ML100360411

University of Utah, Request for Additional Information, Renewal Application and Power Uprate Review

ML100360411
Person / Time
Site:	University of Utah
Issue date:	02/22/2010
From:	Geoffrey Wertz Research and Test Reactors Licensing Branch
To:	Jevremovic T Univ of Utah
Wertz G, NRR/DPR/PRTA, 434-326-1086
References
TAC ME1599
Download: ML100360411 (23)

Text

February 22, 2010 Dr. Tatjana Jevremovic Director, Utah Nuclear Engineering Program Joseph Merrill Engineering Building 50 Central Campus Drive, Room 2298 University of Utah Salt Lake City, UT 84112

SUBJECT:

UNIVERSITY OF UTAH - REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE RENEWAL AND POWER UPRATE APPLICATION REVIEW (TAC NO. ME1599)

Dear Dr. Jevremovic:

The U.S. Nuclear Regulatory Commission (NRC) is continuing the review of your application for renewal and power uprate of Facility Operating License No. R-126, for the University of Utah TRIGA Reactor, dated March 25, 2005, and superseded in its entirety by an Updated Safety Analysis Report, dated June 1, 2009 (a redacted version is available on the NRCs public website, www.nrc.gov, in the Agencywide Documents Access and Management System, Accession No. ML092090027).

During our review, questions have arisen for which we require additional information and clarification. Please provide responses to the enclosed request for additional information within 60 days after the date of this letter. In accordance with Title 10 of the Code of Federal Regulations Part 50.30(b), your response must be executed in a signed original document under oath or affirmation.

If you have any questions regarding this review, or need additional time to respond to this request, please contact me at 301-415-0893 or by electronic mail at: Geoffrey.wertz@nrc.gov.

Sincerely,

/RA By Kathryn M. Brock for/

Geoffrey Wertz, Project Manager Research and Test Reactors Branch A Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-407

Enclosure:

As Stated cc w/encl: See next page

University of Utah TRIGA Reactor Docket No. 50-407 cc:

Mayor of Salt Lake City 451 South State Room 306 Salt Lake City, UT 84111 Dr. Thomas Parks Vice President for Research 201 S. Presidents Circle, Room 210 University of Utah Salt Lake City, UT 84112 Ms. Karen Langley Director, University of Utah Radiological Health 100 OSH, University of Utah Salt Lake City, UT 84112 Dr. Cynthia Furse Associate Vice President for Research 201 Presidents Circle, Room 210 University of Utah Salt Lake City, UT 84112 Test, Research, and Training Reactor Newsletter Universities of Florida 202 Nuclear Sciences Center Gainesville, FL 32611 Director, Division of Radiation Control Dept. Of Environmental quality 168 North 1959 West P.O. Box 144850 Salt Lake City, UT 84114-4850

February 22, 2010 Dr. Tatjana Jevremovic Director, Utah Nuclear Engineering Program Joseph Merrill Engineering Building 50 Central Campus Drive, Room 2298 University of Utah Salt Lake City, UT 84112

SUBJECT:

UNIVERSITY OF UTAH - REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE RENEWAL AND POWER UPRATE APPLICATION REVIEW (TAC NO. ME1599)

Dear Dr. Jevremovic:

The U.S. Nuclear Regulatory Commission (NRC) is continuing the review of your application for renewal and power uprate of Facility Operating License No. R-126, for the University of Utah TRIGA Reactor, dated March 25, 2005, and superseded in its entirety by an Updated Safety Analysis Report, dated June 1, 2009 (a redacted version is available on the NRCs public website, www.nrc.gov, in the Agencywide Documents Access and Management System, Accession No. ML092090027).

During our review, questions have arisen for which we require additional information and clarification. Please provide responses to the enclosed request for additional information within 60 days after the date of this letter. In accordance with Title 10 of the Code of Federal Regulations Part 50.30(b), your response must be executed in a signed original document under oath or affirmation.

If you have any questions regarding this review, or need additional time to respond to this request, please contact me at 301-415-0893 or by electronic mail at: Geoffrey.wertz@nrc.gov.

Sincerely,

/RA By Kathryn M. Brock for/

Geoffrey Wertz, Project Manager Research and Test Reactors Branch A Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-407

Enclosure:

As Stated cc w/encl: See next page DISTRIBUTION:

PUBLIC DPR/PRT r/f RidsNrrDpr RidsNrrDprPrta RidsNrrDprPrtb GWertz, NRR GLappert, NRR ACCESSION NO.:ML100360411

• via e-mail NRR-088 Office PRTA:PM*

PRTA:LA PRTA:BC PRTA:PM*

Name GWertz GLappert KBrock GWertz Date 2/05/10 2/17/2010 2/22/10 2/22/10 OFFICIAL RECORD COPY

ENCLOSURE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION REGARDING THE RENEWAL OF FACILITY OPERATING LICENSE AND POWER UPRATE APPLICATION FOR THE UNIVERSITY OF UTAH TRIGA REACTOR LICENSE NO. R-126; DOCKET NO. 50-407 The U.S. Nuclear Regulatory Commission (NRC) is continuing the review of your application for renewal and power uprate of Facility Operating License No. R-126, for the University of Utah TRIGA Reactor (UUTR), as documented in the UUTR Safety Analysis Report (SAR), dated June 1, 2009 (a redacted version is available on the NRCs public website, www.nrc.gov, in the Agencywide Documents Access and Management System, Accession No. ML092090027). Our review is performed in accordance with the guidance provided in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors. During this review, we have identified areas needing additional information. Please provide responses to the following requests for additional information:

1.

NUREG-1537, Part 1, Section 1.2, Summary and Conclusions on Principal Safety Considerations requests a summary of the principle safety considerations for the UUTR.

The SAR did not provide this information. Please provide a summary of the principle safety considerations as described in NUREG-1537, Part 1, Section 1.2.

2.

NUREG-1537, Part 1, Section 1.3, General Description of the Facility requests a description of the reactors basic design features, operating characteristics, safety systems, and instrumentation and control and electrical systems. SAR Chapter 1.3 does not provide such information. Please provide the basic UUTR design features including the safety systems, their design features, and operating characteristics.

3.

NUREG-1537, Part 1, Section 1.3, General Description of the Facility requests a general description of any engineered safety features (ESFs). In SAR Chapter 1.3.4, the applicant describes a number of UUTR features, but does not specifically identify any ESFs. The heating, ventilation and air conditioning (HVAC) system is described in the SAR Chapter 6, Engineered Safety Features. Please provide a description of any ESFs at UUTR, including the classification system used to determine ESFs.

4.

NUREG-1537, Part 1, Section 1.4, Shared Facilities and Equipment requests the applicant to provide a description of any shared facility systems or equipment and whether the loss of any shared facility systems or equipment would lead to a loss of function that could lead to the uncontrolled release of radioactive material. SAR Chapter 1.4 does not discuss the consequences of the loss of shared facility systems or equipment. Please provide information identifying shared facility systems and equipment and discuss the reasons why any loss of function will not degrade UUTR safety features.

5.

NUREG-1537, Part 1, Section 1.5, Comparison With Similar Facilities requests applicants to describe pertinent similarities to other reactors in their design, including principal design parameters, reactor safety systems, engineered safety systems and instrumentation and control systems. A comparison was not found in the SAR. Please provide a comparison of UUTR to other reactor facilities so as to characterize the degree to which generic information or operational experience from other sites is applicable.

6.

NUREG-1537, Part 1, Section 1.8, Facility Modifications and History requests an overview of facility changes not requiring NRC approval under Title 10 of the Code of Federal Regulations (10 CFR) Part 50.59, and a complete facility history. The SAR did not provide a list of facility changes or history.

6.1.

Please provide a list of significant changes to the facility since it was originally licensed as described in the original SAR. Indicate which changes were reviewed against 10 CFR 50.59.

6.2.

Please provide a brief history of the facility including dates of significant events, such as the issuance of the construction permit and operating license and initial criticality.

6.3.

Please provide a list of fuel shipments into and out of the UUTR facility, as well as the delivery of components received from other reactor sites.

7.

NUREG-1537, Part 1, Section 2.1 Geography and Demography requests information regarding reactor location by longitude and latitude, and population distribution using the 1, 2, 4, 6, and 8 mile demarcations. SAR Chapter 2.2 did not contain this information.

Please provide information regarding reactor location and population distributions. Also, the SAR did not identify the closest temporary housing to the research reactor. Please provide the distance from Merrill Engineering Building to temporary housing in meters.

8.

NUREG-1537, Part 1, Section 2.2, Nearby Industrial, Transportation and Military Facilities requests information regarding descriptions of nearby industrial, transportation, and military facilities. The SAR did not provide a detailed description of roads, railroads, and the potential for hazards to the facility. Please provide information regarding roads, railroads, and the potential for hazards to the facility, and a site map showing distance scales.

9.

NUREG-1537, Part 1, Section 2.3, Meteorology requests information regarding historical weather phenomena and meteorological conditions. The SAR did not provide historical weather phenomena and meteorological conditions, including site meteorology in terms of air flow, temperature, atmospheric water vapor, precipitation, fog, atmospheric stability, and air quality. Please provide information pertaining to historical weather phenomena and meteorological conditions as noted.

10.

NUREG-1537, Part 1, Section 3.2, Meteorological Damage requests a description of the potential for meteorological damage due to wind, snow, and ice loads and the facilitys design features intended to cope with these conditions. The SAR did not provide this information. Please provide a description of the UUTR design features that provide protection from wind, snow, and ice loads including the use of any applicable of building codes and standards.

11.

NUREG-1537, Part 1, Section 3.3, Water Damage requests a description of the potential for flooding and to demonstrate that such flooding will not prevent the proper functioning of safety systems. SAR Chapter 3.3 states, even if flooding occurred, reactor safety would not be an issue since the core is located in a water pool. However, this does not address the potential for flooding to impact any of the facilitys systems as described in NUREG-1537, Part 1, Section 3.3. Please provide a description of the potential for flooding and any measures necessary to ensure the proper functioning of facility systems.

12.

NUREG-1537, Part 1, Section 3.5, Systems and Components requests the design bases for the systems and components required for safe operation or shutdown. The SAR Chapter 3.5 lists several systems, but does not describe their safety classification or design bases. Please provide the design basis for any systems required for safe operation or shutdown. Include electromechanical systems and components required for the safe operation of the UUTR, including responses to transients and potential accidents analyzed in the SAR, and technical considerations such as dynamic and static loads, number of duty cycles, vibration, wear, friction, strength of materials, and the effects of the operating environment including radiation and temperature.

13.

NUREG-1537, Part 1, Section 4.2.1, Reactor Fuel requests a detailed description of the fuel elements to be used in the reactor, including detailed design information and relevant references demonstrating that the design basis for non-power reactor fuel maintains fuel integrity under any conditions assumed in the safety analysis. The SAR should also include all information necessary to establish the limiting characteristics beyond which fuel integrity could be lost. The SAR Chapters 4.1.2 and 4.1.3, do not describe the fuel design basis as stated above, nor the differences in fuel meat length resulting from the use of standard TRIGA stainless steel (SS) and Aluminum (Al) clad fuel and the implications of these differences.

13.1. Please provide a corrected description of the fuel element physical description, drawings showing the primary components of the fuel elements, and a description of the design basis for the fuel.

13.2. In addition, provide information concerning mechanical forces and stresses, corrosion and erosion of cladding, hydraulic forces, thermal changes and temperature gradients, and internal pressures from fission products and the production of fission gas. Discuss the impact of radiation effects, including the maximum fission densities and fission rates that the fuel is designed to accommodate.

14.

NUREG-1537, Part 1, Section 4.2.1, Reactor Fuel requests a detailed description of the fuel elements to be used in the reactor. SAR Chapter 4.1.2, Moderated Fuel Elements, and Section 4.2.1, Moderated Fuel Elements-Chemical Reactions, states the following: In the pulsed mode of operation, a step insertion of 2.1% delta k/k ($3.00) could result in a reactor peak power of about 2000 MW with a prompt reactor period of 2.8 msec, an energy release of 26 MW-sec, and associated peak fuel temperatures below 6500C. And, The design basis limit (safety limit for reactor operation) is that the temperature of a stainless steel "high hydride" fuel element not exceeds 1000°C, and an aluminum clad "low hydride" full element shall not exceed 530°C under any conditions of operation. Furthermore, if the cladding should fail for other reasons, these same temperature limits will assure that any chemical reaction occurring with the fuel will be minor. Please provide the technical basis supporting these statements (i.e., references or analyses).

15.

NUREG-1537, Part 1, Section 4.2.2, Control Rods requests a detailed description of the control rods used to safely control and shut down the reactor. The SAR did not provide a description of the control rods as outlined in NUREG-1537, Part 1, Section 4.2.2. Please provide this description.

16.

NUREG-1537, Part 1, Section 4.2.3, Neutron Moderator and Reflector requests a description of reflectors and moderators designed into the core, including any special features. The SAR only provided a brief discussion of moderator and reflector fuel elements; there is no information pertaining to the naturally circulating water which represents the primary moderator/reflector. Please provide a revised description of the neutron moderator and reflector.

17.

NUREG-1537, Part 1, Section 4.2.4, Neutron Startup Source requests a description of the neutron source used for reactor startup. SAR Chapter 4.1.5 only provides a description of the source holder. Please provide a description of the neutron source in use at the UUTR, including the neutron strength and spectrum, source type and materials, burnup and decay lifetime, and regeneration characteristics. Include any design features that ensure the function, integrity, and availability of the source.

18.

NUREG-1537, Part 1, Section 4.2.5, Core Support Structure requests a description of the structural performance of the core support structure under all reasonable conditions, including the design-basis operational analysis and safety considerations for each reactor component. SAR Chapter 4.1.1 does not provide this information.

18.1. Please provide information pertaining to the adequacy of the core support structure under flood and empty tank conditions to safely support all required components.

18.2. As referenced in SAR Chapter 4.1.1, Grid Plate, Figure 4.1.1 could not be located. Please provide a photo or schematic of the UUTR grid plate showing the appropriate dimensions.

19.

NUREG-1537, Part 1, Section 4.3, Reactor Tank or Pool requests a description of the reactor tank and associated components including how those components will perform their intended functions free from any problems associated with chemical interactions, penetration and weld failures that could lead to the loss of coolant, and how Technical Specifications (TS) impose the appropriate limiting conditions. The SAR did not provide an analysis or discuss the potential loss of water from a failure of penetrations such as the beam ports or the consequences of such a failure on any accident analyses supplied. Please provide information regarding a loss of water through any and all penetrations, preventative measures, design bases and features used to prevent such occurrences, and if any TS are applicable to these occurrences.

20.

NUREG-1537, Part 1, Section 4.4, Biological Shield requests a description the biological shield employed to ensure dose limits in conformance with 10 CFR Part 20.

The SAR did not provide information for the concrete, tank, or pool water as biological shields. Please provide a description of the biological shielding in use at the UUTR.

21.

NUREG-1537, Part 1, Section 4.5.1, Normal Operating Conditions requests a description of the normal operating conditions for the complete, operable reactor core, including controls on empty spaces (e.g. unoccupied fuel locations) that could be used to add reactivity to the core, and the full spectrum of control rod worths, kinetics parameters, and core excess reactivities for all planned or possible configurations. The SAR description was insufficient in these areas.

21.1. Please provide the core loading for the 250 kW condition including control rod worths, kinetics parameters, and core excess reactivities for all planned or possible configurations.

21.2. Figure 4.1, University of Utah Reactor Core (top view), was not legible, and not clear whether this configuration was for 100 kW or 250 kW. Please provide a legible and clear copy.

22.

NUREG-1537, Part 1, Section 4.5.2, Reactor Core Physics Parameters requests a description of the reactor core physics information that characterizes the reactors performance. This information includes: methods used to neutronically characterize the UUTR; uncertainties required to apply calculated results to UUTR operations; methods used to calculate kinetics parameters; coefficients of reactivity that are applicable to the UUTR; comparisons with measurements to demonstrate the effectiveness of the methods employed; and, changes in reactivity coefficients that result from changes to core configurations. The SAR did not provide this information. Please provide information regarding methods, uncertainties, comparisons, and all required technical parameters.

23.

NUREG-1537, Part 1, Section 4.5.3, Operating Limits requests a description of the operating limits, including those nuclear design features necessary to ensure safe operations and shutdown. These include: temperature coefficients of reactivity, void coefficients, Xe-Sm worths, power coefficients (if not otherwise accounted for), and the influence of experiments; minimum control rod worths and stuck rod worths for all allowed core conditions; transient analysis of an uncontrolled rod withdrawal; shutdown margin calculations for limiting core conditions; and, TS implemented to ensure safe operations. The SAR did not provide sufficient information. Please provide information regarding methods, uncertainties, comparisons, and all required technical parameters as identified in the guidance. This information should be specific to the UUTR and commensurate with the methods described in the response to RAI question 22.

24.

NUREG-1537, Part 1, Section 4.6, Thermal-Hydraulic Design requests a description of operating limits on cooling conditions necessary to ensure that fuel integrity will not be lost under any reactor conditions, including accidents. NUREG-1537, Appendix 14.1, Section 2.1.2, indicates an acceptable limit of two (2) for the departure from nucleate boiling ratio (DNBR), and states that no flow instability should be able to contribute to a loss of fuel cooling under any conditions. SAR Appendix A did not provide sufficient information. Please supply information regarding methods, uncertainties, and results of the departure from nucleate boiling (DNB) analysis showing that the proposed safety limits will be maintained during any and all UUTR operating conditions (including those pertaining to rod position).

25.

NUREG-1537, Part 1, Section 5.1, Summary Description requests a description of the type of secondary coolant system, if present, and the method of heat disposal to the environment. The SAR did not provide this information is sufficient detail. Please provide information regarding the ultimate heat sink for UUTR.

26.

NUREG-1537, Part 1, Section 5.2, Primary Coolant System requests a description of the primary coolant system including information to substantiate the removal of heat from the fuel during reactor operation and decay heat during reactor shutdown. The SAR Chapters 5.1 and 5.2 did not provide an analysis showing the adequacy of the primary system to perform this task. Please provide an analysis showing the adequacy of the primary system to cool the reactor under all anticipated modes of operation.

27.

NUREG-1537, Part 1, Section 5.3, Secondary Coolant System requests a description and analysis of the secondary coolant system. The information in SAR Chapter 5.2 was incomplete. Please provide a schematic drawing of the primary and secondary cooling systems showing all possible modes of operation. Discuss the performance of the secondary cooling system as required by NUREG-1537, Part 1, Section 5.3. Provide the design basis for the secondary cooling system including considerations for concrete dryout, resin bed performance, and evaporation limits. Describe how the appropriate design bases for the system are included in the TS. Discuss the potential for heat exchanger leakage as a source of contamination.

28.

NUREG-1537, Part 1, Section 5.3, Secondary Coolant System requests a description and analysis of the secondary coolant system. Section 4.4 of your Environmental Report dated January 7, 2005, states, The heat exchanger is charged with R-134a.

Heat from the R-134a is transferred to an independent water source that is released to the sanitary sewer system. Based on a site visit, it was observed that the heat from the R-134a was transferred to ambient air. Please confirm how the heat from the R-134a is transferred.

29.

NUREG-1537, Part 1, Section 5.4, Primary Coolant Cleanup System requests a description of how the exposure and release of radioactivity do not exceed the requirements of 10 CFR Part 20 and are consistent with the facilitys as low as reasonably achievable (ALARA) program. The SAR information provided was insufficient. Please provide information describing how the cleanup system design incorporates the commitments of the ALARA Program.

30.

NUREG-1537, Part 1, Section 5.5, Primary Coolant Makeup Water Systems requests a description indicating how the makeup water system plan should include provisions for recording the use of makeup water to detect changes that indicate leakage or other malfunction of the primary coolant system. The information provided in the SAR Chapter 5.5 was insufficient. Please provide information describing any provisions to detect abnormal leakage in the primary system.

31.

NUREG-1537, Part 1, Section 5.6, Nitrogen-16 Control Systems requests a description of the nitrogen control system employed and information indicating that the reduction in personnel exposure to nitrogen-16 (N16) should be consistent with the N16 analyses in the SAR Chapter 11. In addition, the total dose shall not exceed the requirements of 10 CFR Part 20 and should be consistent with the facility ALARA program. The SAR, Appendix B.4, provides a method for calculating N16 concentrations but does not address dose. The SAR does not describe the N16 control system as a design element of the facility. Please provide information describing the N16 dose calculations consistent with the guidance cited above and a description of the N16 control system.

32.

NUREG-1537, Part 1, Section 6, Engineered Safety Features requests a description of any active or passive ESFs designed to mitigate the consequences of accidents. SAR Chapter 6 describes the HVAC system as a confinement system, but does not mention or describe the required system performance in the safety analysis. Additionally, there are no diagrams or explanations of differences in the flow paths between normal and emergency conditions. Please provide a description and a diagram of the confinement system during all operational conditions. Clarify the classification of the HVAC system.

33.

NUREG-1537, Part 1, Section 9, Auxiliary Systems requests as description of the fuel testing and surveillance requirements. SAR Chapter 9.2.3 discusses biannual fuel inspections, and Table 12.3 discusses biennial fuel inspections. Please provide clarification as to what the fuel inspection intervals are and how they are imposed.

34.

NUREG-1537, Part 1, Section 9, Auxiliary Systems requests a description of any auxiliary systems that are not fully described in other sections, but are important to the safe operation and shutdown of the reactor and to the protection of the health and safety of the public, the facility staff, and the environment.

The SAR Chapter 9.2.2.1 discusses the fuel storage racks within the reactor vessel and states: Within a fuel storage rack, control of spacing is not actually required to limit the effective multiplication factor of the array (Keff). The in-tank fuel storage racks are configured such that criticality is not possible. Furthermore, 2 racks of 8.5 w% fuel stored back to back are subcritical (i.e., K <= 0.74 for twice the U-235 mass). In the unlikely event of loss of reactor tank coolant water, the loss of the water moderator would increase the safety margin by reducing the Keff. The in-tank fuel storage racks are made of polyethylene and are designed to withstand a UBC Zone 3 earthquake, when fully loaded. However, there is no supporting analysis, or reference for the supporting analysis, of the storage racks that provide the basis for the statement that the effective multiplication factor will maintain less than a Keff of 0.74 for all storage orientations. Please provide the technical basis (i.e., supporting analysis or a reference for the analysis) for the statement that the racks are designed to withstand a Uniform Building Code Zone 3 earthquake when fully loaded.

35.

NUREG-1537, Part 1, Section 9, Auxiliary Systems requests a description of any auxiliary systems that are not fully described in other sections, but are important to the safe operation and shutdown of the reactor and to the protection of the health and safety of the public, the facility staff, and the environment. SAR Chapter 9.2.2.2 states that all storage pit material (liners, racks, plug casing, and pipes) that may contact either the fuel elements or the pit water are fabricated from aluminum or 304 stainless steel. However, SAR Chapter 9.2.2.1 states that the fuel storage racks are fabricated from polyethylene.

Please clarify the material used for fuel storage racks.

36.

NUREG-1537, Part 1, Section 9, Auxiliary Systems requests a description of any auxiliary systems that are not fully described in other sections, but are important to the safe operation and shutdown of the reactor and to the protection of the health and safety of the public, the facility staff, and the environment. The SAR Chapter 9.2.2.2 discusses the safety characteristics of the three fuel storage pits embedded in the main reactor room. However, the SAR does not provide a technical basis to justify the statements claimed in SAR Chapter 9.2.2.2. Please provide the technical basis (i.e., supporting analysis) for SAR Chapter 9.2.2.2 including information that demonstrates that the effective multiplication factor will be maintained at less than a Keff of 0.8 for all storage orientations, and that the racks are designed to withstand a UBC Zone 3 earthquake when fully loaded.

37.

NUREG-1537, Part 1, Section 9.2, Handling and Storage of Reactor Fuel requests a description of the equipment and administrative procedures used for handling fuel. The SAR Chapter 9 did not describe the overhead crane used to handle fuel into shipping casks, the potential for movement of heavy objects over the reactor, or crane load testing, maintenance and surveillance. Please provide a description of the crane use, any administrative controls used to ensure safe fuel handling, and any crane load testing, maintenance and surveillance requirements.

38.

NUREG-1537, Part 1, Section 9.3, Fire Protection Systems and Programs requests a description of fire prevention, including limiting the types and quantities of combustible materials; methods to detect, control, and extinguish fires; and facility design features and protective systems that exist to ensure a safe reactor shutdown and to prevent the uncontrolled release of radioactive material if a fire should occur. The SAR did not discuss this information. Please provide the following:

38.1. Briefly discuss potential causes and consequences of fires at the facility.

38.2. Discuss fire protection plans and protective equipment used to limit the consequences of a fire, including defense-in-depth in the event of an escalation of a fire.

38.3. List the objectives of the fire protection program and discuss the organizations, methods, and equipment necessary for attaining the objectives.

38.4. Identify all passive designs or protective barriers planned for limiting the consequences of a fire, including features of the facility that could affect a safe reactor shutdown or release radioactive material in the event of a continuing fire.

38.5. Discuss the source of facility fire protection brigades and their training and summarize the more detailed discussions of these personnel and offsite fire protection forces in the facility emergency plan.

38.6. Demonstrate compliance with local and national fire and building codes applicable to fire protection.

39.

NUREG-1537, Part 1, Section 10, Experimental Facilities and Utilization requests a discussion on the potential interactions between the core and the experimental facilities.

The SAR did not provide any information with regard to the potential interactions between the core and experimental facilities as a result of the requested power uprate to 250 kW. Please provide a discussion of any potential interactions between the experimental facilities and the core at 250 kW. Also, include any power uprate required changes to any experimental equipment, procedures, or TS.

40.

NUREG-1537, Part 1, Section 10.1, Summary Description requests a discussion of: 1) the limiting experimental characteristics (e.g., reactivity, contents); 2) monitoring and controlling the experiments and the interaction between the experiment and the reactor control and safety systems; and 3) design requirements for the experiment and the review and approval process. SAR Chapter 10 provides only general descriptions of the experimental facilities. Please provide information on limitations on experiments, monitoring and controlling experiments, and designing and approving processes for experiments.

41.

NUREG-1537, Part 1, Section 10.2, Experimental Facilities requests a description of the experimental safety system and the functional interface between this system and the reactor protection system. SAR Chapter 10 only discusses physical features and does not provide any statements regarding safety, assurance of independence, or compliance with the guidance cited above. Please provide information regarding the experimental safety system and the functional interface between this system and the reactor protection system.

42.

NUREG-1537, Part 1, Section 11.1.1, Radiation Sources requests a description of the models and assumptions for predicting and calculating the dose rates and accumulative doses from Argon-41 (Ar41) and N16.

42.1. SAR Chapter 11.2.1.1 indicates that the pneumatic system uses compressed helium, and estimates the steady-state production of Ar41 and N16. The discussion also states that use of air to drive the pneumatic system also satisfies 10 CFR 20, without showing an analysis. Observations made during the site visit are that this system appears to use compressed nitrogen instead of helium or air.

Please provide confirmation of the pneumatic system gas currently in use, and if appropriate provide an update to the calculation of the Ar41 and N16 production rates.

42.2. During the site visit, an experiment was described using fast neutrons that were generated in a submerged box, voided with air or inert gas, and located near the core. Logically, this void would contribute to the Ar41 and N16 source terms, yet there is no mention of this experiment or other similar experiments in the SAR that would potentially alter the Ar41 and N16 production. Please provide a description of the Ar41 and N16 production rates from experiments conducted in the reactor. If these results are bounded by the steady-state analysis, provide confirmatory models, assumptions, and calculations to support this conclusion.

42.3. SAR Chapters 11.2.1.1 through 11.2.1.3 discuss Ar41 production and the resulting potential radiation doses to workers in the facility. The information is insufficient to permit independent verification and contains inconsistencies regarding equilibrium concentrations for restricted areas, selection of dose conversion factors from the literature, and calculations of resulting facility worker radiation doses. Please provide the details of the calculations of radiation doses to individual workers in the reactor room from airborne Ar41 and N16 during a full range of operations showing that the doses are within the applicable limits of 10 CFR Part 20.

42.4. SAR Chapter 11.2.1.1 discusses Ar41 production and the resulting potential radiation doses to members of the public. The information is insufficient to permit independent verification and contains inconsistencies regarding concentrations for unrestricted areas, selection of locations of the maximum exposed individuals and nearest permanent residence, and calculations of conservative best estimates of radiation doses to members of the public. Please provide the details of the best estimates of calculated radiation doses to members of the public from Ar41 and N16 during a full range of operations showing that the doses are within the applicable limits of 10 CFR Part 20. This information should include conservative best estimates of annual total doses to individuals in unrestricted areas: (1) the maximum exposed individual, (2) the nearest permanent residence, and (3) any location of special interest such as a classroom or campus dormitory.

42.5. SAR Chapter 11.2.1.3 indicates that the Ar41 discharge is from the UUTR facility exhaust stack, which is stated to be 40 feet above ground level. Yet during the site visit, UUTR staff described how the discharge system had recently been modified. Please provide a description of the Ar41 and N16 stack releases that updates the information, to include the actual stack configuration.

43.

NUREG-1537, Part 1, Section 11.1.1, Radiation Sources requests that liquid effluent volumes and radionuclide concentrations are within the limits of 10 CFR Part 20. SAR Chapter 11.2.2 indicates that, under normal operating conditions, there is no liquid released from the reactor pool or the cooling loop. The SAR then states that mop water from reactor room floor cleaning is collected in a sub-grade holding tank, characterized, and released or transferred based on the activity of the water, and that spent liquid samples are also characterized and transferred to Radiological Health for disposal. SAR Chapter 11.3.5.6 indicates that liquid releases are disposed of in accordance with University of Utah regulations and 10 CFR Part 20 release to sanitary sewer, and all releases are accompanied by written procedures and radioactive disposition records.

43.1. Please provide an explanation of how the mop water from the sub-grade holding tank, if discharged, is not considered a liquid release per SAR Chapter 11.2.2, when it is noted as a release in SAR Chapter 11.3.5.6. The information is not included in the annual reports, please provide (1) the activity level of the subgrade holding tank released on an annual basis for the past 6 years; (2) an evaluation of the effluent volumes, radionuclide concentrations, and resulting radiation doses to the unrestricted area consistent with 10 CFR Part 20; and (3) a description of how this source may change with the requested power upgrade.

43.2. Please provide an analysis of the production of tritium in heavy-water elements.

Provide an analysis as to whether tritium should be considered a source term for the UUTR.

44.

NUREG-1537, Part 1, Section 11.1.1, Radiation Sources requests a description of all solid sources of radiation at the facility in sufficient detail to permit an evaluation of all significant radiological exposures related to normal operation, utilization, maintenance, and radioactive management including processing and shipment. SAR Chapter 11.2.3 indicates the solid radioactive sources associated with the UUTR program are summarized in. This text is incomplete (i.e., it is missing the referenced summary).

Please provide the following information:

44.1. The missing referenced summary with an expanded discussion of the solid radioactive sources; 44.2. An evaluation of all significant radiological exposures under normal operation and whether they are controlled under the radiation protection program; and 44.3. If radioactive sources described in the SAR are radioactive waste, how is the waste is transferred, processed, and shipped for disposal under the license.

45.

NUREG-1537, Part 1, Section 11.1.1, Radiation Sources requests a best estimate of the maximum annual dose and the collective doses for major radiological activities during the full range of normal operations for facility staff in order to show that the doses are within the applicable limits of 10 CFR Part 20 and should include estimates of the direct radiation dose rates.

45.1. SAR Chapter 11.2.4.2 provides an equation for estimating the gamma dose rate from the core operating at 250 kW, without providing a definition of or assumed values for the parameters used in performing the dose estimate. The information is insufficient to permit independent verification of the estimated gamma dose rate. Please provide a reference for the equations used to estimate the gamma dose from the core through the pool water. Also provide the assumed parameter definitions and assigned values for performing the calculations for the requested power upgrade.

45.2. SAR Chapter 11.2.4.3 indicates that the gamma dose rate to the room directly above the reactor would be approximately 0.1 mrem/hr, without providing a description of the analysis or measurements used to reach this conclusion.

Please provide the details of the best estimate calculations of the gamma ray radiation dose to an individual in the room directly above the reactor, including all equations, parameter definitions, parameter values, and assumptions for the requested power upgrade. Show that the doses are within the applicable limits of 10 CFR Part 20.

45.3. SAR Chapter 11.2.4.4 provides the estimated gamma dose rates from experimental facilities, including the dry tube irradiator and the pneumatic transfer tube. The information references Appendix B.3 (actually found in Appendix B.8) for details of the calculations. However, the information in Appendix B.8 is simply a table of parameter values by an energy group, with a total dose rate of 0.485 without providing units, or citing equations, models, or methods used to perform the calculations. Please provide the details of the best estimate calculations of the gamma ray radiation dose to an individual from experimental facilities, including all equations, parameters definitions, parameter values, and assumptions if based on modeling, for the requested power upgrade.

Demonstrate that the doses are within the applicable limits of 10 CFR Part 20.

46.

NUREG-1537, Part 1, Section 11.1.2, Radiation Protection Program requests a detailed description of the Radiation Protection Program, as required by 10 CFR 20.1101. SAR Chapter 11.3 provides general statements regarding some of the aspects of the radiation protection program, but lacks sufficient detail. Please provide a description of the UUTR Radiation Protection Program for normal and emergency conditions addressing the following aspects:

46.1. Implementation of the regulations to ensure compliance with the requirements for radiation protection; 46.2. The organizational structure within which the radiation protection program will be administered; 46.3. Interfaces and interrelationships of the radiation protection organization with other facility safety organizations and reactor facility operations; 46.4. The policy governing the program and the allocation of policy-making responsibilities and how the organization, policy, and program are designed for effective radiation protection; 46.5. The radiation protection training program (scope and content) and training requirements for all categories of personnel and visitors; 46.6. Full descriptions of any committees with responsibilities for radiation protection; 46.7. The program for conducting reviews and audits of all functional elements of the radiation protection program; 46.8. The system for evaluating experience, including lessons learned, identification of root causes, and implementation of effective corrective actions; 46.9. The radiation work permit or similar program to control tasks with significant radiation hazards that are not described in the SAR; and 46.10. The recordkeeping program.

47.

NUREG-1537, Part 1, Section 11.1.3, ALARA Program requests a description of the evaluation of provisions for maintaining worker and public doses and radiological releases ALARA, including the ALARA program for the facility; the methods to establish and change policy for the ALARA program; how the program is implemented for all activities at the facility; and, criteria for considering economic factors in the ALARA analysis. SAR Chapter 11.3.1 provided an overview of the ALARA program but lacked sufficient detail. Please provide a description of the ALARA program for the facility and include the methods to establish and change policy for the ALARA program, how this program is implemented for all activities at the facility, criteria for considering economic factors in the ALARA analysis, and a demonstration of management commitment and involvement in the ALARA program.

48.

NUREG-1537, Part 1, Section 11.1.4, Radiation Monitoring and Surveying requests a description of the methods and procedures used for detecting contaminated areas, material, and components, and the methods and procedures used for monitoring exposures of personnel, including those working in radiation and high-radiation areas.

Furthermore, Section 11.1.4 requires recordkeeping in accordance with 10 CFR Part 20.

SAR Chapters 11.3.2, 11.3.3 through 11.3.5.6, and 7.7 provide an overview, but lacks information on the methods and procedures used to detect contaminated areas and materials, and record keeping. Please provide a description of the methods and procedures for detecting radioactive contamination, and the radiation records that document the applicability, quality, and accuracy of monitoring and sampling methods, techniques, procedures, and results.

49.

NUREG-1537, Part 1, Section 11.1.5, Radiation Exposure Control and Dosimetry requests a description of the following information: 1) the design of the facility to prevent uncontrolled radiation releases; 2) the design of entry control devices; 3) the design bases of radiation shielding, ventilation, remote handling, and decontamination equipment; 4) personnel protective equipment and materials used in the facility; 5) administrative radiation exposure and dose limits for all accessible locations; 6) the basis for ALARA limits and how they are enforced; 7) selection of applicable and suitable dosimetry for external radiation monitoring; 8) treatment of measurement uncertainty; and 9) the records maintained to record individual radiation exposures. SAR Chapter 11.3.5 lacked sufficient detail. Please provide a description of the UUTR radiation control and dosimetry as described above.

50.

NUREG-1537, Part 1, Section 11.1.6, Contamination Control requests a description of the following: 1) the scope of the contamination control program; 2) the bases of procedures; 3) provisions within the program to avoid, prevent, and remedy the occurrence and spread of contamination; 4) explicit training on contamination control; and 4) recordkeeping regarding the occurrence and spread of contamination. SAR Chapter 11.3.5.5 lacked sufficient detail. Please provide a description of the UUTR contamination control program as described above.

51.

NUREG-1537, Part 1, Section 11.1.7, Environmental Monitoring requests information on the environmental monitoring program, including: 1) compliance with commitments;

2) a review of the effectiveness of the program; 3) the facility policy on environmental monitoring; 4) written plans and the technical bases of procedures for implementing the environmental monitoring operations; and 5) the environmental surveillance program.

SAR Chapter 11.3.5.6 lacked sufficient detail. Please provide a description of the UUTR environmental monitoring program as described above.

52.

NUREG-1537, Part 1, Section 11.2.1, Radioactive Waste Management Program, requests a description of: 1) the organizational philosophy of and approach to management of radioactive waste; 2) organization of the management function; 3) program staffing, position descriptions, and program personnel with responsibilities for and qualifications of radioactive waste management; 4) review or audit committees related to radioactive waste management; 5) training for staff; 6) plans for shipping, disposal, and long-term storage; 7) audits of the effectiveness of the program and the bases for procedures; and, 8) the bases of the Technical Specifications. SAR Chapter 11.4 lacks sufficient detail. Please provide a description of the UUTR radioactive waste management program as described above.

53.

NUREG-1537, Part 1, Section 11.2.2, Radioactive Waste Controls requests the information concerning: 1) how all processes and procedures that could produce radioactive waste material are evaluated; 2) a demonstration that appropriate monitoring and sampling will be performed and sufficient analyses will be completed to assess the extent of the radiation exposures from waste products; 3) methods to avoid inadvertent exposures of personnel or uncontrolled escapes of the radioactive materials; 4) methods to define and maintain continuous control of radioactive materials that require treatment and management as radioactive waste; and, 5) methods for decreasing quantities of radioactive waste. SAR Chapter 11.4 lacked sufficient detail. Please provide a description of the UUTR radioactive waste controls as described above. Include any changes resulting from the requested power uprate.

54.

NUREG-1537, Part 1, Section 11.2.3, Release of Radioactive Waste requests the following information: 1) a description of the methods used to identify and characterize radioactive liquid and gaseous waste effluents; 2) a listing of the radionuclides by quantity and by other relevant characteristics such as, release points and relevant environmental parameters; 3) a demonstration through appropriate calculations or references that all releases of radioactive effluents would be managed, controlled, and monitored; 4) a demonstration that procedures are in place for the transfer of solid waste to other parties in accordance with all applicable regulations; and, 5) a discussion of methods to verify that releases have not exceeded applicable regulations or guidelines.

SAR Chapter 11.4 lacked sufficient detail. Please provide a description of the UUTR process for the release of radioactive waste as described above.

55.

Technical Specification 6.6 requires all replacements, modifications, and changes to systems having a safety related function shall be subjected to a QA review. The power upgrade of UUTR represents a change that potentially involves safety-related systems and functions. Please provide information regarding the quality measures to be employed to for any changes to systems having a safety related function.

56.

NUREG-1537, Part 1, Section 12.11, Startup Plan requests a complete description of a startup plan. SAR Chapter 12.11 Startup Plan for the power uprate lacked sufficient detail. Please provide a startup plan which addresses the specific information requested in NUREG-1537, Part 1, Section 12.11.

57.

NUREG-1537, Part 1, Section 13, Accident Analyses requests information on safety considerations and functional requirements that ensure safe reactor operation and protection of the public and environment.

57.1. SAR Chapter 13.1.1 and Appendix C.2.1 provide different time estimates for draining the entire reactor tank through a sandy underground media. In Appendix C.2.1 the time estimate is 23.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, whereas in Section 13.1.1 it is 19.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. In addition, Appendix C.2.1 includes information on the properties of the assumed sandy condition under the reactor tank, without proper references or indications as to why the cited values are considered conservative. Please provide the reference for the cited values and clarify the final results with revised information pertaining to these two sections of the SAR.

57.2. SAR Chapter 13.1.3 and Appendix C.2.3 provide information on the method and cite dose rate results from the exposed core. Appendix C.2.3 provides three expressions for estimated decay power, gamma flux, and dose rate. The following issues were observed:

57.2.1. The expression for the decay heat power does not appear to follow any standard expressions in the literature (e.g., Glasstone and Sesonske).

57.2.2. In addition, the discussion in Appendix C.2.3 is very limited and does not provide comprehensive information for the staff to perform a confirmatory analysis (i.e., references for the expressions cited and supporting information on numerical values for the parameters and data).

57.2.3. Furthermore, the dose rate of 8 roentgen per hour (R/hr) from a 1 milligram cobalt - 60 (Co60) contamination of the core structure has been described as occurring at two different distances from the exposed core; Appendix C.2.3 refers to a distance of 24 centimeters, whereas Section 13.1.3 refers to a distance of 18 inches.

Please provide clarifications and references for the expressions cited and the numerical data used in these expressions, and correct the inconsistencies in the cited dose rates.

57.3 SAR Appendix C.2.4 provides an expression for estimating decay heat during a shutdown. This expression is different from the expression in Appendix C.2.3.

Also, it appears that there is an error in the decay heat expression, where the constant factor should be 0.065 rather than 0.65. In addition, in an expression for average temperature increase, the time after the shutdown is referred to as the only variable. However, the fuel surface temperature is expected to rise as a function of time. Therefore, this variable should also be indicated. Please provide references for the stated expressions and update the decay heat equation and calculations related to SAR Appendix C.2.4.

57.4 SAR Appendix C.1.1 provides a set of expressions for reactor power and fuel temperature following a prompt excursion. The results provided in Appendix C.1.1 using the cited reactor parameters were unable to be reproduced, e.g.,

using the methodology in Appendix C.1.1, a prompt excursion of $1.5 resulted in a fuel temperature increase of 40 °C, whereas the table in Appendix C.1.1 indicates an increase of 178 °C. Please provide the input data (reactor parameters) that were used in SAR Appendix C.1.1.

57.5 SAR Chapter 13.3 and Appendices B.9, and C.3 provide the method and data for determining an accidental release of fission products from a mishandling or malfunction of fuel. A confirmatory analysis was unable to be performed. Please provide the following:

57.5.1 Information on the method used to estimate the fission product inventory of a bounding fuel element; 57.5.2 Inventory for the nuclides that are expected to be released in the reactor room, assuming a direct release from a fuel cladding failure in the air; 57.5.3 Nuclide concentrations in the reactor room and comparisons to Annual Limit on Intake (ALI) and Derived Air Concentration (DAC) in 10 CFR 20 Appendix B, Table 1; estimates of the doses to workers during the evacuation; and clarifications of assumptions, if applicable; and, 57.5.4 Method and analysis information for determining dose to the public, indicating the distance at which the dose is calculated, supporting information for assumptions made on parameters such as wind speed, stack height, air flow, and diffusion factor.

58.

NUREG-1537, Part 1, Section, Accident Analyses requests information on safety considerations and functional requirements that ensure safe reactor operation and protection of the public and environment. SAR Chapter 13 provides a summary of accident analyses results supported by Appendices A, B, and C. However, the information in these appendices cannot be confirmed due to the following:

58.1. SAR Appendix C.1 provides an equation for estimating maximum core temperature, Tmax, along with a table summarizing the results. However, the cited numerical estimates and results cannot be reproduced; 58.2. SAR Appendix A.6: a) Ii is undefined; b) (a/)air is cited as a mass attenuation coefficient for air; this should be a mass energy-absorption coefficient; and, 58.3. The values cited for the mass energy-absorption coefficients are low by an order of magnitude.

The Maximum Hypothetical Accident (MHA) for the UUTR is not specifically defined in the SAR. Please provide a description of the MHA, and provide the information noted above to allow independent validation of calculations in SAR Appendix A.6.

59.

NUREG-1537, Part 1, Section 14, Technical Specifications requests TS that will provide reasonable assurance that the facility will function as analyzed in the SAR, without endangering the environment or the health and safety of the public and the facility staff.

59.1. According to the core plan observed during the NRC site visit and described in SAR Chapter 7.4, Fuel Temperature Channels, the UUTR is equipped with two independent instrumented fuel elements. SAR Chapter 4.5.3, Temperature Sensors states that one is located in a fuel element. However, SAR Table 7.4.1 and TS 2.2, Limiting Safety System Settings, references four instrumented fuel elements. Please provide information regarding the number of instrumented fuel elements to be used in the UUTR core, and there locations.

59.2. The current UUTR core contains no Al clad fuel elements located in the B-Hexagonal Ring. However, a review of TS found no restrictions from placing an Al clad fuel element in the B-Hexagonal Ring. In this configuration, it does not appear that appropriate temperature monitoring would be provided. Please provide clarification as to why the C and D rings were selected for fuel temperature monitoring. Please provide clarification as to how the UUTR can assure that the Limiting Safety System Settings (LSSS) criteria specified in TS 2.1 for fuel in the B-and E-Hexagonal Ring locations will not be exceeded.

59.3. TS do not clearly state how to establish the trip setpoints to ensure that operation at power levels will not exceed the licensed power. Please provide the Safety Limit and associated LSSS for licensed power levels for the UUTR.

59.4. TS 3.6, Limitations on Experiments, allows as much as 25 milligrams of explosive materials. Regulatory Guide 2.2, Development of Technical Specifications for Experiments in Research Reactors, references 25 milligrams of TNT equivalent explosive materials. Please provide a basis for the acceptability of the more general statement in the SAR that does not use the phrase, TNT equivalent.

59.5. TS 3.7, As Low As Reasonably Achievable (ALARA) Radioactive Effluent Releases, does not contain any required actions and completion times if the limiting condition for operation (LCO) is exceeded. Please provide the required action statements and requisite completion times for those instances when an LCO is exceeded.

59.6. SAR Table 4.6.2, Safety System Setpoints, provides a list of the SCRAM functions and setpoints, which states the percent power channel setpoint is established at 110% of licensed power (275 kW). However, the table in TS 3.3.3, states that the SCRAM is at 120% of full licensed power. Please provide a description of how the trip setpoints were determined and provide consistent information for the power related SCRAM settings.

59.7. SAR Chapter 7.1.2.1, Water Level Meter, states that the float trips the SCRAM function when the water level drops 24 inches from the top of the tank. However, TS 3.3.3, Reactor Safety System, states that the reactor will SCRAM at one foot below normal operating level. Please provide a description of how the noted trip setpoints were determined and provide the desired trip setting.

59.8. TS 5.7 requires that the reactor shall not be operated unless the water level is at least 18 feet above the top of the reactor core. However, TS 3.8 and 4.5, Primary Coolant Conditions, only describe the LCOs and associated Surveillance Requirements for conductivity and pH. Please provide information demonstrating the consistent and complete application of limitations appropriate to the TS listed.

59.9. SAR Chapter 3.1.2, Protection by Multiple Fission Product Barriers - Criterion 13, describes the instrumentation and control systems that have associated SCRAM capabilities as follows: 1) safety channel (percentage of power with scram); 2) redundant, multi-range, and linear power safety channel (source level to full power with SCRAM); and 3), two fuel temperature channels (C and D rings with scram). Additional SCRAMS are triggered by a loss of pool water level, a high voltage interrupt tripping the external security system, and airborne radiation levels in excess of the high area radiation monitor's limits. However, SAR Chapter 7.5.1, Design Criteria, describes the instrumentation and control systems that have associated SCRAM capabilities as having: 1) Manual and key SCRAMS; 2) Loss of console power SCRAM; and, 3) Loss of high voltage SCRAM.

These section narratives do not match the descriptions in TS 3.3.3, Reactor Safety System, nor does the text match what is stated in TS 2.2, Limiting Safety System Setting. Please provide revised narratives for the noted sections of the SAR and TS.

59.10. SAR Chapter 7.7.1, Design Criteria, states that an alarm will sound at the console and at the continuous air monitor (CAM) if the LSSS for this unit are exceeded. However, there are no LSSS established in the TS for this function.

Please provide safety limits and LSSS that pertain to the CAM.

Additionally, stack CAMs are typically set to alarm based on the concentration of activity in the air that the instrument is monitoring, for the most conservative radionuclide (beta-gamma or alpha or both) identified in the facility characterization. Please clarify whether the stack CAM alarm initiates a signal to close the dampers and change the circulation path for the HVAC system.

59.11. The SAR provides conflicting statements about area radiation monitors (ARMs) and CAMs, e.g., the number of alarm locations, alarm levels, if they support a SCRAM function, if alarm levels are monitored at the reactor console, if TS have been set for ARM or CAM alarms, and if instrument and action levels control the HVAC system when a CAM or ARM alarms. Please provide a description of the number of CAMs and ARMS, locations, trip setpoints, whether or not they are scrammable channels, functions, and any auxiliary systems affected when setpoints are tripped.

59.12. TS 6.10 (5) requires the annual report to list all experiments performed. NRC staff reviewed the UUTR annual reports for the years 2005 through 2009 and found no mention of experiments performed in the reports. The facility staff indicated, during the site visit, that experiments are routinely performed. A review of the inspection reports for the time period 2003 through 2009 revealed that no new experiments had been approved during that time period. Please provide information pertaining to experiments performed during the period of 2003 through 2009. Confirm if any new experiments have been approved for the period 2005 through 2009.

59.13. As specified in NUREG-1537, Part 1, Appendix 14.1, Section 1.2.2, sections of the TS should be numbered as indicated in Section 1.2.2 of ANSI/ANS 15.1. In the Technical Specifications section of the SAR, LCOs are not so numbered.

For example, if an LCO is established in TS 3.3, which provides the safety basis for control rod insertions, it is expected that a TS 4.3 is established that provides the requisite surveillance requirements for the LCO. Please provide TS that have a one-to-one correlation between the LCOs in TS 3.0 with the associated surveillance requirements in TS 4.0.

60.

NUREG-1537, Part 1, Section 16.1, Prior Use of Reactor Components requests information on prior use of items significant to safety, such as fuel cladding, reactivity control system, engineered safety features, and radiation monitoring systems. In the case of the UUTR, this means evaluating the continued serviceability of the originally supplied UUTR components (e.g., for aging and wear, etc.); and also to consider the suitability of items supplied by other facilities.

60.1. The SAR provides a rather detailed explanation of why certain items are appropriate for continued use. However, statements concerning the reactor tank relied on claims about aluminum exposure limits and current exposure levels that are not discussed in the SAR. Furthermore, information on the reactor fuel concerned expected operating temperatures for fuel at 250 kW that in fact are not estimated or characterized in the SAR. Please provide a characterization of allowed exposure limits for the aluminum tank and calculations indicating current exposure of the tank. Provide a characterization of expected fuel operating temperatures at 250 kW.

60.2. SAR Chapter 16.1 states that the ion chambers are deemed adequate for use at an upgraded power level of 250 kW. However, SAR Appendix D.2, Power Monitoring Channels, indicates that the useful range for the percentage power and log power channels is from 100 W to 100 kW and from 100 mW to 100 kW, respectively. Please provide information as to support use of these instruments up to the power uprate limit of 250 kW.