names, and additional information. I have scheduled a 1600 EDT telecon today for us to review and identify next steps regarding the Low Level Radioactive Waste challenges.
Telephone: 703-(695-4042 / DSN
Lines available: 11 total
Facilitator: Deanne J. Idar
We are limited to only 11 lines available at that time, so l’ve identified slots for 8 representatives, with 3 lines remaining. For unidentified participants, please e-mail me if you will be unable to join us on the telecon.
For others, please also e-mail if you would like to join us on the call.
Address the 4 questions posed by the DoD LLRW Disposition Advisory Committee and identify next steps, as follows:
1. Will radwaste generated outside the hot, warm or plume zones be returned to Japan or treated as US generated waste?
2. Can this waste be declared 91b, “national defense” waste or do we need to treat as commercial low level radwaste?
3. If commercial, will we need import permit from NRC?
4. Can we access DOE disposal sites?
Unit 1 Rx: Shutdown 3/11. 70% core damage. Cooling with 30 gpm. Significant salt deposits in vessel, core spay plugged. Primary pressure 65 psig. Drywell pressure 25 psig. Secondary containment destroyed. Containment has been vented at least once since fuel damage occurred. Attempting to establish Nitrogen purge prior to resuming
Unit 2 Rx; Shutdown 3/11. 30% core damage. Significant salt deposits in vessel/drywell, Assumed RPV breach, with at least some core ex-vessel that ocurred approximately 3/15, Primary containment breached in the torus. Secondary containment breached. Significant release of volatile fission products has occurred through both airborne release and also via water drainage out of the Rx building,
Unit 3 Rx: same assumptions as Unit 2, but do not assume RPV failure and location of primary containment breach may be the drywell.
SFP 1: 292 bundles. Pool intact. All fuel at least 12 years old. No secondary containment, Rubble on top of pool. Water can be added through external spray. Now at saturation temperature.
SFP 2: 587 bundles. Pool intact, Water added to the point of pool over-flow, Pool had reached saturation temperature at one time.
SFP 3: 548 bundles. Y, core offload previous refueling. No checker boarding of hotter fuel. Structural damage to pool area suspected. Pool leakage possible. External addition of water has been made repeatedly, but flooding of pool may not be possible due to damage.
SFP 4: 1331 bundles. Full core offload about 120 days ago. No checker boarding of hotter fuel. Structural damage to pool area is known to exist, and structure may not support a full pool weight load. Pool leakage likely, requiring addition of water periodically. Pool was likely dry enough to have cladding/water reaction which produced enough hydrogen to lead to catastrophic explosion that destroyed secondary containment.
USAF bases at Misawa and Yokota began sampling (3X daily) on Monday, March 14th, 2011
USAF expanded the locations testing to more AFBs and some ANG installations in Korea(3X), Guam(3X), Alaska, Washington(IX daily), Oregon (1X), and California (1X) on Wednesday, March 16th, 2011.
We have seen some elevated readings at Yokota and Misawa.
AF Radiological Sampling Plan and Isotropic Information(FOUO) – Pages From C142015-02A]]>
Catawba did for a cycle or two back in 2006, I think.
TVA has expressed interest, and Columbia was recently revealed to be in discussions with DOE about it. But nobody has signed up or committed to using it.
From: Sheehan, Neil
Sent: Monday, March 28, 2011 3:05 PM
To: Burnell, Scott; Brenner, Eliot; Hayden, Elizabeth; Harrington, Holly; McIntyre, David
Subject: RE: Media question on plants using MOX fuel
Let me rephrase: Have any plants been testing the use of MOX or signed up to use it in the future?
From: Burnell, Scott
Sent: Monday, March 28, 2011 2:50 PM
To: Sheehan, Neil; Brenner, Eliot; Hayden, Elizabeth; Harrington, Holly; McIntyre, David
Subject: RE: Media question on plants using MOX fuel
From: Sheehan, Neil
Sent: Monday, March 28, 2011 2:49 PM
To: Brenner, Eliot; Hayden, Elizabeth; Harrington, Holly; Burnell, Scott; McIntyre, David
Subject: Media question on plants using MOX fuel
A question from a reporter is how many and which U.S. plants are using MOX fuel.
Does anyone have that information handy?
Pis follow up with RES (or leave a note for next Team Dose assessor to check the status of this request to RES Since it was sent on Thursday.
Ideally since RES completed the worst case runs yesterday or Friday, they can run this to verify the source term from RASCAL.
This is a non-urgent request. When RES is done with the previous request for a “pessimistic” case for Tokyo. We would like to have MELCOR calculate source terms that would be related to the scenarios used previous RASCAL runs that we had sent to NARAC.
Please let us know if this is something that you can do, and when we may see the source terms. To repeat, this is not urgent, but is something we would like to have in the operations center in order to determine if the 50-mile evacuation recommendation should be reduced.
Related Articles on Page 2…
0500 SPEEDI data, unzipped.
This email is UNCLASSIFIED
U.S. Embassy Tokyo
1-10-5, Akasaka 1-Chome, Minato-Ku, Tokyo 107
JACK: Yes, Mike, I don’t know that we’ve had anybody say that the fuel is covered with water.
What I can tell you is there’s clear evidence of a very significant hydrogen explosion. The only source of hydrogen that could feed that explosion is the spent fuel pool, so there must have been very, very high temperature zirconium interacting with water.
There is no visible vapors emanating from Unit 4 spent fuel pool area, which would be indicative of no water. It could also be indicative of a fully cooled core. That does not there is no source of cooling water going into the spent fuel pool, so to have a very significant hydrogen explosion, and then to think about the fuel being covered, those are kind of non sequitur concepts.
We do know that there were parts of debris, that the areas of debris around Unit 4 after the explosion which were contributing to very significantly high dose rates, and I understand that bulldozers were used to bulldoze that debris under some soil shielding, and the dose rates went down dramatically. That would be an indication that were
fragments of fuel since there’s no other source of substantial radioactive material which would have been involved in that explosion.
So, there’s indication of a very significant hydrogen explosion. I want to make sure that it’s clear that we don’t know this. We are just interpreting this from the visual evidence that we have, as well as the radiological measurements.
There’s evidence of a very significant hydrogen explosion. There’s evidence of fuel or some very highly radioactive material outside of the building after that explosion. And there’s no evidence of water vapor, which would tell us that the spent fuel pool is dry.