And even more are in the pipeline!
"Possessing a cadre of new, well-trained judges, we are far more prepared to handle new license applications than we were just a few short years ago."
NRC Commissioner Jeffrey S. Merrifield
Presentation at NRC's Regulatory Information Conference
March 13-15, 2007
Merrifield asserted, "... I believe our mandatory hearing process is broken. ... I believe Congress should repeal the requirement for a mandatory hearing. Absent this change, the commission should take direct responsibility for these reviews. ... I believe the requirements for a mandatory hearing could be fulfilled by a single three to four hour meeting of the commission."
Wednesday, June 27, 2007
Saturday, June 23, 2007
Fouling and Reactivity Insertion Accidents
On Tuesday, April 3, 2007, the ACRS Subcommittee on Fuels considered several aspects of Reactivity Insertion Accidents. However, they continued to ignore the impact of heavy fouling (crud). As I've mentioned many times in this blog, fouling is ubiquitous among the worldwide fleet of LWRs. And I've presented the following slide previously; the heavy textured crud is clearly a very significant thermal resistance.
Sunday, June 10, 2007
I've been in this business since 1950
Robert H. Leyse in One Minute
I've been in this business since 1950 on several tasks
including the FLECHT tests that are referenced in
Appendix K. If you check ADAMS under Leyse, you will
find at least 172 entries. These include documents that I
have submitted to the NRC, related public comments, NRC
evaluations, and other matters.
My succinct discussion of fouling in the range of light water reactors over the decades may be found on GOOGLE by entering unmet relap. That is
U-N-M-E-T space R-E-L-A-P. You will find my slide presentation to the 2003 RELAP5 Users Conference under the title Unmet Challenges for SCDAP/RELAP5.
The impact of fouling on LOCAs or reactivity insertion accidents has not been evaluated although extensive fouling of fuel elements is widespread in the U.S.A. and elsewhere. In the U.S.A., ultrasonic fuel cleaning has been applied at some units. In Europe, chemical cleaning has been applied.
The impact of fouling has not been included in the wide range of international test
programs that address reactor accidents. The U.S.A. FLECHT Program never covered this, LOFT did not, and the present day work at Penn State does not.
I may update the 2003 presentation and call it Unmet Challenges for TRACE. Anyway, there are more examples that I would cite in such an update; however,
the bottom line is unchanged.
I've been in this business since 1950 on several tasks
including the FLECHT tests that are referenced in
Appendix K. If you check ADAMS under Leyse, you will
find at least 172 entries. These include documents that I
have submitted to the NRC, related public comments, NRC
evaluations, and other matters.
My succinct discussion of fouling in the range of light water reactors over the decades may be found on GOOGLE by entering unmet relap. That is
U-N-M-E-T space R-E-L-A-P. You will find my slide presentation to the 2003 RELAP5 Users Conference under the title Unmet Challenges for SCDAP/RELAP5.
The impact of fouling on LOCAs or reactivity insertion accidents has not been evaluated although extensive fouling of fuel elements is widespread in the U.S.A. and elsewhere. In the U.S.A., ultrasonic fuel cleaning has been applied at some units. In Europe, chemical cleaning has been applied.
The impact of fouling has not been included in the wide range of international test
programs that address reactor accidents. The U.S.A. FLECHT Program never covered this, LOFT did not, and the present day work at Penn State does not.
I may update the 2003 presentation and call it Unmet Challenges for TRACE. Anyway, there are more examples that I would cite in such an update; however,
the bottom line is unchanged.
Saturday, June 2, 2007
Upper Head Injection: ULTRA HIGH RISK (Update on May 3, 2008)
Following is an old NRC Information Notice. The history and risks of Upper Head Injection is extensive and I'll have a lot to say about this in future entries. Duke did not immediately abandon its UHI, but it did so in due time. TVA kept its system but ultimately got rid of it. The Japanese kept theirs at OHI for a long time, but they also got rid of it.
Upper Head Injection was installed at Westinghouse Ice Condenser Plants to provide a rapid injection of cooling water in the event of a large break LOCA. Ice condenser plants have a lower back pressure and UHI compensates for this. American electric Power never bought into this for its Cook units.
In Simpson's book, "Nuclear Power from Underseas to Outer Space," he spends one line on UHI on apge 199, "Where this couldn't be done or wasn't enough we proposed an upper head injection system,"
This is my opening salvo that will include my close involvement in a long and interesting history, please stay tuned.
PolInformation Notices > 1985 > IN 85-02
SSINS No.: 6835
IN 85-02
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
January 11, 1985
Information Notice No. 85-02: IMPROPER INSTALLATION AND TESTING OF DIFFERENTIAL PRESSURE TRANSMITTERS
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice provides notification of a potentially significant
problem pertaining to the improper installation and inadequate functional
testing of differential pressure transmitters.
Such conditions occurred at the McGuire Nuclear Station, Unit 1, when the Barton differential pressure switches utilized to control the isolation valves of the upper head injection (UHI) system were replaced with Rosemont differential pressure transmitters. It is expected that recipients will review the information contained in this notice for applicability to their facilities and consider actions, if appropriate, to preclude similar problems from occurring at their facilities. However, suggestions contained in this notice do not constitute NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances:
On November 1, 1984, Duke Power Company informed the NRC that the UHI isolation valves failed to close when the UHI water accumulator was drained at its McGuire Nuclear Station, Unit 1. At that time, the plant was shut down because the nitrogen content of the water in the UHI accumulator exceeded the limit permitted by its technical specifications. Subsequent
investigations revealed that the four differential pressure transmitters used to sense the level of water in the UHI accumulator and initiate automatic closure of the isolation valves on a predetermined level had been improperly installed. As a result, the isolation valves did not automatically close when the water level in the UFII accumulator reached the set point.
The McGuire UHI system design includes a separate nitrogen accumulator that supplies pressurized nitrogen to force the water from the UHI water accumulator into the reactor vessel during the initial phase of a design-basis loss-of-coolant accident (LOCA). Thus, if the UHI isolation valves fail to close during the course of a design-basis LOCA, nitrogen could be injected into the reactor vessel. To prevent such an event, the differential pressure transmitters are designed to initiate automatic closure of the UHI isolation valves when the water in the UHI accumulator reaches a predetermined level.
8501080502
.
IN 85-02
January 11, 1985
Page 2 of 2
During April of 1984, the McGuire Nuclear Station, Unit 1, Barton
reverse-acting differential pressure switches were replaced with Rosemont
direct-acting differential pressure transmitters to improve the accuracy and
repeatability of the UHI water accumulator level sensing system. However,
the Rosemont differential pressure transmitters were not properly installed
in that the impulse lines were not connected to the appropriate transmitter
ports.
Several factors contributed to the improper installation, including
inadequate installation instructions. The major contributor was inadequate
functional testing of the UHI system after it had been modified in that the
post-modification tests were limited to calibration tests of the differential pressure transmitters. These calibration tests were performed with the transmitters isolated from the impulse lines. Consequently, the tests only verified that the transmitters would provide the required output
signal for a given differential pressure, but they lid not demonstrate that
the transmitters sensed the differential pressures associated with water
level changes in the UHI water accumulator. Thus, the differential pressure
transmitters were not only improperly installed, but the error was not
detected until this event. If a design-basis LOCA had occurred during this
period, the UHI system would have been actuated, but the UHI isolation
valves would not have closed when the water in the UHI water accumulator had
been depleted and nitrogen gas could have been injected into the reactor
vessel during the course of the LOCA.
Similar installation errors have been addressed in Information Notice No. No.
84-45, "Reversed-Differential Pressure Instrument Lines." However, the
majority of events described in that information notice occurred in boiling
water reactors during plant construction and were detected by functional
tests performed before commencing power operation.
In contrast, the event described in this infomation notice occurred at a pressurized water reactor and was undetected during approximately 5 months of power operation.
No specific action or written response is required by this information notice; however, if you have any question regarding this notice, please contact the Regional Administrator of the appropriate NRC regional office or the technical contact listed below.
Edward L. Jordan Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: I. Villalva, IE
(301) 492-9007
F. R. McCoy, RII
(404) 221-2689
Attachment:
List of Recently Issued IE Information Notices
Today, May 3, 2008, I found the following via GOOGLE. This is another example of loads of dollars that have gone into useless thermal hydraulic testing. These tests were reported during 1979 and apparently the reporting system caught up with this during 2001. The investigators recommend higher temperature for the UHI water. Of course, UHI has been abandoned, however, the recommendation came in 1979 and was never adopted. It is another matter that could drive plant operators nuts.
ROSA (Rig Of Safety Assessment)
Title
Performance test of the upper head injection system at the ROSA-II test facility
Creator/Author
Tasaka, K. ; Adachi, H. ; Sobajima, M. ; Soda, K. ; Suzuki, M. ; Okazaki, M. ; Shiba, M.
Publication Date
1979 Sep 01
OSTI Identifier
OSTI ID: 5425745
Other Number(s)
CODEN: NUTYB
Resource Type
Journal Article
Resource Relation
Nucl. Technol. ; Vol/Issue: 45:2
Research Org
Japan Atomic Energy Research Inst., Tokai, Ibaraki
Subject
220900 -- Nuclear Reactor Technology-- Reactor Safety ;210200 -- Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled; ;ECCS-- PERFORMANCE TESTING;PWR TYPE REACTORS-- ECCS; HYDRODYNAMICS;TEST FACILITIES
Related Subject
ENGINEERED SAFETY SYSTEMS;FLUID MECHANICS;MECHANICS;REACTOR PROTECTION SYSTEMS;REACTORS;TESTING;WATER COOLED REACTORS;WATER MODERATED REACTORS
Description/Abstract
To evaluate upper head injection system (UHIS) performance during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), ten UHIS tests were conducted at the ROSA-II test facility.^The experimental results were different from the expected UHIS performance in the following points.^First, flashing took place in the upper head and a mixture level was formed before UHIS actuation.^Second, emptying of the upper head was observed immediately after UHIS shut off.^Third, part of the water which flowed down from the upper head, penetrated into the core and contributed to core cooling at the top part of the core, however, most of the water flowed out through the broken loop hot leg.^In the case of higher injection water temperature (approx.^120/sup 0/C), the fluid behavior in the pressure vessel differed significantly from the results for the low injection water temperature (approx.^20/sup 0/C), and the core cooling was remarkably improved.^Therefore, high-temperature UHIS water is recommended for effective core cooling.^The results described above are due to the following physical phenomena: (1) fluid mixing in the upper head is not good; (2) subcooled water, which flows into places such as the upper plenum where steam exists, causes strong condensation-depressurization which affects the flow behavior and core cooling.^Although the magnitude of the thermal-hydraulic effects observed in the ROSA-II/UHI tests may be unique to this facility, the above two physical phenomena observed are applicable to all PWRs with a UHIS.^Therefore, these two phenomena must be included in a LOCA analysis of a PWR with a UHIS.
Country of Publication
United States
Language
English
Format
Pages: 121-139
System Entry Date
2001 May 13
Upper Head Injection was installed at Westinghouse Ice Condenser Plants to provide a rapid injection of cooling water in the event of a large break LOCA. Ice condenser plants have a lower back pressure and UHI compensates for this. American electric Power never bought into this for its Cook units.
In Simpson's book, "Nuclear Power from Underseas to Outer Space," he spends one line on UHI on apge 199, "Where this couldn't be done or wasn't enough we proposed an upper head injection system,"
This is my opening salvo that will include my close involvement in a long and interesting history, please stay tuned.
PolInformation Notices > 1985 > IN 85-02
SSINS No.: 6835
IN 85-02
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
January 11, 1985
Information Notice No. 85-02: IMPROPER INSTALLATION AND TESTING OF DIFFERENTIAL PRESSURE TRANSMITTERS
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice provides notification of a potentially significant
problem pertaining to the improper installation and inadequate functional
testing of differential pressure transmitters.
Such conditions occurred at the McGuire Nuclear Station, Unit 1, when the Barton differential pressure switches utilized to control the isolation valves of the upper head injection (UHI) system were replaced with Rosemont differential pressure transmitters. It is expected that recipients will review the information contained in this notice for applicability to their facilities and consider actions, if appropriate, to preclude similar problems from occurring at their facilities. However, suggestions contained in this notice do not constitute NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances:
On November 1, 1984, Duke Power Company informed the NRC that the UHI isolation valves failed to close when the UHI water accumulator was drained at its McGuire Nuclear Station, Unit 1. At that time, the plant was shut down because the nitrogen content of the water in the UHI accumulator exceeded the limit permitted by its technical specifications. Subsequent
investigations revealed that the four differential pressure transmitters used to sense the level of water in the UHI accumulator and initiate automatic closure of the isolation valves on a predetermined level had been improperly installed. As a result, the isolation valves did not automatically close when the water level in the UFII accumulator reached the set point.
The McGuire UHI system design includes a separate nitrogen accumulator that supplies pressurized nitrogen to force the water from the UHI water accumulator into the reactor vessel during the initial phase of a design-basis loss-of-coolant accident (LOCA). Thus, if the UHI isolation valves fail to close during the course of a design-basis LOCA, nitrogen could be injected into the reactor vessel. To prevent such an event, the differential pressure transmitters are designed to initiate automatic closure of the UHI isolation valves when the water in the UHI accumulator reaches a predetermined level.
8501080502
.
IN 85-02
January 11, 1985
Page 2 of 2
During April of 1984, the McGuire Nuclear Station, Unit 1, Barton
reverse-acting differential pressure switches were replaced with Rosemont
direct-acting differential pressure transmitters to improve the accuracy and
repeatability of the UHI water accumulator level sensing system. However,
the Rosemont differential pressure transmitters were not properly installed
in that the impulse lines were not connected to the appropriate transmitter
ports.
Several factors contributed to the improper installation, including
inadequate installation instructions. The major contributor was inadequate
functional testing of the UHI system after it had been modified in that the
post-modification tests were limited to calibration tests of the differential pressure transmitters. These calibration tests were performed with the transmitters isolated from the impulse lines. Consequently, the tests only verified that the transmitters would provide the required output
signal for a given differential pressure, but they lid not demonstrate that
the transmitters sensed the differential pressures associated with water
level changes in the UHI water accumulator. Thus, the differential pressure
transmitters were not only improperly installed, but the error was not
detected until this event. If a design-basis LOCA had occurred during this
period, the UHI system would have been actuated, but the UHI isolation
valves would not have closed when the water in the UHI water accumulator had
been depleted and nitrogen gas could have been injected into the reactor
vessel during the course of the LOCA.
Similar installation errors have been addressed in Information Notice No. No.
84-45, "Reversed-Differential Pressure Instrument Lines." However, the
majority of events described in that information notice occurred in boiling
water reactors during plant construction and were detected by functional
tests performed before commencing power operation.
In contrast, the event described in this infomation notice occurred at a pressurized water reactor and was undetected during approximately 5 months of power operation.
No specific action or written response is required by this information notice; however, if you have any question regarding this notice, please contact the Regional Administrator of the appropriate NRC regional office or the technical contact listed below.
Edward L. Jordan Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: I. Villalva, IE
(301) 492-9007
F. R. McCoy, RII
(404) 221-2689
Attachment:
List of Recently Issued IE Information Notices
Today, May 3, 2008, I found the following via GOOGLE. This is another example of loads of dollars that have gone into useless thermal hydraulic testing. These tests were reported during 1979 and apparently the reporting system caught up with this during 2001. The investigators recommend higher temperature for the UHI water. Of course, UHI has been abandoned, however, the recommendation came in 1979 and was never adopted. It is another matter that could drive plant operators nuts.
ROSA (Rig Of Safety Assessment)
Title
Performance test of the upper head injection system at the ROSA-II test facility
Creator/Author
Tasaka, K. ; Adachi, H. ; Sobajima, M. ; Soda, K. ; Suzuki, M. ; Okazaki, M. ; Shiba, M.
Publication Date
1979 Sep 01
OSTI Identifier
OSTI ID: 5425745
Other Number(s)
CODEN: NUTYB
Resource Type
Journal Article
Resource Relation
Nucl. Technol. ; Vol/Issue: 45:2
Research Org
Japan Atomic Energy Research Inst., Tokai, Ibaraki
Subject
220900 -- Nuclear Reactor Technology-- Reactor Safety ;210200 -- Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled; ;ECCS-- PERFORMANCE TESTING;PWR TYPE REACTORS-- ECCS; HYDRODYNAMICS;TEST FACILITIES
Related Subject
ENGINEERED SAFETY SYSTEMS;FLUID MECHANICS;MECHANICS;REACTOR PROTECTION SYSTEMS;REACTORS;TESTING;WATER COOLED REACTORS;WATER MODERATED REACTORS
Description/Abstract
To evaluate upper head injection system (UHIS) performance during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), ten UHIS tests were conducted at the ROSA-II test facility.^The experimental results were different from the expected UHIS performance in the following points.^First, flashing took place in the upper head and a mixture level was formed before UHIS actuation.^Second, emptying of the upper head was observed immediately after UHIS shut off.^Third, part of the water which flowed down from the upper head, penetrated into the core and contributed to core cooling at the top part of the core, however, most of the water flowed out through the broken loop hot leg.^In the case of higher injection water temperature (approx.^120/sup 0/C), the fluid behavior in the pressure vessel differed significantly from the results for the low injection water temperature (approx.^20/sup 0/C), and the core cooling was remarkably improved.^Therefore, high-temperature UHIS water is recommended for effective core cooling.^The results described above are due to the following physical phenomena: (1) fluid mixing in the upper head is not good; (2) subcooled water, which flows into places such as the upper plenum where steam exists, causes strong condensation-depressurization which affects the flow behavior and core cooling.^Although the magnitude of the thermal-hydraulic effects observed in the ROSA-II/UHI tests may be unique to this facility, the above two physical phenomena observed are applicable to all PWRs with a UHIS.^Therefore, these two phenomena must be included in a LOCA analysis of a PWR with a UHIS.
Country of Publication
United States
Language
English
Format
Pages: 121-139
System Entry Date
2001 May 13
Friday, June 1, 2007
An adulterated McAdams burnout correlation and the Reactor of Sorts ( Updated May 3, 2008)
I've discussed the Reactor of Sorts (ROS) in today's prior entry. And as I mention there, Orwellian False Memories plague the universe. Nevertheless, please consider the following brief discussion.
It is a fact that McAdams of MIT published a straightforward correlation for burnout heat flux in relatively cold flowing water.
(Q/A)* = [400.000 + 4800(Tsat – Tlocal)] [cube root of V]
Where
(Q/A)*is burnout heat flux Btu/(square foot)(hr)
Tsat is water saturation temperature, degrees Fahrenheit, at local pressure
Tlocal is water temperature, degrees Fahrenheit, at location of burnout
V is water velocity, feet/second, at location of burnout
Now, in my world of false memories I began work at the ROS during January 1960. It was a very interesting machine. However, the corporate heat transfer experts may have published a paper that discussed burnout heat flux for the plate type fuel of the ROS. They might have taken the McAdams correlation and multiplied it by 1/3 to get the burnout heat flux for the ROS. I might have read their report and after a while I could have checked a bit further with one of the corporate heat transfer experts. The paper showed a rectangular heat transfer assembly that was electrically heated. In all cases the burnouts occurred in a corner. Looking at the square geometry of the corner, it became apparent, that the local heat flux was substantially greater than the average around the test section.
If I indeed was promoting a power level increase for the ROS, then I had to expose this error. In my wildest imagination, I succeeded in that, and this obstacle to power level increase was removed. And today, April 27, 2007, I am disclosing that if I succeeded in the power level increase (for which the GEnii would have gotten bonuses if indeed it happened) I would have used the unadulterated McAdams correlation and also posted results using the Bernath correlation, if indeed a licensing report was submitted somewhere.
And now, here are the pages from GETR licensing report, APED-5000-A,Class 1, July 1965. Bernath is there as well as the unadulterated McAdms and also Mirshak. Click these pages for enlargement.
So, that is the documentation added on May 3, 2008. Maybe I'll say more later.
It is a fact that McAdams of MIT published a straightforward correlation for burnout heat flux in relatively cold flowing water.
(Q/A)* = [400.000 + 4800(Tsat – Tlocal)] [cube root of V]
Where
(Q/A)*is burnout heat flux Btu/(square foot)(hr)
Tsat is water saturation temperature, degrees Fahrenheit, at local pressure
Tlocal is water temperature, degrees Fahrenheit, at location of burnout
V is water velocity, feet/second, at location of burnout
Now, in my world of false memories I began work at the ROS during January 1960. It was a very interesting machine. However, the corporate heat transfer experts may have published a paper that discussed burnout heat flux for the plate type fuel of the ROS. They might have taken the McAdams correlation and multiplied it by 1/3 to get the burnout heat flux for the ROS. I might have read their report and after a while I could have checked a bit further with one of the corporate heat transfer experts. The paper showed a rectangular heat transfer assembly that was electrically heated. In all cases the burnouts occurred in a corner. Looking at the square geometry of the corner, it became apparent, that the local heat flux was substantially greater than the average around the test section.
If I indeed was promoting a power level increase for the ROS, then I had to expose this error. In my wildest imagination, I succeeded in that, and this obstacle to power level increase was removed. And today, April 27, 2007, I am disclosing that if I succeeded in the power level increase (for which the GEnii would have gotten bonuses if indeed it happened) I would have used the unadulterated McAdams correlation and also posted results using the Bernath correlation, if indeed a licensing report was submitted somewhere.
And now, here are the pages from GETR licensing report, APED-5000-A,Class 1, July 1965. Bernath is there as well as the unadulterated McAdms and also Mirshak. Click these pages for enlargement.
So, that is the documentation added on May 3, 2008. Maybe I'll say more later.
SL-1, A Ladyfinger
The following is copied from my original blog, http://nuclearenergyblog.blogspot.com/
February 24, 2007
Suppose I had false memories
Of course, one has to be careful about swearing to the truth of lots of stuff. What if I had an imagination, or far worse, some Orwellian false memories? My first entry in this blog briefly mentions the SL-1 explosion and that is pure documented fact.So let us suppose that near the west coast there was a nuclear test reactor of sorts. And the downtown office sends a front man to Idaho to attend some SL-1 explosion briefings. Upon his return, if he went, he comes out in the country and gives us a sanitized briefing at the reactor of sorts (ROS).
He describes the central control rod at SL-1 and before he gets to his next sentence, a sneering commenter might have bellowed out, "It would take a team from Argonne to put a control rod right in the center of that core."
Now we can move ahead a few years. The ROS might have been built with control rods that used boron stainless as structural material as well as poison. The control rod structure could have cracked after moderate use, leading to binding and other bad scenes. So, a cadmium assembly could have been designed that would have superior life and equivalent control strength (with thermal neutrons, black is black).
For a bit of further background, the ROS might have had six control rod assemblies with fuel followers, and if these assemblies existed, they were similar to the assemblies of the Engineering Test Reactor ETR that was in Idaho. The ETR poison sections were about three feet long and they included fuel followers so that fuel was added to the control rod location as control rods were withdrawn.
So, a new cadmium assembly might have been built, but how to test it for worth? The easiest thing would be to place it in the very center position of the ROS, go critical, shut down, remove it, and replace it with an old boron assembly and go critical again. Then a comparison of the positions of criticality would be a great proof test.But there might have been a restriction on such a procedure because the amount of reactivity (plus or minus) that was allowed in the center of ROS might have been limited by its operating license, if it had such.
However, under an AEC rule, 50.59, it might have been stipulated that the restriction on reactivity only applied to long term operations and not a simple field test of only several minutes at the most. So, such a comparison might have proceeded. And if it did proceed, the sneering commentator of the second paragraph above might have had the task of predicting the amount of withdrawal of the six control rods that would yield criticality. And if such a prediction was made, it might have been that criticality would be reached with about half of full withdrawal, about 18 inches.If the test proceeded, it might have been found that criticality was not reached as predicted. And then a guy in the control room might have telephoned the sneering commentator (who could have been elsewhere). The sneering commentator might have provided assurance that the situation was no big deal, that with such an unusual geometry with such a vast amount of poison in the most reactive location in the core, an accurate prediction was likely out of the question.So, the test may have proceeded. And if it did, the careful slow withdrawal of the bank of six control rods might have proceeded with several stops along the way to criticality. And criticality might have been reached with the gang withdrawal at 33 inches out of the maximum 36 inches that was possible with the ETR design. And that would have been very interesting since fuel followers would have added to the reactivity worth of the center of the core. And if the test had proceeded it might have been found that the cadmium section also led to 33 inches of gang withdrawal. After all, black is black.
And so, if in the game of inserting the control rod in location E-5, going critical, shutting down, removing the control rod, placing the new design, going critical, shutting down, removing the new design; if, then it might have been very fortunate that nothing happened that would have made SL-1 look like a ladyfinger.* And maybe, in contrast to SL-1, the ROS, in my wildest false memory could have had targets for isotope production, like cobalt stuff.
*During the mid 1930s, if a kid could get his money onto the counter, he could buy firecrackers. If he saved up, he could buy three ten inchers for a dime. For a nickel he could by a package of zillions of ladyfingers. Ladyfingers were weak little things that would damage nothing even if held held between the kid's fingers.
February 24, 2007
Suppose I had false memories
Of course, one has to be careful about swearing to the truth of lots of stuff. What if I had an imagination, or far worse, some Orwellian false memories? My first entry in this blog briefly mentions the SL-1 explosion and that is pure documented fact.So let us suppose that near the west coast there was a nuclear test reactor of sorts. And the downtown office sends a front man to Idaho to attend some SL-1 explosion briefings. Upon his return, if he went, he comes out in the country and gives us a sanitized briefing at the reactor of sorts (ROS).
He describes the central control rod at SL-1 and before he gets to his next sentence, a sneering commenter might have bellowed out, "It would take a team from Argonne to put a control rod right in the center of that core."
Now we can move ahead a few years. The ROS might have been built with control rods that used boron stainless as structural material as well as poison. The control rod structure could have cracked after moderate use, leading to binding and other bad scenes. So, a cadmium assembly could have been designed that would have superior life and equivalent control strength (with thermal neutrons, black is black).
For a bit of further background, the ROS might have had six control rod assemblies with fuel followers, and if these assemblies existed, they were similar to the assemblies of the Engineering Test Reactor ETR that was in Idaho. The ETR poison sections were about three feet long and they included fuel followers so that fuel was added to the control rod location as control rods were withdrawn.
So, a new cadmium assembly might have been built, but how to test it for worth? The easiest thing would be to place it in the very center position of the ROS, go critical, shut down, remove it, and replace it with an old boron assembly and go critical again. Then a comparison of the positions of criticality would be a great proof test.But there might have been a restriction on such a procedure because the amount of reactivity (plus or minus) that was allowed in the center of ROS might have been limited by its operating license, if it had such.
However, under an AEC rule, 50.59, it might have been stipulated that the restriction on reactivity only applied to long term operations and not a simple field test of only several minutes at the most. So, such a comparison might have proceeded. And if it did proceed, the sneering commentator of the second paragraph above might have had the task of predicting the amount of withdrawal of the six control rods that would yield criticality. And if such a prediction was made, it might have been that criticality would be reached with about half of full withdrawal, about 18 inches.If the test proceeded, it might have been found that criticality was not reached as predicted. And then a guy in the control room might have telephoned the sneering commentator (who could have been elsewhere). The sneering commentator might have provided assurance that the situation was no big deal, that with such an unusual geometry with such a vast amount of poison in the most reactive location in the core, an accurate prediction was likely out of the question.So, the test may have proceeded. And if it did, the careful slow withdrawal of the bank of six control rods might have proceeded with several stops along the way to criticality. And criticality might have been reached with the gang withdrawal at 33 inches out of the maximum 36 inches that was possible with the ETR design. And that would have been very interesting since fuel followers would have added to the reactivity worth of the center of the core. And if the test had proceeded it might have been found that the cadmium section also led to 33 inches of gang withdrawal. After all, black is black.
And so, if in the game of inserting the control rod in location E-5, going critical, shutting down, removing the control rod, placing the new design, going critical, shutting down, removing the new design; if, then it might have been very fortunate that nothing happened that would have made SL-1 look like a ladyfinger.* And maybe, in contrast to SL-1, the ROS, in my wildest false memory could have had targets for isotope production, like cobalt stuff.
*During the mid 1930s, if a kid could get his money onto the counter, he could buy firecrackers. If he saved up, he could buy three ten inchers for a dime. For a nickel he could by a package of zillions of ladyfingers. Ladyfingers were weak little things that would damage nothing even if held held between the kid's fingers.
Subscribe to:
Posts (Atom)
