Post # 74: The Canary, The Ostrich, and the Black Swan

I've been pretty busy lately.  I just returned prior to Thanksgiving from the American Nuclear Society's Winter Meeting in San Diego California, where I participated in the embedded International Meeting on Severe Accident Assessment and Management: Lessons Learned from Fukushima Daiichi.  As I mentioned in Post # 73, I presented a paper, "The Canary, The Ostrich, and the Black Swan – An Historical Perspective on Our Understanding of BWR Severe Accidents and Their Mitigation."  The full paper (22 pages with over 60 references) is contained in the proceedings of the meeting (available from the ANS).  Here are the viewcharts from my presentation...

    Canary Ostrich Black Swan SLIDES 13Nov12     

Cheers!

Sherrell

Post # 73: The Canary, The Ostrich, and The Black Swan

November is an important month for those interested in the nuclear power industry's response to the March 2011 accident at  Fukushima Dai ichi.  The  American Nuclear Society's Winter Meeting will be the setting for the embedded "International Meeting on Severe Accident Assessment and Management: Lessons Learned from Fukushima Daiichi".  This four-day event will bring together experts from around the world to discuss the current status of the Fukushima Dai ichi plants, what we have learned / are learning from Fukushima, and plans for future activities related to Fukushima.  Specific topics to be covered include: severe accident phenomenology, severe accident sequence progression, severe accident management and mitigation strategies, severe accident simulation, and a number of other very relevant topics.

On Monday afternoon, Nov. 12, I will be presenting my recently-completed paper entitled, "The Canary, the Ostrich, and the Black Swan: An Historical Perspective on Our Understanding of BWR Severe Accidents and Their Mitigation."  This twenty-two page paper (with some 65 technical references) is my attempt to chronicle the evolution of our knowledge of commercial boiling water reactor severe accidents since the landmark Reactor Safety Study (WASH-1400) in 1975.  Here's the abstract of the paper...

" Between 1980 and 1995, Oak Ridge National Laboratory (ORNL) was engaged in an intense effort to understand commercial boiling water reactor (BWR) severe accident phenomenology, severe accident progression, and the potential role of the reactor operator in severe accident mitigation. This paper presents a summary of the major findings and conclusions from that period. Both detailed accident- and plant-specific results are discussed.  The author, who was a member of the ORNL research team who performed the work, offers an historical perspective on lessons learned, lessons ignored, and lessons forgotten from that period. The relevancy of these findings in the post-Fukushima world is addressed.  Finally, the author discusses the evolution of the current risk-informed regulatory framework; and identifies some key questions to be addressed, and critical steps to be taken to inform the development of the new nuclear safety construct required in the wake of the Fukushima Dai-ichi accident." 

The title of the paper is, as you might guess, is inspired by the peculiar characteristics of three birds:

The Canary – which once served as an "early warning system" to miners of dangerous conditions associated with carbon monoxide and other dangerous gases;

The Ostrich – which is (incorrectly) known for sticking its head in the sand to avoid obvious imminent danger; and

The Black Swan – the symbol adopted in recent times for a major event which is deemed a surprise to virtually everyone when it occurs, but after its occurrence, is viewed as something that could/should have been expected or predicted.

The paper was motivated by my conviction that accidents like that which occurred at Fukushima Dai ichi are unacceptable (regardless of their cause) and preventable – if the nuclear industry truly commits itself to going beyond the expedient in it's response to the accident.

The full paper will be published in the Proceedings of the meeting and available on CD-ROM at the meeting and thereafter from the ANS. 

Post # 72: NTTF Recommendation 2

Earthquakes and Floods

 Available evidence now indicates that even though the Great East Japan Earthquake of March 11, 2011 was estimated to be a magnitude 9.0 earthquake (exceeding the design basis of the Fukushima Dai ichi plant), the direct damage to the plant from the quake was minor compared to that resulting from the large tsunami that struck the plant some 40 minutes later.  (The plant was designed to weather an 8 meter high tsunami wave.  However, the maximum height of the waves that struct the plant exceeded 14 meters.)  The wave impact damage and the flooding resulting from the multi-tsunami waves that struck the plant accounted for the majority of the damage that lead directly to the ensuing station blackout and, ultimately, to the severe accident. 

And if the Fukushima experience was not sufficient to spawn an international re-look at commercial nuclear power plant vulnerability to seismic and flooding events, the urgency of the matter was heightened in June 2011 due to the partial flooding of the Fort Calhoun Nuclear Station in Fort Calhoun, Nebraska.  That event occurred due to historic flooding of the Missouri River following heavy rain falls and rapid snowpack melting in the Rocky Mountains. (The Fort Calhoun station is still shutdown.)

It was with all of the above as background that the NRC's July 2011 Near Term Task Force (NTTF) report presented Recommendation 2:

NTTF Recommendation 2

The Task Force recommends that the NRC require licensees to reevaluate and upgrade as necessary the design-basis seismic and flooding protection of SSCs for each operating reactor. 

This overall recommendations was further expanded in the NTTF report ...

The Task Force recommends that the Commission direct the following actions to ensure adequate protection from natural phenomena, consistent with the current state of knowledge and analytical methods. These should be undertaken to prevent fuel damage and to ensure containment and spent fuel pool integrity:

2.1 Order licensees to reevaluate the seismic and flooding hazards at their sites against current NRC requirements and guidance, and if necessary, update the design basis and SSCs important to safety to protect against the updated hazards.

2.2 Initiate rulemaking to require licensees to confirm seismic hazards and flooding hazards every 10 years and address any new and significant information. If necessary, update the design basis for SSCs important to safety to protect against the updated hazards.

2.3 Order licensees to perform seismic and flood protection walkdowns to identify and address plant-specific vulnerabilities and verify the adequacy of monitoring and maintenance for protection features such as watertight barriers and seals in the interim period until longer term actions are completed to update the design basis for external events.

The term "SSC" in the above language refers to structures, systems, and components – the basic building blocks of a nuclear power plant.  The basic idea behind Recommendations 2.1-2.3 is to ensure that (regardless of the original seismic and flooding design basis for a commercial nuclear power plant) the most current and accurate data, information, and methods are used to re-examine the plant's vulnerability to such events.

Subsequent Actions Relevant to Recommendations 2.1 & 2.3

In October 2011, Recommendations 2.1 and 2.3  were designated in SECY-11-0137 as "Tier-1" recommendations – "... recommendations which the staff determined should be started without unnecessary delay and for which sufficient resource flexibility, including availability of critical skill sets, exists."

The NRC subsequently issued a Request for Information (the so-called "50.54(f) Generic Letter) to its nuclear power plant licensees on March 12, 2012.  The letter requested a two-phase re-evaluation of seismic and flooding hazards at each plant.  (The letter also addressed Recommendation 9.3, but I'll discuss that in a future posting.)   The Phase I re-evaluation was to consist of a re-look at the seismic and floods hazards at each plant based on the latest seismic and flooding hazard information (expected frequency and magnitude of potential earthquakes and floods) and present-day regulatory guidance and methodologies for evaluation of these hazards.  The NRC noted in the request that the Phase I evaluations would not revise the design basis of the plant (the technical analysis under which it was originally licensed).  Phase II of the hazards evaluation will follow after the NRC reviews the Phase I analyses and decides what, if any, regulatory actions (including plant and procedure modifications) will be required.   The Generic Letter, together with its enclosures and attachments, spell out in great detail the precise nature of the information each licensee is to provide the NRC.  Each licensee has 1.5 years to submit a comprehensive written response to the re-evaluation request.

But the NRC doesn't intend to sit on its hands for 1.5 years while it awaits the licensees' response to the Phase I / Phase II re-evaluation information request.  In addition to the information discussed above, the Generic Letter also requested that each licensee conduct seismic and flooding "walkdowns" to "identify and address plant specific degraded, nonconforming, or unanalyzed conditions and verify the adequacy of strategies, monitoring, and maintenance programs such that the nuclear power plant can respond to external events."  Licensees were given 120 days to confirm their intent to use an NRC-endorsed walkdown procedure or to provide a description of their preferred walkdown process.  Following NRC's approval of the licensee's 120-day submittal, each licensee was afforded another 180 days to complete the walkdowns and submit their results to the NRC.  Thus all the walkdowns are to be complete and their results submitted to the NRC no later than 300 days from March 12, 2012 – or January 2013.

In the months since the March 2012 release of the Generic Letter, the nuclear power industry has devoted tremendous effort to excuting the requested walkdowns and documenting their results.  (The Generic Letter noted that the NRC had estimated each licensee would expend 13,300 hr of effort to conduct their evaluations and prepare their response.)

So we are coming to the end of the allotted time for execution of the requested seismic and flooding walkdowns and submission of the results.  In addition, the licensees are preparing for the Phase I hazards analysis.  And if that is not taxing enough, many plants are undergoing their regularly-scheduled refueling and maintenance outages in the coming weeks and remaining months of 2012.

Bottom line... the accident at Fukushima has spawned a wholesale re-examination of the ability of U.S. commercial nuclear power plants to withstand extreme seismic and flooding events.  While some important near-term actions are coming to completion, significant efforts will be expended by the NRC and the Industry over the next couple of years to more fully comply with NTTF Recommendation 2.

Post # 71: NTTF Recommendation 1

So let's dig in to the NRC's Near Term Task Force (NTTF) recommendations as documented in the NTTF Report:

Recommendation 1.  "The Task Force recommends establishing a logical, systematic, and coherent regulatory framework for adequate protection that appropriately balances defense-in-depth and risk considerations.

1.1 Draft a Commission policy statement that articulates a risk-informed defense-in-depth framework that includes extended design-basis requirements in the NRC's regulations as essential elements for ensuring adequate protection

1.2 Initiate rulemaking to implement a risk-informed, defense-in-depth framework consistent with the above recommended Commission policy statement.

1.3 Modify the Regulatory Analysis Guidelines to more effectively implement the defense-in-depth philosophy in balance with the current emphasis on risk-based guidelines.  The Task Force beleives that the Regulatory Analysis Guidelines could be modified by implementing some of the concepts presented in the technology-neutral framework (NUREG-1860) to better integrate safety goals and defense-in-depth.

1.4 Evaluate the insights from the IPE and IPEEE efforts as summarized in NUREG-1560, "Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance," issued December 1997, and NUREG-1742, "Perspectives Gained from the Individual Plant Examination of External Events (IPEEE) Program, " issued April 2002, to identify potential generic regulations or plant-specific regulatory requirements."

There's a lot nuclear safety and NRC jargon in the text I just quoted from the NTTF report.  I could spend several posts drilling down into the nuances of these recommendations and their potential implications.  Of all the recommendations presented by the NTTF, this is the one that has the greatest potential long-term impact for the commercial nuclear power industry.  I think of it as the "Super Recommendation"  I'll limit myself here to what I consider to be the most important high-level implication of Recommendation 1: Increased uncertainty and angst in the commercial power industry...

While Recommendation 1 and its sub-recommendations do not go so far as to state the current U.S. NRC regulatory framework is fatally flawed, the recommendation clearly communicates the NTTF's position that the current regulatory framework does not adequately integrate traditional deterministic reactor safety design approach (such as the use of design basis accidents, Design Criteria, "defense-in-depth", etc.) and the risk-based or probabilistic safety considerations (core melt probability, large early release LER probabilities, expected public health impacts, etc.) that have evolved in the post-TMI-2 era.

This challenge is not new.  During the past twenty years, the NRC has evolved toward increased usage of risk-based criteria in the examination of the requirement of plant backfits, conduct of plant maintenance, and in the articulation of risk-based safety-goals. Still it is the fact that the accident at Fukushima Dai-ichi did NOT violate current NRC public safety goals (because they are expressed solely in terms of expected public fatalities as a direct result of an accident, and there were no such fatalities at Fukushima). This fact obviously raises the question of whether the NRC's currently risk-based public safety goals are sufficiently comprehensive to protect society from Fukushima-like accidents.  (For instance, many individuals from within and outside of the nuclear industry are now advocating the addition of additional risk-based public impact criteria – such as land contamination.)

The NRC staff is currently considering Recommendation 1 and options for response to it.  We have been told to expect a "SECY" paper early in calendar year 2013.  That paper will  provide an initial roadmap for a (no-doubt long-term) re-examination and revision of the fundamental regulatory framework of commercial nuclear power plants in the U.S.

The potential implications of Recommendation 1 are huge.  The fundamental design of future nuclear power plants, the siting options and policies for new nuclear power plants, and the day-to-day operations of existing and future nuclear power plants might be affected in ways we cannot predict at this moment.

All of this comes, of course, at a time when the nuclear industry is laboring mightily to respond to the suite of "Tier-1" recommendations the NRC identified in SECY-11-0124 as requiring immediate or near-term action.   I'll talk more about Tier-1 recommendations in my next post.

Cheers,
Sherrell

Post # 70: The Structure of the U.S. Fukushima Response

With this post I begin the promised series in which I hope to provide a structured and concise chronicle of the U.S. nuclear industry's ongoing response to the Fukushima Dai-ichi accident and the the evolving "lessons-learned" from it.

In the wake of the Fukushima Dai-ichi incident, the U.S. Nuclear Regulatory Commission (NRC) convened a "Near-Term Task Force" or "NTTF" to examine the events at Fukushima Dai-ichi and offer recommendations regarding actions the NRC should take to enhance commercial nuclear power safety in the U.S. in the light of evolving lessons learned from the event.  The NTTF report,   "Recommendations for Enhancing Reactor Safety in the 21st Century – the Near-Term Task Force Review of Insights From the Fukushima Dai-ichi Accident," (SECY-11-0093, 12 July 2011) concluded there was imminent danger from continued operation of U.S. nuclear power plants, but did present "Twelve Recommendations" for actions the NRC and its licensees should take to further enhance the safety of commercial nuclear power.  

Following issuance of the NTTF report, the NRC Commission asked (SRM-SECY-11-0093) the NRC Staff to examine the NTTF's Twelve Recommendations, and to prioritize them in a logical manner based on the urgency of required actions, and the inter-relationships and inter-dependencies of the various issues.  The NRC Staff's initial recommendations regarding the subset of actions requiring the most urgent action were presented in SECY-11-0124.  Subsequently,  SECY-11-0137 presented the NRC Staff's recommendations for a three-tired prioritization.  This three-tiered hierarchy was accepted by the NRC Commissioners, and became the organizational foundation of the U.S. response to the accident at Fukushima.

Before I continue, I should point out that, as NRC Chairman Macfarlane said a few days ago in her remarks at the IAEA in Vienna, the current structure of the NRC's and the Industry's response to Fukushima is likely to evolve away from the simple "12 Recommendations" approach presented in the Near-Term Task Force (NTTF) report.  Here are Chairman Macfarlane's words,

" As we move forward in the evolution of our nuclear safety culture, we must address the fact that the majority of post-Fukushima activities were placed in special categories. In the period immediately following Fukushima, this approach made the most sense. In the United States, for example, we established a task force to address the impacts of the accident on our domestic program, and then a special Fukushima-related directorate to implement the ensuing recommendations. The NRC is now beginning to transition these Fukushima lessons-learned programs from special, segregated actions back to the offices that handle these matters on a routine basis. Far from minimizing these activities’ importance, this approach will ensure that the lessons we have learned are fully integrated into our regulatory work in the United States. We believe that by weaving the lessons learned from Fukushima into nearly all of our regulatory activities, we are ensuring their long-term sustainability, and encourage our international colleagues to do the same." 

So... now for the Twelve Recommendations as presented in SECY-11-0093....

1. The Task Force recommends establishing a logical, systematic, and coherent regulatory framework for adequate protection that appropriately balances defense-in-depth and risk considerations

2. The Task Force recommends that the NRC require licenses to reevaluate and upgrade as necessary the design-basis seismic and flooding protection of structures, systems, and components for each operating reactor

3. The Task Force recommends, as part of the longer term review, that the NRC evaluate potential enhancements to the capability to prevent for mitigate seismically induced fires and floods

4. The Task Force recommends that the NRC strengthen station blackout mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events

5. The Task Force recommends requiring reliable hardened vent designs in boiling water reactor facilities with Mark I and Mark II containments

6. The Task Force recommends, as part of the longer term review, that the NRC identify insights about hydrogen control and mitigation inside containment or in other buildings as additional information is revealed through further study of the Fukushima Dai-ichi accident

7. The Task Force recommends enhancing spent file pool makeup capability and instrumentation for the spent fuel pool.

8. The Task Force recommends strengthening and integrating onsite emergency response capabilities such as emergency operating procedures, severe accident management guidelines, and extensive damage mitigation guidelines.

9. The Task Force recommends that the NRC require that facility emergency plans address prolonged station blackout and multiunit events.

10. The Task Force recommends, as part of the longer term review, that the NRC pursue additional emergency preparedness topics related to multiunit events and prolonged station blackout.

11. The Task Force recommends, as part of the longer term review, that the NRC should pursue emergency preparedness topics related to decision making, radiation monitoring, and public education.

12. The Task Force recommends that the NRC strengthen regulatory oversight of licensee safety performance (i.e., the Reactor Oversight Process) by focusing more attention on defense-in-depth requirements consistent with the recommended defense-in-depth framework.

Things are a bit more complicated than this simple list, because each of the Twelve Recommendations (except for # 3)  were parsed into two or more sub-recommendations (e.g. 1.1, 1.2, 1.3, and 1.4).   Those seeking to monitor U.S. progress in addressing the Twelve Recommendations find their task complicated further because the three-tier prioritization of the Twelve Recommendations was done at the sub-recommendation level rather than at the top level. Thus, Recommendation 5.1 is a Tier-1 priority, while Recommendation 5.2 is a Tier-3 priority.

I will end this post here.  Future posts will discuss the sub-recommendations for each of the Twelve Recommendations, the prioritization of the sub-recommendations, actions taken to date by the U.S. NRC and the nuclear industry to address each sub-recommendation, future directions for continued progress, etc.

Again, my goal in this series of posts is not to provide an exhaustive review of all the safety-related activities in the industry.  I'm simply attempting to "status" U.S. progress in the key areas identified by the NTTF in a structured, clear, and simple manner that enables the non-expert to understand and track the post-Fukushima evolution of U.S. commercial nuclear power safety.

Cheers!
Sherrell

Post # 69: Monitoring the U.S. Response To Fukushima

This blog is about sustainable energy.  Over the past couple of years I've discussed a wide variety of topics that relate to sustainable energy production and use.  I've frequently noted my conviction that nuclear energy has to be the foundation of any sustainable AND ABUNDANT energy future for this small blue planet.

Those of you who have followed me here for any length of time also know I've spent much of my career in the nuclear reactor safety arena.  More specifically, I spent many years working with my colleagues at Oak Ridge National Laboratory (ORNL), the other national laboratories, and the commercial nuclear industry to improve our understanding of severe accident phenomenology, severe accident progression, and severe accident management strategies in commercial boiling water reactors (BWRs).  Along with all of you, I was deeply saddened by the events of March 2011 in Japan and at the Fukushima Dai-ichi plant.   Several of my previous blogs have dealt with BWR severe accident phenomenology and the events at Fukushima.

As you can imagine, I am closely following the post-Fukushima response of the global commercial nuclear industry and regulatory agencies worldwide.  In fact, at EnergX, we're part of that response.  We've assembled an incredibly talented team of nuclear reactor safety and risk experts – some of whom have been heavily involved in the beyond-design-basis accident and severe accident research and regulatory arenas for over forty years (predating the 1975 Reactor Safety Study (WASH-1400).  Members of our team were on-site during and following the accidents at TMI-2 and Chernobyl.  They served on the Advisory Committee on Reactor Safeguards (ACRS), and they led both industry and national laboratory research efforts for the decades following these accidents.  And, like some of you, our team is engaged in the industry's effort to learn from and respond to the Fukushima Dai-ichi accident.

So it is with this background I've decided to initiate a series of updates here to discuss the status of the U.S. industry's and the U.S. Nuclear Regulatory Commission's response to Fukushima Dai-ichi.  I will attempt to strike a balance between technical detail and clarity so that you do not have to be a nuclear engineer, or a nuclear regulatory expert to following along.  My goal in doing this is not to be an evangelist for the nuclear industry.  The industry needs no evangelist.  Rather, my goal is to chronicle as accurately and clearly as possible how the events at Fukushima are changing the present and the future of nuclear power in the U.S. and abroad.

So visit me regularly in the coming weeks and months to following the unfolding story of how the U.S. commercial nuclear power sector is learning from, and responding to the events at Fukushima.

Oh... and not to worry... I don't plan to entirely abandon the broader sustainable energy topic.  I will continue to post on non-Fukushima and non-nuclear energy matters as my interests and evolving events warrant.

Cheers,

Sherrell

Post # 68: Update From the ANS Utility Working Conference

I just returned from the American Nuclear Society's Utility Working Conference in Hollywood, Florida.  This 3-day annual event is focused on the commercial nuclear power industry and the needs of the commercial nuclear power plant owner/operators.  My wife, Rebecca (who is a mechanical engineer) and I staffed the EnergX booth at the vendor's exhibition.

As you might imagine, several of this year's working sessions spotlighted, the evolving industry response to the accident at Fukushima Dai ichi.

Based on the Nuclear Regulatory Commission's early implementation actions on the Near Term Task Force's twelve recommendations (and the "Tier-1" recommendations in particular), the industry is heavily engaged in implementation of the so-called "FLEX" initiative to pre-position emergency response equipment and resources, and in plant walkdowns and other actions related to identification of seismic and flood vulnerabilities.  Additionally, there is much anticipation related to the NRC's Notice of Intended Rulemaking regarding Station Blackout.

One of the central and repeated messages from the conference sessions I attended, was the short supply of experienced engineers and related technical specialists.  There's a shortage of experienced specialists skilled in conducting plant walkdowns, performing safety and risk assessment analyses, and performing plant modifications.  It is clear the industry will face significant challenges in meeting the human resource needs related to timely and adequate response to evolving Fukushima lessons-learned.

Gives the 50's to 60's age group (like me) who lived through TMI-2 and it's aftermath an opportunity to contribute the the industry we all feel is so vital to the long-term interests of humanity.

Cheers,

Sherrell

Post # 67: John Rowe & Mike Simpson On The Nuclear Renaissance

Last week I attended the annual meeting of the American Nuclear Society, held this year in Chicago.  (Before I forget to mention it, let me say how impressed I was with the beauty of Chicago's downtown waterfront / river walk district.  And I'm definitely not a big-city type of guy...)

Those of you who are regular readers of this blog know I am a pro-nuclear energy advocate.  I'm absolutely convinced that access to affordable and reliable electricity is the chief determinant of the quality of life for our fellow Earth-dwellers.  I am, therefore, distressed that billions of people have little or no access to electricity.  I'm also a pro-environment advocate.  These two convictions lie at the foundation of my belief that nuclear power is a key to a sustainable planet.

Now back to the meeting...  One of two "celebrity speakers" in the opening plenary session was John Rowe.  Rowe, in addition to being the retired CEO of Exelon (the largest market-cap electric utility in the country), is the past chairman of both the Nuclear Energy Institute (NEI) and the Edison Electric Institute (EEI).  His message to the assembled group (which appeared to me to number well in excess of a thousand attendees) was basically that the nuclear power industry has to face facts – stare the dragon in the mouth as I would say.  Rowe had two basic points:

1.  Cheap natural gas will probably be with us for at least 10 years - perhaps much longer.  There will be no nuclear renaissance while this is the case.  (Because most utilities will, of course, opt for combined-cycle gas turbine power plant additions.)

2.  When the nuclear renaissance does happen, the nuclear power plant of choice will have to be much simpler than present-day systems and much more passive.  And it will likely be a small modular reactor (SMR).

Another "celebrity" speaker at the plenary was Congressman Mike Simpson (R – ID).  Congressman Simpson, along with Sen. Lamar Alexander (R – TN) is a leading advocate of common-sense energy policies that include a major role for nuclear power and a balanced energy generation portfolio approach.  Congressman Simpson's basic message was, "Congress is broken, and I don't know how to fix it" - my translation, not his precise words.)  He also pointedly bemoaned a lack of consistency in DOE policies from administration to administration, a lack of specificity in DOE budget requests, and the challenges of staying on program across Congressional shift changes (every 2 years in the House and 6 years in the Senate).

What is one to think in the face of these words from two informed, pro-nuclear, pro-energy leaders?

My thoughts...

1.  Natural gas will not stay cheap.  Several companies are working around the clock to build the infrastructure required to export liquified natural gas (LNG).  When this is accomplished, our domestic natural gas will become a "world supply", and its price will move to world market prices.  That is, it's price will increase significantly above current domestic natural gas prices.  I continue to be concerned about the practice of fracking – both in terms of it substantial use of ground water, and the potential of fracked oil and gas wells to leak into and contaminate ground water aquifers.  It is for this reason some countries forbid its use.  A single incident in the U.S. in which a major aquifer is contaminated would evoke major regulatory changes in the fracking business.  I sincerely hope this never happens, but it could.  Even if gas does stay cheap, prudence and experience should motivate the pursuit of mixed energy generation portfolios to avoid the "all the eggs in one basket" vulnerability.

2.  With regard to Congress, I'm as lost as Congressman Simpson in attempting to identify solutions.  Unlike some folks, I do not believe Congress is The Problem.  I believe Congress is a manifestation of The Problem.  Our elected officials are, I believe, representing the views of the folks who elected them. (That's the great thing about a republic.)  So the breakdown in Congress actually mirrors, "The Problem" – a breakdown of the core shared values & world views that have guided our country since 1776.  Choose any major issue of the day – social, economic, energy, defense, etc.  Our nation is clearly split, divided, fractured far beyond anything I've seen during my lifetime.  This worries me because as a (very) amateur student of history, my read is that nations that manifest these characteristics tend to have two destinies.  Either (a) they continue on a downward spiral of disintegration into the faded pages of history, or (b) they are reunited by some external threat that so endangers their existence as to re-set their collective national views and values. The threat can be economic, militaristic, health & welfare, etc.  Neither of these two scenarios are attractive and I hope and pray they do not occur.  But history is history.  Oh... and I've also become a believer in term limits for all elected officials.  Politics should not be a lifetime career.  It should be a citizen-service.  But more about that at another time...

Enough for today.  I've got real work to do :)

Sherrell

Post # 66: ORNL Light Water Reactor Safety Publications

I've been asked by some to present a more complete listing of Oak Ridge National Laboratory's historical reports on light water reactor safety (especially boiling water reactor safety) research.  Luckily, before I retired from ORNL last year I undertook to compile such a list for my own reference use.  So, with gratitude to Ms. Linda Dockery of ORNL, who spent quite a few hours at the keyboard creating this list,  I present an incomplete bibliography here.  One thing of interest to note.  If my count is correct, the ORNL team published almost 250 publications between 1980 and 1995 alone.
Quite a productive group !

ORNL LIGHT WATER REACTOR SAFETY PUBLICATIONS – 
AN INCOMPLETE BIBLIOGRAPHY COVERING THE PERIOD 1975 – 2010

1975

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1975, A. P. Malinauskas, R. A. Lorenz, M. F. Osborne, J. L. Collins, and S. R. Manning, ORNL-TM-5021, Oak Ridge National Laboratory, September 1975.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period July–September 1975, A. P. Malinauskas, R. A. Lorenz, M. F. Osborne, J. L. Collins, and S. R. Manning,ORNL-TM-5143, Oak Ridge National Laboratory, November 1975.

1976

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period October–December 1975, R. A. Lorenz, J. L. Collins, and S. R. Manning, ORNL-TM-5290, Oak Ridge National Laboratory, March 1976.

Knudsen Cell-Mass Spectrometer Studies of Cesium-Urania Interactions, J. L. Collins, M. F. Osborne, A. P. Malinauskas, R. A. Lorenz, and S. R. Manning, ORNL/NUREG/TM-24, Oak Ridge National Laboratory, June 1976.

Behavior of Iodine, Methyl Iodide, Cesium Oxide, and Cesium Iodide in Steam and Argon, R. A. Lorenz, M. F. Osborne, J. L. Collins, S. R. Manning, and A. P. Malinauskas, ORNL/NUREG/TM-25, Oak Ridge National Laboratory, July 1976.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period January–March 1976, R. A. Lorenz, J. L. Collins, S. R. Manning, and A. P. Malinauskas, ORNL/NUREG/TM-30, Oak Ridge National Laboratory, July 1976.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1976, R. A. Lorenz, J. L. Collins, S. R. Manning, O. L. Kirkland, and A. P. Malinauskas, ORNL/NUREG/TM-44, Oak Ridge National Laboratory, August 1976.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period July–September 1976, R. A. Lorenz, J. L. Collins, and O. L. Kirkland, ORNL/NUREG/TM-73, Oak Ridge National Laboratory, December 1976.

1977

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period October–December 1976, R. A. Lorenz, J. L. Collins, and O. L. Kirkland, ORNL/NUREG/TM-88, Oak Ridge National Laboratory, March 1977.

“Fission Product Release from Simulated LWR Fuel,” R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, in Proceedings of the Specialist Meeting on the Behavior of Water Reactor Fuel Elements Under Accident Conditions, Spatind, Norway, Sept. 13–16, 1976 (CSNI Report No. 13), Part Two, Session I, Subsession I.3, March 1977.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1977, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, O. L. Kirkland, and R. L. Towns, ORNL/NUREG/TM-139, Oak Ridge National Laboratory, September 1977.

1978

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period July–September 1977, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, O. L. Kirkland, and R. L. Towns, ORNL/NUREG/TM-110, Oak Ridge National Laboratory, January 1978.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period October–September 1977, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, ORNL/NUREG/TM-186, Oak Ridge National Laboratory, March 1978.

“Fission Product Release from Highly Irradiated Fuel Under SFTA and LOCA Conditions,” J. L. Collins and R. A. Lorenz, presented at the 80th Annual Meeting of the American Ceramic Society, Detroit, MI, May 1978.

“Modeling Fission Product Release from Ruptured LWR Fuel Rods,” R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, presented at the 1978 Annual Meeting of the American Nuclear Society, San Diego, CA, June 1978.

Fission Product Source Terms for Loss-of-Coolant Accident: Summary Report, R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, NUREG/CR-0091 (ORNL/NUREG/TM-206), Oak Ridge National Laboratory, June 1978.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period January–March 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-0116 (ORNL/NUREG/TM-208), Oak Ridge National Laboratory, June 1978.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-0370 (ORNL/NUREG/TM-242), Oak Ridge National Laboratory, September 1978.

Fission Product Release from Simulated LWR Fuel, R. A. Lorenz, J. L. Collins, and S. R. Manning, NUREG/CR-0274 (ORNL/NUREG/TM-154), Oak Ridge National Laboratory, October 1978.

“Fission Product Source Terms for the LWR Loss-of-Coolant Accident,” R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, in Proceedings of the ENS/ANS International Topical Meeting on Nuclear Power Reactor Safety, Brussels, Belgium, Oct. 16–19, 1978, Vol. 3, p. 2293 (October 1978).

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period July–September 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-0493 (ORNL/NUREG/TM-280), December 1978.

1979

“Fission-Product Source Terms for the Light Water-Reactor Loss-of-Coolant Accident,” R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, Nuclear Technology, 46(3), p. 404–410 (1979).

“Fission-Product Release During LWR Loss-of-Coolant Accidents,” A. P. Malinauskas, R. A. Lorenz, and J. L. Collins, Transactions of the American Nuclear Society, 32, p. 651 (1979).

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period October–December 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-0682 (ORNL/NUREG/TM-308), Oak Ridge National Laboratory, April 1979.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period January–March 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, 

NUREG/CR-917 (ORNL/NUREG/TM-332), Oak Ridge National Laboratory, August 1979.

“Fission Product Release from LWR Fuel Defected in Steam in the Temperature Range 500 to 1600°C,” R. A. Lorenz, J. L. Collins, M. F. Osborne, and A. P. Malinauskas, in Proceedings of the Specialists’ Meeting on the Behavior of Defected Zirconium Alloy Clad Ceramic Fuel in Water Cooled Reactors, Chalk River, Canada, Sept. 17–21, 1979, International Atomic Energy Agency, IWGFPT/6, 1979.

“Chemical Behavior of Fission Products Released from Highly Irradiated LWR Fuel Under Accident Conditions,” J. L. Collins, R. A. Lorenz, A. P. Malinauskas, and M. F. Osborne, presented at the Basic Science and Nuclear Divisions of the American Ceramic Society’s Joint Fall Meeting, New Orleans, LA, October 1979.

Quarterly Progress Report on Fission Product Release From LWR Fuel for the Period April–June 1978, A. P. Malinauskas, R. A. Lorenz, J. L. Collins, M. F. Osborne, and R. L. Towns, NUREG/CR-1061 (ORNL/NUREG/TM-348), Oak Ridge National Laboratory, October 1979.

1980

“Fission-Product Release from High Gap-Inventory LWR Fuel Under LOCA Conditions,” R. A. Lorenz, J. L. Collins, M. F. Osborne, et al., Transactions of the American Nuclear Society, 34, pp. 462–463 (1980).

Fission Product Release from Highly Irradiated LWR Fuel, R. A. Lorenz, J. L. Collins, A. P. Malinauskas, O. L. Kirkland, and R. L. Towns, NUREG/CR-0722 (ORNL/NUREG/TM-287/R2), Oak Ridge National Laboratory, February 1980.

Fission Product Source Terms for the LWR Loss-of-Coolant Accident, R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, NUREG/CR-1288 (ORNL/NUREG/CR-321), Oak Ridge National Laboratory, August 1980.

Fission Product Release from Highly Irradiated LWR Fuel Heated to 1300–1600°C in Steam, R. A. Lorenz, J. L. Collins, A. P. Malinauskas, M. F. Osborne, and R. L. Towns, NUREG/CR-1386 (ORNL/NUREG/TM-346), Oak Ridge National Laboratory, November 1980.

1981

“The Adsorption-Desorption Characteristics of Iodine on Graphite,” R. A. Lorenz, Transactions of the American Nuclear Society, 39, pp. 444–445 (1981).

“Sorption of Iodine on Low-Chromium Alloy-Steel,” M. F. Osborne, R. B. Briggs, and R. P. Wichner, Transactions of the American Nuclear Society, 38, pp. 463–464 (June 1981).

Fission Product Release from BWR Fuel Under LOCA Conditions, R. A. Lorenz, J. L. Collins, M. F. Osborne, R. L. Towns, and A. P. Malinauskas, NUREG/CR-1773 (ORNL/NUREG/TM-388), Oak Ridge National Laboratory, July 1981.

Station Blackout at Browns Ferry Unit One—Accident Sequence Analysis, D. H. Cook, R. M. Harrington, S. R. Greene, S. A. Hodge, and D. D. Yue, NUREG/CR-2182, Vol. 1 (ORNL/NUREG/TM-455/V1), Oak Ridge National Laboratory, November 1981.

1982

Station Blackout at Browns Ferry Unit One—Iodine and Noble Gas Distribution and Release, R. P. Wichner, C. F. Weber, R. A. Lorenz, W. Davis, Jr., S. A. Hodge, and A. D. Mitchell, NUREG/CR-2182, Vol. 2 (ORNL/NUREG/TM-455/V2), Oak Ridge National Laboratory, 1982.

“Fission-Product Release from LWR Fuel in Steam,” M. F. Osborne, R. A. Lorenz, and R. P. Wichner, Transactions of the American Nuclear Society, 43, pp. 357–359 (1982).

“Investigation of Drywell Flooding at the Browns Ferry Nuclear Plant,” S. R. Greene, W. A. Condon, and H. L. Dodds, Jr., presented at American Nuclear Society Summer Meeting, Los Angeles, CA, June 1982.

“Application of MARCH to BWR Severe Accident Analysis,” S. R. Greene, in Proceedings of the 10^th Water Reactor Safety Information Meeting, Gaithersburg, MD, October 1982.

“Improvement of MARCH for BWR Applications,” S. R. Greene, in Proceedings of the 10^th Water Reactor Safety Information Meeting, Gaithersburg, MD, October 1982.

SBLOCA Outside Containment at Browns Ferry Unit One–Accident Sequence Analysis, W. A. Condon, R. M. Harrington, S. R. Greene, and S. A. Hodge, NUREG/CR2672, Volume 1 (ORNL/TM-8119/V1), Oak Ridge National Laboratory, November 1982.

“Severe Accident Sequence Analysis (SASA),” S. R. Greene, presented to R. Minogue, Director, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Oak Ridge, TN, December 10, 1982.

1983

SBLOCA Outside Containment at Browns Ferry Unit One—Volume 2. Iodine, Cesium, and Noble Gas Distribution and Release, R. P. Wichner, C. F. Weber, A. L. Wright, S. A. Hodge, R. A. Lorenz, and J. W. Nehls, NUREG/CR-2672, Volume 2 (ORNL/TM-8119/V2), Oak Ridge National Laboratory, 1983.

“Fission-Product Transport in A BWR Small-Break LOCA,” C. F. Weber, R. P. Wichner, S. A. Hodge,
et al., Transactions of the American Nuclear Society, 45, pp. 479–480 (1983).

“Iodine Volatility Under LWR Accident Conditions,” E. C. Beahm and W. E. Shockley, Transactions of the American Nuclear Society, 45, pp. 482–483 (1983).

Loss of DHR Sequences at Browns Ferry Unit One—Accident Sequence Analysis, D. H. Cook, S. R. Greene, R. M. Harrington, and S. A. Hodge, NUREG/CR-2973 (ORNL/TM-8532), Oak Ridge National Laboratory, May 1983.

“Fission Product Release from Fuel Under LWR Accident Conditions,” M. F. Osborne, R. A. Lorenz, K. S. Norwood, J. L. Collins, and R. P. Wichner, p. 4.1-1 in Proceedings of the International Meeting on Light-Water Reactor Severe Accident Evaluation, Cambridge, Mass., Aug. 28–Sept. 1, 1983,
CONF-830816, Vol. 1 (1983).

The Effect of Small-Capacity, High-Pressure Injection Systems on TQUV Sequences at Browns Ferry Unit One, R. M. Harrington and L. J. Ott, NUREG/CR-3179 (ORNL/TM-8635), Oak Ridge National Laboratory, September 1983.

“The Effect of Low Capacity Injection Systems on Transient Initiated Loss of Vessel Water Injection at Browns Ferry Unit One,” L. J. Ott, presented at the International Meeting on Light Water Reactor Severe Accident Evaluation, Cambridge MA, August 28–September 1, 1983.

“The Effect of Small-Capacity, High-Pressure Injection Systems on BWR Transient Initiated Loss of Injection Accident Sequences,” L. J. Ott, presented at the Eleventh Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 24–28, 1983.

Mechanistic Core-Wide Meltdown and Relocation Modeling for BWR Applications, M. Z. Podowski, R. P. Taleyarkhan, and R. T. Lahey, Jr., NUREG/CR-3525 (ORNL/Sub/81-9080/1), Oak Ridge National Laboratory, December 1983.

1984

Realistic Simulation of Severe Accidents in BWRs—Computer Modeling Requirements, S. R. Greene, NUREG/CR-2940 (ORNL/TM-8517), Oak Ridge National Laboratory, 1984.

“Behavior of Cesium, Iodine, and Tellurium in the Fission-Product Release Program at ORNL,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Transactions of the American Nuclear Society, 47, pp. 320–322 (1984).

Noble Gas, Iodine, and Cesium Transport in a Postulated Loss of Decay Heat Removal Accident at Browns Ferry, R. P. Wichner, C. F. Weber, S. A. Hodge, E. C. Beahm, and A. L. Wright, NUREG/CR-3617 (ORNL/TM-9028), Oak Ridge National Laboratory, 1984.

Data Summary Report for Fission Product Release Test HI-4, M. F. Osborne, J. L. Collins, R. A. Lorenz, K. S. Norwood, J. R. Travis, and C. S. Webster, NUREG/CR-3600 (ORNL/TM-9001), Oak Ridge National Laboratory, June 1984.

“Characterization and Chemistry of Fission Products Released from LWR Fuel Under Accident Conditions,” K. S. Norwood, J. L. Collins, M. F. Osborne, R. A. Lorenz, and R. P. Wichner, in Proceedings of the ANS Topical Meeting on Fission Product Behavior and Source Term Research, Snowbird, Utah, July 15–19, 1984.

Noble Gas, Iodine, and Cesium Transport in A Postulated Loss of Decay Heat Removal Accident at Browns Ferry, R. P. Wichner, S. A. Hodge, C. F. Weber, E. C. Beahm, and A. L. Wright, NUREG/CR-3617 (ORNL/TM-9028), Oak Ridge National Laboratory, August 1984.

Design, Construction, and Testing of a 2000°C Furnace and Fission Product Collection System, M. F. Osborne, J. L. Collins, R. A. Lorenz, J. R. Travis, and C. S. Webster, NUREG/CR-3715 (ORNL/TM-9135), Oak Ridge National Laboratory, September 1984.

Pressure Suppression Pool Thermal Mixing, D. H. Cook, NUREG/CR-3471 (ORNL/TM-8906), Oak Ridge National Laboratory, October 1984.

“Chemical Factors Affecting Fission Product Transport in BWR Severe Accidents,” E. C. Beahm, R. P. Wichner, and C. F. Weber, Trans. of 12th Water Reactor Safety Research Information Meeting, U. S. Nuclear Regulatory Commission, Gaithersburg MD, October 23, 1984.

“Effects of Improved Modeling on Best Estimate BWR Severe Accident Analysis,” L. J. Ott and C. R. Hyman, presented at the Twelfth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 23, 1984.

“Behavior of Cs, I, and Te in the Fission Product Release Program at ORNL,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Transactions of 1984 Winter Meeting of American Nuclear Society, 47, pp. 320–22, Washington, DC, Nov. 11–16, 1984.

“Measurement and Characterization of Fission Products Released from LWR Fuel,” M. F. Osborne, J. L. Collins, R. A. Lorenz, K. S. Norwood, and R. V. Strain, in Proceedings of the Fifth International Meeting on Thermal Nuclear Reactor Safety, , Karlsruhe, Federal Republic of Germany, Sept. 9–13, 1984, Vol. 3 (published December 1984 as KfK 3880/1, -/2, -/3) (1984).

1985

“A Chemical-Equilibrium Estimate of the Aerosols Produced in An Overheated Light Water-Reactor Core,” R. P. Wichner and R. D. Spence, Nuclear Technology, 70(3), pp. 376–393 (1985).

BWR–LTAS:  A Boiling Water Reactor Long-Term Accident Simulation Code, R. M. Harrington and L. C. Fuller, NUREG/CR-3764 (ORNL/TM-9163), Oak Ridge National Laboratory, February 1985.

Observed Behavior of Cesium, Iodine, and Tellurium in the ORNL Fission Product Release Program, J. L. Collins, M. F. Osborne, R. A. Lorenz, K. S. Norwood, J. R. Travis, and C. S. Webster, NUREG/CR-3930 (ORNL/TM-9316), Oak Ridge National Laboratory, February 1985.

Highlights Report for Fission Product Release Tests at Simulated LWR Fuel, M. F. Osborne, J. L. Collins, and R. A. Lorenz, ORNL/NRC/LTR-85/1, Oak Ridge National Laboratory, February 1985.

The Modeling of BWR Core Meltdown Accidents—For Application in the MELRPI.MOD2 Computer Code, B. R. Koh, S. H. Kim, R. P. Taleyarkhan, M. Z. Podowski, and R. T. Lahey, Jr., NUREG/CR-3889 (ORNL/Sub/81-9088/2), Oak Ridge National Laboratory, April 1985.

“Tellurium Precursor Effects on Iodine Transport in A BWR Accident,” C. F. Weber, R. P. Wichner, and S. A. Hodge, Transactions of the American Nuclear Society, 49, pp. 257–258 (June 1985).

Data Summary Report for Fission Product Release Test HI-5, M. F. Osborne, J. L. Collins, R. A. Lorenz, K. S. Norwood, J. R. Travis, and C. S. Webster, NUREG/CR-4037 (ORNL/TM-9437), Oak Ridge National Laboratory, June 1985.

The Absorption of Gaseous Iodine by Water Droplets, M. F. Albert, NUREG/CR-4081 (ORNL/TM-9488), Oak Ridge National Laboratory, July 1985. 

Data Summary Report for Fission Product Release Test HI-6, M. F. Osborne, J. L. Collins, R. A. Lorenz, K. S. Norwood, J. R. Travis, and C. S. Webster, NUREG/CR-4043 (ORNL/TM-9443), Oak Ridge National Laboratory, September 1985.

“Station Blackout Calculations for Browns Ferry,” L. J. Ott, C. R. Hyman, and C. F. Weber, presented at the Thirteenth Water Reactor Safety Research Information Meeting, Gaithersburg, MD,
October 22–25, 1985; also in Proc. 13^th Water Reactor Safety Research Information Meeting, NUREG/CP-0071, 1985.

“Chemistry and Transport of Iodine in Containment,” E. C. Beahm, W. E. Shockley, and C. F. Weber, presented at the International Symposium on Source Term Evaluation for Accident Conditions, Columbus, OH, October 28–November 1, 1985.

“Fission Product Release and Fuel Behavior in Tests of LWR Fuel Under Accident Conditions,” M. F. Osborne, J. L. Collins, R. A. Lorenz, and R. V. Strain, in Proceedings of the International Symposium on Source Term Evaluation for Accident Conditions, Columbus, OH, Oct. 28–Nov. 1, 1985.

1986

“Fission Product Release from UO₂ Under LWR Accident Conditions:  Recent Data Compared with Review Values,” M. F. Osborne, J. L. Collins, and R. A. Lorenz, presented at the 87th Annual Meeting of American Ceramic Society, Cincinnati, Ohio, May 5–9, 1985; also in Advances in Ceramics 17, pp. 191–97 (1986).

Chemistry and Transport of Iodine in Containment, E. C. Beahm, W. E. Shockley, C. F. Weber, S. J. Wisbey, and Y.-M. Wang, NUREG/CR-4697 (ORNL/TM-10135), Oak Ridge National Laboratory, 1986.

“Behavior of Fission Product Tellurium Under Severe Accident Conditions,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Proceedings of the International ANS/ENS Topical Meeting on Thermal Reactor Safety, San Diego, CA, February 2–6, 1986, ANS Order No. 700106 (ISBN Order No.
0-89448-121-5), February 1986.

Design and Final Safety Analysis Report for Vertical Furnace Fission Product Release Apparatus in Hot Cell B, Building 4501, M. F. Osborne, J. L. Collins, P. A. Haas, R. A. Lorenz, J. R. Travis, and C. S. Webster, NUREG/CR-4332 (ORNL/TM-9270), Oak Ridge National Laboratory, March 1986.

Highlights Report for Fission Product Release Test VI-1, M. F. Osborne, J. L. Collins, R. A. Lorenz, and T. Yamashita, ORNL/NRC/LTR-86/7, Oak Ridge National Laboratory, March 1986.

Loss of Control Air at Browns Ferry Unit One—Accident Sequence Analysis, R. M. Harrington and S. A. Hodge, NUREG/CR-4413 (ORNL/TM-9826), Oak Ridge National Laboratory, April 1986.

“Calculational Support for the DF-4 Experiment in the ACRR,” L. J. Ott, presented at the Severe Fuel Damage and Source Term Research Program Review Meeting, Oak Ridge, TN, April 1986.

“Effects of Lateral Separation of Oxidic and Metallic Core Debris on the BWR MK I Containment Drywell Floor,” C. R. Hyman, C. F. Weber, and S. A. Hodge, Committee on Safety of Nuclear Installations (CSNI), Specialists Meeting on Core Debris/Concrete Interactions, EPRI NP-5054-SR, Palo Alto CA, September 3–5, 1986.

“Iodine Behavior in Containment Under LWR Accident Conditions,” S. J. Wisbey, E. C. Beahm, W. E. Shockley, et al., Abstracts of Papers of the American Chemical Society, 192, pp. 92–NUCL (September 7, 1986). 

“The Role of BWR MK I Secondary Containments in Severe Accident Mitigation,” S. R. Greene and S. A. Hodge, in Transactions of the 14^th Water Reactor Safety Information Meeting, Gaithersburg, MD, October 1986.

“Effects of Lateral Separation of Oxidic and Metallic Core Debris on the BWR MK I Containment Drywell Floor,” C. R. Hyman and C. F. Weber, Trans. of 14th Water Reactor Safety Information Meeting, NUREG/CP-0081, U. S. Nuclear Regulatory Commission, Gaithersburg, MD,
October 27–31, 1986.

“ORNL MELCOR BWR Assessment Efforts,” S. R. Greene, presented at the MELCOR Users Group Meeting, October 31, 1986.

Thermal-Hydraulic and Characteristic Models for Packed Debris Beds, G. E. Mueller and A. Sozer, NUREG/CR-4689 (ORNL/TM-10117), Oak Ridge National Laboratory, December 1986.

Highlights Report for Fission Product Release Test VI-2, M. F. Osborne, J. L. Collins, R. A. Lorenz, and T. Yamashita, ORNL/NRC/LTR-86/18, Oak Ridge National Laboratory, December 1986.

Response of the BWR Shroud Head Under Severe Accident Conditions, E. W. Barnes, M. Keyhani, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, December 31, 1986.

1987

“The Influence of Environment on Release Behavior and Chemical Forms of Fission Products Released Under LWR Accident Conditions,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Proceedings of Workshop on Chemical Reactivity of Oxide Fuel and Fission Product Release, Berkeley, Gloucestershire, England, April 7–9, 1987. 

“The Basic Chemistry Involved in the Internal-Gelation Method of Precipitating Uranium as Determined by pH Measurements,” J. L. Collins, M. H. Lloyd, and R. L. Fellows, Radiochimica Acta, 42, pp. 121–134 (1987).

“Severe Core Damage and Associated In-Vessel Fission-Product Release,” C. Allison, J. Rest, R. A. Lorenz, et al., Progress in Nuclear Energy, 20(2), pp. 89–132 (1987).

“Calculation of Absorbed Doses to Water Pools in Severe Accident Sequences,” C. F. Weber, Trans. Am. Nucl. Soc., 55, pp. 367–368 (1987).

Calculations of Iodine Source Terms in Support of NUREG-0956, E. C. Beahm, C. F. Weber, T. S. Kress, and R. J. Anderman, ORNL/NRC/LTR-86/17, Oak Ridge National Laboratory, 1987.

Effects of Lateral Separation of Oxidic and Metallic Core Debris on the BWR MK I Containment Drywell Floor, C. R. Hyman and C. P. Weber, NUREG/CR-4610 (ORNL/TM-10057), Oak Ridge National Laboratory, January 1987.

“Fission-Product Tellurium Release Behavior Under Severe Light-Water Reactor Accident Conditions,” J. L. Collins, M. F. Osborne, and R. A. Lorenz, Nuclear Technology, 77(1), pp. 18–31 (April 1987).

“The Absorption of Gaseous Iodine by Water Droplets,” M. F. Albert, J. S. Watson, and R. P. Wichner, Nuclear Technology, 77(2), pp. 161–174 (May 1987).

“Calculational Support for the DF-4 Experiment in the ACRR,” L. J. Ott, presented at the Severe Fuel Damage, Containment Loads, and Source Term Research Program Review Meeting, Silver Spring, MD, May 4–8, 1987.

“The Release and Transport of Fission Product Cesium in the TMI-2 Accident,” R. L. Lorenz and J. L. Collins, pp. 4-69–4-83 in the Proceedings of the Symposium on Chemical Phenomena Associated with Radioactivity Releases During Severe Nuclear Plant Accidents, held at Anaheim, CA,
Sept. 9–12, 1986, also published as NUREG/CP-0078, June 1987.

ORNL Severe Accident Analysis for 25% Power Operation at the Shoreham Nuclear Power Station, S. R. Greene and S. A. Hodge, Oak Ridge National Laboratory, June 18, 1987.

An Assessment of the Shoreham Nuclear Power Station’s Secondary Containment Severe Accident Mitigation Capability, S. R. Greene, ORNL LTR Report to U.S. Nuclear Regulatory Commission, June 26, 1987.

“Organic Iodide Formation During Severe Accidents in Light Water Nuclear-Reactors,” E. C. Beahm, Y. M. Wang, S. J. Wisbey, et al., Nuclear Technology, 78(1), pp. 34–42 (July 1987).

Highlights Report for Fission Product Release Test VI‑3, M. F. Osborne, J. L. Collins, R. A. Lorenz, J. R. Travis, C. S. Webster, S. R. Daish, H. K. Lee, T. Nakamura, and Y. C. Tong, draft letter report to SFD Partners, July 1987.

“Experimental Studies of Fission-Product Release from Commercial Light-Water Reactor Fuel Under Accident Conditions,” M. F. Osborne, J. L. Collins, and R. A. Lorenz, Nuclear Technology, 78(2), pp. 157–169 (August 1987).

“Modeling of the Transport of Vapor Fission Product Species to Aerosols,” T. S. Kress, E. C. Beahm, and C. F. Weber, presented at the Workshop on Water-Cooled Reactor Aerosol Code Evaluation and Uncertainty Assessment, Brussels, Belgium, September 9–11, 1987.

“The Impact of BWR MK I Primary Containment Failure Dynamics on Secondary Containment Integrity,” S. R. Greene, in Transactions of the 15^th Water Reactor Safety Information Meeting, Gaithersburg, MD, October 1987.

“Advanced Severe Accident Response Models for BWR Application,” L. J. Ott, presented at the Severe Fuel Damage, Containment Loads, and Source Term Research Program Review Meeting, Silver Spring, MD, October 19–23, 1987.

“Advanced Severe Accident Response Models for BWR Application,” L. J. Ott, presented at the Fifteenth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 29, 1987.

“Fission Product Iodine and Cesium Release Behavior Under Severe LWR Accident Conditions,” J. L. Collins, M. F. Osborne, R. A. Lorenz, and A. P. Malinauskas, Nucl. Technol. 81(10), pp. 78–94 (November 1987).

“Calculation of Absorbed Doses to Water Pools in Severe Accident Sequences,” C. F. Weber, presented at the Winter Meeting Am. Nucl. Soc., Los Angeles, CA, November 15–19, 1987.

1988

“Considerations for Severe Accident Management Strategies in a Pressurized Water Reactor,” J. J. Carbajo and J. C. Carter, Trans. Am. Nucl. Soc., 57, pp. 159–160 (1988).

“The Xenon Poisoning Problem Previous to the Chernobyl Accident,” J. J. Carbajo, Trans. Am. Nucl. Soc., 56, pp. 404–405 (1988).

Peach Bottom Containment Response Calculations for Unmitigated Short-Term Station Blackout, S. R. Greene, S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, February 1, 1988.

“Fission-Product Iodine and Cesium Release Behavior under Light Water-Reactor Accident Conditions,” J. L. Collins, M. F. Osborne, R. A. Lorenz, et al., Nuclear Technology, 81(1), pp. 78–94 (April 1988).

“Proposed ORNL BWR Mark II Containment Response Considerations,” S. R. Greene, presented to NRC Staff, April 21, 1988.

Modeling of Time-Dependent Emergence of Core Debris from A Boiling Water Reactor Vessel Under Severe Accident Conditions, L. J. Ott and S. A. Hodge, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, April 22, 1988.

Small Scale BWR Core Debris Eutectics Formation and Melting Experiment—An Update, G. W. Parker, S. A. Hodge, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), April 22, 1988.

“Separate Effects Melt Experiments,” G. W. Parker, S. A. Hodge and L. J. Ott, presented at the Severe Accident Research Program Review Meeting, Albuquerque, NM, April 26, 1988.

Primary Containment Response Calculations for Unmitigated Short-Term Station Blackout at Peach Bottom (with Consideration of (1) Separated Metal and Oxide Pours and (2) Degassing of Drywell Concrete), S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, May 2, 1988.

“The Role of BWR Secondary Containments in Severe Accident Mitigation:  Issues and Insights from Recent Analyses,” S. R. Greene, in Proceedings of the 4^th Workshop on Containment Integrity, Arlington, VA, NUREG/CR-0095, June 1988.

“Containment Venting as A Mitigation Technique for BWR MARK-1 Plant ATWS,” R. M. Harrington, Nuclear Engineering and Design, 108(1–2), pp. 55–69 (June 1988).

“Iodine Behavior in Containment,” E. C. Beahm, C. F. Weber, T. S. Kress, S. R. Daish, and
W. E. Shockley, presented at the 2^nd CSNI Workshop of Iodine Chemistry in Reactor Safety, Toronto, Canada, June 2–3, 1988.

“Chemistry and Mass Transport of Iodine in Containment,” E. C. Beahm, C. F. Weber, T. S. Kress, W. E. Shockley, and S. R. Daish, presented at the Meeting Am. Chem. Soc., Toronto, Canada, June 4–9, 1988.

“Chemistry and Mass-Transport of Iodine in Containment,” E. C. Beahm, C. F. Weber, T. S. Kress, et al., Abstracts of Papers of the American Chemical Society, 195, Part 2, pp. 101–NUCL (June 5, 1988).

“The Chemistry and Behavior of Iodine Vapor Species in Nuclear Plant Air-Monitoring Sampling Lines,” A. L. Wright, B. R. Fish, E. C. Beahm, and C. F. Weber, presented at the 20^th DOE/NRC Nuclear Air Cleaning Conf., Boston, MA, August 22–25, 1988.

“Trends Status,” T. S. Kress, E. C. Beahm, W. E. Shockley, and C. F. Weber, presented at Severe Accident Research Program Review Meeting, Bethesda, MD, October 17–21, 1988.

“Experiment-Specific Codes for the DF-4 and NRU Experiments with BWR Geometries,” L. J. Ott, presented at the Severe Accident Research Program Review Meeting, Bethesda, MD, October 17–21, 1988.

 “BWRSAR Calculations of Reactor Vessel Debris Pours for Peach Bottom Short-Term Station Blackout,” L. J. Ott and S. A. Hodge, presented at the Sixteenth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 27, 1988.

“Small Scale BWR Core Debris Eutectics Formation and Melting Experiment,” G. W. Parker, L. J. Ott, and S. A. Hodge, presented at the Sixteenth Water Reactor Safety Research Information Meeting, Gaithersburg, MD, October 27, 1988.

“BWR Mark II and III Parametrics Program,” S. R. Greene and S. A. Hodge, ORNL NRC Program Review for E. Beckjord, Director, Office of Research, U. S. Nuclear Regulatory Commission, November 9, 1988.

Primary Containment Response Calculations for Unmitigated Short-Term Station Blackout at Peach Bottom (with Consideration of (1) Separated Metal and Oxide Pours, (2) Melting of Reactor Vessel Bottom Head, and (3) Degassing of Drywell Concrete, S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, November 28, 1988.

Boiling Water Reactor Severe Accident Technology at Oak Ridge—Purpose and Goals, S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), Oak Ridge National Laboratory, December 6, 1988.

1989

“Station Blackout Severe Accidents in APWRs,” J. J. Carbajo, presented at the International Topical Meeting on Probability, Reliability, and Safety Assessment, PSA 1989, April 2-7, 1989, Pittsburgh, PA, 1989.

Boiling Water Reactor Severe Accident Response (BWRSAR) Code Description and Assessment, S. A. Hodge and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor, Oak Ridge National Laboratory, February 1, 1989.

Primary Containment Response Calculations for Unmitigated Short-Term Station Blackout at Peach Bottom, S. A. Hodge, C. R. Hyman, and L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor, Oak Ridge National Laboratory (1989).

“Time-Dependent Release of Fission Products from LWR Fuel Under Severe Accident Conditions,” M. F. Osborne, R. A. Lorenz, J. L. Collins, and T. Nakamura, Proceedings of International ENS/ANS Conference on Thermal Reactor Safety, Avignon, France, Oct. 3–7, 1988 [published October 1988 (Vols. 1–4) and March 1989 (Vols. 5–6)].

“Boiling Water Reactor Severe Accident Response (BWRSAR) Post-Test Analyses of the ACRR DF-4 Experiment,” L. J. Ott, presented at the Severe Accident Research Program Review Meeting, Idaho Falls, ID, April 11, 1989.

Failure Modes of the BWR Reactor Vessel Bottom Head, S. A. Hodge and L. J. Ott, ORNL/M-1019 (Letter Report to the BWRSAT Program NRC Technical Monitor [Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC]), Oak Ridge National Laboratory, May 10, 1989.

“Iodine Behavior in Containment,” E. C. Beahm, C. F. Weber, T. S. Kress, and W. E. Shockley, presented at ICHMT Seminar on Fission Product Transport Processes in Reactor Accidents, Dubrovnik, Yugoslavia, May 22–26, 1989.

“TRENDS:  A Code for Modeling Iodine Behavior in Containment During Severe Accidents,” C. F. Weber, E. C. Beahm, T. S. Kress, S. R. Daish, and W. E. Shockley, in Proc. ICHMT Int. Seminar on Heat and Mass Transfer Aspects of Fission Product Releases, Dubrovnik, Yugoslavia, May 22–26, 1989.

Data Summary Report for Fission Product release Test VI-1, M. F. Osborne, J. L. Collins, R. A. Lorenz, J. R. Travis, C. S. Webster, and T. Yamashita, NUREG/CR-5339 (ORNL/TM-11104), Oak Ridge National Laboratory, June 1989.

“Advanced Severe Accident Response Models for BWR Application,” L. J. Ott, Nuclear Engineering and Design, 115(2–3), pp. 289–303 (July 1989).

Post-Test Analyses of the DF-4 BWR Experiment Using the BWRSAR/DF4 Code, L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/M-1020, Oak Ridge National Laboratory, August 11, 1989.

Data Summary Report for Fission Product release Test VI-2, M. F. Osborne, J. L. Collins, R. A. Lorenz, J. R. Travis, and C. S. Webster, NUREG/CR-5340 (ORNL/TM-11105), Oak Ridge National Laboratory, September 1989.

“BWR Severe Accident Experiment Needs,” L. J. Ott, presented at the International CORA Workshop 1989, Karlsruhe, FRG, September 25–27, 1989.

“BWR Specific Effects in Severe Accident Behavior,” L. J. Ott, presented at the International CORA Workshop 1989, Karlsruhe, FRG, September 25–27, 1989.

“Experiment-Specific BWR Code Analysis of the ACRR DF-4 Experiment,” L. J. Ott, presented at the International CORA Workshop 1989, Karlsruhe, FRG, September 25–27, 1989.

Description of the NRU FLHT-6 Experiment-Specific Code and Preliminary Pre-Test Predictions, L. J. Ott, Letter Report to the BWRSAT Program NRC Technical Monitor (Dr. T. J. Walker, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/M-1021, Oak Ridge National Laboratory, September 22, 1989.

Thermophoretic Transport of Particles that Act as Volumetric Heat Sources in Natural Convection Flow, J. C. Conklin and R. J. Krane, ORNL-6573, Oak Ridge National Laboratory, October 1989.

The DF-4 Fuel Damage Experiment in ACRR With A BWR Control Blade and Channel Box, R. O. Gauntt, R. D. Gasser, and L. J. Ott, NUREG/CR-4671 (SAND-86-1443), Sandia National Laboratories, Albuquerque, NM, November 1989.

“Behavior of Iodine in Containment,” E. C. Beahm, C. F. Weber, M. L. Brown, and W. E. Shockley Presentation, Oak Ridge National Laboratory, Oak Ridge, TN, December 5, 1989.

1990

The Response of BWR Mark II and Mark III Containments to Short-Term Station Blackout Severe Accident Sequences, S. R. Greene et al., ORNL/NRC/LTR-89/13, Oak Ridge National Laboratory, 1990.

“Small-Scale BWR Core Debris Eutectics Formation and Melting Experiment,” G. W. Parker, L. J. Ott, and S. A. Hodge,” Nuclear Engineering and Design, 121 (1990).

“BWRSAR Calculations of Reactor Vessel Debris Pours for Peach Bottom Short-Term Station Blackout,” S. A. Hodge and L. J. Ott, Nuclear Engineering and Design, 121 (1990).

Debris Bed Behavior in the BWR Lower Plenum—The BWRSAR Approach with Recommended Improvements for Installation in MELCOR, S. A. Hodge, ORNL/NRC/LTR-90/10, Oak Ridge National Laboratory, April 26, 1990.

“BWR Experiment Analyses and Principal BWR Severe Accident Uncertainties,” L. J. Ott, presented at the Severe Accident Research Program Partners Review Meeting, Brookhaven National Laboratory, Upton, NY, May 1, 1990.

Boiling Water Reactor Severe Accident Models for MELCOR, S. A. Hodge, ORNL/NRC/LTR-90/13, Oak Ridge National Laboratory, May 22, 1990.

Data Summary Report for Fission Product Release Test VI‑3, M. F. Osborne, R. A. Lorenz, J. L. Collins, J. R. Travis, C. S. Webster, H. K. Lee, T. Nakamura, and Y.-C. Tong, NUREG/CR-5480 (ORNL/TM-11399), Oak Ridge National Laboratory, June 1990.

“Fission Product Release from Commercial vs Simulated Fuels in LWR Accident Studies,” M. F. Osborne, H. Albrecht, R. A. Lorenz, and J. L. Collins, presented at the 1990 Annual Meeting of the American Nuclear Society, Nashville, TN, June 10–14, 1990; summary published in Trans. 1990 ANS Annl. Mtg., 61, pp. 251–52, American Nuclear Society (June 1990).

“The Role of BWR Secondary Containments in Severe Accident Mitigation: Issues and Insights from Recent Analyses,” S. R. Greene, Nuclear Engineering and Design, 120(1), pp. 75–86 (June 1, 1990).

A Survey of Current Models of BWR Core Plate Failure Used in the Severe-Accident Codes APRIL, BWRSAR, MELCOR, MELPROG, and SCDAP/RELAP, L. J. Ott and W. I. Van Rij, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-90/14, Oak Ridge National Laboratory, June 29, 1990.

Kinetic Modeling and Parameter Optimization of Iodine Hydrolysis Reactions, C. F. Weber, M.S. Thesis, University of Tennessee, August 1990.

“BWRSAR Calculations of Reactor Vessel Debris Pours for Peach Bottom Sort-Term Station Blackout,” S. A. Hodge and L. J. Ott, Nuclear Engineering and Design, 121(3), pp. 327–339 (August 1990).

“Small-Scale BWR Core Debris Eutectics Formation and Melting Experiment,” G. W. Parker, L. J. Ott, and S. A. Hodge, Nuclear Engineering and Design, 121(3), pp. 341–347 (August 1990).

“Contain Calculations of Debris Conditions Adjacent to the BWR MARK-1 Drywell Shell During the Later Phases of a Severe Accident,” C. R. Hyman, Nuclear Engineering and Design, 121(3), pp.
379–393 (August 1990).

“Experiment-Specific BWR Code Analysis of the CORA-17 Experiment,” L. J. Ott, presented at the International CORA Workshop 1990, Karlsruhe, Germany, September 28–October 2, 1990.

“Post-Test Analyses of the CORA-16 Experiment,” L. J. Ott, presented at the International CORA Workshop 1990, Karlsruhe, Germany, September 28–October 2, 1990.

A Survey of Current Models of BWR Lower Head Failure Used in the Severe-Accident Codes APRIL, BWRSAR, MELCOR, MELPROG, and SCDAP/RELAP, L. J. Ott and W. I. Van Rij, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-90/26, Oak Ridge National Laboratory, September 30, 1990.

Description of the CORA BWR Experiment-Specific Code, L. J. Ott, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-90/23, Oak Ridge National Laboratory, September 30, 1990.

Post-Test Analyses of CORA BWR Experiments, L. J. Ott, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-90/24, Oak Ridge National Laboratory, September 30, 1990.

“Results of Recent ORNL Fission Product Release Tests,” R. A. Lorenz, M. F. Osborne, and J. L. Collins, presented at the 18th Water Reactor Safety Meeting at Rockville, MD, October 22–24, 1990.

“Experiment-Specific Analyses in Support of Code Development,” L. J. Ott, presented at the Eighteenth Water Reactor Safety Research Information Meeting, Rockville, MD, October 23, 1990.

1991

“Effect of an In-Containment Refueling Water Storage Tank on Station Blackout Accidents,” J. J. Carbajo, Trans. Am. Nucl. Soc., 63, pp. 311–313 (1991).

“In-Vessel Hydrogen Generation During Station Blackout Severe Accident Sequences in APWRs,” J. J. Carbajo, Trans. Am. Nucl. Soc., 63, pp. 313–314 (1991).

“In-Vessel Hydrogen Generation During a Station Blackout Severe Accident,” J. J. Carbajo, Trans. Am. Nucl. Soc., 63, pp. 327–328 (1991).

“Direct Containment Heating Calculations for the Zion Plant,” J. J. Carbajo, Trans. Am. Nucl. Soc., 63, pp. 328–329 (1991).

“Characterization of Debris/Concrete Interactions for Advanced Research Reactor and Commercial BWR Severe Accidents,” C. R. Hyman, R. P. Taleyarkhan, and S. R. Greene, in Proceedings for the American Nuclear Society Winter Meeting, 1991.

“Severe LWR Accident Studies of Fission Product Release and Fuel Behavior at ORNL,” M. F. Osborne, R. L. Lorenz, and J. L. Collins, presented at the 15th Nuclear Safety Research Reactor Technical Review Meeting, Washington, DC, 1991.

“Characterization of Debris/Concrete Interactions for Commercial BWR Mark II and Research Reactors Severe Accidents,” C. R. Hyman, R. P. Taleyarkhan, and S. R. Greene, presented at the National Heat Transfer Conference, Minneapolis, MN, 1991.

Calculation of Absorbed Doses to Water Pools in Severe Accident Sequences, C. F. Weber, ORNL/TM-11970, Oak Ridge National Laboratory, 1991.

Data Summary Report for Fission Product Release Test VI-4, M. F. Osborne, R. A. Lorenz, J. L. Collins, J. R. Travis, C. S. Webster, and T. Nakamura, NUREG/CR-5481 (ORNL/TM-11400), Oak Ridge National Laboratory, January 1991.

“Identification and Initial Assessment of BWR In-Vessel Accident Management Strategies,” S. A. Hodge, presented at Accident Management Information Exchange Meeting USNRC—BMU/GRS, April 15, 1991.

“In-Vessel Phenomena—CORA,” L. J. Ott and W. I. van Rij, presented at the Cooperative Severe Accident Research Program (CSARP) Semiannual Review Meeting, Bethesda, MD, May 7, 1991.

The Response of BWR Mark III Containments to Short-Term Station Blackout Severe Accident Sequences, S. R. Greene, S. A. Hodge, C. R. Hyman, B. W. Patton, and M. L. Tobias, NUREG/CR-5571 (ORNL/TM-11549), Oak Ridge National Laboratory, June 1991.

“Fission Product Behavior,” T. S. Kress, E. C. Beahm, and C. F. Weber, presented at the Annual Meeting American Nuclear Society, Orlando, FL, June 2–6, 1991; also published in Trans. Amer. Nucl. Soc., 63, pp. 260 (1991).

CCCTF Shakedown Tests, Draft, R. N. Morris, J. L. Collins, W. A. Gabbard, J. C. Whitson, and M. F. Osborne, July 1991. 

“Atmospheric Effects on Fission Product Behavior at Severe Accident Conditions,” M. F. Osborne, R. A. Lorenz, and J. L. Collins, in Proceedings of the American Nuclear Society International Topical Meeting on the Safety of Thermal Reactors, Portland, OR, July 21–25, 1991.

Significant BWR Technical Developments from the CORA Program, L. J. Ott, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. R. W. Wright, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-91/14, Oak Ridge National Laboratory, July 31, 1991.

Report of Foreign Travel by C. F. Weber, September 7–15, 1991, C. F. Weber, ORNL/CSD/FTR-4049, Oak Ridge National Laboratory, September 1991.

Data Summary Report for Fission Product Release Test VI-5, M. F. Osborne, R. A. Lorenz, J. R. Travis, C. S. Webster, and J. L. Collins, ORNL/CR-5668 (ORNL/TM-11743), Oak Ridge National Laboratory, September 1991.

“Iodine Chemical Forms in LWR Severe Accidents,” C. F. Weber, E. C. Beahm, and T. S. Kress, presented at the 3^rd Committee for Safety in Nuclear Installations Workshop on Iodine Chemistry in Reactor Safety, Tokai-mura, Japan, September 11–13, 1991.

“Iodine Chemical Forms in LWR Accidents:  Preliminary Presentation of Results,” E. C. Beahm, C. F. Weber, and T. S. Kress, presented at Committee for Safety in Nuclear Installations Iodine Specialists’ Meeting, Shinhara, Mito-Shi, Japan, September 11–13, 1991; also in Nucl. Technol., 101, pp.
262–269 (1993).

Current Status of the Experimental Data and Phenomenological Modeling in BWR Core Melt Progression, L. J. Ott and W. I. Van Rij, Letter Report to the BWRCMP Program NRC Technical Monitor (Dr. Ali Behbahani, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-91/16, Oak Ridge National Laboratory, September 31, 1991.

Assessment of Two BWR Accident Management Strategies, S. A. Hodge and M. Petek, in
CONF-911079-2, October 1991.

“BWR Lower Plenum Debris Bed Models for MELCOR,” S. A. Hodge and L. J. Ott, presented at the Nineteenth Water Reactor Safety Research Information Meeting, Bethesda, MD, October 28–30, 1991.

“Iodine Chemical Forms in LWR Severe Accidents,” E. C. Beahm, C. F. Weber, T. S. Kress, and G. W. Parker, in Proceedings of the 19^th Water Reactor Safety Information Meeting, 2, pp. 325–342, Bethesda, MD, October 28–30, 1991.

BWR Mark II Ex-Vessel Corium Interaction Analyses, S. R. Greene, A. E. Levin, C. R. Hyman, A. Sozer, and R. P. Taleyarkhan, NUREG/CR-5623 (ORNL/TM-11644), Oak Ridge National Laboratory, November 1991.

“ORNL Studies of Fission Product Release Under LWR Accident Conditions,” F. F. Osborne, R. A. Lorenz, and J. L. Collins, presented at the Second Workshop on LWR Severe Accident Research A JERI, Tokoyo, Japan, November 25–27, 1991 (also published in the proceedings of the meeting).

1992

“Spent-Fuel Decay Heat and Source Term Uncertainties,” J. J. Carbajo, Trans. Am. Nucl. Society, 65, pp. 66–67 (1992).

“Effect of the Timing of Vessel Depressurization on a Short-Term Station Blackout in a BWR-4 Performed with the MELCOR Code,” J. J. Carbajo, Trans. Am. Nucl. Society, 66, pp. 327–328 (1992).

The Core Conduction Cooldown Test Facility: Current Status and Issues, R. N. Morris, C. A. Baldwin, J. L. Collins, L. C. Emerson, W. A. Gabbard, C. M. Malone, B. F. Myers, M. F. Osborne, M. L. Peters, J. C. Whitson, and J. L. Wright, ORNL/NRP-91/7, Oak Ridge National Laboratory, January 1992.

Core Conduction Cooldown Test Facility Shakedown Tests, R. N. Morris, J. L. Collins, W. A. Gabbard, and J. C. Whitson, ORNL/NRP-91/25, Oak Ridge National Laboratory, May 1992.

HRB-17 and HRB-18 HEU TRISO UCO Unbonded Irradiated Particle Core Conduction Cooldown Tests, R. N. Morris, C. A. Baldwin, J. L. Collins, C. M. Malone, W. A. Gabbard, J. R. Travis, C. S. Webster, J. C. Whitson, J. L. Wright, and M. J. Kania, ORNL/NRP-92/9, Oak Ridge National Laboratory, June 1992.

Iodine Chemical Forms in LWR Severe Accidents, Final Report, E. C. Beahm, C. F. Weber, T. S. Kress, and G. W. Parker, ORNL/TM-11861 (NUREG/CP-0119), Oak Ridge National Laboratory, 1992.

Iodine Evolution and pH Control, E. C. Beahm, R. A. Lorenz, and C. F. Weber, ORNL/TM-12242, Oak Ridge National Laboratory, 1992.

Models of Iodine Behavior in Reactor Containments, E. C. Beahm, C. F. Weber, T. S. Kress, and G. W. Parker, ORNL/TM-12202, Oak Ridge National Laboratory, 1992.

“MELCOR Modifications for SBWR Applications,” C. R. Hyman, R. L. Sanders, S. R. Greene, and S. A. Hodge, in Proceedings of the Twentieth Water Reactor Safety Information Meeting, Bethesda, MD, 1992.

“Optimal Determination of Rate Coefficients in Multiple Reactor Systems,” C. F. Weber, E. C. Beahm, and J. S. Watson, Comput. Chem., 16(4), pp. 325–333 (1992).

“Iodine Chemical Forms in LWR Severe Accidents,” E. C. Beahm and C. F. Weber, presentation to Advisory Committee Reactor Safeguards (ACRS), Bethesda, MD, January 7–8, 1992.

“BWR Core Melt Progression Phenomena – Experimental Analyses,” L. J. Ott, presented at the Cooperative Severe Accident Research Program (CSARP) Review Meeting, Bethesda, MD, May 4–8, 1992.

“Iodine Transport in Containment,” E. C. Beahm, C. F. Weber, and T. S. Kress, presented at Cooperative Severe Accident Research Program (CSARP) Meeting, Bethesda, MD, May 4–8, 1992.

An Assessment of MELCOR Modeling Modifications Required for the Simplified Boiling Water Reactor, S. R. Greene, S. A. Hodge, C. R. Hyman, and R. L. Sanders, ORNL/NRC/LTR-92/11, Oak Ridge National Laboratory, May 29, 1992.

BWR Control Blade/Channel Box Interaction and Melt Relocation Models for SCDAP, F. P. Griffin, ORNL/NRC/LTR-92/12, Oak Ridge National Laboratory, June 15, 1992.

Post-Test Analysis of the CORA-16 and CORA 17 BWR Experiments, L. J. Ott, ORNL/NRC/LTR-92/17, Oak Ridge National Laboratory, July 10, 1992.

“ORNL Studies of Fission-Product Release Under LWR Severe Accident Conditions,” M. F. Osborne and R. A. Lorenz, Nuclear Safety, 33(3), pp. 344–365 (July–September 1992).

“Iodine Chemical Forms in LWR Severe Accidents,” C. F. Weber, E. C. Beahm, and T. S. Kress, pp. 414–430 in Proc. 3^rd CSNI Workshop on Iodine Chemistry in Reactor Safety, Tokai-mura, Japan, September 11–13, 1992, NEA/CSNI/R(91)15, JAERI, 1992.

An Assessment of SCDAP/RELAP5 Modeling Modifications Required for the Simplified Boiling Water Reactor, F. P. Griffin, ORNL/NRC/LTR-92/25, Oak Ridge National Laboratory, September 30, 1992.

Thermal Shock Considerations for External Flooding of A BWR Reactor Vessel, S. A. Hodge and W. J. McAfee, ORNL/NRC/LTR-92-23, Oak Ridge National Laboratory, September 30, 1992.

Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, S. A. Hodge, J. C. Cleveland, T. S. Kress, and M. Petek, NUREG/CR-5869 (ORNL/TM-12080), Oak Ridge National Laboratory, October 1992; also published in CONF-921007-31, October 1992.

“ORNL Pre- and Post-Test Analyses of BWR Experiments,” L. J. Ott, presented at the International CORA Workshop 1992, Karlsruhe, Germany, October 6, 1992.

“Core Melt Source Reduction System (COMSORS) for Light-Water Reactors (Draft), presented at COMSORS Program Review Meeting, Tenera Bethesda Office, Bethesda, MD, October 20, 1992.

Report of Foreign Travel to KfK, Germany and Participation in the CORA-33 Experiment, L. J. Ott, letter report to the BWRCMP Program NRC Technical Monitor (Dr. Ali Behbahani, Accident Evaluation Branch, Division of Systems Research, RES USNRC), ORNL/NRC/LTR-92/28, November 20, 1992.

BWR Control Blade/Channel Box Interaction and Melt Relocation Models for SCDAP, Revision 1, F. P. Griffin and K. A. Smith, ORNL/NRC/LTR-92/12/R1, Oak Ridge National Laboratory, December 31, 1992.

Post-Test Analyses of the CORA-31 Slow Heatup BWR Experiment, L. J. Ott, ORNL/NRC/LTR-92/29, Oak Ridge National Laboratory, December 31, 1992.

1993

“Iodine Evolution and pH Control,” E. C. Beahm, R. A. Lorenz, and C. F. Weber, Trans. Am. Nucl. Soc., 59, pp. 387 (1993).

“Iodine Transport in a Severe Accident at the High Flux Isotope Reactor,” C. F. Weber and E. C. Beahm, Trans. Amer. Nucl. Soc., 68A, p. 275 (1993).

“MELCOR Simulation of a Large-Break LOCA at the High Flux Isotope Reactor,” R. H. Morris, S. E. Fisher, and S. R. Greene, in CONF-930601, Transactions of the American Nuclear Society, 68 (1993).

Severe Accident Source Term Characteristics for Selected Peach Bottom Sequences Predicted by the MELCOR Code, J. J. Carbajo, NUREG/CR-5942 (ORNL/TM-12229), Oak Ridge National Laboratory, 1993.

“Vessel Failure Time for a Low-Pressure Short-Term Station Blackout in a BWR-4,” J. J. Carbajo, Trans. Am. Nucl. Society, 67, pp. 306–307 (1993).

“MELCOR Code Analysis of a Severe Accident LOCA at Peach Bottom Plant,” J. J. Carbajo, Trans. Am. Nucl. Society, 69, pp. 319–321 (1993).

“Containment Failure Time and Mode for a Low-Pressure Short-Term Station Blackout in a BWR-4 with Mark-I Containment,” J. J. Carbajo and S. R. Greene, Trans. Am. Nucl. Society, 69, pp. 325–326 (1993).

MELCOR 1.8.2 Assessment:  Comparison of Fuel Fission Product Release Models to ORNL Vifuel Fission Product Release Experiment, S. R. Greene et al., ORNL/NRC/LTR-94/34, Oak Ridge National Laboratory, 1993.

“Fission-Product Transport Behavior,” T. S. Kress, E. C. Beahm, C. F. Weber, and G. W. Parker, Nuclear Technology, 101(3), pp. 262–269 (March 1993).

“Iodine Transport Models for LWR Accidents,” C. F. Weber and E. C. Beahm, in Cooperative Severe Accident Research Program Meeting (CSARP), Bethesda, MD, May 7, 1993.

“Iodine Transport in a Severe Accident at the High Flux Isotope Reactor,” C. F. Weber and E. C. Beahm, Annual Meeting Am. Nucl. Soc., 1993.

Current Status of BWR Experimental Analyses and the Need for Additional BWR Experiments, L. J. Ott, ORNL/NRC/LTR-93/12, Oak Ridge National Laboratory, June 30, 1993.

Post-Test Analyses of the CORA-33 Dry Core BWR Experiment, L. J. Ott, ORNL/NRC/LTR-93/21, Oak Ridge National Laboratory, August 31, 1993.

Post-Test Analyses of the CORA-28 Preoxidized BWR Experiment, L. J. Ott, ORNL/NRC/LTR-93/21, Oak Ridge National Laboratory, August 31, 1993.

“Analysis and Interpretation of the Results of the CORA-33 Dry Core BWR Experiment,” L. J. Ott, presented at the International CORA Workshop 1993, Karlsruhe, Germany, September 27, 1993.

“BWR Control Blade/Channel Box Interaction Models for SCDAP/RELAP5,” F. P. Griffin, presented at the 21st Water Reactor Safety Research Information Meeting, Bethesda, MD, October 25‑27, 1993.

“Interpretation of the Results of the CORA-33 Dry Core BWR Test,” L. J. Ott and S. Hagen, presented at the 21st Water Reactor Safety Research Information Meeting, Bethesda, MD, October 25–27, 1993.

“Containment Failure Time and Mode for a Low-Pressure Short-Term Station Blackout in a BWR-4 with Mark-III Containment,” S. R. Greene, in Proceedings of the American Nuclear Society Winter Meeting, San Francisco, CA, November 1993.

Current Status of ORNL FLHT-6 Experimental Analyses, L. J. Ott and K. A. Smith, ORNL/NRC/LTR-93/30, Oak Ridge National Laboratory, November 30, 1993.

“Iodine Evolution and pH Control,” E. C. Beahm, R. A. Lorenz, and C. F. Weber, in Proceedings of the Winter Meeting of the American Nuclear Society, San Francisco, CA, November 14–18, 1993.

Status of ORNL XR Experimental Analyses, L. J. Ott, ORNL/NRC/LTR-93/34, Oak Ridge National Laboratory, December 30, 1993.

BWR Control Blade/Channel Box Interaction and Melt Relocation Models for SCDAP, Revision 2, F. P. Griffin, ORNL/NRC/LTR-92/12/R2, Oak Ridge National Laboratory, December 30, 1993.

1994

“Optimum Depressurization for a Short-Term Station Blackout in a BWR-6 with High Burnup Fuel,” J. J. Carbajo, Trans. Am. Nucl. Society, 71, pp. 324–326 (1994).

“MELCOR Sensitivity Studies for a Low-Pressure, Short-Term Station Blackout at the Peach Bottom Plant,” J. J. Carbajo, Nucl. Eng. and Des., 152, pp. 287–317 (1994).

Data Summary Report for Fission Product Release Test VI-6, M. F. Osborne, R. A. Lorenz, J. R. Travis, C. S. Webster, and J. L. Collins, ORNL/CR-6077 (ORNL/TM-12416), Oak Ridge National Laboratory, March 1994.

“Assessment of 2 BWR Accident Management Strategies,” S. A. Hodge and M. Petek, Nuclear Engineering and Design, 148(2–3), pp. 185–203 (July 1994).

“Detailed Analysis of the BWR Dry Core Conditions for the XR2 Experiments,” L. J. Ott and F. P. Griffin, presented at the International CORA Workshop 1994, Karlsruhe, Germany,
October 10–12,1994.

Development of the BWR Dry Core Initial and Boundary Conditions for the SNL XR2 Experiments, L. J. Ott and F. P. Griffin, ORNL/NRC/LTR-94/38, Oak Ridge National Laboratory, October 31, 1994.

1995

“MELCOR Calculations for a Low-Pressure, Short-Term Station Blackout in a BWR-6,” J. J. Carbajo, Trans. Am. Nucl. Society, 72, pp. 250–252 (1995).

“Recovery Sequences for a Station Blackout Accident at the Grand Gulf Nuclear Station,” J. J. Carbajo, Trans. Am. Nucl. Society, 72, pp. 226–228 (1995).

“Source Terms Released into the Environment for a Station Blackout Severe Accident at the Peach Bottom Atomic Power Station,” J. J. Carbajo, Proceedings of the Topical Meeting on Safety of Operating Reactors, pp. 259–266 (1995).

“Comparison of MELCOR and SCDAP/RELAP5 Results for a Low-Pressure, Short-Term Station Blackout at Browns Ferry,” J. J. Carbajo, Trans. Am. Nucl. Society, 73, pp. 259–269 (1995).

Comparison of MELCOR Modeling Techniques and Effects of Vessel Water Injection on a Low-Pressure, Short-Term Station Blackout at the Grand Gulf Nuclear Station, J. J. Carbajo, ORNL/TM-12771, Oak Ridge National Laboratory, 1995.

Comparison of MELCOR Results, Including Parametric Variations, to SCDAP/RELAP5 Results for a Station Blackout Accident at Browns Ferry, J. J. Carbajo, ORNL/NRC/LTR-95/25, Oak Ridge National Laboratory, 1995.

Description of the XR/BWR Experiment-Specific Code and Preliminary Post-Test Analyses of the XR1 Tests, L. J. Ott, ORNL/NRC/LTR-95/5, Oak Ridge National Laboratory, February 28, 1995.

BWR Control Blade/Channel Box Interaction and Melt Relocation Models for SCDAP (Revision 3), F. P. Griffin, ORNL/NRC/LTR-92/12/R3, Oak Ridge National Laboratory, March 31, 1995.

Data Summary Report for Fission Product Release Test VI-7, M. F. Osborne, R. A. Lorenz, J. R. Travis, J. L. Collins, and C. S. Webster, NUREG/CR-6318 (ORNL/TM-12937), Oak Ridge National Laboratory, May 1995.

“Recent SCDAP/RELAP5 Improvements for BWR Severe Accident Simulations,” F. P. Griffin, presented at the 23rd Water Reactor Safety Information Meeting, Bethesda, MD, October 23‑25, 1995.

Initial Development of an Upper Plenum Structure Degradation Model for SCDAP/RELAP5, F. P. Griffin, ORNL/NRC/LTR-95/31, Oak Ridge National Laboratory, October 31, 1995.

1996

In-Vessel Core Degradation Code Validation Matrix, T. J. Haste, L. J. Ott, et al., OCDE/GD(96)14, Committee on the Safety of Nuclear Installations, OECD Nuclear Energy Agency, 1996.

“Comparison of HASCAL and MELCOR Source Terms,” J. J. Carbajo, Trans. Am. Nucl. Society, 75, pp. 271–73 (1996).

“MELCOR Calculated In-Containment Source Terms Using Different Release Models,” J. J. Carbajo, Trans. Am. Nucl. Society, 74, pp. 252–253 (1996).

Assessment of the RN and BH Packages of MELCOR 1.8.3, J. J. Carbajo, ORNL/NRC/LTR-96/01, Oak Ridge National Laboratory, 1996.

BWR Control Blade/Channel Box Model for SCDAP/RELAP5: Damage Progression Theory and User Guide, F. P. Griffin, ORNL/NRC/LTR-96/20, Oak Ridge National Laboratory, July 12, 1996.

1997

“Interpretation of the Results of the CORA-33 Dry Core Boiling Water Reactor Test,” L. J. Ott and S. Hagen, Nuclear Engineering and Design, 167, pp. 387–306 (1997).

“Severe Accident Sequences Analyzed for a Two-Loop PWR,” J. J. Carbajo, Trans. Am. Nucl. Society, 77, pp. 259–260 (1997).

“MELCOR-Calculated In-Containment Source Terms With and Without the BH Package,” J. J. Carbajo, Trans. Am. Nucl. Society, 77, pp. 290–292 (1997).

“Modeling the Chernobyl Accident with the HASCAL Code,” J. J. Carbajo, Trans. Am. Nucl. Society, 76, pp. 273–275 (1997).

“Modeling the TMI-2 Accident with the HASCAL Code,” J. J. Carbajo, Trans. Am. Nucl. Society, 76, pp. 275–277 (1997).

Simulation of the PHEBUS FPT-1 Test with the MELCOR Code, J. J. Carbajo, ORNL/NRC/LTR-97/03, Oak Ridge National Laboratory, 1997.

Validation of SCDAP/RELAP5 Against CORA BWR Test Results, F. P. Griffin, ORNL/NRC/LTR-97/1, Oak Ridge National Laboratory, February 28, 1997.

SCDAP/RELAP5 Mod 3.2 Simulations for the Browns Ferry BWR Design, F. P. Griffin, ORNL/NRC/LTR-97/27, Oak Ridge National Laboratory, December 16, 1997.

“Interpretation of the XR2-1 Experiment and Characteristics of the BWR Lower Plenum Debris Bed,” S. A. Hodge and L. J. Ott, presented at the Second International Conference on Advanced Reactor Safety (ARS’97), Orlando, FL, June 1–4, 1997.

“Applicability of BWR SFD Experiments and Codes for Advanced Core Component Designs,” L. J. Ott, presented at the Technical Session on Thermal Hydraulics of Severe Accidents at the ANS Winter Annual Meeting, Albuquerque, NM, November 1997.

SCDAP/RELAP5 MOD 3.2 Simulation for the Browns Ferry BWR Design, F. P. Griffin, ORNL/NRC/LTR-97/27, Oak Ridge National Laboratory, December 16, 1997.

1998

“Accident Sequences Simulated at the Juragua Nuclear Power Plant,” J. J. Carbajo, pp. 213–220 in Proceedings of the International Topical Meeting on Safety of Operating Reactors, San Francisco, CA, 1998.

“Severe Accident Sequences Simulated at a VVER 440-213,” J. J. Carbajo, Trans. Am. Nucl. Society, 78, pp. 198–199 (1998).

Technical Assistance in Review of Source Term-Related Issues of Advanced Reactors, E. C. Beahm, C. F. Weber, and T. A. Dillow, ORNL/TM-13144, Oak Ridge National Laboratory, 1998.

Iodine Volatility and pH Control in the AP-600 Reactor, C. F. Weber and E. C. Beahm, ORNL/TM-13555, Oak Ridge National Laboratory, 1998.

SCDAP/RELAP5 Modifications for New BWR Control Blade Designs, F. P. Griffin, ORNL/NRC/LTR-98/18, Oak Ridge National Laboratory, September 30, 1998.

1999

MELCOR Small Break LOCA Calculations, J. J. Carbajo, ORNL/NRC/LTR-98/21, Oak Ridge National Laboratory, 1999.

MELCOR DBA LOCA Calculations, J. J. Carbajo, ORNL/NRC/LTR-97/21, Oak Ridge National Laboratory, 1999.

“Severe Accident Sequences Simulated at the Grand Gulf Nuclear Station,” J. J. Carbajo, Trans. Am. Nucl. Society, 80, pp. 174–176 (1999).

Iodine Revolatizitation in a Grand Gulf LOCA, C. F. Weber and E. C. Beahm, ORNL/M-6544, Oak Ridge National Laboratory, 1999.

SCDAP/RELAP5 Lower Core Plate Model, E. W. Coryell and F. P. Griffin, INEEL/EXT-99-01029, Final Design Report, September 1999.

2000

Review of Grand Gulf Methodology and Calculations for pH Control and Iodine Volatility, C. F. Weber, Letter Report to USNRC, ORNL/NRC/LTR-00/07, Oak Ridge National Laboratory, 2000.

2005

Fission Product and Chemical Speciation Test Plan and Cost Estimate, D. D. Lee, J. L. Collins, B. E Lewis, J. L. Binder, S. C. Marschman, C. F. Weber, and R. N. Morris, Oak Ridge National Laboratory (February 28, 2005).

A Review and an Analysis of the ORNL and Chalk River Fission Product Release Tests and Their Relevance to Hypothetical Loss–of–Coolant Spent–Fuel Storage Pool Accidents, J. L. Collins and J. L. Binder, ORNL/TM-2004/20, Oak Ridge National Laboratory, April 2005.

2010

“Spray Sensitivity Study Performed with the MELCOR Code,” J. J. Carbajo and A. Drozd, Trans. Am. Nucl. Society, 102, pp. 324–326 (June 2010).