ML101160266

University of Missouri at Columbia - Request for Additional Information License Renewal, Safety Analysis Report, 45-Day Response Questions

ML101160266
Person / Time
Site:	University of Missouri-Columbia
Issue date:	06/01/2010
From:	Alexander Adams Research and Test Reactors Licensing Branch
To:	Rhonda Butler Univ of Missouri - Columbia
ADAMS A, NRC/NRR/DPR/PRLB 415-1127
References
TAC ME3034
Download: ML101160266 (22)

Text

June 1, 2010 Mr. Ralph Butler, Director Research Reactor Center University of Missouri-Columbia Research Park Columbia, MO 65211

SUBJECT:

UNIVERSITY OF MISSOURI AT COLUMBIA C REQUEST FOR ADDITIONAL INFORMATION RE: LICENSE RENEWAL, SAFETY ANALYSIS REPORT, 45-DAY RESPONSE QUESTIONS (TAC NO. MD3034)

Dear Mr. Butler:

We are continuing our review of your renewal request for Amended Facility Operating License No. R-103 for the University of Missouri - Columbia Research Reactor which you submitted on August 31, 2006, as supplemented. During our review of your renewal request, questions have arisen for which we require additional information and clarification. Enclosed is a partial request for additional information. We are requesting a response within 45 days of the date of this letter.

An additional request for information containing our remaining questions of a complex nature has been sent under separate cover. In accordance with Title 10 of the Code of Federal Regulations Section 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 renewal request.

If you have any questions regarding this review, please contact me at (301) 415-1127.

Sincerely,

/RA/

Alexander Adams, Jr., Senior Project Manager Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-186

Enclosure:

As stated cc w/encl: See next page

University of Missouri-Columbia Docket No. 50-186 cc:

University of Missouri Associate Director Research Reactor Facility Columbia, MO 65201 Homeland Security Coordinator Missouri Office of Homeland Security P.O. Box 749 Jefferson City, MO 65102 Planner, Dept of Health and Senior Services Section for Environmental Public Health 930 Wildwood Drive, P.O. Box 570 Jefferson City, MO 65102-0570 Deputy Director for Policy Department of Natural Resources 1101 Riverside Drive Fourth Floor East Jefferson City, MO 65101 A-95 Coordinator Division of Planning Office of Administration P.O. Box 809, State Capitol Building Jefferson City, MO 65101 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611

ML101160266 OFFICE PRLB:PM PRPB:LA PRLB:ABC PRLB:PM NAME AAdams GLappert AAdams /L.Tran for/ AAdams DATE 5/20/2010 5/12/2010 5/28/2010 6/1/2010

OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE RENEWAL REQUEST FOR THE UNIVERSITY OF MISSOURI-COLUMBIA RESEARCH REACTOR LICENSE NO. R-103 DOCKET NO. 50-186 The purpose of the following questions is to assist the staff in determining that the renewal application from the University of Missouri meets the requirements of the regulations, in particular the regulations in Title 10 of the Code of Federal Regulations (10 CFR) Parts 20 and

50. The questions are based on a review of your application for license renewal using the NRC staffs standard review plan in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Part 2, Standard Review Plan and Acceptance Criteria.

We have split our questions into two groups, complex questions that have a 120-day requested response time and the remaining questions with a 45-day response time. Because of this, the numbering of the questions below is not consecutive. Both groups of question considered together result in a complete set of consecutively numbered questions.

GENERAL 0.1 Your renewal Safety Analysis Report (SAR) is dated August 18, 2006. Please describe changes that have occurred to the facility since the SAR was submitted to the U. S.

Nuclear Regulatory Commission (NRC) that impact the safety conclusions of the SAR.

CHAPTER 1 1.1 Section 1.6, Compliance With the Nuclear Waste Policy Act of 1982, Page 1-22. Provide details on the Missouri University Research Reactor (MURR) contract for disposition of fuel. Include the contract number and with whom the contract exists.

CHAPTER 2 2.1 Section 2.1.1, Site Location and Description, Page 2-1. The statement is made that personnel located within 1,500 feet of the facility can be rapidly evacuated if required.

What is the basis of this statement?

CHAPTER 3 3.1 Section 3.2, Meteorological Damage, Page 3-28. The statement is made that the containment structure has been designed for area wind loads. Discuss the design wind loads and why the design of the containment is sufficient?

ENCLOSURE

CHAPTER 4 4.1 Section 4.2.1.1, Reactor Fuel System, Page 4-5. The conclusive statement in the text no aluminde fuel element failures have occurred during the past thirteen years does not appear to be supported by Reference 4.46, dated 1990. Discuss and provide a reference for MURR fuel performance more recent than 1990.

4.2 Section 4.2.1.2, Fuel Element Description, Page 4-10. Table 4-3 gives innermost and outermost fuel plate center radii which differ from what is stated in the text. Please discuss.

4.3 Section 4.2.1.2, Fuel Element Description, Page 4-8. This section includes a statement about assemblies withstanding severe hydraulic tests as a basis that the fuel assemblies will remain intact under hydraulic conditions in MURR. Provide a reference or basis for this statement.

4.4 Section 4.2.1.3, Fuel Element Material of Construction, Page 4-11. Provide a reference or a discussion to how using other materials such as Al alloy 1100 for fuel cladding affects fuel performance.

4.5 Section 4.2.2.1, Control Blade Description.

a. Page 4-15. Provide a further description of the BORAL plate and its design to withstand the hydraulic and radiation environment of the reactor.

b. Provide a discussion of the control blade drive and control independence that prevents a malfunction in one blade from affecting the insertion or withdrawal, or both, of any other blade.

c. Provide a discussion of the decision criteria used for replacement of control blades.

4.6 Section 4.2.2.2, Evaluation of the Control Blades, Page 4-18. A coefficient of linear expansion for aluminum is given in the text. Please confirm that the proper reference for this property is 4.19.

4.7 Section 4.2.2.3, Evaluation of Control Blade Distortion, Page 4-18. How 860 BTU/hr/ft2 was calculated as the fuel element thermal conductivity is not discussed. Back-calculating we performed results in a thermal conductivity which differs from other published sources (ORNL-981). Please provide additional detail for the calculation of this value.

4.8 Section 4.2.3, Neutron Moderator and Reflector, Page 4-19.

a. Discuss why the impact of radiation damage and thermal expansion on the rigid structure of the reflector materials is acceptable.

b. Discuss if potential experiment facility malfunctions in the reflector could affect reactor core components.

c. No technical specification (TS) is proposed for the 26,000 Megawatt-Day (MWD) period between replacements of the Beryllium reflector. Propose a TS or justify why this is not needed.

d. Discuss Wigner energy in the graphite reflector and its potential hazard, if any, to the safe operation of the reactor.

4.9 Section 4.2.4, Neutron Startup Source, Page 4-20. No TS for the administrative control of the irradiation of the regenerative neutron source has been proposed. Discuss actions to ensure that the neutron source licensed activity limit will not be exceeded. Propose TS wording or justify why it is not needed.

4.10 Section 4.2.5, Core Support Structure, Page 4-23. Discuss the impact of radiation damage, coolant chemistry and reactivity effects on core support structure materials over the renewal period of the reactor license.

4.11 Section 4.3, Reactor Pool, Page 4-24.

a. Discuss how the tank material and reactor position limits radiation damage to the pool liner.

b. Provide details of plans to assess irradiation or chemical damage of the pool liner over the term of the license renewal, or justify why this is not needed.

c. What is the minimum detectable amount of pool leakage and the length of time a leak could exist before detection? In the event of a pool leak, describe the probable path of the pool water, the potential for the pool water to enter the environment, and potential for radiological impact on members of the public.

4.12 Section 4.4, Biological Shield, Page 4-25. Describe how heating of the biological shield from interaction with radiation is controlled. Describe any potential for unacceptable damage from this heating if it occurs.

4.13 Section 4.5, Nuclear Design.

a. Page 4-30. Figure 4.8 appears to show details of AMPX-II and not the computational system BOLD VENTURE IV as stated in the text. Provide a figure similar to Figure 4.8 which illustrates components of BOLD VENTURE.

b. Page 4-35, Figure 4.11. Figure 4.11 does not contain error bars or a legend, and the burnup step being compared is not clear. In light of this, discuss the conclusion that the R-Z MURR model agrees with the observed measurements from fuel element 775-F3.

c. Page 4-37, Figure 4.9. Provide further detail of the methodology used to generate Table 4.9 and how reactivity due to burnup of fuel was calculated when control rod position varies at every MWD step.

d. Provide details of the Theta-R 2D calculations and how they are incorporated into the 3D MURR model.

e. Page 4-36. An error larger than the 0.43% error listed in the text can be calculated in the same manner as the 0.43% error between the blades-all-out 2D and 3D model discussion in the text. Discuss the error in keff for all blades full-in and how this impacts reactor control.

f. Page 4-37. The result keff =1.109 appears to be calculated from the R-Z model (1.1124 - 0.0034), not from the 3D MURR model as suggested in the text.

Provide a clarification for the comparison of the 3D MURR model to the measured value of keff =1.095.

g. Page 4-38, Table 4-11. The R-Z model total blade worth differs from what is stated in the text, 0.1809 on pg 4-36. Please explain this difference.

h. Page 4-36. Neither the flux trap model nor the benchmark study are well described in the text of this section. Provide additional clarification.

i. Page 4-37, Table 4-10. Provide a clarification of the results presented in the table. How is the peak power density for the control blades full-in case calculated from the 3D model?

4.14 Section 4.5.3, Operating Limits.

a. Page 4-39. The core temperature reactivity coefficient is given as -6.0x105.

Confirm that this is typographical error regarding the positive exponent.

b. Page 4-37. The cold clean core keff in Table 4-12 does not agree with pg 4-37 first paragraph. Discuss this difference.

c. Page 4-40. Discuss how the limits on shutdown margin are met given the TS limits on excess reactivity and experiment worth.

4.17 Section 4.6.2, Steady State Forced Cooling.

a. It appears that the SAR uses two sets of peaking factors for two different thermal analyses. Provide clarification as to the reason for reporting both analysis results.

Are the inputs used bounding for current and all fuel cycles?

b. Explain the differences between the core coolant and pool coolant flow rates given in Table 4-13 and those given in other sections of Chapter 4.

4.18 Section 4.6.3, Safety Limit Analysis.

a. Page 4-57. Cobra3C is stated as being used to verify outdated BOLERO results.

Discuss the Cobra3C analysis and results, and the verification methodology.

b. Page 4-51. The NUS study results in three sets of safety limit curves (figures 4.12, 4.13, and 4.14), which effectively define the conditions at which the Departure from Nucleate Boiling Ratio (DNBR) is equal to 1.2. However, this is not the typical DNBR at normal operating conditions (or transients). Discuss the thermal margin by showing the DNBR at normal operating conditions.

c. Page 4-59. Explain the impact of measurement uncertainty and instrument response time on the results of Case One.

d. Present the results of the calculation from the equation at the top of page 4-60 or explain why the results are not given.

CHAPTER 5 5.1 Section 5.2.2, General Operating Conditions, Page 5-2. The inlet temperature to the reactor is listed as 120 degrees F, however it is listed as 140 degrees F in Table 4-13.

Explain this difference in values.

5.2 Section 5.2, Primary Coolant System.

a. American National Standards Institute/American Nuclear Society Standard ANSI/ANS 15.1 recommends TS limits on conductivity and pH of the primary coolant water. Propose TS wording for primary water chemistry (pH and conductivity) and an appropriate surveillance frequency or justify why it is not needed. Provide a history of typical values for primary water pH.

b. Describe how oxygen and hydrogen from the radiolysis of primary coolant is controlled.

5.3 Section 5.2.4, Heat Exchangers. Provide a discussion of pressures in the primary, pool and secondary coolant system during operation and shutdown conditions and the potential for primary and pool coolant to enter the secondary system. Discuss the sensitivity of the secondary coolant monitoring system to detect radionuclides in the secondary coolant. In the event of leakage into the secondary system, what is the minimum detectable leak rate, the length of time a leak could go undetected, and the radiological impact to the unrestricted environment?

5.4 Section 5.3, Pool Coolant System.

a. Propose TS wording for surveillance of gamma-emitting isotopes in the reactor pool water or justify why it is not needed.

b. Propose TS wording for pool water chemistry (pH and conductivity) and an appropriate surveillance frequency or justify why it is not needed.

5.5 Section 5.5.6, Reactor Coolant Cleanup System, General Description. The statement is made that any malfunctions or leaks in the reactor coolant cleanup system will not lead to radiation exposure to personnel or releases to the environment that exceed the regulatory requirements or facility ALARA Program guidelines. Discuss the bases for reaching this conclusion.

5.6 Section 5.7.1, Nitrogen-16 Control System, Introduction. The statement is made that the design of the nitrogen-16 control system provides reasonable assurance that the system will not interfere with reactor cooling, cause an uncontrolled loss or release of primary coolant, to prevent safe reactor shutdown. Explain the bases for reaching this conclusion.

CHAPTER 6 6.1 Section 6.2.2, Reactor Containment Building, Pages 6-2 to 6-5. Describe the containment building isolation set points.

6.2 Section 6.3.8, Anti-Siphon System, Design Analysis. The analysis of the system using RELAP5 assumes the anti-siphon valves will open 85 msec after the primary coolant rupture occurs. Discuss the basis for this assumption. Provide additional TS wording on anti-siphon valve performance or justify why this is not needed.

CHAPTER 7 7.1 Section 7.2, Design of Instrumentation and Control System. In Section 3.1.2, Overall Requirements. For Criterion 3: Fire Protection, the SAR states that fire protection is not required to accomplish a safe shut down of the reactor. Discuss the basis for this statement.

7.2 Section 7.7, Table 7-8, Reactor Scrams, and TS 3.3, Reactor Safety System. The scrams listed in the table do not appear to match the supporting TS. Discuss TSs for scrams initiated by building plenum and bridge high activity and low primary heat exchanger (HX) differential pressure or justify why they are not needed.

7.3 Section 7.9.3.3, Surveillance. Propose TS wording to support the SAR statement that the Fuel Element Failure Monitoring Systems high radiation annuciator alarm is tested before reactor startup or justify why this is not needed.

CHAPTER 9 9.1 Section 9.3, Fire Protection and Programs. The SAR discusses why one fire exit from the containment building is acceptable according to National Fire Protection Association (NFPA) 101 Life Safety Code. Provide a reference for the 300 ft used in the evaluation and the issue date of NFPA 101 that supports these statements.

9.2 Please describe the areas of the facility that are under the jurisdiction of the reactor license. Section 9.5.2, Rooms, Spaces, and Equipment states that any room or area designated by MURR management and approved by Health Physics may be used for radioactive materials.

9.3 Section 9.5, Possession and Use of Byproduct, Source, and Special Nuclear Material.

a. Describe how licensed radioactive material is kept segregated between the broad-scope materials license and the reactor license. Also, describe the process for transferring materials between the licenses.

b. It is unclear from your application if the current license possession limits are to be carried over to the renewal license? Please state and justify the license material possession limits.

9.4 Section 9.1.3, Surveillance. Propose TS wording to support the SAR statement, Fan Failure Alarm Panel is periodically tested for operability, or justify why this is not needed.

9.5 Section 9.8, Pool Skimmer System. Discuss if loss of pool water caused by a failure in the pool skimmer system could effect reactor operations and personnel.

9.6 Section 9.14, Compressed Air System. The SAR states that there are no functions or malfunctions of the compressed air system that could initiate a reactor accident, prevent safe reactor shutdown, or initiate the uncontrolled release of radioactive material.

Discuss the basis for this statement.

CHAPTER 10 10.1 Section 10.3.2.3, General Requirements. Discuss how the TSs protect against the production of unacceptable levels ozone from liquid nitrogen.

10.2 Section 10.3.3.2, Description. The SAR discusses a rod withdrawal inhibit if the thermal column door is in the open position. Discuss the actions that occur if the door is opened with the reactor in operation.

10.3 Section 10.4, Experiment Review.

a. This section of the SAR lists criteria that are considered during review and approval of reactor experiments. Explain the relationship between these criteria and the TSs.

b. The SAR states that the Reactor Safety Subcommittee and the Reactor Advisory Committee are involved in the review process if required. Please discuss when these committees would be involved in the review process.

10.4 Section 10.4.2, Reactor Utilization Request. The SAR states that fueled experiments are an exception to the experiment failure analysis. Provide a discussion of the analysis performed for the failure of fueled experiments.

10.5 Section 10.3.1, Center Test Hole. Related to license amendment request of August 6, 2009.

a. Describe any occurrences where the center test hole canister failed to properly latch or when a canister was removed from the reactor when not intended.

b. Describe any changes to procedures for sample handling or irradiation in the center test hole necessitated by the installation of the flux-trap irradiations reactivity safety trip (FIRST) system.

c. Describe the possible failure modes of the switches used on the FIRST system.

d. The SAR states that the cable jacket will display good radiation resistance for about 10 years of operation. Discuss the consequences of any potential failure of the FIRST cabling.

e. Discuss the proposed methodology that will be used to determine sample reactivity when the FIRST system is in use.

f. Provide a discussion of how the design of the FIRST system meets the requirements of IEEE-279.

g. There does not appear to be a surveillance requirement to help ensure proper operation of the FIRST system. Propose and justify an appropriate surveillance requirement for the FIRST system or explain why a surveillance is not needed.

CHAPTER 11 11.1 Section 11.1.1.1.1, Argon-41 from the Pneumatic Tube System. Discuss any potential failures of this system that could release Argon-41 to the containment or laboratory building and the potential radiological consequence to facility staff.

11.2 Section 11.1.2.5, Radiation Protection Training. The SAR states that Class I training is for materials under the MURR broad scope materials license. Please discuss training for persons involved with material under the MURR reactor license.

11.3 Section 11.1.5.6.1, Restricted Area. The SAR and Table 11-18 indicate that film badges are used for area radiation monitoring throughout the facility. Confirm that film badges are being used or provide a description of the dosimetry currently in use.

11.4 Section 11.1.1.1.3, Argon-41 Release to the Unrestricted Area. Table 11-2 appears to have a typographical error in the title indicating that the doses were calculated at 50 meters when the text of Chapter 11 and Appendix B indicate the doses were calculated at 150 meters. Confirm the appropriate title for the table.

11.5 Section 11.1.3, ALARA Program. Table 11-11 lists Investigation Levels I and II, but provides no detail regarding the significance of this distinction. Discuss the significance of Level I and II regarding the investigative rigor applied to personnel exposures.

11.6 Section 11.1.2, Radiation Protection Program. Provide typical staffing of the Reactor Health Physics Branch.

11.7 Section 11.1.2.4, Radiation Safety Officer. Clarify the relationship between the Radiation Safety Officer and the Reactor Health Physics Manager.

11.8 Section 11.2, Radioactive Waste Management. Please discuss continued access to solid waste disposal sites for the 20-year period of the license renewal. Please discuss planning for material that may not have a disposal path 11.9 Section 11.2.1, Radioactive Waste Management Program. This section of the SAR states the responsibility for safe disposal of radioactive waste from materials licensed under the MURR Broad Scope Material License but not the reactor license. Please address.

CHAPTER 12 12.1 Section 12.1.2, Responsibility, Figure 12-1 and TS Figure 6.0. SAR Figure 12-1 and TS Figure 6.0 indicate that the Reactor Health Physics Manager reports to the MURR Director. However, Section 12.1.2.4 of the MURR SAR indicates that the Reactor Health Physics Manager reports to the Associate Director, Regulatory Assurance Group.

Please clarify the reporting relationship for the Reactor Health Physics Manager and list the responsibilities of the Associate Director, Regulatory Assurance Group and show the position on the organizational chart. Show the Chief Operating Officer position on the organizational chart.

12.2 Section 12.1.3, Staffing and TS 6.1, Organization.

a. The minimum staffing during reactor operation includes an individual knowledgeable of the facility. Please describe the extent of this knowledge and the actions this person should be prepared to take in an emergency.

b. It appears that during operation or when the reactor is not considered secure, that the minimum staffing requirement can be one reactor operator. Please discuss how you will meet the requirements of 10 CFR 50.54(m)(1) for senior reactor operator coverage and actions for which the presence of a senior reactor operator at the facility is required.

12.3 Section 12.1.4, Selection and Training of Personnel. The SAR refers to reference 12.2, ANSI/ANS 15.4, Selection and Training of Personnel for Research Reactors, 1988.

This standard was revised in 2007. Please address the use of the revised standard.

12.4 Section 12.2, Review and Audit Activities and TS 6.2, Review and Audit.

a. The information in the SAR and proposed TS should, at a minimum, meet the recommendations of ANSI/ANS 15.1, "The Development of Technical Specifications for Research Reactors," 2007 (ANS 15.1). Please discuss the minimum number of members of the Reactor Advisory Committee (RAC) and

limitations on members of the MURR staff serving as members and meeting quorum requirements.

b. Please discuss the dissemination, review and approval of committee minutes in a timely manner.

c. Please discuss how your proposed review functions meet the recommendations of ANS 15.1. There are no audit functions listed for the RAC as recommended by ANS-15.1. Please propose audit requirements or justify why they are not needed.

12.5 Section 12.3, Procedures, and TS 6.3, Procedures. The information in the SAR and proposed TS should, at a minimum, meet the recommendations of ANS 15.1. Please discuss how the SAR and proposed TS meet the recommendations of ANS 15.1 or justify why meeting the recommendations is not needed.

12.6 Chapter 12, Conduct of Operations. The information in the SAR and proposed TS should, at a minimum, meet the recommendations of ANS 15.1. While Section 10.4 contains information on initial experiment review, the SAR and the proposed TS do not completely meet the recommendations of ANS 15.1. Please propose changes to the SAR and TS to meet the recommendations of ANS 15.1 or justify your TS and SAR as proposed.

12.7 Section 12.4, Reportable Events and Required Actions and TS 6.5 a., Reportable Events and Required Actions, Safety Limit Violation. Reporting events to the NRC under proposed TS 6.5 specifies notification of the NRC Project Manager for MURR rather than the 24-hour NRC Operations Center. Propose changes to the TS to have NRC notification go to the NRC Operations Center.

12.8 Section 12.4.3, Other Reportable Occurrences, TS 1.1, Abnormal Occurrences and TS 6.5.c, Other Reportable Occurrences.

a. The information in the SAR and proposed TS should, at a minimum, meet the recommendations of ANS 15.1. The TS lists actions to take in the event of Abnormal Occurrences. However, the text does not require the return to a safe condition or reactor shutdown. Please address this discrepancy.

b. Some of the proposed abnormal occurrence definitions are different than the recommendations of ANS 15.1. For example, TS 1.1 e only considers it an occurrence if radiation exposure limits are exceeded. The NRC needs to be informed promptly regarding any abnormal and significant degradation of the fuel, cladding, primary coolant boundary or containment boundary regardless of the radiological impact. TS 1.1 e. only considers the primary coolant boundary.

Provide justification for not also considering the pool system coolant boundary.

Propose appropriate changes to the SAR and TS to meet the recommendations of ANS 15.1 or justify your TS and SAR as proposed.

12.9 Section 12.5, Records and TS 6.4, Records. The information in the SAR and proposed TS should, at a minimum, meet the recommendations of ANS 15.1. In addition to the

record recommendations of ANS 15.1, 10 CFR 50.36 requires reports of violations of safety limits, limiting safety system settings, and limiting conditions for operation to be retained until the Commission terminates the facility license. Propose appropriate changes to the SAR and TS to meet the recommendations of ANS 15.1 and the requirements of 10 CFR 50.36, or justify your TS and SAR as proposed.

CHAPTER 13 13.1 Section 13.2.1.2, Accident Analysis and Consequences.

a Page 13-3. The mass of Al used here is 436 g. Explain the difference in mass numbers between this analysis and the containment pressure analysis of section 6.2.2.2. (Chapter 6 states the core contains 29.4 kg of aluminum while Chapter 13 states 33.56 kg).

b Page 13-4. Provide rationale for not including semi-volatiles such as Cs and Sr in the analysis.

c Page 13-8 and 13-9. Please clarify the units on the Air Submersion Dose conversion. Is it Sv Bq-1 m3 sec-1, not Sv/Bq-m3 as it is indicated on page 13-8 and Sv/Bq-sec-m3 on page 13-9?

d Page 13-13. Please clarify how the concentrations are derived. Should they be 25 percent of the average concentrations listed on page 13-8, not 25 percent of those listed on page 13-6?

13.5 Section 13.2.3.2, Accident analysis, Page 13-29. Provide a reference for the 900 degree F fuel blister temperature.

13.7 Section 13.2.4.2, page 13-55. Explain why assuming longer rather than shorter close times for the isolation valves is a conservative assumption for this accident.

13.8 Section 13.2.5.2.1, Damage to a Fuel Element Due to Mishandling. Please discuss the steps taken to minimize the possibility of damage to the reactor or fuel while moving the shipping cask into or out of the reactor pool.

13.10 Section 13.2.9.5, Failure of the Neutron Startup Source. The SAR discusses detection sensitivity for a leaking neutron source. It appears that the measurement is conducted weekly. If the source failed between measurements, what level of antimony could build up in the pool coolant without detection?

CHAPTER 14, TECHNICAL SPECIFICATIONS 14.1 Section 14.2, Format and Content and Introduction section of TSs. The SAR states that Section 5 of the TSs, Design Features, only contains specifications. The regulations in 10 CFR 50.36(a)(1) requires bases for design features TSs. Please amend your proposed design feature TSs to include bases.

CHAPTER 16, OTHER LICENSE CONSIDERATIONS 16.2 Section 16.1.2, Primary Coolant System Pressure Boundary. Your estimates in this section are based on a license term until 2026. Please update assuming a license term of 2030.

APPENDIX A, TECHNICAL SPECIFICATIONS The regulations in 10 CFR 50.36, Technical Specifications, contains the requirements for proposed TS submitted as part of a license application. Important general requirements in 10 CFR 50.36 include:

...Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section (10 CFR 50.36).

...summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

...technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to § 50.34. The Commission may include such additional technical specifications as the Commission finds appropriate.

ANSI/ANS-15.1, 2007, Development of Technical Specifications for Research Reactors, and NUREG 1537 (Part 1 and 2), Guidance for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, provide guidance for satisfying the requirements of 10 CFR 50.36. The TSs should include all parameters necessary to define the safe envelope for operations of the facility and should be internally consistent.

A.1 Definitions, General. The information in the SAR and proposed TS should, at a minimum, meet the recommendations of ANS 15.1. Definitions for channel calibration, channel check, measured value, operating, protective action, reactivity worth of experiment, reference core condition, research reactor, research reactor facility, shall, should and may, and unscheduled shutdown are missing from the proposed TSs.

Please justify not needing these definitions or add them to the proposed TSs.

A.2 Definition 1.2, Calibration or Testing Interval. The intervals given in this definition are for operational flexibility and are not long term frequencies. Please revise the definition to clarify this or justify the use of the proposed frequencies for the long-term.

A.3 Definition 1.5, Excess Reactivity. Why is the regulating rod not included in the definition of excess reactivity?

A.4 Definition 1.6, Exclusion Area. Explain the relationship between the exclusion area and the area under the reactor license.

A.5 Definition 1.7, Experiment. The definition provided differs from that given in ANS 15.1.

In TS 1.7 a. please explain what is considered significant radiation. In TS 1.7 b. an operation to monitor reactor parameters is an experiment. The reactor operators in the control room are monitoring reactor parameters during reactor operation. Is routine reactor operation considered an experiment? Please clarify.

A.6 Definition 1.11, Irradiated Fuel. Discuss the basis for this definition as used in TS 3.5 a. 2 and 3.8 e. Confirm that this definition is not related to the transportation of fuel.

A.7 Definition 1.13, Movable Experiment. ANS 15.1 includes moving experiments in or near the core as part of the definition. Justify why this part of the definition is not needed at MURR or propose a definition consistent with ANS 15.1.

A.8 Definition 1.21, Reactor Safety System. The definition provided differs from ANS 15.1.

Why are systems that provide information for initiation of manual protective action not considered reactor safety systems?

A.9 Definition 1.23, Reactor Secured. The definition provided differs from ANS 15.1. In TS 1.23 (1) why are optimum available conditions of moderation and reflection not considered. In TS 1.23 (2) a. why is the regulating rod not included. Explain the use of dummy load connectors in this definition. There appears to be no limits on experiment movement with the reactor secured. Please justify.

A.10 Definitions 1.25, Regulating Blade (Rod), and 1.30, Shim Blade (Rod). Please clarify if the control rods have in-run capability referred to in TS 3.4 c.

A.11 Definition 1.28, Secured Experiment. The definition provided differs from ANS 15.1.

Please discuss.

A.12 TS 3.1 e., Reactivity Limitations. Shutdown margin measurements are normally conducted with the core and experimental configuration in their most reactive conditions.

Please discuss and justify measurement conditions for the reactor (e.g., reference core condition) and experiments (e.g., moveable experiments in their most reactive condition) to help ensure that changes in core or experiment conditions will not lead to violation of shutdown margin requirements (See RAI 4.14 c.).

A.13 TS 3.1 f., Reactivity Limitations. The TS states that core excess reactivity is limited above cold, clean critical. Define what is considered cold, clean critical (Also see RAI A.1 for definition of reference core condition, and A.12).

A.14 TS 3.1 g.,3.1 j. and 3.1 k., Reactivity Limitations. Please explain why these experiment reactivity values are not the absolute values.

A.15 TS 3.1, Reactivity Limitations. The TS does not contain a limit on the sum of the absolute values of the reactivity worths of all experiments. Propose appropriate TS wording or justify why it is not needed.

A.16 TS 3.1 provides Limiting Conditions for Operation (LCOs) for Reactivity Limitations.

However, there are no corresponding Surveillance Requirements for these LCOs.

Propose appropriate TS Surveillance Requirements for TS 3.1 or justify why none are needed. There appears to be no requirement for the periodic determination of the worth of control blades. Propose and justify a TS or discuss why no TS is needed.

A.17 TS 3.2.b requires that above 100 kW, the maximum distance between the highest and lowest shim blade is one inch. However, there is no Surveillance Requirement for this LCO. Propose an appropriate TS Surveillance Requirement for TS 3.2.b or justify why one is not needed.

A.18 TS 3.3, Reactor Safety System. Please explain note 1 to TS 3.3 a. The note and trip set points in the table appear to define forced convection operation which is not consistent with the bases for the safety limits and LSSS for Mode III. Please explain the significance of the natural convection flange and pressure vessel cover being removed.

Why is this not stated as a separate limiting condition for operation? Please explain the basis of the reactor being subcritical by 0.015 k/k and the need for the channels required by the TS.

A.19 TS 3.3, Reactor Safety System. The pool low water level scram is only required for Mode III operation. The bases state the purpose of the scram is to control radiation levels above the reactor pool with Section 11.1.5.1.1 of the SAR stating that a minimum water depth of 23.6 feet is required over the fuel region at an operating power level of 10 MW. Please explain why this scram is not needed for all modes of operation. It is not clear what the baseline of the 23 foot scram set point is. Where is the 23 feet measured from and how does it relate to the minimum water depth discussed in the SAR?

A.20 TS 3.3, Reactor Safety System. The number of instrumentation channels required for primary coolant flow appears to imply a minimum of two instrument channels per pump without specifying a pump must be operating. Propose TS wording to clarify the intent of the TS is to have two instrument channels per operating pump.

A.21 TS 3.3, Reactor Safety System. The set point for the pool coolant flow scram is a minimum flow rate of 850 gpm. However, the bases discuss a flow rate of 425 gpm.

Please explain this difference.

A.22 TS 3.4, Reactor Instrumentation. There does not appear to be requirements for any interlocks in the TSs. Please discuss the interlocks provided by the reactor I&C system and proposed TSs or explain why they do not need to be TS requirements.

A.23 TS 3.4 a., Reactor Instrumentation.

a. The TSs require three radiation monitors be operable. However, the SAR discusses additional monitors such as beamport floor, south, west, east and north wall, fuel storage vault, mechanical equipment room, fuel element failure monitoring system, and secondary coolant monitoring system. Please proposed TS requirements for these radiation monitors or explain why these monitors need not be TS requirements.

b. The SAR refers to two bridge monitors while the TS require one. Clarify which monitor is required by the TSs and justify why the other monitor isnt required.

c. The TS require one exhaust plenum monitor while the SAR refers to two monitors. Clarify which monitor is required by the TSs and justify why the other monitor isnt required.

d. The TS refer to the off-gas radiation monitor while the SAR discusses the off-gas radiation monitor system which contains three-channels. Please clarify if this is the same instrumentation.

e. Please provide maximum alarm set points and the bases for any TS required radiation monitor not covered in TS 3.7 or justify why TSs are not needed.

f. The off-gas radiation monitor may be out of service for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. In the unlikely event of a release, discuss what other radiation monitoring equipment would alert the staff to the condition.

A.24 TS 3.4 b., Reactor Instrumentation. Please describe the instrumentation considered sufficient to assure that the stated limits are not exceeded and why the instrumentation is considered sufficient. Justify why this instrumentation should not be a TS requirement.

A.25 TS 3.4 c., Reactor Instrumentation. Please explain note 1 to TS 3.4 c. The note is inconsistent with Mode III operation which is limited to 50 kW(t). Please explain the significance of the natural convection flange and pressure vessel cover being removed.

Why is this not stated as a separate limiting condition for operation? Please explain the basis of the reactor being subcritical by 0.015 k/k and the channel being required by the TS.

A.26 TS 3.3 and 3.4. Please discuss any reactor safety and control system permitted bypasses and the condition under which the system may be bypassed. Please add permitted bypassing to the TSs or justify why a TS requirement on bypasses is not needed.

A.27 TS 3.5 a. (2), Reactor Containment Building. Please provide a technical basis for not needing containment integrity when handling fuel with a decay time of greater than sixty days.

A.28 TS 3.5 b., Reactor Containment Building. Please provide a technical basis for containment isolation at 10 times above previously established levels.

A.29 TS 3.5, Reactor Containment Building. ANS 15.1, Section 3.5 suggests requirements for ventilation system TSs. Please propose TSs for ventilation during reactor operation or justify why such TSs are not needed.

A.30 TS 3.6 c., Experiments. This TS controls occupational exposures. Please discuss limits on exposure for members of the public from experiment failure.

A.31 TS 3.6 d., Experiments. Please define what constitutes an explosive material.

A.32 TS 3.6 l., Experiments. Please provide examples of controls on the use or exclusion of corrosive, flammable and toxic materials and explain how the controls help ensure the safety of the reactor.

A.33 TS 3.7, Facility Airborne Effluents. Please discuss the instantaneous release concentration limit of 3,500 Air Effluent Concentration (AEC). At what time duration do the short burst releases impact doses?

A.34 TS 3.8 c., Reactor Fuel. Please provide a technical basis for operation with less than eight fuel elements at power levels up to 100 Watts(t).

A.35 TS 3.9, Reactor Coolant System. ANS 15.1 Section 3.3 suggests TS limits on water chemistry requirements. Please propose TSs on primary and pool water chemistry or justify why they are not needed (See RAIs 5.2 a and 5.4 b).

A.36 TS 3.9, Reactor Coolant System. ANS 15.1 Section 3.3 suggests TS limits on secondary coolant activity. Please propose TSs on secondary coolant activity or justify why they are not needed.

A.37 TS 3.9 b. (1), Reactor Coolant System. Please discuss the action set point for the fuel element failure monitor. Please add the set point to the TSs or justify why the set point is not needed in the TSs. Please discuss the basis for the four-hour interval between analyses.

A.38 TS 3.9 c., Reactor Coolant System. Please discuss the basis for limiting the Iodine-131 concentration of the primary coolant to 5 x 10-3 Ci/ml.

A.39 TS 4.0, Surveillance Requirements. There appears to be no requirement for surveillance testing on any TS required system after replacement, repair or modification to declare the system operable and returned to service. Please propose and justify a TS or justify why such a specification is not needed.

A.40 TS 4.4, Reactor Instrumentation. This TS refers to reactor instrumentation whose LCOs are detailed in TS 3.4. The wording of the TS does not make it clear if TS 4.4 also encompasses TS 3.3, Reactor Safety System. Propose appropriate TS wording to ensure that TS 3.3 is covered by surveillance requirements or justify why a TS is not needed. What is the basis for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit? For example, does it apply if the reactor has been secured?

A.41 TS 4.0, Surveillance Requirements. ANS-15.1 suggests a surveillance requirement for thermal power verification. Provide appropriate TS wording to incorporate a surveillance requirement for thermal power verification or justify why it is not needed.

A.42 TS 4.0, Surveillance Requirements. The TSs do not appear to contain surveillance requirements for interlocks. Provide appropriate TS wording to incorporate surveillance requirements for interlocks or justify why they are not needed.

A.43 TS 4.1, Containment System. TS 1.17 contains the definition of containment integrity.

Discuss how TS 4.1 assures that containment integrity will be maintained when needed.

A.44 TS 4.0, Surveillance Requirements. TS 3.10.b requires the operability of the Primary Coolant Makeup Water System as well as a connection to a minimum of 2,000 gallons of primary grade water. However, there is no surveillance requirement for this LCO.

Propose an appropriate TS surveillance requirement for TS 3.10.b or justify why one is not needed.

A.45 TS 4.0, Surveillance Requirements. TS 3.9.a (3) requires the operability of the incore convection coolant system. However, there is no surveillance requirement for this LCO.

Propose an appropriate TS surveillance requirement for TS 3.9.a (3) or justify why no TS is needed.

A.46 TS 4.2, Reactor Coolant System. TS 3.9 a. (1) requires the anti-siphon system to be operable. TS 4.2 b. contains testing requirements for the anti-siphon valves. Clarify how this surveillance confirms operability of the system. Provide a justification why the primary coolant isolation valves and the anti-siphon isolation valves (system) need not be operable if the reactor is shut down.

A.47 TS 4.0, Surveillance Requirements. TS 3.8 provides limitations on use and storage of reactor fuel at MURR. However, there are no surveillance requirements for this LCO.

Propose appropriate surveillance requirements for TS 3.8 or justify why they are not needed.

A.48 TS 5.1, Site Description. Clearly describe the area under the authority of the reactor license.

A.49 TS 5.3 e., Reactor Coolant Systems. Explain what constructed principally of aluminum alloys or stainless steel means. Explain the use of exception b. and the evaluation that would be used to determine what other materials would be acceptable.

A.50 TS 5.3 i., Reactor Coolant Systems. Describe the natural convection flow path consistent with the evaluation in the SAR.

A.51 TS 5.4 b., Reactor Core and Fuel. The term fully enriched may not be clear. Provide a nominal enrichment level for the reactor fuel.

A.52 TS 5.4 d., Reactor Core and Fuel. Given that fuel can also be stored in racks, clarify this TS.

A.53 TS 6.1, Organization. ANS 15.1, Section 6.1, Organization and Section 6.1.2, Responsibility, recommends TSs contain information on functions, assignments, and responsibilities of key organization staff. The standard also recommends that TSs contain information that individuals at the various management levels, in addition to having responsibility for the policies and operation of the reactor facility, shall be responsible for safeguarding the public and facility personnel from undue radiation

exposures and for adhering to all requirements of the operating license and technical specifications. In all instances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications.

Please include requirements similar to these in the TSs or justify why they are not needed.

A.54 TS 6.1, Organization. ANS 15.1, Section 6.1.3 (2), Staffing, recommends TSs contain a requirement for control room contact information. Propose appropriate TS wording to include requirements similar to these in the TSs or justify why they are not needed.

A.55 TS 6.0, Administrative Controls. ANS 15.1, Section 6.3, Radiation Safety, recommends TSs for the implementation of radiation safety. Propose appropriate TS wording to include requirements similar to these in the TSs or justify why they are not needed.

APPENDIX B B.2 Section B.6 compares the calculated offsite dose to the annual average radiation dose of 360 mrem for the US population as reported in NCRP Report 94. This report has been superseded by NCRP Report 160 (2009). Further, the 360 mrem annual dose includes several components that are regional in nature (i.e., radon dose from granitic formations) and may not be appropriate for comparison with MURR conditions. Justify the use of 360 mrem/yr for comparison purposes or provide other comparison data.

APPENDIX C C.1 Section C.1, Introduction. Please describe any benchmarking that has been performed on the RELAP model and conclusions as to the accuracy of the models results.