Post # 49: That's 1 step forward and 2 steps back...

One of my favorite sources of global energy data is BP's Statistical Review of World Energy, published annually.  The 2011 version (which actually reports data from 2010) can be found here.

Some highlights:

1. Global energy consumption grew by 5.6% – the strongest growth since 1973.

2. China's share of global energy consumption was 20.3%.  The U.S. share was 19.0%

3. Total global oil consumption grew by 3.1% – the smallest consumption growth of all the fossil fuels.

4.  Global natural gas consumption grew by 7.4% – the strongest growth since 1984.

5.  Coal accounted for 29.6% of global energy consumption – the highest percentage since 1970.  China's share of global coal consumption was 48.2%

6.  Renewable energy consumption accounted for just 1.8% of global energy consumption, but U.S. renewable energy consumption grew by 16.3% from 2009 to 2010.

7.  Global hydroelectricity consumption grew by 5.3%, but U.S. hydroelectricity consumption dropped by 6.0%.

8.  Global nuclear energy consumption grew by only 2.0%, and just 1.0% in the U.S.

So...

We are burning more coal, damming more rivers, burning more natural gas, and straining to maintain our nuclear energy production.

Burning more coal is not a good thing until we can find a way to do it cleanly - from both the carbon capture and storage, and ash management perspectives.

As a conservationist, I'm rarely in favor of damming free-flowing streams and rivers.  I've seen first-hand the archaeological, cultural, and agricultural damage this can do.  I'm in favor of exploring kinetic hydropower.  The jury is still out, but the prospects appear promising.

Natural gas burns more cleanly than coal.  That's good.  The U.S. has significant natural gas reserves.  Problem is, it's down there in a geological "lock box" and a controversial process called "hydrofracking" is currently required to recover it.  See here and here for additional information and views on hydrofracking.  Hydrofracking is definitely one of the technologies at the center of the "energy-water nexus".

I've often said I believe there is no sustainable solution to our global energy problem that doesn't rest upon a major expansion of nuclear energy.  Energy conservation is an appropriate focus for those of us who are part of the 1.5 billion inhabitants of this globe who live in the developed world.  But energy production, and lots of it, is the only option available for the over 4 billion people on this small blue planet who have little or no access to energy - especially electricity.

So there you have it... the view from and Sherrell's sound-bite interpretation of the latest BP Statistical Review of World Energy.

Just thinking,

Sherrell

Post # 48: TV-Asahi Interview on BWR Station Blackout

The Fukushima Dai-ichi accident in March of this year, has stimulated a great deal of interest in Japan in the BWR severe accident work done in the U.S. over the past few decades.  In late June, the Japanese television network, tv-asahi sent a team over to interview a number of individuals in the U.S. about that work.  TV-Asahi requested an interview with me to discuss the work done at ORNL on BWR station blackout severe accidents in the early 1980s, with particular interest in the 1981 station blackout report I've discussed in prior posts.  The interview was conducted at ORNL on June 28.  This posting is my prepared pre-interview Q&A script, based on the questions tv-asahi provided me in advance of the interview.  The interview session closely followed this Q&A dialog...

QUESTION: When did you first work on an SBO study?

ANSWER: 

·      The U.S. Nuclear Regulatory Commission launched the Cooperative Severe Accident Research Program (CSARP) in the wake of the 1979 accident at TMI-2.  CSARP included a variety of R&D activities.  Among them was a focus on detailed systems-level accident sequence analysis for the risk dominant sequences previously identified by the WASH-1400 Reactor Safety Study in 1975.  ORNL was selected by the NRC in 1980 as the lead lab for analysis of Boiling Water Reactor severe accident sequences in a program called the BWR Severe Accident Sequence Analysis (SASA) Program.  Station blackout (which had been identified by WASH-1400 as an risk dominate sequence for BWRs) was selected as the first accident to be analyzed.

·      Independent of NRC activities, the industry launched the Industrial Degraded Core (IDCOR) Rulemaking Program in 1981.  IDCOR ran through about 1989, and included independent code and model development, plant analyses, and experiments funded directly by the nuclear industry. 

QUESTION: What kind of study was it?

ANSWER:

·      The 1981 study was the first detailed analysis of the Browns Ferry Unit-1 unmitigated station blackout scenario.  BNFP-1 is a 3440 MWt / 1152 MWe BWR-4/Mk-I plant, larger than Fukushima Dai-Ichi Units 1-5.  We focused on identifying the sequence of events in the accident, the timing of these events, analyzing the role of operator actions and various plant systems and equipment, and identifying uncertainties and unknowns in the analysis.

·      It is important to understand that our analyses were conducted with computer models that were extremely primitive compared with the computer simulation tools we have available today.  Nevertheless, the overall sequence of events developed in the 1981 analysis has held-up as being basically valid through many subsequent analyses since that time.

QUESTION: What was the result of that study? What did you find out?

ANSWER:

·      The major accident sequence events and their timings were estimated based on an assumed 4-hr station battery life and some specific assumptions regarding operator actions.  The potential role of the operator in delaying the onset of core damage was identified, along with some system hardware modifications that would be beneficial.

QUESTION: In an SBO event, what are the primary risks a plant eventually faces if not resolved?

ANSWER:

·      PROVIDED THERE IS NO OTHER PLANT DAMAGE associated with the event that led to loss of off-site power and on-site power, the plant can recover normally without damage, if on-site or off-site power is restored before the station batteries are exhausted.

·      Assuming a 4-hr battery lifetime, and no special operator actions, core damage would begin between 1 and 2 hours after the batteries are exhausted (~ 6 hours after the initiating event).

·      Unless reactor cooling is restored, the accident sequence would progress through core oxidation, melting and relocation to the lower regions of the reactor vessel.  During the period, a few hundred kg of hydrogen would be produced due to interaction between steam and the over-heated fuel assemblies.  Eventually, the lower head of the reactor vessel would fail, allowing hot core debris to fall onto the concrete floor of the primary containment “drywell”, where it would interact with the concrete, releasing more hydrogen, other non-condensible gases, and steam.  Along the way, various radioactive materials would be released into the primary containment (drywell and wetwell).  Eventually, the primary containment would fail due to over-pressure and temperature, releasing its mixture of combustible gases and radioactive material into the surrounding reactor building.

QUESTION: In the case of this hypothetical SBO in the study, what was the sequence of events?  (i.e. _ occurs after 10 minutes, after 30 minutes, _ long until containment failure, _ hours until core meltdown?)

ANSWER:

  First few seconds

·      Recalling that our assumption was that other than losing off-site power and on-site diesel generators, the plant was undamaged…

·      Within the first few seconds, the plant senses the loss of power, and vital plant functions transfer their load to the station batteries, which in our case were assumed to last for four hours.

·      The reactor “scrams” or shuts down.  It’s power level drops from 100% power to a few % within seconds and down 2% or so, slowly decaying for hours and days after this point.

·      Normal cooling water to the reactor ceases, and steam flow to the electrical turbines is terminated.  The reactor is “isolated”.

·      The reactor’s safety/relieve valves (SRVs) function automatically as designed to control reactor pressure, but venting steam from the upper regions of the reactor into the pressure suppression pool

  First few minutes

·      The plant’s HPCI and RCIC systems trigger and begin injecting water into the reactor vessel. (The plant we analyzed did not have an isolation condenser.) These systems draw water from a large water storage tank called the “condensate storage tank” which sits outside the reactor building.  Everything is working as designed during this period.

  First 4 hours

·      During the next four hours, up until the station batteries are exhausted, the combined effects of HPCI/RCIC system injection, coupled with SRV actuation, keep the reactor core covered and undamaged.  However, both the pressure suppression pool and the drywell atmosphere are heating up and slowly pressurizing.  The plant could recover normally without damage if power is restored during this period.

 @ 4 hours

·      Station batteries are depleted, and all water injection to the reactor ceases at 4 hours due to the assumed 4-hr battery lifetime.

 4-5 hours

·      The water level in the reactor drops as water is boiling off and the steam is dumped to the pressure suppression pool through the safety relief valves.  The water level drops to the top of the core in approximately 1-hr after battery depletion.

 5-6.5 hours

·      The core begins to overheat as the water level in the reactor continues to drop below the top of the fuel.  The fuel heats up, and various fuel assembly and control plate materials interact with the hot steam releasing a few hundred kg of hydrogen into the reactor and (via the safety/relief valves) into the pressure suppression pool.  The different core components and materials overheat and melt at different temperatures, but the overall effect is for the core to melt and relocate downward into the lower regions of the reactor vessel – ultimately interacting with the lower head of the reactor vessel and the various penetrations in the lower head.

 @ 7 hr

·      The bottom head of the reactor vessel fails due attack from the hot core debris inside the reactor vessel

 7 – 8.5 hr

·      Molten core debris escapes the reactor vessel and falls upon the concrete floor of the primary containment drywell.

·      The hot core debris interacts with the drywell floor concrete, releasing steam, a variety of non-condensible gases, and radioactive aerosols (smoke) into the drywell atmosphere.

·      The primary containment drywell pressure and temperature increase due to the effects of the core-concrete interactions

@ 8.5 hr

·      The flexible seals in the primary containment drywell electrical penetrations fail due to the combined effect of high pressure and high temperature.

 It should be noted that the timings of major sequence events (such as core uncovery, reactor vessel failure, etc.) are very sensitive to the assumed battery life and key assumptions about operator actions to depressurize the reactor.  I recall a later re-analyses of the SBO sequence with an assumed battery life of 6-hr rather than 4-hr, and with an assumption the operators take steps to depressurize the reactor.  Reactor core uncovery time for that sequence was estimated to be delayed until ~ 10.5 hrs (rather than 5 hrs), and reactor vessel failure was estimated to be delayed till ~ 15.5 hrs after the initiation of the event (rather than 7 hrs).

QUESTION: What system safeguards are critical to deal with an SBO event?  What do you think about what countermeasures or coping measures are needed in this scenario?

ANSWER:

·     The specific answers to the question will vary from plant to plant.  In general, the functions that must be maintained are reactor core and containment cooling to avoid core damage and containment failure.

·      In general, the following types of countermeasures can help assure the key reactor and containment cooling functions are maintained:

o  Assured power supply: multiple off-site power feedlines to the plant, multiple emergency diesel generators with secure fuel sources, and multiple, long-life station batteries, combined with the ability to physically import units from off-site when/if needed.  The physical placement of these resources on-site, and the manner and location in which they are interfaced with and connect to the power plant are also important.

o  Assured cooling water supply: a secure, large condensate storage tank capable of supplying water to the HPCI/RCIC system for extended periods (this was not an issue in our SBO analyses).  Alternatively, dedicated diesel-powered portable pumps can be staged to provide this function from other water sources.

o  Assured reactor vessel pressure control: Reactor vessel depressurization has been shown to be a very useful severe accident mitigation technology.  SRV operability is essential to accomplish such depressurizations.  An assured control air supply is required to maintain SRV operability for the long periods of time involved in an SBO.  This can be provided by secure bottled gas systems.

o  Assured containment cooling: A secure means of cooling the primary containment pressure suppression pool and drywell atmosphere under SBO conditions. A number of options are possible, but the use of diesel-driven RHR pumps and drywell coolers powered by backup power systems are options.

o  Assured containment pressure control: the ability to vent the primary containment if necessary to relieve containment pressure while scrubbing radioactive material from the vented gas and avoiding hydrogen explosions is very important.

o  Remedial reactor vessel cooling and debris cooling: the ability to flood the primary containment with water up to about 2/3 the height of the reactor vessel, coupled with simple modifications to the reactor support skirt, can be effective in preventing reactor vessel failure and cooling any core debris that escapes the reactor.

o  Assured station condition monitoring: Instrumentation that continues to function in an SBO to provide the operators critical information about the state of the reactor, containment, and critical systems.

o  Assured operator preparedness: The power plant operators can play critical roles in managing the accident.  Realistic, focused training of the operators to cope with the real-life circumstances to be expected in an SBO is essential.

 All of these insights emerged from the work performed 1975 and 2000.

QUESTION: How do you feel about the dangers of an SBO?

ANSWER:

·       Numerous studies have shown the importance of SBO as a contributor to the overall risk profile of commercial BWR nuclear plants.  During the past thirty years, nuclear plant designers, regulators, operators here in the U.S. have devoted a great deal of attention to this fact and have taken a number of actions to respond accordingly.  Nevertheless, the Fukushima Dai-Ichi accident demonstrates there’s more work to be done.  The nuclear industry must and will learn and improve from this unfortunate event.

Post # 47: The Race for Rare Earths

Back in Post # 27, I discussed the relevance of rare earths to our high-tech society.  Rare earths are used in an astonishing variety of today's high-tech items.  Computer displays, windmill turbines, solar PV arrays, loud speakers, cell phones, and a plethora of other gadgets we rely on daily would not exist, or would not be nearly as functional, without their rare earth ingredients.

Though the U.S. was the major producer of rare earths for many years, China is currently the source for ~ 97% of rare earth production.  (There's an interesting short video piece on this topic here.

All of this could change if the recent news out of Japan plays-out.  It seems the Japanese have been searching the Pacific sea bed floor for rare-earth-rich ocean sediments and have apparently hit the "jackpot".  According to recent media postings here, the sediment in 1 square km of ocean sediment (found at 3,500 to 6,000 meters below the surface of the ocean) in some locations, could provide 20% of the current total annual world production of rare earths.  The total estimated inventory could be as high as 100 billion tons of these valuable natural resources.  If these repositories can be economically recovered, the potential exists to reduce China's monopoly on rare earths and possibly even reduce their cost.

There's just one wee tiny challenge.... how does one mine these deposits in the dark depths of the ocean, without creating unintended and unacceptable consequences?  The dialog is beginning.  See here and here.  The easiest approach appears to involve leaching the minerals out of the ocean sediment with acid.  Does one do that on the sea-bed floor, or can the sediments be mined and transported to the surface for this treatment?  And what else does one bring up with these sediments?  How are the residues and "waste streams" from this process treated and returned to the environment?  What about the damage to the deep ocean eco-system? All of these questions and many more make for an intense debate - complicated by the fact that no one owns the sea bed in international waters where most of these deposits have been found.

So, some interesting news on this Fourth of July.

Wishing all of you... especially those serving us in harm's way and their families here at home, a peaceful and happy Independence Day.

Sherrell

Post # 46: TBS Interview – ORNL 1981 BWR Station Blackout Analysis (NUREG/CR-2182)

 Since the unfortunate event at Fukushima, there's been a significant amount of attention focused on the work performed at ORNL in the 1980's and 1990's on BWR severe accidents.  On June 23, I sat with a crew from the Tokyo Broadcasting System to discuss ORNL's 1981 BWR Station Blackout severe accident sequence analysis study (NUREG/CR-2182, Vol. 1).


This (very long) post is a compilation of my prepared notes from the interview Q&A session.

QUESTION # 1: Please provide us with a brief background of this project. How did ORNL see the necessity to conduct such studies in the early 1980s?

·       MY ANSWER: 

      

         What we now call “severe accidents” were first studied in 1957, when the AEC released WASH-740.  They looked at a hypothetical Maximum Credible accident in which 50% of the core of a (small by today’s standards) nuclear power plant were released into the atmosphere 30 miles from a major city.  They did not estimate the probability of such accidents.

·       F. Reginald Farmer of the UK pioneered PRA methodologies and published a paper in 1965 in which he defined many of the fundamental approaches to reactor safety risk assessment that are still used today.

·       The AEC published the WASH-1400 “Reactor Safety Study” in 1975.  It was the first systematic attempt to assess and compare the risk of reactor operations to other common activities within society.  Among other things, the RSS concluded

(1) the risks to the public of commercial nuclear power plant operations (expressed as expected fatalities, injuries, and property damage from reactor accidents) was small compared to other normal activities of life such as driving a car, commercial aviation, tornadoes and hurricanes, etc.

(2) the risk of reactor operations was dominated by transients such as station blackout, loss of decay heat removal, and failure to shut down the reactor; and small break loss of coolant accidents (LOCAs) – rather than the large-break LOCAs.

·       The accident at Three Mile Island Unit-2 occurred in March 1979, and involved an accident sequence of the generic type the WASH-1400 study concluded was risk-dominant.

·       Immediately following the TMI-II accident, the U.S. Nuclear Regulatory Commission launched the Cooperative Severe Accident Research Program (CSARP), which included a variety of R&D activities.  Among them, was a focus on detailed systems-level accident sequence analysis for the risk dominant sequences identified by WASH-1400.  ORNL was selected by the NRC to conduct the Boiling Water Reactor Severe Accident sequences in a program called the BWR Severe Accident Sequence Analysis (SASA) Program.  Station blackout (which had been identified by WASH-1400 as an risk dominate sequence for BWRs) was selected as the first accident to be analyzed.

·       Independent of NRC activities, the industry launched the Industrial Degraded Core (IDCOR) Rulemaking Program in 1981.  IDCOR ran through about 1989, and included independent code and model development, plant analyses, and experiments funded directly by the nuclear industry. 

·       I joined ORNL in December 1978 immediately after graduate school (three months before the TMI-2 accident), and joined the SASA team in 1980.  

QUESTION # 2: How were the studies conducted and what were the major findings?  Please walk us through the accident sequences resulting in meltdown that were observed in the study.

·       MY ANSWER:

         First, it is important to understand that severe or “core melt” accident progression in nuclear reactors is generally broken into “phases”:

o   Pre-core damage phase – from the triggering event to the beginning of core uncovery or damage.  

o   In-vessel degraded core phase – from beginning of core damage to failure or melt-through of the reactor vessel and escape of core debri into the primary containment

o   Ex-vessel phase – from failure of the reactor vessel to failure of the primary containment, secondary containment, and release of radioactivity material into the environment.

o   Post-containment failure phase – continued release of radioactivity from the reactor and it’s transport and deposition beyond the reactor site.

·       Second, it is important to understand that our analyses were conducted with computer models that were primitive compared with the computer simulation tools we have available today.  Nevertheless, the overall sequence of events developed in the 1981 analysis has held-up as being basically valid through many subsequent analyses since that time.

·       As far as the accident progression goes:

ACCIDENT INITIATION:

a.    the station blackout analysis we analyzed focused on the so-called “long-term station blackout”.  The analysis we did in 1981 assumed a simple sequence in which the station battery lifetime was assumed to be only 4 hours, and we further assumed the operators did not take action to depressurize the reactor vessel (more about this later);

b.    The reactor (BWR-4 / MK-I containment) was assumed to be operating at full power (3440 MWt / 1152 MWe), pressure (~1020 psig / 7 MPa), and temperature (530 ºF / 275 ºC);

c.    An un-specified event was assumed to occur that resulted in complete loss of off-site power;  

d.    In addition, it was assumed that the station emergency diesel generators failed to start and load (as they normally would) upon loss of off-site power.  Thus the station emergency diesel generators were assumed to never play a role in the plant’s response to the loss of off-site power;

e.    Because neither off-site AC power or the emergency diesel generators are available, all power is supplied by the station batteries, which in our case, were assumed to be available for 4 hours.

f.     It is very important to understand that no other equipment damage was assumed as part of the accident initiation.  All other plant systems and equipment were assumed to be undamaged.

FIRST FEW MINUTES:

g.    The plant “scrams” or shuts down in a first few seconds, and the reactor power level drops to just a few percent of full power, to ~2% within 1 hour, ~1% at 10 hours, and ~ 0.5% within a day.  However, reactor power is still at ~ 0.3% after 1 week.  For a large modern 1 GW-class BWR, this is still 50-70 MW of energy being released an hour after the reactor is "scramed". 

h.    The main steam isolation valves close, and the reactor is isolated behind the main steam isolation valves.

i.      Steam-turbine driven reactor coolant systems act automatically to draw water from a large water storage tank outside of the reactor and inject it into the reactor to keep the core covered, adequately cooled, and in a safe condition.

j.      The relieve valves in the reactor cycle automatically to maintain safe reactor pressure by venting reactor vessel steam into the million-gallon pressure suppression pool. Operators can also manually actuate these relieve valves to control reactor pressure and assure the pressure pool is not over-heat due to repeated venting of the same relieve valve.

FIRST 4 HOURS:

k.    Barring reactor operator actions to the contrary, the situation described above continues for four hours, at which time the station batteries were assumed to be exhausted, and all water injection into the reactor ceases.  Up until this time, the reactor would recover with no damage if AC power or emergency diesels were restored.

l.      Following exhausting of station batteries, it is no longer possible to inject water into the reactor vessel or to depressurize the reactor vessel.  The reactor water level gradually decreases as the reactor pressure relief valves dump steam into the suppression pool. Thus station battery lifetime is a key determinant of the time to initial core uncovery.

HOURs 4-6:

m.  5 hrs: The reactor core begins to uncover at 5 hours ( ~ 1 hour after the station batteries are exhausted) and the reactor fuel assemblies begin to heat up.

n.    The water level in the rector continues to drop and the reactor fuel assemblies temperatures rise, eventually passing 1000 ºC, at which point the zircaloy fuel assemble structures and fuel cladding being to react with the steam, releasing more energy and hydrogen (which is being vented to the pressure suppression pool through the relief valves).  Can generate 500-600 kg of hydrogen or more.

HOURs 6 - 6.5:

o.    6 hrs: The reactor water level drops below the core and stainless steel control blades and fuel melting begins ~ 2 hours after core uncovery. A large BWR such as Browns Ferry can have over 250 MT of fuel in it – not including control blades and other structures. Molten control plate material, zirconium structure, fuel pellets, and melted fuel move downward or “candle” down to the reactor core plate.

Material	Melting Temperature (ºC)
Stainless Steel	1450
Zircalloy	1850
Boron Carbide	2440
UO2	2800

p.    6.5 hrs: The core collapses into the lower head @ 2.5 hrs after cover uncovery

HOURs 6.5 - 7:

q.     7 hrs:  The reactor vessel fails and core debris leaves the reactor and falls in the containment floor 3 hrs after core uncovery.  

HOURs 7 - 8.5:

r.    8.5 hrs: The hot core debris escaping the reactor vessel begins interacting with the concrete floor of the drywell, releasing steam, and non-condensible gases, accelerating the heatup and pressurization of the primary containment.


s.    The drywell (primary containment) fails at ~ 8.5 hr as the seals in drywell electrical penetrations are blown out, allowing radioactive gases, aerosols and hydrogen to flow into the surrounding reactor building.

Our understanding of the details of the station blackout severe accident sequence discussed above has evolved since 1981, as various experiments were conducted, better computer models became available, and more analyses were completed.  Still, there are some areas of uncertainty in the details of important phenomena.  However, the basic sequence of events developed in the 1981 analysis has been confirmed by several subsequent analyses.  It is also important to note that ORNL re-ran the accident analyses in the early 1990s with an assumed 6 hour battery life (rather than 4 hour) and with the assumption the operators DO take action to depressurize the reactor vessel as we recommended.  Under these assumptions, the timing of major events roughly doubled (e.g. ~ 10 hours to core uncovery and reactor vessel failure not occurring until over 15 hours after the start of the event).

QUESTION # 3: Based on these studies, what suggestions did ORNL make to the NRC? 

·       MY ANSWER:

    A few important observations were documented in the 1981 report...


    Some changes in automatic plant system configurations could be helpful in delaying the accident event progression timing (such as adjusting the rules for switchover of the High Pressure Coolant Injection System water source from the condensate storage tank to the pressure suppression pool).

·         Pro-active operator actions could be helpful in delaying accident event times.  Chief among these was the benefit of depressurizing the reactor vessel before station batteries are exhausted.  Depending on the timing of this action, it can have several beneficial consequences.  If done early in the accident, it significantly reduces the rate at which heat is transferred from the reactor vessel to the drywell (primary contaiment) atmosphere – delaying primary containment temperature and pressure increases than can lead to containment failure.  If done later in the sequence (just before or early in the core damage phase, this action quenches the core, drops the vessel water level below the fuel, limits the steam source available to drive exothermic heatup of the core due to steam/zirconium reactors, (and the resultant hydrogen production), and reduces the heatup rate of the drywell. 

·        A second recommended operator action was to take manual control of the reactor vessel pressure relieve valves to rotate SRV operation around the pressure pool, and prevent localized heatup and over pressurizing of the pressure suppression pool.

·       In subsequent studies the ORNL team evaluated:

o   drywell flooding as  potentially important severe accident mitigation procedures to prevent reactor vessel melt-through

o   containment venting to prevent containment failure (necessary for drywell flooding)

o   borating condensate storage tank water to prevent recriticality during core reconfiguration

·       And in general, it was learned that plant-specific design details (such as the lifetime of the batteries, the size of the condensate storage tank and pressure suppression pools, etc.) and operator actions can have significant impact on the event sequence timing.

QUESTION # 4: What kind of impact have these studies had on the safety measures and accident-management procedures adopted by US nuclear plants?

·       MY ANSWER:

         First, it is important to understand that ORNL’s analysis was but the first of many such analyses subsequently conducted between 1980 and 2000 by several different organizations (governmental and industrial).  Both the Industry and the federal sector were mobilized.  A great deal of effort was put into understanding and learning from the TMI-II event, expanding our fundamental understanding of severe accident phenomena, and evaluating what could be done to reduce the plants' vulnerability to such accidents.   ORNL's work in the 1980s focused on analyzing the major BWR risk-dominant accidents from WASH-1400 (station blackout, small-break loss of coolant accidents, loss of decay heat remove events, and transients without scram events).  The work continued through the 1990s, shifting to a focus on identifying accident prevention, coping, and mitigation procedures. There was a significant collaborative effort with the nuclear industry to improve their severe accident management procedures.  As previously mentioned there were major parallel and complementary severe accident programs at the NRC and in Industry.  All during this period, the both programs included an array of experimental efforts, computer code development activities, and plant-specific accident sequence and PRA analyses.

·       In 1988, the NRC required every operating nuclear plant in the U.S. to conduct a so-called Individual Plant Examination (IPE) that was designed to understand how each plant fit within the generic risk profile established by the Reactor Safety Study and the so-called NUREG-1150 risk study. 

·       So during the period from 1980 through 2000, all of the U.S. plants were evaluated with respect to their risk of severe accidents, and many plant-specific full-scope probabilistic risk assessments were performed.

·       There was a multi-year effort within industry to update their severe accident management guidelines, though I cannot say that all of the ORNL recommendations developed during the 1980s and 1990s were implemented.

·       Following the Sept. 11, 2001 attack, the NRC also issued a directive that all plants again analyze and address any severe accident vulnerabilities to terrorist activities resulting in large scale fires or explosions."

END OF INTERVIEW NOTES....

Well, that's all for now.  Take care everyone and have a wonderful Independence Day!

Sherrell