ML080390035
| ML080390035 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 02/08/2008 |
| From: | NRC/NRR/ADRO/DPR/RTRBA |
| To: | |
| Shared Package | |
| ML080390044 | List: |
| References | |
| TAC MA6084 | |
| Download: ML080390035 (5) | |
Text
MITR RAIs Rev 0B Draft, January 29, 2008
RAI-1
Technical Evaluation Report - MITR CHAPTER 2 2.1 Section 2.2.2 mentioned an initiative from Logan Airport to construct a second easterly approach runway. Describe the actions that Logan Airport has taken since 2000 to approve or construct this runway. Describe any additional risk to MITR from this runway if it has been or will be constructed.
CHAPTER 4 4.1 Section 4.2, Reactor Core, page 4-4. Provide a description of the design and specifications for the boron impregnated stainless steel absorbers installed in the core.
Include a mechanical description of how these absorbers are installed, replaced (including frequency) and fixed in place. Discuss the safety implications of the absorbers. See RAI 4.11.
4.2 Section 4.2, Reactor Core, page 4-4. Provide a more detailed description of the flow path within the core tank. Include in the description the flow distribution to the fuel elements, bypass flow, the location of the three exit ports, purpose of the flow guide, and location of the outlet plenum on Figure 4-10 or similar drawing.
4.3 Deleted.
4.4 Section 4.2.1, Reactor Fuel, page 4-5. The second paragraph states the fuel meat is 2.082 inches wide or 5.288 cm. In RAI Response # 17, the width dimension is 5.588 cm. Explain the apparent contradiction.
4.5 Section 4.2.1, Reactor Fuel. Provide a discussion of and references for the fuel development and testing program for the MITR-III fuel, including the limiting characteristics of the specific fuel used in the reactor (NRC to search for documentation for approval of the fuel design).
4.6 Section 4.2.1, Reactor Fuel, page 4-6. Provide references 4-1 and 4-2 and any other references necessary to provide the design bases for the requested increase in the fission density and void fraction limits as indicated in RAI Response # 13.
4.7 Provide the following reference regarding corrosion thickness referred to in RAI # 99 Response, C. deWalsche, "Prediction of the Oxidation of the Fuel Clad and Consequences for the MIT Research Reactor, June, 1997".
4.8 Section 4.2.1, Reactor Fuel, page 4-7. Provide the reference(s) for the fuel heat capacity and thermal conductivity. Explain how the uncertainties in these values are incorporated into the safety margin for the thermal limits.
MITR RAIs Rev 0B Draft, January 29, 2008
RAI-2
Technical Evaluation Report - MITR 4.9 Section 4.2.2.6, Scram Logic and Circuitry, page 4-15. This section states One relay cuts off all power to the paired magnet power supply (blade 4 for channel 1, etc.)
dropping that shim blade, while the other relay opens the withdraw permit circuit dropping the remaining shim blades. Explain the blade 4 and channel 1 correlation.
4.10 Section 4.5, Nuclear Design, page 4-37. Clarify the code usage for nuclear safety analysis power density distributions. Sections 4.5 b) and 4.5 c) indicate that CITATION is currently in use and MCNP is being evaluated for this purpose. However, response to RAI # 25 indicates that this is a misinterpretation of section 4.5 b) and that MCNP is under development for use in a fuel management program. Section 4.6.1.2 also suggests that MCNP is used for power density distribution calculations. If MCNP is to be used for safety analysis purposes, provide benchmarking and validation documentation, e.g.,
critical position, flux distributions, reactivity coefficients.
4.11 Section 4.5.1.3, Reactor Operating Characteristics, page 4-39. Clarify the statement, These studies have shown that as shim blades and/or fixed absorbers are raised..., i.e.,
explain how the fixed absorbers are raised. See RAI 4.1.
4.12 Section 4.5.1.4, Effect of Fuel Burnup, page 4-39. Discuss the typical historical fuel element peak fission density achievement. Include a discussion of the method of determining the approach to the fission density limit.
4.13 Section 4.5.1.5, Kinetic Behavior/Requirements and Features of Control Devices, page 4-
- 40. Provide a reference calculation showing the equilibrium xenon worth of 4.2 stated in this section.
4.14 Section 4.5.2.1, Neutron Lifetime and Effective Delayed Neutron Fraction, page 4-47.
Provide an estimate of the accuracy for the prompt neutron lifetime and a comparison with similar reactor facilities and calculational methods.
4.15 Section 4.5.3.3, Shutdown Margin, page 4-51. For the three types of experiments, movable, non-secured, and secured, only the secured experiment is defined as having sufficient restraining forces greater than the normal operating environment or as a result of credible malfunctions. Justify why the shutdown margin requirement includes only the movable experiments (or samples in the text).
4.16 Section 4.6.6.2, Calculation of the Safety Limits for Forced Circulation, page 4-71.
Explain how the OFI correlation uncertainty is incorporated into the forced convection safety limits.
4.17 Section 4.6.6.3, Calculation of the Safety Limits for Natural Circulation, page 4-72.
Explain why a uniform axial power distribution (i.e., no Fa is used) is most appropriate/limiting and why the enthalpy rise engineering hot channel factor is used instead of the one for heat flux. Provide a calculation showing the result of 2.353 x 104 W/m2 (units typo in text) and clarify if the ratio of densities term (f/g) is correctly stated in equation (4-30) and in SAR reference 6-1. Also, since reference 4-20 states an
MITR RAIs Rev 0B Draft, January 29, 2008
RAI-3
Technical Evaluation Report - MITR experimental range for L/De of 8-240, justify the use of the correlation for L/De of 260 (RAI # 29).
4.18 Section 4.6.7, Calculation of the Limiting Safety System Settings, page 4-75. Provide the heat transfer correlation used for the finned plates in equation 4-35. In the last sentence on the page, explain how the hot channel subcooling of 10º C is used in determining TONB. Identify and explain any uncertainty in the LSSS associated with the Bergles-Rohsenow correlation.
4.19 Section 4.6.7.1 and 4.6.7.2, Calculation of the Limiting Safety System Settings for Forced/Natural Convection, page 4-76. Explain how the core flow distribution factors are incorporated into the LSSS.
4-20. Section 4.2.3.3, Neutron Reflector - Graphite, page 4-18. Provide a discussion of the Wigner effect with the graphite in the reflector and its potential hazards.
CHAPTER 5 5.1 Section 5.2.1.3, Heat Removal Considerations, page 5-7. Explain how core outlet temperature is monitored for the LSSS during natural circulation operation.
5.2 Provide a description of the Primary Coolant Makeup Water System in accordance with Section 5.5 of NUREG-1537, Part 1.
CHAPTER 6 6.1 Section 6.2, Natural Convection Valves, page 6-3. Explain how the statement that either three of the four natural convection valves or the anti-siphon valves alone are enough to remove decay heat from 6 MW steady-state operation is demonstrated in reference 6-1.
6.2 Deleted.
CHAPTER 8 8.1 Deleted.
CHAPTER 11 11.1 The text does not describe the typical staffing level for personnel implementing the Radiation Protection Program. Aside from the Reactor Radiation Protection Officer and Assistant, provide the number of Health Physicists and Technicians that form the typical staff. Provide a description of the typical staffing available to implement the Radiation Protection Program.
11.2 Training is specified in Section 11.1.2.2. Describe the retraining requirements for facility personnel.
MITR RAIs Rev 0B Draft, January 29, 2008
RAI-4
Technical Evaluation Report - MITR 11.3 Deleted.
CHAPTER 12 12.1 Chapter 12 does not address a startup plan. Provide documentation that addresses MITR III changes such as the effects of power increase from 5 to 6 MW (e.g., Shielding, Instrumentation ranges for power, primary (forced and natural convection), secondary and shield coolant flow and temperature, operating procedure to run at higher power level.
CHAPTER 13 13.1 Section 13.1, Accident Initiating Events and Scenarios, page 13-1. Based on the responses to RAIs #33 and #78, provide a summary of the accident analyses that were redone, including the initial conditions and results.
13.2 Section 13.2.1.5, Conclusion for the Maximum Hypothetical Accident, page 13-16.
Provide an analysis and discussion of the maximum effective doses to the reactor staff for the MHA.
13.3 Section 13.2.2.1, Step Reactivity Insertion, page 13-18. Provide a calculation that includes all of the assumptions, inputs, outputs, and a discussion of the analysis results and justification of the limits chosen based on the results. Explain the relationship between the results and the SL criteria prescribed in Chapter 4.
From reference [13-14], The step reactivity insertion analyses previously made by Dutto and Evo using PARET and Gaborieau using RELAP5 concluded that the step reactivity insertion limit is about $1.35 in order to avoid fuel softening (450º C). The reference memo claims these previous results are incorrect and too low due to a time-step instability in PARET. Explain why the previous similar result from the Gaborieau using RELAP5 is invalid, and justify the use of PARET if there is a time-step instability in the code. Discuss what other instances of this time-step instability are reported in the literature or by ANL or RSICC and describe the remedies.
13.4 Section 13.2.4, Loss of Primary Coolant Flow, page 13-25. Describe the uncertainties associated with the decay power estimate, the flow coast-down data, and the results from the MULCH-II code. Describe what the safety margins are to the approach to the Safety Limits (or the critical heat flux ratios) for the LOF transient. See RAI response # 92(d).
13.5 General. In section 4.6.3, it is stated that the hot channel has the combination of the most limiting conditions including least coolant flow. In RAI response # 28, 24 assemblies are assumed in presenting the worst-case channel flow, and in RAI response # 92, App. D, it appears that 23 assemblies are used in the MULCH-II code. Explain how the number of fuel assemblies is chosen for determining thermal-hydraulic limits and performing accident analyses and explain why this is not represented by the maximum number of assemblies for the worst-case flow.
MITR RAIs Rev 0B Draft, January 29, 2008
RAI-5
Technical Evaluation Report - MITR 13.6 General. The fuel assemblies have 15 plates with 14 internal, two heated side flow channels and 2 external, single heated flow channels. Describe the flow characteristics of these external flow channels and explain how these are modeled in the thermal-hydraulic limits and accident analyses.
13.7 Section 13.2.9.1, Operation with Shim Blades in a Non-Uniform Bank Position, page 13-
- 38. This section states that the subcritical interlock blocks blade withdrawal beyond four inches unless all blades are first brought to the four inch position. TS 3.2.4, Control System Interlocks, describes the subcritical interlock withdrawal restriction at 5.0 inches.
Please explain the discrepancy.
CHAPTER 14 14.1 The definition for containment in the Technical Specifications matches the wording from ANSI/ANS-15.1-1990 verbatim with the exception of the clause that is in the normally closed configuration. Correct or justify why this wording was omitted from the Technical Specifications.
14.2 TS 3.3.5, Cooling Radioactivity Limits, contains both LCOs and Surveillance Requirements. Correct or justify why the surveillance requirements should not be removed from Section 3 and relocated to Section 4 with other surveillance requirements.
14.3 TS 3.4, Reactor Containment Integrity and Pressure Relief System correctly lists the conditions under which containment integrity is required. Provide justification for not listing the minimum equipment for operability in accordance with the guidance of ANSI/ANS-15.1-1990.
14Property "ANSI code" (as page type) with input value "ANSI/ANS-15.1-1990.</br></br>14" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..4 TS 3.5, Ventilation System, correctly lists the conditions under which the ventilation system is required. Provide justification for not listing the minimum equipment for operability in accordance with the guidance of ANSI/ANS-15.1-1990.
14Property "ANSI code" (as page type) with input value "ANSI/ANS-15.1-1990.</br></br>14" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..5 TS 3.6, Emergency Power, correctly lists the conditions under which emergency power is required. Provide justification for not listing the minimum equipment for operability in accordance with the guidance of ANSI/ANS-15.1-1990.
14Property "ANSI code" (as page type) with input value "ANSI/ANS-15.1-1990.</br></br>14" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..6 TS 3.7.2, Effluents specifies that the site will comply with 10 CFR 20, but details specific dilution factors for use. This specification does not list release limits as required by ANSI/ANS-15.1-1990, Section 3.7.2. Include more specificity as to which table and column from Part 20 is applicable or justify this omission.