Sunday, August 31, 2014

Rickover and Rigged Research

Obviously, adverse data could have had serious consequences.

And, an effective way to avoid adverse data is to rig the research.  One set of rigged research is 
the Baker-Just equation which has its roots at Rickover's Bettis.  This is documented on pages 345 and 346 of:

Nuclear Power from Underseas to Outer Space
American Nuclear Society, 1995 - Technology & Engineering - 467 pages

John Simpson, former president of Westinghouse Power Systems Company and past president of the American Nuclear Society, provides a vibrant account of the events associated with the birth of the nuclear industry. Simpson's account of his career and the many turns it took is formidable. 

Following are excerpts from pages 345 and 346:

The question was raised whether ignition of zirconium, even if locally initiated, could spread in an uncontrollable manner to a major fraction of the zirconium contained in a core. The accident condition under which such a reaction was considered most likely to occur was a loss-of-coolant accident in which a coolant pipe would suffer a major break.

If it could be shown that Zircaloy-4 reacted with water in a predictable manner at all temperatures, up through its melting point, this should assuage fears of unpredictable and uncontrollable reactions.  These data were required in less than three weeks.  An experiment was devised by Bill Bostrom in which samples of zirconium tubing similar to those used in Shippingport were heated  inductively while immersed in water; the released volume of  hydrogen was collected and used to monitor the rate  of reaction. Barricades were set up around the water container and a system of mirrors was devised so engineers could control metal temperature during the reaction.  Another engineer 
 observed the change in hydrogen volume as it bubbled  up through the water column.  Temperatures were gradually 
increased in consecutive experiments until, as a grand finale, zirconium was melted under water and the amount of reaction monitored during melting. 

These experiments, completed during the required time  period, demonstrated that the occurrence of high temperature reaction rates could be extrapolated from those measured at lower temperatures, and that no new or unexpected phenomena intervened that would endanger reactor plant safety. Crude as these initial experiments were, the  kinetic data derived from them and the conclusions 
drawn have been supported by subsequent experiments
and analyses.  Obviously, adverse data could have had serious consequences.

Please read page 1 of 8, ignore the rest.

Saturday, August 30, 2014

Enformable reports spent fuel pool accident at Fukushima Daiichi Reactor 3

Posted: 29 Aug 2014 06:11 AM PDT
At the crippled Fukushima Daiichi nuclear power plant, workers accidentally dropped a large piece of debris into the Unit 3 spent fuel pool on Friday, a little after noon.
The workers were carrying out operations to remove debris with a large remote controlled crane.  At the time of the accident, workers were manipulating the control console for the refueling machine, a piece of equipment that weighs almost a thousand pounds.
Tokyo Electric, who is in charge of cleanup operations at Fukushima Daiichi, told reporters that they have not detected any change in radiation levels around the spent fuel pool after the accident.
TEPCO is working to check the 566 spent fuel assemblies in the Unit 3 spent fuel pool to see if any of them have been damaged by the most recent accident.  According to decommissioning plans, the utility is scheduled to start removing spent fuel rods from the Unit 3 spent fuel pool in the first half of 2015 at the earliest.
This is not the first time that debris and large objects have been accidentally dropped, pulled, or pushed into the Unit 3 spent fuel pool.  Between 2012 and 2013, TEPCO workers used the remote control cranes to remove debris from atop the Unit 3 reactor building, and multiple instances were recorded where operators moving cranes via remote control knocked debris into the spent fuel pool or dislodged other materials on the roof.
In February 2013, workers accidentally knocked the 1.5 ton fuel handling machine mast into the Unit 3 spent fuel pool, and it was later found to have come to rest on top of the spent fuel racks after it narrowly avoided damaging the liner of the spent fuel pool.

Saturday, August 23, 2014

Mark Edward Leyse’s Comments on the Proposed Rule: 10 C.F.R. § 50.46(c)

Mark Edward Leyse’s Comments on the Proposed Rule for Performance-Based Emergency Core Cooling Systems Cladding Acceptance Criteria: 10 C.F.R. § 50.46(c)
Mark Edward Leyse’s Comments on the Proposed Rule for Performance-Based Emergency Core Cooling Systems Cladding Acceptance Criteria: 10 C.F.R. § 50.46(c)…………………...………3
I. Statement of Commenter’s Interest…………………..…………………………………………3
II. Considering the Thermal Resistance of Crud and/or Oxide Layers in Loss-of-Coolant Accident Analysis.……………………………………………………………………….………..4
III. Results of the PHEBUS B9R-2 Test Pertain to How Breakaway Oxidation Could Affect a LOCA ….……………………………………………………………………………...…………..6
IV. Results of the PHEBUS B9R-2 Test Pertain to the 2200°F Peak Cladding Temperature Limit….………………………………………………………………………………..…………..9

In these comments, Mark Edward Leyse responds to the U.S. Nuclear Regulatory Commission’s (“NRC”) solicitation of public comments—published in the Federal Register on March 24, 2014—on a proposed rule, revising the acceptance criteria for the emergency core cooling system (“ECCS”) for light-water nuclear power reactors.
I. Statement of Commenter’s Interest
On March 15, 2007, Mark Edward Leyse submitted a 10 C.F.R. § 2.802 petition for rulemaking, PRM-50-84,1 to NRC. PRM-50-84 requested that NRC make new regulations: 1) to require licensees to operate light water reactors under conditions that effectively limit the thickness of crud (corrosion products) and/or oxide layers on fuel cladding, in order to help ensure compliance with 10 C.F.R. § 50.46(b) ECCS acceptance criteria; and 2) to stipulate a maximum allowable percentage of hydrogen content in fuel cladding.
Additionally, PRM-50-84 requested that NRC amend Appendix K to Part 50—ECCS Evaluation Models I(A)(1), “The Initial Stored Energy in the Fuel,” to require that the steady-state temperature distribution and stored energy in the fuel at the onset of a postulated loss-of-coolant accident (“LOCA”) be calculated by factoring in the role that the thermal resistance of crud and/or oxide layers on cladding plays in increasing the stored energy in the fuel. PRM-50-84 also requested that these same requirements apply to any NRC-approved best-estimate ECCS evaluation models used in lieu of Appendix K to Part 50 calculations.
In 2008, NRC decided to consider the issues raised in PRM-50-84 in its rulemaking process.2 And in 2009, NRC published “Performance-Based Emergency Core Cooling System Acceptance Criteria,” which gave advanced notice of a proposed rulemaking, addressing four objectives: the fourth being the issues raised in PRM-50-84.3 In 2012, the NRC Commissioners voted unanimously to approve a proposed rulemaking—revisions to Section 50.46(b), which will
1 Mark Leyse, PRM-50-84, March 15, 2007 (ADAMS Accession No. ML070871368).
2 Federal Register, Vol. 73, No. 228, “Mark Edward Leyse; Consideration of Petition in Rulemaking Process,” November 25, 2008, pp. 71564-71569.
3 Federal Register, Vol. 74, No. 155, “Performance-Based Emergency Core Cooling System Acceptance Criteria,” August 13, 2009, pp. 40765-40776.
become Section 50.46(c)—that was partly based on the safety issues Petitioner raised in PRM-50-84.4
Leyse also coauthored a paper, “Considering the Thermal Resistance of Crud in LOCA Analysis,”5 which was presented at the American Nuclear Society’s 2009 Winter Meeting.
II. Considering the Thermal Resistance of Crud and/or Oxide Layers in Loss-of-Coolant Accident Analysis
As published in SECY-12-0034, regarding the thermal effects of crud and oxide layers, the proposed rule for Section 50.46(c), Paragraph (g)(2)(ii), states:
The thermal effects of crud and oxide layers that accumulate on the fuel cladding during plant operation must be evaluated. For the purposes of this paragraph, crud means any foreign substance deposited on the surface of fuel cladding prior to initiation of a LOCA.6
Paragraph (g)(2)(ii) needs to be augmented with additional instructions, explaining that licensees are required to conservatively evaluate the thicknesses and thermal conductivities of crud and/or oxide layers for each fuel cycle.
Licensees will be better able to conservatively evaluate how the thermal effects of crud and oxide layers would increase the peak cladding temperature in the event of a LOCA, if they first conservatively evaluated the thicknesses and thermal conductivities of the crud and/or oxide layers present during each fuel cycle.
At the end of some fuel cycles, it would be helpful to examine the thicknesses and thermal conductivities of crud and oxide layers that had accumulated on fuel cladding; and estimate the quantity of non-tenacious crud released from fuel rods during the refueling outage. This would help to provide valid data for benchmarking ECCS evaluation models.
Clearly, there are a number of factors that could play a role in how much crud would be present in any forthcoming operating cycle. Ultrasonic fuel cleaning could be used to remove a portion of the existing crud; and of course more crud would accumulate on the fuel cladding.
4 NRC, Commission Voting Record, Decision Item: SECY-12-0034, Proposed Rulemaking—10 CFR 50.46(c): Emergency Core Cooling System Performance During Loss-of-Coolant Accidents (RIN 3150-AH42), January 7, 2013, (ADAMS Accession No. ML13008A368).
5 Rui Hu, Mujid S. Kazimi, Mark Leyse, “Considering the Thermal Resistance of Crud in LOCA Analysis,” American Nuclear Society, 2009 Winter Meeting, Washington, D.C., November 15-19, 2009.
6 NRC, “Proposed Rulemaking: 10 CFR 50.46c: Emergency Core Cooling System Performance During Loss-of-Coolant Accidents,” SECY-12-0034, March 1, 2012, p. 84.
But it’s important to note that there are models of crud and oxide deposition that have been developed that are intended to predict of the thicknesses of crud deposits and oxide layers on the fuel cladding. (For example, such information is mentioned in a 2003 paper titled “Taming the Crud Problem: The Evolution.”7)
In the question-and-answer session after a presentation I gave at NRC headquarters on June 24, 2014 regarding the Section 50.46c proposed rule, Paul Clifford made what I believe is an important point. He pointed out that during a fuel cycle there could be “fluffy,” non-tenacious crud on the fuel rods, which would not be observed at the end of the fuel cycle. Non-tenacious crud can be released from fuel rods during refueling outages; this is sometimes termed a “crud burst.” Regarding some observed crud bursts, Electric Power Research Institute (“EPRI”) states that “[s]everal PWRs have experienced anomalous crud releases during refueling outages, characterized by unexpectedly high particulate crud releases followed by deposition or by abnormally high activity releases after peroxide addition (or release after floodup).”8
It is pertinent that EPRI has developed a BWR crud-deposition model called the Crud DepOsition Risk Assessment ModeL (“CORAL”),9 “[t]o facilitate improved management of any crud-related fuel performance risk.”10
Regarding CORAL, EPRI states:
The BWR Fuel Crud Model provides a prediction of the crud deposition and tenacious crud layer thickness both radially and axially within a selected fuel assembly throughout the entire operating history of the assembly. The analysis inputs include the fuel assembly geometry and actual or projected operating history, as well as crud inputs determined from a reactor system mass balance. A boiling deposition model, in conjunction with mechanistic crud release models, defines the deposited inventory along the length of all fuel rods within the fuel assembly. The deposited crud material is separated between an outer loose, fluffy layer and an inner tenacious layer. The thickness of the tenacious layer is determined [emphasis added].
7 Yovan D. Lukie, Jeffrey S. Schmidt, “Taming the Crud Problem: The Evolution,” Advances in Nuclear Fuel Management 2003 Hilton Head Island, South Carolina, USA, October 2003.
8 EPRI, “Product Abstract: High Activity Crud Burst Impacts and Responses,” available at:
9 EPRI, “Product Abstract: Technical Basis and Benchmarking of the Crud Deposition Risk Assessment Model (CORAL),” available at:
10 EPRI, “Product Abstract: Fuel Reliability Program: BWR Fuel Crud Modeling,” available at:
The BWR Fuel Crud Model has been extensively validated through benchmarking of the model using data taken on fuel rods operated in actual commercial BWRs. These benchmarking measurements include (1) poolside fuel deposit sampling of both total and tenacious crud layer, (2) laboratory examination of crud flakes and tenacious deposits on irradiated fuel rods to determine composition, thickness, density, and structure, and (3) poolside eddy current lift-off and cladding diametral profilometry data. This successful benchmarking activity provides confidence in the model’s ability to capture and quantitatively describe crud deposition behavior, as well as quantifying the inherent variability in the crud deposition and release processes and resultant tenacious crud layer thickness11 [emphasis added].
Such models need to be used to predict the thicknesses of crud deposits—for both outer loose, fluffy layers and inner tenacious layers—that would be present on the fuel cladding during each fuel cycle. And such models need to be used for both PWRs and BWRs. Clearly, the thermal effects of crud—for fluffy layers and tenacious layers—and oxide layers need to be evaluated in LOCA analysis; and the thicknesses and thermal conductivities of such layers need to be modeled conservatively.
III. Results of the PHEBUS B9R-2 Test Pertain to How Breakaway Oxidation Could Affect a LOCA
As published in SECY-12-0034, the proposed rule for Section 50.46(c) states:
Breakaway oxidation, for zirconium-alloy cladding material, means the fuel cladding oxidation phenomenon in which weight gain rate deviates from normal kinetics. This change occurs with a rapid increase of hydrogen pickup during prolonged exposure to a high-temperature steam environment, which promotes loss of cladding ductility.12
And Draft Regulatory Guide 1261 states that “breakaway oxidation is an instability phenomenon that can spread rapidly in the axial and circumferential directions” of fuel rods and that there is a criterion of 200-weight parts per million (wppm) for hydrogen pickup. It says that
11 Id.
12 NRC, “Proposed Rulemaking: 10 CFR 50.46c: Emergency Core Cooling System Performance During Loss-of-Coolant Accidents,” SECY-12-0034, March 1, 2012, p. 78.
fuel-cladding ductility is maintained as long as the average hydrogen content is below 435 wppm.13
Draft Regulatory Guide 1261 states:
[T]he 200-wppm hydrogen pickup criterion is conservative by a factor of at least two. However, it is not overly conservative for high oxidation temperatures because the time needed to increase from 200 wppm to >400 wppm hydrogen pickup could be as low as 100 seconds.14
When high burnup and other fuel rods are discharged from the reactor core, the fuel cladding can have local zirconium dioxide (ZrO2) “oxide” layers that are up to 100 microns (μm) thick (or greater); there can also be local crud layers on top of the oxide layers. And according to NUREG/CR-6851, medium to high burnup fuel cladding typically has a “hydrogen concentration in the range of 100-1000 wppm;” it adds that “[z]irconium-based alloys, in general, have a strong affinity for oxygen, nitrogen, and hydrogen…”15
NRC’s conclusions on how hydrogen content affects fuel cladding ductility are based on the results of isothermal experiments conducted with small specimens. These were experiments in which a tiny section of a fuel rod was held at a constant temperature. I think most of this program was done at Argonne; and there were some tests done with pre-oxidized fuel cladding.
The PHEBUS B9R-2 test is an integral experiment conducted with pre-oxidized fuel cladding, which I recommend the NRC study. PHEBUS B9R-2 was conducted in a light water reactor—as part of the PHEBUS severe fuel damage program—with an assembly of 21 UO2 fuel rods.16 A 1996 European Commission report states that the B9R-2 test had an unexpected fuel-cladding temperature escalation in the mid-bundle region; the highest temperature escalation rates were from 20°C/sec (36°F/sec) to 30°C/sec (54/°F/sec).17
13 NRC, “Conducting Periodic Testing for Breakaway Oxidation Behavior,” Draft Regulatory Guide 1261, Undated, Appendix B: Rational for the 200-wppm Hydrogen Pickup Criterion for Breakaway Oxidation,” (ADAMS Accession No: ML111100300), p. B-1.
14 Id.
15 K. Natesan, W.K. Soppet, Argonne National Laboratory, “Hydrogen Effects on Air Oxidation of Zirlo Alloy,” NUREG/CR-6851, October 2004, (ADAMS Accession No: ML042870061), p. iii, 3.
16 G. Hache, R. Gonzalez, B. Adroguer, Institute for Protection and Nuclear Safety, “Status of ICARE
Code Development and Assessment,” in NRC “Proceedings of the Twentieth Water Reactor Safety
Information Meeting,” NUREG/CP-0126, Vol. 2, 1992, (ADAMS Accession No: ML042230126), p. 311.
17 T.J. Haste et al., “In-Vessel Core Degradation in LWR Severe Accidents,” European Commission,
Report EUR 16695 EN, 1996, p. 33.
Discussing PHEBUS B9R-2, the 1996 European Commission report states:
The B9R-2 test…illustrates the oxidation in different cladding conditions representative of a pre-oxidized and fractured state. … During B9R-2, an unexpected strong escalation of the oxidation of the remaining Zr occurred when the bundle flow injection was switched from helium to steam while the maximum clad temperature was equal to 1300 K [1027°C (1880°F)]. The current oxidation model was not able to predict the strong heat-up rate observed even taking into account the measured large clad deformation and the double-sided oxidation (final state of the cladding from macro-photographs).
… No mechanistic model is currently available to account for enhanced oxidation of pre-oxidized and cracked cladding.18
As stated, the cladding-temperature escalation commenced at approximately 1027°C (1880°F). That is thermal runaway. The fact that PHEBUS B9R-2 was conducted with a preoxidized test bundle makes its results pertinent to the cladding of medium and high burnup fuel rods.
The hydrogen content of the cladding of PHEBUS B9R-2 test bundle most likely played a role in the test results. I think the results indicate what could happen in a LOCA; the test was conducted under conditions far more representative of LOCA conditions than the Argonne tests with tiny specimens.
As quoted above, Draft Regulatory Guide 1261 states that breakaway oxidation deviates from normal kinetics. It seems that normal oxidation kinetics are supposed to be those observed in the tests with tiny zirconium specimens held at a constant temperature. In such tests the rate of steam flow is controlled. And different adjustments can influence oxidation rates. This is discussed in a paper by Gerhard Schanz titled “Recommendations and Supporting Information on the Choice of Zirconium Oxidation Models in Severe Accident Codes.” The Schanz paper states that an investigator “reached an important improvement of the specimen temperature homogeneity by only optimizing the geometry of the specimen and registered considerably increased reaction rates.”19
I think that any honest, objective study of the results of the PHEBUS B9R-2 test and those of a number of other integral experiments conducted with multi-rod bundles of fuel rod simulators or real fuel rods with UO2 fuel would reveal many deviations from so-called normal
18 Id., p. 126.
19 Gerhard Schanz, “Recommendations and Supporting Information on the Choice of Zirconium
Oxidation Models in Severe Accident Codes,” FZKA 6827, 2003, p. 5.
oxidation kinetics. The reaction rates have been rapid in a number of the large-scale, integral experiments. In those cases thermal runaway is more of an issue than cladding embrittlement. Preventing thermal runaway could be a more important safety issue than preventing excessive cladding embrittlement.
IV. Results of the PHEBUS B9R-2 Test Pertain to the 2200°F Peak Cladding Temperature Limit
The results of the PHEBUS B9R-2 test should be reviewed, along with other integral experiments to help determine if the proposed rule for Section 50.46(c), Paragraph (g)(1)(i) is non-conservative. That is, the test results of integral experiments may indicate that the 2200°F peak cladding temperature limit needs to be lowered. In a large break LOCA there could be steam-binding conditions that would not allow much coolant to be injected. This could cause the fuel-cladding temperature to increase at a rate of approximately 10°F per second or greater—mainly from the stored energy (heat) in the fuel, at the beginning of the LOCA. And if the peak fuel-cladding temperature were to reach approximately 1832°F in a steam environment; and there were little or no coolant injection, there would probably be results similar to those of the PHEBUS B9R-2 test, in which thermal runaway commenced when peak fuel-cladding temperatures were lower than 2200°F.
Respectfully submitted,
Mark Edward Leyse
P.O. Box 1314
New York, NY 10025

Friday, August 22, 2014

The year 2007 ACRS manipulation to relocate Appendix K to a regulatory guide

The year 2007 ACRS manipulation to relocate Appendix K to a regulatory guide

There are current manipulations to relocate Appendix K to a regulatory guide. The following three bullets are copied from ML14204A009, July 23, 2014.

• Bert Dunn from AREVA provided a presentation recommending relocating the requirements in Appendix K to a regulatory guide (ADAMS Accession No.ML14175A094).

• There was a significant amount of follow-on discussion on this topic, and the NRC staff noted that this was the first time this recommendation had been presented. Industry noted that this recommendation will likely come to the NRC in the form of a formal written comment, associated with the request for comment on renumbering the regulations.

• In a later discussion on the Appendix K topic, NRC’s Office of General Counsel (OGC) noted that there is no legal reason why this information cannot be moved to a regulatory guide, but that both NRC and industry need to be aware of the implications.

AREVA (and likely others) seek to relocate the requirements in Appendix K to a regulatory guide because regulatory guidance is not enforceable.

Following is relevant history.   During 2007, the ACRS encouraged the Chairman of the NRC to move the requirements in Appendix K to a regulatory guide.  NRC staff responded that since regulatory guidance is not enforceable it cannot substitute for regulatory requirements.

Following are excerpts from those letters:

The ACRS recommended that, “The staff should complete planned research needed to provide a sound basis for a regulatory guide applicable to current and future zirconium-based cladding alloys.”

On July 11, 2007, NRC staff replied to ACRS, PROPOSED TECHNICAL BASIS FOR THE REVISION TO 10 CFR 50.46, ML071640115.
The NRC staff pointed out that, “However, an important regulatory challenge in developing a performance-based regulatory requirement for any type of fuel cladding is the need to develop objective and legally-enforceable acceptance criteria. Since regulatory guidance is not enforceable it cannot substitute for regulatory requirements.”

Clearly, AREVA (and likely others) seek to relocate the requirements in Appendix K to a regulatory guide because regulatory guidance is not enforceable.

In its letter of May 23, 2007, the ACRS erroneously declares that, “The correlation specified for the rates of steam reaction with the cladding is viewed by the technical community as an anachronism.”  The fact is that while the correlation specified for the rates of steam reaction with the cladding is indeed an anachronism, it is not viewed as such by the NRC staff as well as by the Nuclear Energy Institute (NEI). In fact, the correlation specified for the rates of steam reaction with the cladding is fiercely defended by the NRC staff as well as by the Nuclear Energy Institute.
This defense is well documented by the NEI in its comment 16 opposing PRM-50-93, “The Baker-Just correlation, using the current range of parameter inputs, has been shown to be conservative and adequate to assess Appendix K ECCS performance. Data published since the Baker-Just correlation was developed has clearly demonstrated the conservatism of the correlation above 1800°F,” refer to ML101040678, Industry Comments on Petition for Rulemaking (PRM-50-93); Multi-Rod (Assembly) Severe Fuel Damage Experiments. Docket ID NRC-2009-0554, April 12, 2010.

The NRC also fiercely defends Baker-Just. In its analysis of PRM-50-76, “The Baker-Just correlation (Reference 4) using the current range of parameter inputs is conservative and adequate to assess Appendix K ECCS performance. Virtually every data set published since the Baker-Just correlation was developed has clearly demonstrated the conservatism of the correlation above 1800°F.” refer to Memo to Matthews/Black-Technical Safety Analysis of PRM-50-76, A Petition for Rulemaking to Amend Appendix K to 10 CFR Part 50 and Regulatory Guide 1.157 - ML041210109, April 29, 2004.

It appears that the above erroneous declaration by the ACRS was designed to augment its recommendation to relocate the requirements in Appendix K to a regulatory guide that would not be enforceable.

Tuesday, August 19, 2014

TEPCO to abandon AREVA system amid contaminated water crisis

Following is from ENFORMABLE.  It is revealing.  What if Pacific Northwest, formerly called WPPSS, had to cope with a situation like this on the Columbia River?

Tokyo Electric has determined that it will cease use of AREVA’s decontamination system, which uses chemicals to remove radioactive materials from water, as it has not lived up to expectations since it was installed.  The utility will file an application with the Nuclear Regulation Authority in order to scrap the system.
The decontamination system was set up in June 2011, three months after the onset of the Fukushima Daiichi nuclear disaster.  The design was so complicated that it took 50 welders more than a month to put the system together.  In the first three months, the system processed 76,000 tons of contaminated water, but was repeatedly forced to be shut down by a variety of problems.
For the last three years the system has been unused and kept out of operations and in the meantime, TEPCO has introduced a new system to process the ever-accumulating amounts of contaminated water at the crippled plant.
One of the problems will be dismantling and disposing of the decontamination system, as it has become contaminated itself after processing the radioactive materials.  The radiation levels measured in the system posed a risk to workers during operation and maintenance of the system.
TEPCO is still facing an uphill battle when it comes to dealing with water storage at the Fukushima Daiichi plant.  Underneath the plant, an enormous amount of ground water continues to flow directly into the ocean.
The utility has a limited amount of water it can store in storage tanks on site, and need to both stem the accumulation of contaminated water on site and the amount of contaminated water reaching the Pacific Ocean.
Three months ago, TEPCO attempted to divert some of the flow of ground water with an underground bypass, but failed to achieve any meaningful results.
The utility filed another application with the NRA this week, which if approved, would allow them to construct a new system for the collection and discharge of contaminated groundwater after it has been processed in order to remove some radioactive materials back into the Pacific Ocean.
This new groundwater dumping plan will likely face fierce opposition from local fisherman, who feel like they have been making continual concessions for TEPCO – which will undoubtedly affect the future of their own ability to operate, with little or no results.

Saturday, August 9, 2014

Sensors within fuel rod assemblies: W developing with academics

20 June 2014
Nuclear power plant operators may soon be able to monitor a reactor core by using sensors within fuel rod assemblies that literally scream the type and location of a problem, Michael Heibel, technical programme manager at Westinghouse, told World Nuclear News.
Westinghouse has applied for a patent on the device it is developing with academics from the Pennsylvania State University and Idaho National Laboratory. The team is building a prototype they plan to test at the Breazeale research reactor early in January 2015 and Westinghouse expects to bring the device to market by 2019, Heibel said.
Based on the thermo-acoustic engine principle – turning heat into sound – these devices do not require electric power supply. That fact has safety and cost benefits, Heibel said.
Using a thermo-acoustic neutron sensor, or an array of them, in the reactor to monitor the core power distribution and the temperature distribution, removes the need for tubing, wiring and vessel penetrations that are required to support existing surveillance instruments. That reduces the costs associated with maintaining such equipment, Heibel said. Plant operators will be able to monitor the core much more accurately, allowing them to produce more electricity from the same amount of uranium, he said.
Asked whether the device could be seen as a potential passive safety measure developed in the wake of Fukushima, Heibel said it would indicate "when melting is imminent".
"The more heat input you have, which you get when you have lost cooling, the louder it's going to scream," he said.
The "scream" is the frequency of sound wave produced according to the length of each device's resonance chamber.
Ideally, each fuel rod would have its own particular frequency, so plant operators would be able to determine by listening outside the reactor which fuel rods have an issue, if any, Heibel said.
Although these devices could be built into the fuel rods themselves, the Westinghouse team is looking at loading a number of them into a single tube that goes into the instrument thimble located inside the centre of a fuel assembly.
"We would be able to monitor different axial positions in every fuel assembly in the core and from that we could get fission rate and temperature information. Fission rate is probably the most important for the commercial nuclear power industry outside of extreme accident conditions," Heibel said.
Each device is somewhere between five and eight inches (13-20 centimetres) long, depending on the length of its resonance chamber. "We want each one to have a slightly different frequency, so that when we're listening, we can tell which assembly contains that particular device and we can then infer what the power distribution in that assembly is doing," Heibel said. Up to seven of these devices will go into each fuel assembly, he said.
Asked whether the device is being designed specifically for Westinghouse fuel rods, Heibel said, "We do have proprietary information associated with this specific application. It is not something off the shelf," he said. In its patent search, Westinghouse did not find any comparable devices being patented at this time, he said.
The sensors require a "signal processing system" to generate the core condition information in a form an operator can work with, he said.
"This investment generates the best return if the sensors are integral to the fuel. Ideally this system will be capable of being back-fit into existing fuel assemblies and be made integral to new fuel assembly designs," he said.
Researched and written
by World Nuclear News