Accession Number
ML082660533
Title
Comment (11) of James H. Riley on Behalf of Nuclear Energy Institute on Technical Basis for New Performance-Based Emergency Core Cooling System Requirements.
Document Date
09/11/2008
Author Name
Riley J H
Author Affiliation
Nuclear Energy Institute (NEI)
Date Declared
09/22/2008 1:17:58 PM
Document/Report Number
Estimated Page Count
17
File Size
2113221
Keyword
exb3 stt
SUNSI Review Complete
Availability
Publicly Available to be released 09/30/08
Accession Number
ML082660611
Title
Comment (10) of Robert H. Leyse Opposing Research Information Letter (RIL) 0801, "Technical Basis for New Performance-Based Emergency Core Cooling System Requirements."
Document Date
09/06/2008
Author Name
Leyse R H
Author Affiliation
- No Known Affiliation
Date Declared
09/22/2008 2:47:30 PM
Document/Report Number
Estimated Page Count
3
File Size
115399
Keyword
jef1 lxk1
SUNSI Review Complete
Availability
Publicly Available to be released 09/30/08
Accession Number
ML082700230
Title
Comment (12) of James F. Harrison on Behalf of GE Hitachi Nuclear Energy (GEH) and Global Nuclear Fuel-Americas (GNF) on Documents Under Consideration to Establish the Technical Basis for New Performance-Based Emergency Core Cooling. System Requirement.
Document Date
09/11/2008
Author Name
Harrison J F
Author Affiliation
GE-Hitachi Nuclear Energy Americas, LLC
Date Declared
09/26/2008 11:42:03 AM
Document/Report Number
Estimated Page Count
12
File Size
557390
Keyword
jef1 laf1
SUNSI Review Complete
Availability
Publicly Available to be released 10/06/08
ML082660533
Title
Comment (11) of James H. Riley on Behalf of Nuclear Energy Institute on Technical Basis for New Performance-Based Emergency Core Cooling System Requirements.
Document Date
09/11/2008
Author Name
Riley J H
Author Affiliation
Nuclear Energy Institute (NEI)
Date Declared
09/22/2008 1:17:58 PM
Document/Report Number
Estimated Page Count
17
File Size
2113221
Keyword
exb3 stt
SUNSI Review Complete
Availability
Publicly Available to be released 09/30/08
Accession Number
ML082660611
Title
Comment (10) of Robert H. Leyse Opposing Research Information Letter (RIL) 0801, "Technical Basis for New Performance-Based Emergency Core Cooling System Requirements."
Document Date
09/06/2008
Author Name
Leyse R H
Author Affiliation
- No Known Affiliation
Date Declared
09/22/2008 2:47:30 PM
Document/Report Number
Estimated Page Count
3
File Size
115399
Keyword
jef1 lxk1
SUNSI Review Complete
Availability
Publicly Available to be released 09/30/08
Accession Number
ML082700230
Title
Comment (12) of James F. Harrison on Behalf of GE Hitachi Nuclear Energy (GEH) and Global Nuclear Fuel-Americas (GNF) on Documents Under Consideration to Establish the Technical Basis for New Performance-Based Emergency Core Cooling. System Requirement.
Document Date
09/11/2008
Author Name
Harrison J F
Author Affiliation
GE-Hitachi Nuclear Energy Americas, LLC
Date Declared
09/26/2008 11:42:03 AM
Document/Report Number
Estimated Page Count
12
File Size
557390
Keyword
jef1 laf1
SUNSI Review Complete
Availability
Publicly Available to be released 10/06/08
Some reminders:
Plesset remarks to ACRS following witnessing the inadvertent LOFT uncovery following the TMI meltdown. File is somewhere. Also hydrogen impact proposal.
http://www.nea.fr/html/nsd/docs/2008/csni-r2008-2.pdf
1 INTRODUCTION
On the 10th April 2003 severe damage of fuel assemblies took place during an incident at
Unit 2 of Paks Nuclear Power Plant in Hungary. The assemblies were being cleaned in a
special tank below the water level of the spent fuel storage pool in order to remove crud
buildup. That afternoon, the chemical cleaning of assemblies was completed and the fuel rods
were being cooled by circulation of spent fuel storage pool water. The first sign of fuel failure
was the detection of some fission gases released from the cleaning tank during that evening.
The cleaning tank cover locks were released after midnight and this operation was followed
by a sudden increase in activity concentrations. The visual inspection revealed that all 30 fuel
assemblies were severely damaged. The first evaluation of the event showed that the severe
fuel damage happened due to inadequate coolant circulation within the cleaning tank. The
damaged fuel assemblies were removed from the cleaning tank in 2006 and are stored in
special canisters in the spent fuel storage pool of the Paks NPP.
the formation of water level in the cleaning tank took only some minutes after saturation was
reached. The fuel behaviour calculations proved that the loss of fuel integrity in 5 hours after
the initiation of intermediate cooling mode was a result of fuel rod ballooning and burst [18].
The fuel rods could heat up above 800 ºC by that time and the high internal pressure in some
rods resulted in plastic deformation and burst. The calculation of zirconium oxidation pointed
out that the volume of produced hydrogen could be much larger, than the volume of the
The validity of safety analyses performed earlier was reviewed. The Russian firm Hydropress
performed the analysis of the operation with magnetite deposits burdened fuel assemblies [1]
for both normal and incidental conditions of the reactor units. These investigations have
shown that the deformations of the fuel assembly shrouds might be attained much easier if the
shroud temperature exceeds 355°C.
The Hydropress report suggested new operational limits: the relative mass flow of the
assemblies could not fall under 86% of the average value, because at lower mass flow rates,
the shroud temperatures might exceed the problematic temperature value of 355°C. For
controlling the mass flow limitations, the modification of the VERONA reactor core
monitoring system was also necessary. In addition, the Hydropress suggested to plan the
refuelling with extreme caution.
As a consequence of the above drafted events, the plant management decided to clean the
affected fuel assemblies, first time on the turn of the millennium. In the years 2000 and 2001,
the chemical cleaning of a total sum of 170 fuel assemblies was performed successfully in an
instrument designed by Siemens KWU, which was capable for housing 7 fuel assemblies
simultaneously [15]. In accordance with the preliminary expectations, this instrument
operated effectively. (From the viewpoint of the events discussed below, it is important to
mention that the cleaned fuel assemblies were removed from the core two years before the
cleaning process started and they had low residual heat. Furthermore, this equipment would
have been able to ensure the cooling of fuel assemblies of much higher thermal power. This
cleaning tank had inlet junction in the bottom and outlet junction on the top of the vessel. The
safety criteria specified by the Paks NPP experts were fulfilled for this equipment as it was
proved by the main constructor Hydropress.)
In November of 2002, the Paks NPP commissioned a reputed Western-European nuclear
company to design, construct and operate a new chemical cleaning equipment of larger
dimensions. With this new cleaning tank the operators were able to perform the chemical
cleaning of 30 fuel assemblies simultaneously with the same permanganate-oxalic-acid
method. The instrument arrived to Hungary in the beginning of 2003 and was placed into the
refuelling pit beside the reactor of Unit 2 in the Phase 1 of the Paks NPP (see Figure 2.2). The cleaning of the first batch of 30 fuel assemblies was started on 20th of March. Prior to the
incident, five batches were cleaned successfully. The 1st, 2nd, and 4th batches consisted of fuel
assemblies removed a few years earlier, but the 3rd and 5th batches contained fuel assemblies
with significant residual power (removed from the reactor on those days).
References to Chapter 2.
[1] OKB Hydropress: Analysis of operation of fuel assemblies with reduced flow rate due
to deposit in the Paks NPP (final technical report), U-213-TI-1762, 2003, in Hungarian
2.3.2 Plastic deformation of cladding and high temperature oxidation
Continuous temperature increase started in the tank, when the upper part of the fuel rods was
not cooled by water. The heat removal from the tank to the surrounding water was very low,
because the vessel was isolated by the double wall system. The temperature increase led to the
increase of pressure inside of the fuel rods. At 800-900 ºC the internal pressure could reach
30-40 bars. In this range of pressure and temperature plastic deformation of the cladding can
take place and the ballooning can lead to bursts and activity release from the fuel. It is very
probable that this type of fuel failure was responsible for the activity release measured by the
85Kr detectors. (Very long ballooned areas were found in the later visual inspection of the
fuel.) The first fuel rod ruptures were detected at 21:50, when the 85Kr measurement of the
AMDA system showed an unexpected jump in the signal (see Fig. 2.8) and a few minutes
later the noble gas detectors of the unit were also alerted.
1 INTRODUCTION
On the 10th April 2003 severe damage of fuel assemblies took place during an incident at
Unit 2 of Paks Nuclear Power Plant in Hungary. The assemblies were being cleaned in a
special tank below the water level of the spent fuel storage pool in order to remove crud
buildup. That afternoon, the chemical cleaning of assemblies was completed and the fuel rods
were being cooled by circulation of spent fuel storage pool water. The first sign of fuel failure
was the detection of some fission gases released from the cleaning tank during that evening.
The cleaning tank cover locks were released after midnight and this operation was followed
by a sudden increase in activity concentrations. The visual inspection revealed that all 30 fuel
assemblies were severely damaged. The first evaluation of the event showed that the severe
fuel damage happened due to inadequate coolant circulation within the cleaning tank. The
damaged fuel assemblies were removed from the cleaning tank in 2006 and are stored in
special canisters in the spent fuel storage pool of the Paks NPP.
the formation of water level in the cleaning tank took only some minutes after saturation was
reached. The fuel behaviour calculations proved that the loss of fuel integrity in 5 hours after
the initiation of intermediate cooling mode was a result of fuel rod ballooning and burst [18].
The fuel rods could heat up above 800 ºC by that time and the high internal pressure in some
rods resulted in plastic deformation and burst. The calculation of zirconium oxidation pointed
out that the volume of produced hydrogen could be much larger, than the volume of the
The validity of safety analyses performed earlier was reviewed. The Russian firm Hydropress
performed the analysis of the operation with magnetite deposits burdened fuel assemblies [1]
for both normal and incidental conditions of the reactor units. These investigations have
shown that the deformations of the fuel assembly shrouds might be attained much easier if the
shroud temperature exceeds 355°C.
The Hydropress report suggested new operational limits: the relative mass flow of the
assemblies could not fall under 86% of the average value, because at lower mass flow rates,
the shroud temperatures might exceed the problematic temperature value of 355°C. For
controlling the mass flow limitations, the modification of the VERONA reactor core
monitoring system was also necessary. In addition, the Hydropress suggested to plan the
refuelling with extreme caution.
As a consequence of the above drafted events, the plant management decided to clean the
affected fuel assemblies, first time on the turn of the millennium. In the years 2000 and 2001,
the chemical cleaning of a total sum of 170 fuel assemblies was performed successfully in an
instrument designed by Siemens KWU, which was capable for housing 7 fuel assemblies
simultaneously [15]. In accordance with the preliminary expectations, this instrument
operated effectively. (From the viewpoint of the events discussed below, it is important to
mention that the cleaned fuel assemblies were removed from the core two years before the
cleaning process started and they had low residual heat. Furthermore, this equipment would
have been able to ensure the cooling of fuel assemblies of much higher thermal power. This
cleaning tank had inlet junction in the bottom and outlet junction on the top of the vessel. The
safety criteria specified by the Paks NPP experts were fulfilled for this equipment as it was
proved by the main constructor Hydropress.)
In November of 2002, the Paks NPP commissioned a reputed Western-European nuclear
company to design, construct and operate a new chemical cleaning equipment of larger
dimensions. With this new cleaning tank the operators were able to perform the chemical
cleaning of 30 fuel assemblies simultaneously with the same permanganate-oxalic-acid
method. The instrument arrived to Hungary in the beginning of 2003 and was placed into the
refuelling pit beside the reactor of Unit 2 in the Phase 1 of the Paks NPP (see Figure 2.2). The cleaning of the first batch of 30 fuel assemblies was started on 20th of March. Prior to the
incident, five batches were cleaned successfully. The 1st, 2nd, and 4th batches consisted of fuel
assemblies removed a few years earlier, but the 3rd and 5th batches contained fuel assemblies
with significant residual power (removed from the reactor on those days).
References to Chapter 2.
[1] OKB Hydropress: Analysis of operation of fuel assemblies with reduced flow rate due
to deposit in the Paks NPP (final technical report), U-213-TI-1762, 2003, in Hungarian
2.3.2 Plastic deformation of cladding and high temperature oxidation
Continuous temperature increase started in the tank, when the upper part of the fuel rods was
not cooled by water. The heat removal from the tank to the surrounding water was very low,
because the vessel was isolated by the double wall system. The temperature increase led to the
increase of pressure inside of the fuel rods. At 800-900 ºC the internal pressure could reach
30-40 bars. In this range of pressure and temperature plastic deformation of the cladding can
take place and the ballooning can lead to bursts and activity release from the fuel. It is very
probable that this type of fuel failure was responsible for the activity release measured by the
85Kr detectors. (Very long ballooned areas were found in the later visual inspection of the
fuel.) The first fuel rod ruptures were detected at 21:50, when the 85Kr measurement of the
AMDA system showed an unexpected jump in the signal (see Fig. 2.8) and a few minutes
later the noble gas detectors of the unit were also alerted.
However, with heavy crud and oxide layers on the cladding (the conditions of cycle 8)
the ECCS design basis for River Bend is substantially non-conservative in at least the
following aspects: 1) the cladding surface temperature (at some locations) at River Bend
Cycle 8 has been reported to have reached temperatures approaching 1200°F; therefore,
the starting temperature in the event of a LOCA would be almost 1200°F, not the
licensing basis for temperatures around 578°F; 2) the stored energy in the fuel with
cladding that had surface temperatures approaching 1200°F (at some locations) would be
substantially greater than that of fuel with cladding surface temperatures in the range of
578°F at the onset of a LOCA; 3) the amount of coolant in the vicinity of cladding with
heavy crud and oxide layers at the onset of a LOCA would be substantially less than if
the cladding were clean; 4) during blowdown and also during reflood the amount of
coolant flow past cladding with heavy crud and oxide layers would be substantially less
than the flow past clean cladding; 5) the increased quantity of the stored energy in the
fuel and the delay in the transfer of that stored energy to the coolant caused by a heavy
crud layer (mentioned by Klapproth in his letter to the NRC) would cause the cladding to
be subjected to extremely high temperatures for a substantially longer time duration than
the time duration used in the licensing basis, providing more time for heatup and
degradation of the fuel and cladding; 6) the severity of the fuel and cladding degradation
occurring in the event of a LOCA and its effect on obstructing coolant flow would be
substantially greater than those calculated by an ECCS design based on clean cladding; 7)
the increased quantity of the stored energy in the fuel and the delay in the transfer of that
stored energy to the coolant would increase the time until quench; 8) at the onset of a
LOCA, there would already be severe cladding degradation, massive oxidation and
absorption of hydrogen at some locations, which would contribute to a loss of cladding
ductility. (These same deficiencies in the design basis for the ECCS at River Bend-for
situations where cladding is heavily crudded and oxidized-also apply to the design basis
for the ECCS at other nuclear power plants.)
Because the ECCS design basis for River Bend is substantially non-conservative
when it comes to calculating the PCT for a postulated LOCA for conditions where there
are heavy crud and oxide layers on the cladding, there is reason to believe that with high
probability the PCT in the event of a LB LOCA at River Bend Cycle 8 would have
exceeded 2200°F (and that the plant would have violated other requirements of 10 C.F.R.
9 50.46(b)).
the ECCS design basis for River Bend is substantially non-conservative in at least the
following aspects: 1) the cladding surface temperature (at some locations) at River Bend
Cycle 8 has been reported to have reached temperatures approaching 1200°F; therefore,
the starting temperature in the event of a LOCA would be almost 1200°F, not the
licensing basis for temperatures around 578°F; 2) the stored energy in the fuel with
cladding that had surface temperatures approaching 1200°F (at some locations) would be
substantially greater than that of fuel with cladding surface temperatures in the range of
578°F at the onset of a LOCA; 3) the amount of coolant in the vicinity of cladding with
heavy crud and oxide layers at the onset of a LOCA would be substantially less than if
the cladding were clean; 4) during blowdown and also during reflood the amount of
coolant flow past cladding with heavy crud and oxide layers would be substantially less
than the flow past clean cladding; 5) the increased quantity of the stored energy in the
fuel and the delay in the transfer of that stored energy to the coolant caused by a heavy
crud layer (mentioned by Klapproth in his letter to the NRC) would cause the cladding to
be subjected to extremely high temperatures for a substantially longer time duration than
the time duration used in the licensing basis, providing more time for heatup and
degradation of the fuel and cladding; 6) the severity of the fuel and cladding degradation
occurring in the event of a LOCA and its effect on obstructing coolant flow would be
substantially greater than those calculated by an ECCS design based on clean cladding; 7)
the increased quantity of the stored energy in the fuel and the delay in the transfer of that
stored energy to the coolant would increase the time until quench; 8) at the onset of a
LOCA, there would already be severe cladding degradation, massive oxidation and
absorption of hydrogen at some locations, which would contribute to a loss of cladding
ductility. (These same deficiencies in the design basis for the ECCS at River Bend-for
situations where cladding is heavily crudded and oxidized-also apply to the design basis
for the ECCS at other nuclear power plants.)
Because the ECCS design basis for River Bend is substantially non-conservative
when it comes to calculating the PCT for a postulated LOCA for conditions where there
are heavy crud and oxide layers on the cladding, there is reason to believe that with high
probability the PCT in the event of a LB LOCA at River Bend Cycle 8 would have
exceeded 2200°F (and that the plant would have violated other requirements of 10 C.F.R.
9 50.46(b)).
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