ML080710352
| ML080710352 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 03/12/2008 |
| From: | Pierce S NRC/NRR/ADRA/DPR/PRTA |
| To: | Bernard J Massachusetts Institute of Technology (MIT) |
| PIERCE S, NRR/DPR/PRTA 415-2261 | |
| References | |
| TAC MA6084 | |
| Download: ML080710352 (8) | |
Text
March 12, 2008 Dr. John A. Bernard, Director Nuclear Reactor Laboratory Massachusetts Institute of Technology 138 Albany Street Cambridge, MA 02139-4296
SUBJECT:
MASSACHUSETTS INSTITUTE OF TECHNOLOGY - REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL REQUEST (TAC NO. MA6084)
Dear Dr. Bernard:
We are continuing our review of your July 8, 1999 application, response to RAI submission dated January 29, 2004 and associated supplements, for the renewal of the Massachusetts Institute of Technologys (MIT) Research Reactor Amended Facility Operating License No. R-37. During our review of your request, questions have arisen for which we require additional information and clarification. Please provide responses to the enclosed request for additional information within 90 days of the date of this letter. In accordance with 10 CFR 50.30(b), your response must be executed in a signed original under oath or affirmation.
Following receipt of the additional information, we will continue our evaluation of your application request. NRC staff is conducting this review in accordance with the guidance of NUREG 1537.
Sincerely,
/RA/
Stephen Pierce, Project Manager Research and Test Reactors Branch A Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No.50-020 License No. R-37
Enclosure:
As stated cc: See next page
Massachusetts Institute of Technology Docket No.50-020 cc:
City Manager City Hall Cambridge, MA 02139 Department of Environmental Protection One Winter Street Boston, MA 02108 Director Radiation Control Program Department of Public Health 90 Washington Street Dorchester, MA 02121 Nuclear Preparedness Manager Massachusetts Emergency Management Agency 40 Worcester Road Framingham, MA 01702-5399 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611 Bill Watkins Washington Safety Management Solutions 2131 S. Centennial Avenue Aiken, SC 29803
ML080710352 OFFICE DPR/PRTA DPR/PRTA:LA DPR/PRTB:PM DPR/PRTA:BC DPR/PRTA NAME SPierce sp CHart for EBarnhill DHughes deh DCollins dsc SPierce scp DATE 3/11/08 3/11/08 3/12/08 3/12/08 3/12/08
REQUEST FOR ADDITIONAL INFORMATION MASSACHUSETTS INSTITUTE OF TECHNOLOGY RESEARCH REACTOR DOCKET NO. 50-20 The following questions pertain to the facility and site characteristics as described in the Safety Analysis Report (SAR) and are necessary to verify compliance with 10 CFR Part 50.34(a)(10),
Appendix E Emergency Planning and Preparedness for Production and Utilization Facilities.
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.
The following questions pertain to neutronics, thermal hydraulics, and reactor design as described in the SAR and previous RAI responses. They are necessary to verify compliance with 10CFR Part 50.34 Contents of Applications and 10 CFR 50.36 Technical Specifications.
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 4.7 below.
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 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.
4.4 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.
4.5 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.
4.6 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.
4.7 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...,
Explain how the fixed absorbers are raised. See RAI 4.1 above.
4.8 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.
4.9 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.
4.10 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 calculational methods.
4.11 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.)
4.12 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.
4.13 Please provide the deWalsche report referenced in RAI response #99.
5.1 Provide a description of the Primary Coolant Makeup Water System.
6.1 Section 6.2, Natural Convection Valves, page 6-3. Explain how the statement 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.
The following questions apply to potential accident or radiological release scenarios or conditions as described in the SAR and previous RAI responses. These questions are necessary to verify compliance with 10 CFR 50.36, Technical Specifications, 10 CFR Part 20 Subpart C, Occupational Dose limits, and 10 CFR Part 20 Subpart D, Radiation Dose Limits for Individual Members of the Public.
4.14 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.
4.15 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 experimental range for L/De of 8-240, justify the use of the correlation for L/De of 260 RAI #29.
4.16 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.
4.17 Sections 4.6.7.1 and 4.6.7.2, Calculation of the Limiting Safety System Settings for Forced Convection, and Calculation of the Limiting Safety System Settings for Natural Convection page 4-76. Explain how the core flow distribution factors are incorporated into the LSSS.
11.1 Section 11.1.1 of the MIT SAR describes radiation sources included in the scope of the MITR license. The principal hazards are the reactor, its fuel, and beams of radiation from the reactor or fission converter. Please provide descriptions of any experiments or routine operations and the controls used to minimize personnel dose.
13.1 Section 13.2.1.5, Conclusion for the Maximum Hypothetical Accident (MHA), page 13-
- 16. Provide an analysis and discussion of the maximum effective doses to the reactor staff for the MHA.
13.2 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.
13.3 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).
13.4 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.
13.5 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.
The following questions pertain to technical specification (TS) and are necessary to verify compliance with 10 CFR Part 50.36 Technical Specifications.
5.2 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.
12.1 Chapter 12 does not address a startup plan. Using the guidance of NUREG 1537 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.
14.1 TS 3.3.5, Cooling Radioactivity Limits, contain both LCOs and Surveillance Requirements. Correct or justify why the surveillance requirements should not be relocated to Section 4 with other surveillance requirements.
14.2 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..3 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..4 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..5 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 recomended 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.
These questions pertain to personnel staffing levels, training and qualification. These questions are necessary to verify compliance with 10 CFR 55 and 10 CFR Part 20 Subpart C Occupational Dose limits.
11.2 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.
11.3 Training is specified in Section 11.1.2.2. Describe the retraining requirements for facility personnel.
12.2 The Safety Analysis Report submitted on February, 10, 2000, presents a requalification program that reflects 10 CFR 55 regulations. Please provide the requalification program.
The following questions are editorial in nature.
4.18 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.
13.6 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.
14.6 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.