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{{#Wiki_filter:3. NUREG-1 537, Part 1, Sectioti 4.6 recommends that an | {{#Wiki_filter:3. NUREG-1 537, Part 1, Sectioti 4.6 recommends that an applicationshouldjustify the assumptions and methods, and validate results. bi/scuss the BLOOST code transientanalysis in additionaldetail including the quantitativevalues of the rfajor parametersthat enter the analysis. This should include the flux peakingfactors used to represehf the hottestfuel rod. | ||
This should include the flux | Please see the attached "BLOOST Code Validation Report" which provides more detail regarding the assumptions and methods of th'e BLOOST code. The "BLOOST Code Validation Report" also discusses a benchmarking study performed at the Sandia TRIGA Annular Core Pulse Reactor (ACPR). | ||
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: 5. NUREG-1537, Part 1, Section 13.1.2 provides guidance to the licensee to identfi,5 all potential methods whereby excess reactivity could be accidentally inserted into the | : 5. NUREG-1537, Part 1, Section 13.1.2 provides guidance to the licensee to identfi,5 all potential methods whereby excess reactivity could be accidentally inserted into the reactorto cause an excursion. Please discuss these potential methods and evaluate those found to be credible. At a minimum, analyze a ramp insertion of reactivity of the most reactive control rod (not the transient rod) at its maximum insertion rate, startingfrom the most Jim iting power level. Evaluationsshould include discussion of the model, assumptions, and calculationresults. | ||
Please discuss these potential methods and evaluate those found to be credible. | A ramp insertion of reactivity resulting from the withdrawal of a control rod may be caused by operator error or instrument malfunction. In order to determine power as a function of time during a linear reactivity increase, a single delayed neutron group model with a prompt jump approximation is used: | ||
At a minimum, analyze a ramp insertion of reactivity of the most reactive control rod (not the transient rod) at its maximum insertion rate, | *where P(t) is the power at time t, P0 is the initial power level, f3 is the total delayed neutron fraction (0.007), X,is the one group decay constant (0.405 sec-'), t is time, and y is the linear insertion rate of reactivity. | ||
In order to determine power as a function of time during a linear reactivity increase, a single delayed neutron group model with a prompt jump approximation is used:*where P(t) is the power at time t, | Control rod data for the AFRRI TRIGA reactor is shown in Table 1. | ||
Control rod data for the AFRRI TRIGA reactor is shown in Table 1.Table 1. Control rod data for the AFRRI TRIGA reactor in core position 500.Rod Total Worth Total Withdrawal Average Insertion ($). Time (sec) Rate ($/sec)Transient 2.89 29.0 0.0997 Safety. 2.65 39.4 0.0673 Shim 2.74 36.1 0.0760 Regulating 3.01 34.8 0.0865 For the reactivity insertion accident, starring power levels of 100 W and 1.0 MW were considered. | Table 1. Control rod data for the AFRRI TRIGA reactor in core position 500. | ||
The SCRAM set point was assumed to b~e 1.09 MW, and a 0.5 second delay time was assumed between reaching the SCRAM set point and of the control rods. In addition to the single control rod withdrawal scenario, the simultaneous withdrawal of all four control rods was also analyzed as a worst case. | Rod Total Worth Total Withdrawal Average Insertion | ||
Table 2. Summary of ramp insertion of reactivity for AFRRI TRIGA control rods.'Rod Withdrawn Starting Power Time until Release of Total Reactivity Control Rods (sec) Inserted at SCRAM ($)Transient 100 W 9.77 0. | ($). Time (sec) Rate ($/sec) | ||
: 6. NUREG-1537 | Transient 2.89 29.0 0.0997 Safety. 2.65 39.4 0.0673 Shim 2.74 36.1 0.0760 Regulating 3.01 34.8 0.0865 For the reactivity insertion accident, starring power levels of 100 W and 1.0 MW were considered. The SCRAM set point was assumed to b~e 1.09 MW, and a 0.5 second delay time was assumed between reaching the SCRAM set point and a*tual-elease of the control rods. In addition to the single control rod withdrawal scenario, the simultaneous withdrawal of all four control rods was also analyzed as a worst case. | ||
Section 13.1.5 of the SAR presents the results oalan analysis al a reactivity insertion of $0.51.Justify the magnitude of this assumed reactivity insertion in comparison with the maximum reactivity insertion | |||
The following analysis will replace the reference to a reactivity insertion of $0.51 of the SAR: The failure of an experiment or experiments could result in instantaneous insertion of reactivity. | Table 2. Summary of ramp insertion of reactivity for AFRRI TRIGA control rods. | ||
The worst possible case would be the prompt addition of $3.00 (2.1% Ak/k) within the reactor core. The Technical Specifications establish that the sum of the absolute reactivity worths of all experiments in the reactor and in the associated experimental facilities shall not exceed $3.00 (2.1% Ak/k). The instantaneous insertion of $3.00 (2.1% Ak/k) to the reactor core as a result of a worst case reactivity insertion is bounded by the an~lysis of the $3.50 (2.45% Ak/k) pulse limit and would not result in any adverse safety conditions withili the AFRRI TRIGA core.}} | 'Rod Withdrawn Starting Power Time until Release of Total Reactivity Control Rods (sec) Inserted at SCRAM ($) | ||
Transient 100 W 9.77 0.93 | |||
___________1.0 MW 1.23 0.13 Safety 100 W 13.5 0.92 | |||
___________1.0 MW 1.53 0.11 Shim 100 W 12.23 0.93 1.0 MW .1.43 0.11 Regulating "100 W 10.99 0.95 1.0 MW 1.32 0.12 All Rods 100 W 3.51 1.16 1.0 MW 0.74 0.25 In all cases, including the simultaneous withdrawal of all control rods, the total reactivity insertion is well below the pulse reactivity insertion limit of $3.50 (2.45% Ak/k), thus safety of the reactor would not be adversely impacted. | |||
: 6. NUREG-1537 Part1, Section 13 provides guidance to the licensee to discuss potential accident scenarios. Section 13.1.5 of the SAR presents the results oalan analysis al a reactivity insertion of $0.51. | |||
Justify the magnitude of this assumed reactivity insertion in comparison with the maximum reactivity insertion associatedwith any sin~gle experiment. | |||
The following analysis will replace the reference to a reactivity insertion of $0.51 of the SAR: | |||
The failure of an experiment or experiments could result in instantaneous insertion of reactivity. The worst possible case would be the prompt addition of $3.00 (2.1% Ak/k) within the reactor core. The Technical Specifications establish that the sum of the absolute reactivity worths of all experiments in the reactor and in the associated experimental facilities shall not exceed $3.00 (2.1% Ak/k). The instantaneous insertion of $3.00 (2.1% Ak/k) to the reactor core as a result of a worst case reactivity insertion is bounded by the an~lysis of the $3.50 (2.45% Ak/k) pulse limit and would not result in any adverse safety conditions withili the AFRRI TRIGA core.}} | |||
Latest revision as of 05:20, 31 October 2019
| ML15296A451 | |
| Person / Time | |
|---|---|
| Site: | 05000704, Armed Forces Radiobiology Research Institute |
| Issue date: | 04/20/2012 |
| From: | Sherman R General Atomics, Hitachi America, Ltd, Marubeni Canada, Ltd, TRIGA Technologies |
| To: | Govt of Thailand, Office of Atomic Energy for Peace, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME1587 21C025, Rev. 0 | |
| Download: ML15296A451 (4) | |
Text
3. NUREG-1 537, Part 1, Sectioti 4.6 recommends that an applicationshouldjustify the assumptions and methods, and validate results. bi/scuss the BLOOST code transientanalysis in additionaldetail including the quantitativevalues of the rfajor parametersthat enter the analysis. This should include the flux peakingfactors used to represehf the hottestfuel rod.
Please see the attached "BLOOST Code Validation Report" which provides more detail regarding the assumptions and methods of th'e BLOOST code. The "BLOOST Code Validation Report" also discusses a benchmarking study performed at the Sandia TRIGA Annular Core Pulse Reactor (ACPR).
por~-A_*ieddw 3-t
- 5. NUREG-1537, Part 1, Section 13.1.2 provides guidance to the licensee to identfi,5 all potential methods whereby excess reactivity could be accidentally inserted into the reactorto cause an excursion. Please discuss these potential methods and evaluate those found to be credible. At a minimum, analyze a ramp insertion of reactivity of the most reactive control rod (not the transient rod) at its maximum insertion rate, startingfrom the most Jim iting power level. Evaluationsshould include discussion of the model, assumptions, and calculationresults.
A ramp insertion of reactivity resulting from the withdrawal of a control rod may be caused by operator error or instrument malfunction. In order to determine power as a function of time during a linear reactivity increase, a single delayed neutron group model with a prompt jump approximation is used:
- where P(t) is the power at time t, P0 is the initial power level, f3 is the total delayed neutron fraction (0.007), X,is the one group decay constant (0.405 sec-'), t is time, and y is the linear insertion rate of reactivity.
Control rod data for the AFRRI TRIGA reactor is shown in Table 1.
Table 1. Control rod data for the AFRRI TRIGA reactor in core position 500.
Rod Total Worth Total Withdrawal Average Insertion
($). Time (sec) Rate ($/sec)
Transient 2.89 29.0 0.0997 Safety. 2.65 39.4 0.0673 Shim 2.74 36.1 0.0760 Regulating 3.01 34.8 0.0865 For the reactivity insertion accident, starring power levels of 100 W and 1.0 MW were considered. The SCRAM set point was assumed to b~e 1.09 MW, and a 0.5 second delay time was assumed between reaching the SCRAM set point and a*tual-elease of the control rods. In addition to the single control rod withdrawal scenario, the simultaneous withdrawal of all four control rods was also analyzed as a worst case.
Table 2. Summary of ramp insertion of reactivity for AFRRI TRIGA control rods.
'Rod Withdrawn Starting Power Time until Release of Total Reactivity Control Rods (sec) Inserted at SCRAM ($)
Transient 100 W 9.77 0.93
___________1.0 MW 1.23 0.13 Safety 100 W 13.5 0.92
___________1.0 MW 1.53 0.11 Shim 100 W 12.23 0.93 1.0 MW .1.43 0.11 Regulating "100 W 10.99 0.95 1.0 MW 1.32 0.12 All Rods 100 W 3.51 1.16 1.0 MW 0.74 0.25 In all cases, including the simultaneous withdrawal of all control rods, the total reactivity insertion is well below the pulse reactivity insertion limit of $3.50 (2.45% Ak/k), thus safety of the reactor would not be adversely impacted.
- 6. NUREG-1537 Part1, Section 13 provides guidance to the licensee to discuss potential accident scenarios. Section 13.1.5 of the SAR presents the results oalan analysis al a reactivity insertion of $0.51.
Justify the magnitude of this assumed reactivity insertion in comparison with the maximum reactivity insertion associatedwith any sin~gle experiment.
The following analysis will replace the reference to a reactivity insertion of $0.51 of the SAR:
The failure of an experiment or experiments could result in instantaneous insertion of reactivity. The worst possible case would be the prompt addition of $3.00 (2.1% Ak/k) within the reactor core. The Technical Specifications establish that the sum of the absolute reactivity worths of all experiments in the reactor and in the associated experimental facilities shall not exceed $3.00 (2.1% Ak/k). The instantaneous insertion of $3.00 (2.1% Ak/k) to the reactor core as a result of a worst case reactivity insertion is bounded by the an~lysis of the $3.50 (2.45% Ak/k) pulse limit and would not result in any adverse safety conditions withili the AFRRI TRIGA core.