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t G WASHINGTON, D. C. 20555 MEMORANDUM FOR:
Jack Rosenthal, Team Leader Nine Hile 2 Incident Investigation Team FROM:
SUBJECT:
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data TRANSMITTAL OF POSSIBLE RE-CRITICALITY TRANSIENT ANALYSIS As per your request, I have conducted a primary analysis of the possible consequences of a re-criticality event during the cooldown of the Nine Mile 2 nuclear reactor following the August 13, 1991 event.
This analysis should be considered as a general scoping study of possible consequences of an event of this kind.
The analysis is limited to the information available at the time and modeling of a generic BWR/5-6.
The analysis was performed using a high fidelity thermal hydraulic code, Real-Time Advanced Core and Thermohydraulic (RETACT) to model the Nine Mile 2 reactor.
This computation work was supplemented by hand calculations and discussion with other reactor safety experts and operational personnel.
The report includes best estimate analysis of the consequence of the event as well as some possible consequences of additional system failures.
The primary conclusions of this preliminary analysis of the event are that:
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the rapid injection of feedwater into the reactor during cooldown, concurrent with a partial failure of all rods to fully insert into the core, will likely lead to a re-criticality transient; this re-criticality will lead to a rapid repressurization of the reactor vessel, a power excursion, and possible localized fuel damage; due to the power excursion at a low reactor core flow rate, the core could enter LaSalle type flow and power oscillations.
This is of particular concern because under these conditions the plant operator has no control of the control rods to shut the reactor
- down, and because this type of power and flow behavior is not well understood; E
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C the primary governing parameters in this event ar'e the effective multiplication factor at the time of the feedwater injection, which is a function of core life and degree of rod insertion, the amount of feedwater injected and its temperature, and the ability of the operator to control and relieve reactor pressure.
The overall conclusion of this study is that the injection of feedwater into the vessel could have produced significant but not severe damage to reactor components including reactor fuel, although no gross fuel damage is predicted.
Additionally, the possibility of entering a poorly understood region of plant operation during a time period when a large portion of plant instrumentation was not available is 'of concern.
However, the likelihood of an event of this kind is not high and the Anticipated Transient with Scram (ATWS) Emergency Operating Procedures (EOP) should give appropriate guidance in methods of dealing with the consequences of such an event.
A short report, detailing this analysis and its finding will be submitted to you in the near future.
If additional information or analysis is needed, please contact me at x29004.
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 MEMORANDUM FOR:
Jack Rosenthal, Team Leader Nine Mile 2 Incident Investigation Team FROM:
SUBJECT:
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data TRANSMITTAL OF POSSIBLE RE-CRITICALITY TRANSIENT ANALYSIS As per your request, I have conducted a primary analysis of the possible consequences of a re-criticality event during the cooldown of the Nine Mile 2 nuclear reactor following the August 13, 1991 event.
This analysis should be considered as a general scoping study of possible consequences of an event of this kind.
The analysis is limited to the information available at the time and modeling of a generic BWR/5-6.
The analysis was'erformed using a high fidelity thermal hydraulic code, Real-Time Advanced Core and Thermohydraulic (RETACT) to model the Nine Mile 2 reactor.
This computation work was supplemented by hand calculations and discussion with other reactor safety experts and operational personnel.
The report includes best estimate analysis of the consequence of the event as well as some possible consequences of additional system failures.
The primary conclusions of this preliminary analysis of the event are that:
the rapid injection of feedwater into the reactor during cooldown, concurrent with a partial failure of all rods to fully insert into the core, will likely lead to a re-criticality transient; this re-criticality will lead to a rapid repressurization of the reactor vessel, a power excursion, and possible localized fuel damage; due to the power excursion at a low reactor core flow rate, the core could enter LaSalle type flow and power oscillations.
This is of particular concern because under these conditions the plant operator has no control of the control rods to shut the reactor
- down, and because this type of power and flow behavior is not well understood;
P-the primary governing parameters in this event are the effective multiplication factor at the time of the feedwater injection, which is a function of core life and degree of rod insertion, the amount of feedwater injected and its temperature, and the ability of the operator to control and relieve reactor pressure.
The overall conclusion of this study is that the injection of feedwater into the vessel could have produced significant but not severe damage to reactor components including reactor fuel, although no gross fuel damage is predicted.
Additionally, the possibility of entering a poorly understood region of plant operation during a time period when a large portion of plant instrumentation was not available is of concern.
However, the likelihood of an event of this kind is not high and the Anticipated Transient with Scram (ATWS) Emergency Operating Procedures (EOP) should give appropriate guidance in methods of dealing with the consequences of such an event.
A short report, detailing this analysis and its finding will be submitted to you in the near future.
If additional information or analysis is needed, please contact me at x29004.
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 HEHORANDUM FOR:
Jack Rosenthal, Team Leader Nine Mile 2 Incident Investigation Team FROM:
SUBJECT:
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data TRANSMITTAL OF POSSIBLE RE-CRITICALITY TRANSIENT ANALYSIS As per your request, I have conducted a primary analysis of the possible consequences of a re-criticality event during the cooldown of the Nine Mile 2 nuclear reactor following the August 13, 1991 event.
This analysis should be considered as a general scoping study of possible consequences of an event of this kind.
The analysis is limited to the information availabl.e at the time and modeling of a generic BWR/5-6.
The analysis was performed using a high fidelity thermal hydraulic code, Real-Time Advanced Core and Thermohydraulic (RETACT) to model the Nine Mile 2 reactor.
This computation work was supplemented by hand calculations and discussion with other reactor safety experts and operational personnel.
The report includes best estimate analysis of the consequence of the event as well as some possible consequences of additional system failures.
The primary conclusions of this preliminary analysis of the event are that:
the rapid injection of feedwater into the reactor during cooldown, concurrent with a partial failure of all rods to fully insert into the core, will likely lead to a re-criticality transient; this re-criticality will lead to a rapid repressurization of the reactor vessel, a power excursion, and possible localized fuel damage; due to the power excursion at a low reactor core flow rate, the core could enter LaSalle type flow and power oscillations.
This is of particular concern because under these conditions the plant operator has no control of the control rods to shut the reactor
- down, and because this type of power and flow behavior is not well understood;
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the primary governing parameters in this event are the effective multiplication factor at the time of the feedwater injection, which is a function of core life and degree of rod insertion, the amount of feedwater injected and its temperature, and the ability of the operator to control and relieve reactor pressure.
The overall conclusion of this study is that the injection of feedwater into the vessel could have produced significant but not severe damage to reactor components including reactor fuel, although no gross fuel damage is predicted.
Additionally, the possibility of entering a poorly understood region of plant operation during a time period when a large portion of plant instrumentation was not available is of concern.
However, the likelihood of an event of this kind is not high and the Anticipated Transient with Scram (ATWS) Emergency Operating Procedures (EOP) should give appropriate guidance in methods of dealing with the consequences of such an event.
A short report, detailing this analysis and its finding will be submitted to you in the near future.
If additional information or analysis is needed, please contact me at x29004.
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 MEMORANDUM FOR:
Jack Rosenthal, Team Leader Nine Mile 2 Incident Investigation Team FROM:
SUBJECT:
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data TRANSMITTAL OF POSSIBLE RE-CRITICALITY TRANSIENT ANALYSIS As per your request, I have conducted a primary analysis of the possible consequences of a re-criticality event during the cooldown of the Nine Mile 2 nuclear reactor following the August 13, 1991 event.
This analysis should be considered as a general scoping study of possible consequences of an event of this kind.
The analysis is limited to the information available at the time and modeling of a generic BWR/5-6.
r The analysis was performed using a high fidelity thermal hydraulic code, Real-Time Advanced Core and Thermohydraulic (RETACT) to model the Nine Mile 2 reactor.
This computation work was supplemented by hand calculations and discussion with other reactor safety experts and operational personnel.
The report includes best estimate analysis of the consequence of the event as well as some possible consequences of additional system failures.
The primary conclusions of this preliminary analysis of the event are that:
the rapid injection of feedwater into the reactor during cooldown, concurrent with a partial failure of all rods to fully insert into the core, will likely lead to a re-criticality transient; this re-criticality will lead to a rapid repressurization of the reactor vessel, a power excursion, and possible localized fuel damage; due to the power excursion at a low reactor core flow rate, the core could enter LaSalle type flow and power oscillations.
This is of particular concern because under these conditions the plant operator has no control of the control rods to shut the reactor
- down, and because this type of power and flow behavior is not well understood;
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the primary governing parameters in this event are the effective multiplication factor at the time of the feedwater injection, which is a function of core life and degree of rod insertion, the amount of feedwater injected and its temperature, and the ability of the operator to control and relieve reactor pressure.
The overall conclusion of this study is that the injection of feedwater into the vessel could have produced significant but not severe damage to reactor components including reactor fuel, although no gross fuel damage is predicted.
Additionally, the possibility of entering a poorly understood region of plant operation during a time period when a large portion of plant instrumentation was not available is of concern.
- However, the likelihood of an event of this kind is not high and the Anticipated Transient with Scram (ATWS) Emergency Operating Procedures (EOP) should give appropriate guidance in methods of dealing with the consequences of such an event.
A short report, detailing this analysis and its finding will be submitted to you in the near future.
If additional information or analysis is needed, please contact me at x29004.
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data
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UNITED STATES NUCLEAR REGULATORY COMMISSlON WASHINGTON, D. C. 20555 MEMORANDUM FOR:
Jack Rosenthal, Team Leader Nine Mile 2 Incident Investigation Team FROM:
SUBJECT:
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data TRANSMITTAL OF POSSIBLE RE-CRITICALITY
- TRANSIENT ANALYSIS As per your request, I have conducted a primary analysis of the possible consequences of a re-criticality event during the cooldown of the Nine Mile 2 nuclear reactor following the August 13, 1991 event.
This analysis should be considered as a general scoping study of possible consequences of an event of this kind.
The analysis is limited to the information available at the time and modeling of a generic BWR/5-6.
The analysis was performed using a high fidelity thermal hydraulic code, Real-Time Advanced Core and Thermohydraulic (RETACT) to model the Nine Mile 2 reactor.
This computation work was supplemented by hand calculations and discussion with other reactor safety experts and operational personnel.
The report includes best estimate analysis of the consequence of the event as well as some possible consequences of additional system failures.
The primary conclusions of this preliminary analysis of the event are that:
the rapid injection of feedwater into the reactor during cooldown, concurrent with a partial failure of all rods to fully insert into the core, will likely lead to a re-criticality transient; this re-criticality will lead to a rapid repressurization of the reactor vessel, a power excursion, and possible localized fuel damage; due to the power excursion at a low reactor core flow rate, the core could enter LaSalle type flow and power oscillations.
This is of particular concern because under these conditions the plant operator has no control of the control rods to shut the reactor
- down, and because this type of power and flow behavior is not well understood;
I the primary governing parameters in this event are the effective multiplication factor at the time of the feedwater injection, which is a function of core life and degree of rod insertion, the amount of feedwater injected and its temperature, and the ability of the operator to control and relieve reactor pressure.
The overall conclusion of this study is that the injection of feedwater into the vessel could have produced significant but not severe damage to reactor components including reactor fuel, although no gross fuel damage is predicted.
Additionally, the possibility of entering a poorly understood region of plant operation during a time period when a large portion of plant instrumentation was not available is of concern.
However, the likelihood of an event of this kind is not high and the Anticipated Transient with Scram (ATWS) Emergency Operating Procedures (EOP) should give appropriate guidance in methods of dealing with the consequences of such an event.
A 'short report, detailing this analysis and its finding will be submitted to you in the near future.
If additional information or analysis is needed, please contact me at x29004.
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 MEMORANDUM FOR:
Jack Rosenthal, Team Leader Nine Mile 2 Incident Investigation Team FROM:
SUBJECT:
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data TRANSMITTAL OF POSSIBLE RE-CRITICALITY TRANSIENT ANALYSIS As per your request, I have conducted a primary analysis of the possible consequences of a re-criticality event during the cooldown of the Nine Mile 2 nuclear reactor following the August 13, 1991 event.
This analysis should be considered as a general scoping study of possible consequences of an event of this kind.
The analysis is limited to the information available at the time and modeling of a generic BWR/5-6.
The analysis was performed using a high fidelity thermal hydraulic code, Real-Time Advanced Core, and Thermohydraulic (RETACT) to model the Nine Mile 2 reactor.
This computation work was supplemented by hand calculations and discussion with other reactor safety experts and operational personnel.
The report includes best estimate analysis of the consequence of the event as well as some possible consequences of additional system failures.
The primary conclusions of this preliminary analysis of the event are that:
the rapid injection of feedwater into the reactor during cooldown, concurrent with a partial failure of all rods to fully insert into the core, will likely lead to a re-criticality transient; this re-criticality will lead to a rapid repressurization of the reactor
- vessel, a power excursion, and possible localized fuel damage; due to the power excursion at a low reactor core flow rate, the core could enter LaSalle type flow and power oscillations.
This is of particular concern because under these conditions the plant operator has no control of the control rods to shut the reactor
- down, and because this type of power and flow behavior is not well understood.
the primary governing parameters in this event are the effective multiplication factor at the time of the feedwater injection, which is a function of core life and degree of rod insertion, the amount of feedwater injected and its temperature, and the ability of the operator to control and relieve reactor pressure.
The overall conclusion of this study is that the injection of feedwater into the vessel could have produced significant but not severe damage to reactor components including reactor fuel, although no gross fuel damage is predicted.
Additionally, the possibility of entering a poorly understood region of plant operation during a time period when a large portion of plant instrumentation was not available is of concern.
However, the likelihood of an event of this kind is not high and the Anticipated Transient with Scram (ATWS) Emergency Operating Procedures (EOP) should give appropriate guidance in methods of dealing with the consequences of such an event.
A short report, detailing this analysis and its finding will be submitted to you in the near future.
If additional information or analysis is needed, please contact me at x29004.
Steven Amdt Division of Operational Assessment Office for Analysis and Evaluation of Operational Data